ML20073C255
| ML20073C255 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 03/28/1983 |
| From: | Vassallo D Office of Nuclear Reactor Regulation |
| To: | Vermont Yankee |
| Shared Package | |
| ML20073C262 | List: |
| References | |
| DPR-28-A-077 NUDOCS 8304130091 | |
| Download: ML20073C255 (11) | |
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k UNITED STATES NUCLEAR REGULATORY COMMISSION D
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i VERMONT YANKEE NUCLEAR POWER CORPORATION DOCKET NO. 50-271 VERMONT YANKEE NUCLEAR POWER STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 77 License No. DPR-28 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Vermont Yankee Nuclear Power Corporation l
(the licensee) dated February 22, 1983 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 compliance with the Comission't regulations; D.
The 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 Comission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifi-cations 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:
B.
Teclinical Specifications The Technical Specifications contained in Appendix A, as revisid through Amendment No. 77 are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
8304130091 83032G PDR ADOCK 05000271 PDR p
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3.
This license amendment is effective as of the date of issuance.,
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FOR THE NUCLEAR REGULATORY COMMISSION A
Domenic B. Vassallo, Chief Operating Reactors Branch #2 1
Division of Licensing
Attachment:
Changes to the Technical Speci fications Date of Issuance:
March 28,1983 b
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ATTACHMENT TO LICENSE AMENDMENT NO. 77 FACILITY OPERATING LICENSE NO. DPR-28 DOCKET NO. 50-271 Revise the Technical Specifications by removing the following pages and inserting identically numbered pages.
183 185 185-a 185-b 186 186-a 187 187-a 1
e 5
1 i
VYNPS 3.12 LIMITING CONDITION FOR OPERATION 4.12 SUkVEILLANCE HEQUIREMENT
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moved and one in an adjacent quadrant.
Fo r a n SRM Thereafter, the SRMs shall be checked daily to be considered operable the f ollowirg conditions f or response.
shall be satist f ed:
1.
The SRM shall be inserted to the normal operating level.
(Use of special movable, dunkirs type detectors during initial f uel loading and major core alternations in place of normal detector is permissible as long as the detectors is connected into the proper circuitry which contain the required rod bloc ks).
"2 The SRM shall have a minimum of 3 cps with all rods fully inserted in the core.
3.
Pr. lor to spiral unloadirg, the SRMs shall be Prior to spiral unloadirg or Eeloading, t'I4 3.12.5.1 and 3.12.B.2 a bove, however, during SRMs shall be functionally tested. Prior to proven operable as stated in Sections spiral reloading, t he SRMs shall be checked spiral unloading the count rate may drop f or neutron response, below 3 cps.
4 Prior to spiral reloading, two diagonally adjace nt fuel assemblies, which have previously accumulated exposure in the reactor, shall be loaded into their designated core positions next to each of the I
4 SRMs to obtain the required 3 eps.
Until these eight bundles have been loaded, the 3 cps requirement is not necessary, C.
Fuel Storage' Pool Water Level C.
Fuel Storage Pool Water Level Whenever irradiated fuel is stored in the Whenever irradiated f uel is stored in the f uel storage pool the pool water level shall f uel storage pool, the pool level shall be be maintained at a level of at least 36 f eet, recorded daily.
183 n-enaman+ % am 77
VYNPS 3.12 LIMITING CONDITION FOR OPERATION 4.12 SURVEILLANCE REQUIREMENT k
1.
T)g " reactor mode switch shall be locked in 1.
This surveillance requirement is the same as the Refuel" position. The refueli ng that given in Specification 4.12.A.
Interlock which prevents more than one control rod f rom bei ng wi t hd rawn may be bypassed on a withdrawn control rod af ter the fuel assemblies in the cell containing (controlled by) that control rod have been removed from the reactor core. All other refueling interlocks shall be operable.
2.
SRMs shall be operable in the core quadrant 2.
This surveillance requirement is the same as where fuel or control rods are being moved, that given in Specification 4.12.B
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and in an adjacent quadrant. The requirements f or 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:
a.
Prior to spiral unload and reload, the SRMs shall be proven operable as stated in Specification 3.12.B1 and 3.12.B2.
However, during spiral unloading, the count rate may drop below 3 eps.
b.
