ML20236M974
| ML20236M974 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 07/10/1998 |
| From: | Chandu Patel NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20236M976 | List: |
| References | |
| NUDOCS 9807140346 | |
| Download: ML20236M974 (11) | |
Text
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UNITED STATES g
g NUCLEAR REGULATORY COMMISSION
%,...../e WASHINGTON, D.c. 20E66 4001 ENTERGY OPERATIONS. INC.
DOCKET NO. 50-382 WATERFORD STEAM ELECTRIC STATION. UNIT 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 144 License No. NPF-38 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Entergy Operations, Inc. (the licensee) dated March 27,1997, as supplemented by letters dated April 3, July 21, October 23, November 13, and December 12,1997, January 21, January 29, March 23, May 1, May 19, May 21, May 28, and June 12,1998, complies with the standards and requirements of the Atomic Energy Act of1954, 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 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 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.
9807140346 980710 DR ADOCK 05000382 p
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2 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C(2) of Facility Operating License No. NPF-38 is hereby amended to read as follows:
(2)
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No.144, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of its date of issuance to be implemented no later than completion of the reracking modification or prior to the movement of cycle 9 fuel assemblies into the Cask Storage Pit or Spent Fuel Pool.
FOR THE NUCLEAR REGULATORY COMMISSION
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Chandu P. Patel, Project Manager Project Directorate IV-1 Division of Reactor Projects lil/IV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of issuance: July 10,1998
ATTACHMENT TO LICENSE AMENDMENT NO. 144 TO FACILITY OPERATING LICENSE NO. NPF-38 1
DOCKET NO. 50-382 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 areas of change. The corresponding overleaf pages are also provided to maintain document completeness.
REMOVE PAGES INSERT PAGES 3/4 9-7 3/4 9-7 55 55 5-6 5-6 54a 5-6b 5-6c B 3/4 9-2 B 3/4 9-2 l
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' REFUFlING OPERATIONS 3/4.9.7 CRANE TRAVEL - FUEL HANDLING BUILDING LIMF mlG CONDITION FOR OPERATION 3.9.7 Cranes in the fuel handling building shall be restricted as follows:
a.
The spent fuel handling machine shall be used* for the movement of fuel assemblies (with or without CEAs) and shall be OPERABLE with:
1.
A minimum hoist capacity of 1800 pounds, and 2.
An overload cutoff limit of less than or equal to 1900 pounds, and, b.
Loads in excess of 2000 pounds shall be prohibited from travel over irradiated fuel assemblies in the Fuel Handling Building.
APPLICABILITY: During movement of irradiated fuel assemblies in the fuel handling building, or with irradiated fuct assemblies in the Fuel Handling Building.
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ACTION:
a.
With the spent fuel handling machine inoperable, suspend the use of the spent fuel handling machine for movement of fuel assemblies and place the crane load in a safe position.
b.
With loads in excess of 2000 pounds over irradiated fuel assemblies in the Fuel Handling Building, place the crane load in a safe position.
c.
The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS l
4.9.7.1 The spent fuel handling machine shall be demonstrated CPERABLE within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> l
prior to the start of fuel assembly movement and at least once per 7 days thereafter by i
performing a load test of at least 1800 pounds and demonstrating the automatic load cutoff l
when the hoist load exceeds 1900 pounds.
4.9.7.2 The electrical interlock system which prevents crane main hook travel over irradiated fuel assemblies in the Fuel Handling Building shall be demonstrated OPERABLE within 7 days prior to crane use and at least once per 7 days thereafter during crane operation.#
4.9.7.3 Administrative controls which prevent crane auxiliary hook travel with loads in excess of 2000 pounds over the irradiated fuel assemblies in the Fuel Handling Building shall be enforced i
during crane operations.
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- Not required for movement of new fuel assemblies outside the spent fuel pool and Cask Storage Pit.
