ML19339A383
| ML19339A383 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 10/09/1980 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19339A379 | List: |
| References | |
| NUDOCS 8011030700 | |
| Download: ML19339A383 (9) | |
Text
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UNITED STATES O I'
i "uc'e^a aEGULATORY COMMISSION f
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. E W ASHING TON, D. C. 20555 f
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CONNECTICUT LIGHT AND POWER COMPANY THE HARTFORD ELECTRIC LIGHT COMPANY WETTERN MASSACHUSETTS ELECTRIC COMPANY NORTHEAST NUCLEAR ENERGY COMPANY DOCKET NO. 50-245 MILLSTONE NUCLEAR POWER STATION UNIT NO. 1 AMENDMENT TO PROVISIONAL OPERATING LICENSE Amendment No.69 License No. DPR-21 1.
The Nuclear Regulatory Comission (the Comission? has found that:
A.
The application for amendment by Connecticut Light and Power Company, The Hartford Electric Light Company, Western Massachusetts Electric Company and Northeast Nuclear Energy Company (the licensees) dated September 4,1980, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (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.
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 requirw ' have been satisfied.
8011030 700
o 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and Paragraph 3.8 of Provisional Operating License No. DPR-21 is hereby amended to read as follows:
B.
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 69, are hereby incorporated in the license.
Northeast Nuclear Energy Company shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION i
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. Crutchfield, Ch Operating Reactors Branch e5 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: October 9, 1980 l
l
i ATTACEMENT TO LICENSE AMENDMENT N0.69__
PROVISIONAL OPERATING LICENSE NO. DPR-21 DOCKET NO. 50-245 Replace the attached pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by the captioned amend-ment number and contain vertical lines indicating the area of change.
i PAGES 3/4 3-3 1
3/4 10-1 3/4 10-2 l
l B 3/410-1 B 3/4 10-2 i-i l
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f Overleaf page 3/4 3-4 is included for document completeness.
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LIMITING CON 0! TION FOR OPERATION SURVEILLANCE REQUIDF"ENT 3.3.8 Control Rod Withdrawal 4.3.B Control Rod Withdrawal r
3.
Whenever the reactor is in the startup
- 3. (a) To consider the rod worth minimizer or run mode below 20% rated thermal operable, the following steps must power, no control rods shall be moved be performed:
unless the rod worth minimizer is operable or a second independent (1) The control rod withdrawal operator or engineer verifies that sequence for the rod worth the operator at the reactor console minimizer computer shall be is fol?9 wing the control rod pro-verified as correct, gram. The second operator may be used as a substitute for an inoper-(11) The rod worth minimizer compute able rod worth minimizer during line diagnostic test shall be a startup only if the rod worth successfully completed.
minimizer fails after withdrawal of at least twelve control rods.
(iii) Proper annunciation of the select error of at least one 4.
Control rods shall not be withdrawn out-of-sequence control rod in 4
l for startup or during refueling unless:
each fully inserted group shall at least two source range channels be verified.
2 have an observed count rate equal to or greater than three counts *
(iv) The rod block function of the rod I
per second ; or all fuel bundles have worth minimizer shall be verified been removed from the reactor vessel, by attempting to withdraw an out-of-sequence control rod be-yond the block point.
(b) If the rod worth minimizer is inoperable while the reactor is in the startup or run mode below 10% rated thermal parar, and a second independent operator or engineer is being used, he shall verify that all rod positions are correct prior j
tc commencing withdrawal of each rod j
- group, i
knendment No. 72, #9, 69 3/4 3-3
- A aURV[lLLANCE REQUIRIMENT LIM [ TING CONDITION FOR OPLRATION I
4.
Prior to control rod withdrawal for 5.
During operation with limiting control startup or during refueling, verify rod patterns, as detennined by the that at least two source range thannels reactor engineer, either:
have an observed count rate of at l
least three counts per second.
Both RUM channels shall be operable; 4.
or 5.
When a limiting control rod pattern exists, an instrument functional test b.
Control rod withdrawal shall be of the RBM shall be performed in ior blocked; or to withdrawal of the designated rod (s) and daily thereafter The operating power level shall be c.
limited so that the MCPR will C.
Scram insertion Times remain above 1.06 assuming a single error that results in complete During cach operating cycle, each opeiable withdrawal of any single operable control rod shall be subjected Io scram control rod.
time tests frun the fully withdrawn position.
if testing is not accomplished during Scram Insertion Times reactor power operation, the measured C.
c l'
scram insertion times shall be extrapolated 1.
The average scram insertion time, based to the reactor power operation condition on the deenergization of the scram pilot utilizing previously detennined correlat ions.
valve solenoids as time zero, of all operable control rods in the reactor power operation condition shall be no greater than:
% Inserted From Average Scram Fully Withdrawn Insertion Times (jtecl S
0.375 20 0.900 1
50 2.000 90 3.500 J
a 3/4 3-4 Amendment No.16, 47
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 4.10 REFUELING AND SPENT FUEL HANDLING 3.10 REFUELING AND SPENT FUEL HANDLING Applicability:
Applicability:
Applies to the periodic testing of those inter-Applies to fuel handling, core reactivity limitations, locks and instruments used during refueling and spent fuel handling.
and spent fuel handling.