The core may be spirally reloaded to either the original configuration or a dif ferent configuration in the reverse sequence of that used to unload, with the exception that two (2) diagonally adjacent fuel assemblies, which have previously accumulated exposure in the reactor, shall be loaded into their designated core positions next to each
- of the four (4) SRMs to obtain the required 3 cyn.
Until these eight (8) bundles have been loaded, the 3 cps j
requirement is not neces sa ry.
185 Amondment Nn. 5977
.VYNPS 3.12 LIMITING CONDITION FOR OPERATION 4.12 SURVEILLANCE REQUIREHENT
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-# t Following insertion of the initial eight (8) bundles, the reactor will be spirally reloaded from the center cell outwards, until the core is fully loaded.
c.
At least 50% of the fuel assemblies to be reloaded into the core shall have previously accumulated a minimum exposure of 1000 Mwd /T.
F.
Fuel Movement F.
Fuel Movement Fuel shall not be moved or handled in the Prior to any fuel handling or movement in the reactor core for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following reactor reactor core, the licensed operator shall
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shutdown to cold shutdown conditions.
verify that the reactor has been in the cold shutdown condition for a minimum of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
G.
Crane Operability C.
Crane Operability 1.
The Reactor Building crane shall be l.
a.
Within one month prior to spent operable when the crane is us d for fuel cask handling operations, an handling of a spent fuel cask, inspection of crane cables, sheaves, hook, yoke, and cask lifting trunnions will be made.
These inspections chall meet the requirements of ANSI Standard 185-a.
Amendment
- o. 59 77
VYNPS 3.12 LIMITING CONDITION FOR OPERATION 4.12 SURVEILLANCE REQUIREMENT B30.7, 1967. k crane rope shall be replaced if any of the replacement criteria given in ANSI B30.2.0-1967 are met, b.
No-load mechanical and electrical tests will be conducted prior to lif ting the empty cask from its transport vehicle to verify proper operation of crane controls, brakes and lif ting speeds. A functional test of the crane brakes will be conducted each time an empty cask is lifted clear of its transport vehicle.
2.
Crane Travel 2.
Crane Travel Spent fuel casks shall be prohibited Crane travel limiting mechanical stops i
f rom travel over irradiated f uel shall be installed on the crane trolley a ssembl ie s.
rails prior to cask handling operations to prohibit cask travel over irradiated l
fuel assemblies.
H.
Spent Fuel Pool Water Temperature H.
Spent Fuel Pool Water Temperature Whenever irradiated f uel is stored in the Whenever irradiated fuel is in the spent fuel spent fuel pool, the pool water temperature pool, the pool water temperature shall be l
shall be maintained below 1500F.
recorded daily. If the pool water temperature reaches 1500F, all ref ueling operations tending to raise the pool water
. temperature shall cease and measures taken immediately to reduce the pool water I
temperature below 1500F.
ISS-b Amendment No. 5p 77 i
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VYNPS Bases:
3.12 & 4.12 REFUELING A.
During refueling operations, the reactivity potential of the core is being altered. It is necessary to require certain interlocks and restrict certain refueling procedures such that there is assurance that inadvertent critidelity does not occur.
To minimize the possibility of loading fuel into a cell containing no control rod, it is required that all control rods are fully inserted when fuel is being loaded into the reactor core. This requirement assures that during refueling the refueling interlocks, as designed, will prevent inadvertent criticality. The core reactivity limitation of Specification 3.2 limits the core alterations to assure that the resulting core loading can be controlled with the Reactivity Control System and interlocks at any time during shutdown or the following operating cycle.
The addition of large amounts of reactivity to the core is prevented by operating procedures, which are in turn backed up by refueling interlocks on rod withdrawal and movement of the refueling platform. When the mode switch is in the " Refuel" position, interlocks prevent the refueling platform f rom being moved over the core if a control rod is withdrawn and fuel is on a hoist.
Likewise, if the refueling platform is over the core with fuel on a hoist, contral rod motion is blocked by'the
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4 interlocks. With the mode switch in the refuel position, only one control rod can be withdrawn.
R.