- The crane main book electrical interlocks may be replaced with administrative controls for activities associated with the implementation of the spent fuel rerack project being performed prior to Refueling Outage 9. While these administrative controls are in place, the performance of SR 4.9.7.2 is not required.
l WATERFORD - UNIT 3 3/4 9-7 AMENDMENT NO. 6,144 l
DESIGN FEATURES 5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The reactor core shall contain 217 fuel assemblies with each fuel assembly containing a maximum of 236 fuel rods clad with Zircaloy-4. Each fuel rod shall have a nominal active fuel length of 150 inches and contain a nominal total weight of 1830 grams uranium. The initial core loading shall have a maximum enrichment of 2.91 weight percent U-235. Reload fuel shall be similar in physical design to the initial core loading. Assemblies shall have a maximum nominal i
enrichment of 5.0 weight percent U-235 and meet the final storage requirements described in Section 5.6.
CONTROL ELEMENT ASSEMBLIES 5.3.2 The reactor core shall contain 83 full-length and 8 part-length control element assemblies.
5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE
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5.4.1 The Reactor Coolant System is designed ar.d shall be maintained; a.
In accordance with the code requirements specified in Section 5.2 of the FSAR with allowance for normal degradation pursuant of the applicable Surveillance Requirements, b.
For a pressure of 2500 psia, and c.
For a temperature of 650 F, except for the pressurizer and surge line which is 700*F.
VOLUME 5.4.2 The total water and steam volume of the reactor coolant system is 11,800 +600, -0 cubic feet at a nominal T of 582.1 *F.
m 5.5 METEOROLOGICAL TOWERS LOCATION 5.5.1 The primary and backup meteorological towers shall be located as shown on Figure 5.1-1.
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WATERFORD - UNIT 3 5-5 AMENDMENT NO. 408,144
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5.6 FUEL STORAGE CRITICALITY 5.6.1 The spent fuel storage racks are designed and shall be maintained with:
l A normal k,, of less than or equal to 0.95 when flooded with unborated water, a.
which includes a conservative allowance for uncertainties.
b.
A nominal a0.185 inch center-to-center distance between fuel assemblies placed in Region 1 (cask storage pit) spent fuel storage racks.
A nominal 8.692 inch center-to-center distance between fuel assemblies in the c.
Region 2 (spent fuel pool and refuelling canal) racks, except for the four southern-most racks in the spent fuel pool which have an increased N-S center-to-center nominal distance of 8.892 inches.
d.
New or partially spent fuel assemblies may be allowed unrestricted storage in Region 1 racks.
New fuel assemblies may be stored in the Region 2 racks provided that they are e.
stored in a " checkerboard pattern" as illustrated in Figure 5.6-1.
f.
Partially spent fuel assemblies with a discharge burnup in the "accceptable range" of Figure 5.6-2 may be allowed unrestricted storage in the Region 2 racks.
Partially spent fuel assemblies with a discharge burnup in the " unacceptable range" g.
of Figure 5.6-2 may be stored in the Region 2 racks provided that they are stored in a " checkerboard pattem", as illustrated in Figure 5.6-1, with spent fuel in the
" acceptable range" of Figure 5.6-3.
- 5.6.2 The k,, for new fuel stored in the new fuel storage racks shall be less than or equal to 0.95 when flooded with unborated water and shall not exceed 0.98 when aqueous foam moderation is assumed.
DRAINAGE 5.6.3 The spent fuel pooiis designed and shall be maintained to prevent inadvertent draining of the pool below elevation +40.0 MSL. When fuelis being stored in the cask storage pit and/or the refueling canal, these areas will also be maintained at +40.0 MSL.
CAPACITY 5.6.4 The spent fuel pool is designed and shall be maintained with a storage capacity limited to no more than 1849 fuel assemblies in the main pool,255 fuel assemblies in the cask storage pit and after permanent plant shutdown 294 fuel assemblies in the refueling canal. The heat load from spent fuel stored in the refueling canal racks shall not exceed 1.72 X 10E6 BTU /Hr. Fuel shall not be stored in the spent fuel racks in the cask storage pit or the refueling canal unless all of the racks are installed in each respective area per the design.
fi.LCOMPONENT CYCLIC OR TRANSIENT LIMITS 5.7.1 The components identified in Table 5.7-1 are designed and shall be maintained within the cyclic or transient limits of Table 5.7-1.
