ML20129F773
| ML20129F773 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 09/30/1996 |
| From: | Bajwa S NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20129F779 | List: |
| References | |
| NUDOCS 9610070034 | |
| Download: ML20129F773 (6) | |
Text
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NUCLEAR REEULATORY COMMISSION 2
WASHINGTON, D.C. 20066 0001 5
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VERMONT YANKEE NUCLEAR POWER CORPORATION DOCKET NO. 50-271 VERMONT YANKEE NUCLEAR POWER STATIO_N AMENDMENT TO FACILITY OPERATING LICENSE l
Amendment No. 149 License No. DPR-28 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment filed by the Vermont Yankee Nuclear Power Corporation (the licensee) dated April 4, 1996, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended j
(the Act), and the Comission'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 Comission; 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 compliance with the Comission's regulations; D.
!L. issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and 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-tions as indicated in the attachment to this license amendment, and paragraph 3.B of Facility Operating License No. DPR-28 is hereby amended to read as follows:
9610070034 960930 PDR ADOCK 05000271 P
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.149, 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 C0lWISSION
/
S. Singh Bajwa, Acting Director Project Directorate I-l Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Sp2cifications Pue of issuance: September 30, 1996
i ATTACHMENT TO LICENSE AMENDMERT NO. 149 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 83 83 89 89 90 90 l
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VYNPS 3.3 LIMITING CONDITIONS FOR 4.3 SURVEILLANCE REQUIREMENTS OPERATION positive coupling and the results of each test shall be recorded. The l
drive and blade shall be coupled l
and fully withdrawn. The position and over-travel lights shall be observed.
2.
The Control Rod Drive 2.
The Control Rod Drive Housing Support System.
Housing Support System
- .411 be in place when shall be inspected after
'he Reactor Coolant reassembly and the system is pressurized results of the above atmospheric inspection recorded.
pressure with fuel'in the reactor vessel unless all operable control rods are fully inserted.
3.
While the reactor is 3.
Prior to control rod below 20% power, the Rod withdrawal for startup Worth Minimizer (RWM) the Rod Worth Minimizer shall be operating while (RWM) shall be verified moving control rods as operable by except that performing the following:
I (a)
If after withdrawal (a)
The Reactor of at least 12 Engineer shall control rods during verify that the a startup, the RWM control rod fails, the startup withdrawal sequence may continue for the Rod Worth provided a second Minimizer computer licensed operator is correct.
verifies that the operator at the reactor console is following the control rod program; or (b)
If all rods, except (b) The Rod Worth those that cannot Minimizer be moved with diagnostic test control rod drive shall be performed.
Amendment No. 49,149 83
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l l
VYNPS BASES:
3.3 & 4.3 CONTROL ROD SYSTEM A.
Reactivity Limitations l.
Reactivity Marcin - Core Loadino i
The core reactivity limitation is a restriction to be applied principally to the design of new fuel which may be loaded in the i
core or into a particular refueling pattern. Satisfaction of the i
limitation can only be demonstrated at the time of loading and must be such that it will apply to the entire subsequent fuel cycle. At each refueling the reactivity of the core loading will be limited so the core can be made suberitical by at least R + 0.25% Ak with the highest worth control rod fully withdrawn and all others inserted. The value of R in % Ak is the amount by a
which the calculated core reactivity, at any time in the operating cycle, exceeds the reactivity at the time of the 7
demonstration. R must be a positive quantity or zero. The value of R shall include the potential shutdown margin loss assuming full B C settling in all inverted poison tubes present in the 4
core. The 0.25% Ak is provided as a finite, demonstrable, sub-criticality margin.
2.
Reactivity Marcin - Inocerable Control Rods Specification 3.3.A.2 requires that a rod be taken out of service if it cannot be moved with drive pressure. If a rod is disarmed 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 centrol rods valved out of service will be available to the reactor
)
operator. The number of rods permitted to be inoperable could be i
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 genex-ic 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.
Circumferential cracks resulting from stress assisted 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.
1.
Control rod dropout accidents as discussed in the FSAR can lead to significant core damage. If coupling integrity is raintained, 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 Amendment No. G4, MVY 07 131,109 89
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l VYNPS 1
1.*
3& GEE:
3.3 & 4.3 (Cont *d) 1 coupling integrity since only an uncoupled CRD can reach the i
over-travel position. The verification is required to be performed when a control rod is. fully withdrawn after each refueling outage (since work on the control rod or CRD System may have affected coupling), and after each uncoupling.
I 2.
The control rod housing support restricts the outward movement of l'
a control rod to less than 3 inches in the extremely remote event of a housing failure. The enount of reactivity which could be-added by this saml1 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 4
in Subsection 3.5.2 of the FSAR, and the design evaluation is 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.
i 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 ro6 dropout accidents which might lead to significant core damage, cannot occur if this sequence of rod withdrawals or insertions is 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 startup without having the RWM operable would entail unnecessary risk, continuing to withdraw rods if the RMM fails subsequently is acceptable if a second licensed operator verifies the withdrawal oequence. Continuing the startup increases core power, reducts the rod worth and reduces the consequences of dropping any rod. Withdrawal of rods for testing is permitted 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 anything but minor consequences.
4.
Refer to the Vermont-Yankee Core Performance Analysis report.
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 counts per second assures that any transient should it occur, begins at or above the initial value of 10-e,of rated power used in the analyses of transients from cold conditions. One operable SRM channel is adequate to monitor the approach to criticality, i
therefore, tw6 operable SRM's are specified for added j
conservatism.
j 6.
The Rod Block Monitor (RBM) is designed to automatically prevent fuel damage in the event of erroneous rod withdrawal from loca-cions 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. G6, 49, 64, 44,109 90
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