The SRMs are provided to monitor the core during periods of station shutdown and to guide the operator during refueling operations and station startup. Requiring two operable SRMs in or adjacent to any core quadrant where fuel or control rods are being moved assures adequate monitoring of that quadrant during such alterations. The requirement of 3 counts pet second provides assuranc'e that neutron flux is being monitored. Under the special condition of complete spiral core unloading, it is expected that the count rate of the SRMs will drop below 3 cps before all the fuel is unloaded. Since there will be no reactivity additions, a lower number of counts will not present a hazard. When all of the fuel has been removed to the spent fuel storage pool, the SRMs will no longer be required. Requiring the SRMs to be operational prior to fuel removal assures that the SRMs are operable and can be relied on even when the count rate may go below 3 cps, t
j Prior to spiral reload, two diagonally adjacent fuel assemblies, which have previously accumulated exposure in j
the reactor, will be loaded into their designated core positions next to each of the 4 SRMs to obtain the required 3 cps. Exposed fuel continuously produces neutrons by spontaneous fission of certain plutonium isotopes, photo fission, and photo disintegration of deuterium in the moderator. This neutron production is normally great enough to meet the 3 cps minimum SRM requirement, thereby providing a means by which SRM response may be demonstrated before the spiral reload' begins. During the spiral reload, the fuel will be loaded in the I
reverse sequence that it was unloaded with the exception of the initial eight (8) fuel assemblies which are loaded next to the SRMs to provide a means of SRM response.
186 Amendment
- o. pp 77 m
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VYNPS 3.12 & 4.12 (Continued)
C.
To assure that there is adequate water to shield and cool the irradiated 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).
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i 186-a i
Amendment No. 59 77
VYNPS 3.12 & 4.12 (Continued)
D.
During certain periods, it is desirable to perf orm maintenance on two control rods and/or control rod drives at the same time. This specification provides assurance that inadvertent criticality does not occur during such mai nte na nce. (
.< t The maintenance is perf ormed with the mode switch in the " Refuel" position to provide the refueling interlocks normally available during refueling operctions as explained in Part A of these Bases. In order to withdraw a second control rod af ter withdrawal of the first rod, it is necessary to bypass the refueling interlock on the first control rod which prevents more than one control rod f rom being withdrawn at the same time. The requirement that an adequate shutdown margin be demonstrated with the control rods remaining in service ensures that inadve rt ent criticality cannot occur during this maintenance. The shutdown margin is verified by demonstrating that the core is shut down even if the strongest control rod remaining in service is fully withdrawn. 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 significant portion of the reactor core for I
such purposes as inservice inspection requirements, examination of the core support plate, etc.
This specification provides assurance that inadvertent criticality does not occur duri ng such operation.
This operation is performed with the mode switch in the " Refuel" position to provide the refueling interlocks normally available during refuelirs 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 f, rom 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 f uel in the outermost cells (f our fuel bundles surrounding a control rod) is removed first. Unloading continues by unloading the remaining outermost fuel by cell spiralling inward towards the I
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 I
spirals outward until the core is fully loaded. Additionally, at least 50% of the fuel assemblies to be 181 Amendmen5No.5977
VYNPS 3.12 & 4.12 (Continued) reloaded into the core shall have previously accumulated a minleum exposure of 1000 Mwd /T to ensure the presence of a minimum neutron flux as described in Barcs Section 3.12.B.
i The 1 tpnt of this specification is to assure that the reactor core has been in the cold shut down condition for F.
9 at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> f ollowing power operation and prior to fuel hanJ11ng or movement. The safety analysis for the postulated refuellog accident assumed that the reactor had been shut down for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for fission product decay prior to any fuel handling which could result in dropping of a fuel assembly.
C.
The operability requirements of the reactor building crane ensures that the redundant features of the crane have been adequately inspected just prior to using it for handling of a spent fuel cask. The redundant hoist system ensures that a load will not be dropped for any postulated credible single component f ailures. Details of the design of the redundant features of the crane and specific testing requirements for the crane are delineated in the Vermont Yankee document entitled " Reactor Building Crane Modification" (December 1975).
H.,
The Spent Fuel Pool Cooling System is designed to maintain the pool water temperature below 1250F durirg normal refueling operations.
If the reactor core is completely discharged, the temperature of the pool water may increase to greater than 1250F. The RHR System supplemental fuel pool cooling may be used under these conditions to maintain the pool water temperature less than 150 F.
187-a Amendmert Ao. p 77
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