WATERFORD - UNIT 3 5-6 AMENDMENT NO. 408,144
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3/4.9 REFUELING OPERATIONS
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BASES 3/4.9.1 BORON CONCENTRATION The limitations on reactivity conditions during REFUELING ensure that:
(1) the reactor will remain subcritical during CORE ALTERATIONS, and (2) a uniform boron concentration is maintained for reactivity control in the water volume having direct access to the reactor vessel. These limitations are con-sistent with the initial conditions assumed for the boron dilution incident in the safety analyses. The K value specified in the COLR includes a 1%
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delta k/k conservative allowance for uncertainties.
Similarly, the boron concentration value specified in the COLR also includes a conservative l
uncertainty allowance of 50 ppe boron.
3/4.9.2-INSTRUMENTATION The OPERABILITY of the source range neutron flux monitors-ensures that redundant monitoring capability is available to detect changes in the reactivity condition of the core.
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3/4.9.3 DECAY TIME The min'aum requirement for reactor sabcriticality prior to movement of Jeradiated fuel assemblies in the reactor pressure vessel ensures that sufficient time has elapsed to allow the radioactive decay of the short lived fission products. This decay time is consistent with the assumptions used in the safety analyses.
3/4.9.4 CONTAINMENT BUILDING PENETRATIONS
.The requirements on containment penetration closure and OPERABILITY ensure that a release of radioactive material within containment will be restricted from leakage to the environment. The OPERABILITY and closure restrictions are sufficient to restrict radioactive material release from a fuel element rupture based upon the lack of containment pressurization potential while in the REFUELING MODE.
3/4.9.5 COMMUNICATIONS The requirement for communications capability ensures that refueling station persor.nel can be promptly informed of significant changes in the facility status or core reactivity condition during CORE ALTERATIONS.
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-WATERFORD - UNIT 3 B 3/4 9-1 AMENDMENT NO.102 MAR 1 1995
l REFUELING OPERATIONS l
BASES j
l 3/4.9.6 REFUELING MACHINE l
l The OPERABILITY requirements for the refueling machine ensure that: (1) the refueling j
machine will be use.d for movement of CEAs and fuel assemblies, (2) each hoist has sufficient l
load capacity to lift a CEA or fuel assembly, and (3) the core internals and pressure vessel are protected from excessive lifting force in the event they are inadvertently engaged during lifting l-operations.,
l-3/4.9.7 CRANE TRAVEL - FUEL HANDLING BUILDING Tne restriction on movement of loads in excess of the nominal weight of a fuel assembly, CEA, and associated handling tool over other irradiated fuel assemblies in the Fuel Handling Building ensures that in the event this load is dropped (1) the activity release will be limited to that contained in a single fuel assembly, and (2) any possible distortion of fuel.in the storage racks will not result in a critical array. This assumption is consistent with the activity release assumed in the safety analyses.
3/4.9.8 SHUTDOWN COOLING AND COOLANT CIRCULATION The requirement that at least one shutdown cooling train be in operation ensures that l
(1) sufficient cooling capacity is available to remove decay heat and maintain the water in the i
reactor pressure vessel below 140*F as required during the REFUELING MODE, and j
(2) sufficient coolant circulation is maintained through the reactor core to minimize the effects of j
a boron dilution incident and prevent boron stratification, j
The requirement to have two shutdown cooling trains OPERABLE when there is less than t
23 feet of water above the reactor pressure vessel flange, ensures that a single failure of the
' operating shutdown cooling train will not result in a complete loss of decay heat removal capability. When there is no irradiated fuel in the reactor pressure vessel, this is not a consideration and only one shutdown cooling train is required to be OPERABLE. Witn the
^i reactor vessel head removed and 23 feet of water above the reactor pressure vessel flange, a large heat sink is available for core cooling, thus in the event of a failure of the operating shutdown cooling train, adequate time is provided to initiate emergency procedures to cool the core.
I 3/4.9.9 CONTAINMENT PURGE VALVE ISOLATION SYSTEM l
The OPERABILITY of this system ensures that the containment purge valves will be
. automatically isolated upon detection of high radiation levels within the containr..ent. The OPERABILITY of this system is required to restrict the release of radioactive material from the j
containment atmosphere to the environment.
WATERFORD - UNIT 3 B 3/4 9-2 AMENDMENT NO. 144
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