ML20134N937
| ML20134N937 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 11/19/1996 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20134N940 | List: |
| References | |
| NUDOCS 9611270162 | |
| Download: ML20134N937 (3) | |
Text
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3.3 & 4.3 CONTROL ROD SYSTEM A.
Reactivity Limitations 1.
Reactivity Haroin - Core Loadina j
i The specified~ shutdown margin (SDM) limit accounts-for the j
uncertainty in the demonstration of SDN by testing. Separate SDM limits are provided for testing where the highest worth control
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rod is determined analytically or by measurement. This is due to the reduced uncertainty in the SDM test when the' highest worth control rod is determined by measurement (e.g., SDN may be 4
demonstrated by an in-sequence control rod withdrawal, in which j
the highest worth control rod is analytically determined, or by j
local criticals, where the highest worth rod is determined by I
testing).
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Following a refueling, adequate SDM must be demonstrated to ensure that the reactor can be made subcritical at any point i
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, i
which is the difference between the calculated value of f
'R*,
l maximum core reactivity during the operating cycle and the I
calculated BOC core reactivity.
If the value of "R" is negative j
(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 j
B C settling in all inverted poison tubes. present in the core.
i 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 4
5 provide a reasonable amount of time to perform the required calculations and have appropriate verification.
I When SDM is demonstrated by calculations not associated with a l
to confirm SDM during the tuel loading sequence),
test (e.g.,
additional margin must be included to account for uncertainties I
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 L'
is required to ensure adequate SDM is maintained during core) 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 SDN for the most reactive configurations during the w
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refueling may be performed to demonstrate acceptability of the These bounding analyses include entire fuel movement sequence.
additional margins to account for the associated uncertainties Ln w*
the calculation.
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$2 2.
Reactivity Marcin - Inonerable Control Rods
-g Specification 3.3. A.2 requires that a rod be taken out of service O
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if it cannot be, moved with drive pressure.
If a rod is disarmed electrically, its position shall be consistent with the shutdown This reactivity limitation stated in Specification 3.3.A.l.
assures that the core can be shutdown at all times with the remaining control rods, assuming the highest worth, operable etLn.
control rod does rod insert.
An allowable pattern for control rods valved out of service will be available to the reactor The number of rods permitted to be inoperable could be operator.
89 Amendment No. 44, :n"? f 7 131.148
VYNPS y
ESES:
3.3 & 4.3 (Cont'd) many more than the six allowed by the Specification, particularly late in the operation cycle, 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 affecting a number of drives cannot be ruled out.
I Circumferential cracks resulting from stress assisted intergranular corrosion have occurred in the collet housing of drives at several BWRs.
This type of ; racking could occur in a j
number of drives and if the cracks propagated until severance of j
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 to significant core damage.
If coupling integrity is maintained, the possibility of a rod dropout accident is eliminated.
Neutron instrumentation response.to rod movement provides a verification that the rod is following its drive.
Coupling verification is performed to ensure the control rod is connected to'the control rod drive mechanism and will perform its intended function when necessary. The surveillance requires verifying a control rod does not go to the withdrawn over-travel position.
The over-travel position feature provides a positive check on the coupling integrity since only an uncoupled CRD can reach the over-travel position. The verification is required to be
-i performed when a control rod is fully withdrawn after each
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refueling outage (since work on the control rod or CRD System may have affected coupling), and after each uncoupling.
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89a Amendment No. 146, 149
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VYNPS E!,ggs, 3. 3
- t. 4.3 (Cont'd) i 2.
The control rod housing support restricts the outward movement of a control rod to less than 3 inches in the extremely remote event of a housing failure.
The amount of reactivity which could be added by this small amount of rod withdrawal, which is less than a normal single withdrawal increment, will not contribute to any damage of the primary coolant system.
The design basis is given in Subsection 3.5.2 of the FSAR, and the design evaluation is 3
l.
given in Subsection 3.5.4.
This support is not required if the reactor coolant system is at atmospheric pressure since there would then be no driving force to rapidly eject a drive housing.
3.
In the course of performing normal startup and shutdown procedures, a pre-specified sequence for the withdrawal or insertion of control rods is followed.
Control rod dropout accidents which might lead to significant core damage, cannot occur if this sequence of rod withdrawals or insertions is j
followed. The Rod Worth Minimizer restricts withdrawals and insertions to those listed in the pre-specified sequence and provides an additional check that the reactor operator is following prescribed sequence. Although beginning a reactor j
startup without having the RWM operable would entail unnecessary risk, continuing to withdraw rods if the RWM fails subsequently 1
is acceptable if a second licensed operator verifies tt.e withdrawal sequence.
Continuing the startup increases core i
power, reduces the rod worth and reduces the consequences of j
dropping any rod.
Withdrawal of rods for testing is permitted j
-with the RWM inoperable, if the reactor is suberitical and all other rods are fully inserted.
Above 20% power, the RWM is not needed since even with a single error an operator cannot withdraw a rod with sufficient worth, which if dropped, would result in i
j anything but minor consequences.
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4.
Refer to the Vermont Yankee Core Performance Analysis report.
i 5.
The Source Range Monitor (SRM) system has no scram functions.
It does provide the operator with a visual indication of neutron level. The consequences of reactivity accidents are a function of the initial neutron flux. The requirement of at least three l
countspersecondassuresthatanytransieng,ofratedpowerused should it occur, begins at or above the initial value of 10-in the analyses of transients from cold conditions.
One operable SRM channel is adequate to monitor the approach to criticality, i
therefore, two operable SRM's are specified for added
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conservatism.
4 6.
The Rod Block Monitor (RBM) is designed to automatically prevent fuel damage in the event of erroneous rod withdrawal from locations of high power density during high power level operation.
During reactor operation with certain limiting control rod patterns, the withdrawal of a designated single control rod could result in one or more fuel rods with MCPR less than the fuel cladding integrity safety limit.
During use of such patterns, it is judged that testing of the RBM system prior to withdrawal of such rods will provide added assurance that improper withdrawal does not occur.
It is the responsibility of the Nuclear Engineer to identify these limiting patterns and the designated rods either when the patterns are initially established or as they develop due to the occurrence of inoperable control rods.
Amendment No. 45, 49, G+,
70 90
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