Objective:
Objective:
To assure core reactivity is within capability of the To verify the operability of instrumentation and control rods, to prevent criticality during refueling, interlocks used in refueling and spent fuel and to assure safe handling of spent fuel casks.
handling.
Speci fica tion:
Spect fication:
A.
Refueling Interlocks A.
Refueling Interlocks The reactor mode switch shall be locked in the Prior to any fuel handling, with the head
" Refuel" position during core alterations and the off the reactor vessel, the refueling inter-refueling interlocks shall be operable.
locks shall be functionally tested. They shall also be tested at weekly intervals B.
Core Monitoring thereafter until no longer roquired and following any repair work associated with During core alterations two SRM's shall be oper-the interlocks.
able, one in the core quadrant where fuel or control rods are being moved and one in an B.
Core Monitoring l
adjacent quadrant, except as specified in Para-graphs 3 and 4 below.
For an Sim to be considered 1
Prior to making any alterations to the core, operable, the following conditions shall be the SRM's shall be functionally tested and satisfied:
checked for neutron response. Therea f ter,
1.
The SRM shall be inserted to the normal the SRM's will be checked daily for response operating level.
(Use of special moveable when core alterations are being made.
l dunking type detectors during fuel loading or major core alterations in 2.
Prior to spiral unloading or reloading, the place of normal detectors are permissible SRM's shall be functionally tested.
Prior as long as the detector is connected into to spiral unloading, the SRM's should also the normal SRM circuit.)
be checked for neutron response.
Amendment No. / 69 3/4 10-1
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 2.
The SRM shall have a minimum neutron C.
Fuel Storage Pool Water Level induced count rate of three per second with a61 rods fully inserted in the core.
Whenever irradiated fuel is stored in the fuel _ storage pool, the pool level shall 3.
Prior to unloading, the SRM's shall be be recorded daily.
1 proven operable as stated above, however, during spiral unloading, the count rate D.
Crane Operability may drop below 3 cps.
Within 4 days prior to Spent Fuel Cask 4.
Special movable dunking type detectors will handling operations, a visual inspection of be inserted into the core, prior to reload-crane cables, sheaves, hook, yoke, and cask ing fuel assemblies into the central core lifting trunnions will be made. Following region (with all control rods inserted).
these inspections, no-load mechanical and Bafore the ninth fuel assembly is loaded electrical tests will be conducted to into the core in the close proximity of verify proper operation of crane controls, the movable dunking chambers or the SRM s brakes and lifting speeds. A load test Paragraph 3.10.B.1 and 2 apply.
will then be conducted by lifting the enpty cask C.
Fuel Storage Pool Water Level out of the pivot cradle. The above inspections and pre-lifting procedure shall meet the Whenever irradiated fuel is stored in the fuel requirements of ANSI Standard B30.2,1967.
storage pool, the pool water level shall be maintained at a level greater than or equal to E.
Crane Interlocks and Switches 33 feet.
Crar.e interlocks and limit switches which prevent D.
Crane Operability crane travel over irradiated fuel assemblies shall be demonstrated OPERABLE within seven The 110-ton redundant crane shall be operable days prior to handling of all spent fuel when the crane is used for handling of a spent casks and every seven days thereafter during fuel cask.
spent fuel cask handling.
E.
Crane Travel With a Soert Fuel Cask Spent fuel casks shall be prohibited from travel over irradiated fuel assemblies. When handling a spent fuel cask, the crane mode switch shall be in the " Mode 2" position and the mode switch key removed.
Amendment No. 27 69 3/4 10-2
3.10 Bases A.
Refueling Interlocks 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 criticdlity does not occur.
To minimize the possibility of loading fuel into a cell containing no control red, 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 cere 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.
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 from 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 the fuel on a hoist, control rod motion is blocked by the interlocks. With the mode switch in the refuel position, only one control rod can be withdrawn.
For a new core, the dropping of a fuel assembly into a vacant fuel location adjacent to a withdrawn control rod does not result in an excursion or a critical configuration, thus, adequate margin is provided.
B.
Core Monitoring The SRM's 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 SRM's, one in and one 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 three neutron induced count per second provides assurance that neutron flux is being monitored.
During unloading, it is not necessary to maintain 3 cps because core alterations will involve only reactivity removal and will not result in criticality.
During the loading of an empty sourceless core, the special movable dunking type fission detectors will not detect a neutron count because of the lack of neutron sources. Therefore, fuel must be placed in the core to establish a neutron count rate. The restriction of eight fuel assemblies will minimize the probability of an inadvertent criticality prior to achieving 3 cps, while providing the flexibility to load fuel bundles araund the portable detectors and SRM's.
Amendment No. 69 83/410-1
C.
Fuel Storage _ Pool Water Level 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. The minimum water level of 33 feet is established because it would be a significant change from the normal level (37' 9"), well above a level to assure adequate cooling (just above active fuel).
t D.
Crane Operability The operability requirements of the crane used for handling of spent fuel casks ensures that the re-dundant features of the crane have beer. adequately inspected. The redundant hoist system ensures that a load will not be dropped for all postulated credible single-component failures.
E.
Crane Travel The restriction of movement of spent fuel casks over irradiated fuel ensures (in addition to the redundancy features) that a cask cannot be dropped on irradiated fuel assemblies.
Amenciment No. J4' 69 8 3/4 10-2
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