ML20081K526
| ML20081K526 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 06/19/1991 |
| From: | Wessman R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20081K527 | List: |
| References | |
| NUDOCS 9106280073 | |
| Download: ML20081K526 (6) | |
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WASHINGTON. D.C. 20666 gv f BOSTON EDISON COMPANY DOCKET.NO, 50-293 PILGRIM.NUCtr*R POWER-STATION AMENDMENT T0 FACILITY OPERATING-LICENSE 1
Amendment No. 137 License No. DPR-35 1.
The Nuclear Regulatory Comission (the Comission or the NRC) has found that 4
A.
The application for amendment filed by the Boston Edison Company (the licensee) dated February 6,1991, 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 set forth in 10 CFR Chapter I; 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 Specifica-tions as indicated in the attachment to this license amendment, and paragraph 3.8 of Facility Operating License No. DPR-35'is hereby amended to read as follows:
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.137, are hereby incorporated in the license.
The' licensee shall operate the facility in accordance with the Technical Specifications.
91 6280073 910619 D
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3.
This license amendment is effective as of its date of issuance and shall be implemented within 30 days.
FOR THE NUCLEAR REGULATORY COMMISSION w
Richard Wessman, Director Project Directorate I-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
June 19, 1991
ATTACHMENT-TO. LICENSE. AMENDMENT NO, J37 FACILITY 0PERATING-LICENSE.NO, DPR-35 DOCKET NO.-50-293
. 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 204A 204A 205 205 217a 217a i
t
s 3.10 BASES:
8.
Core Monitorina 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 in'or adjacent to any core quadrant where fuel or control rods are being moved ensures adequate monitoring of that quadrant during such alterations. The requirement of 3 counts per second provides assurance that neutron flux is being monitored and ensures startup is conducted only if the source range flux level is above the minimum assumed in the control rod drop accident.
The limiting conditions for operation of the SRM subsystem of the Neutron Monitoring System Arc derived from the Station Nuclear Safety Operational Analysis (Appendix G) and a functional analysis of the neutron monitoring system. The specification is based on the Operational Nuclear Safety Requirements in subsection 7.5.10 of the Safety Analysis Report.
A spiral unloading program is one by which the fuel is in the outermost cells (four fuel bundles surrounding a control blade) is removed first. Unloading continues by removing the rema The center cellwillbethelastremoved.gngoutermostfuelcellbycell.
A spiral loading program is one by which fuel is loaded on the periphery of the previou;1y loaded fueled region beginning around a single SRM. Spiral unloading c,nd reloading will preclude the creation of flux traps (moderator filled cavities sur.ounded on all sides by fuel).
During spiral unloading, the SRH's shall have an initial count rate of 13 cps with all rods fully inserted.
The count rate will diminish during fuel removal. Under the special condition of complete spiral core unloading, it is expected that the count rate of the SRH's will drop below 3 cps before all of 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 SRH's will no longer be required.
Requiring the SRM's to be i
operational prior to-fuel removal assures that the SRM's are operable-and can be relied on even when the count rate may go below 3 cps.
During spiral. reload, SRM operability will be verified by using a portable external source every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> until the required amount of fuel is loaded to maintain 3 cps. As an alternative to the above, up to two fuel assemblies will be loaded in different cells containing control blades around each SRM to obtain the required 3 cps. Until these assemblies have been, loaded, the 3 cps requirement is not necessary.
N During selected refueling outages, ~ prior to initiating spiral unloading, the central controlled cell-will be removed to facilitate inspection of the Core Spray Spargers.
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l-204A Amendment No. #,137, l
s 3.10 MSES:
C.
Seent fuel Pool Hater Level To ensure 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 vould be a significant change from the normal level (-l foot)
.d is well above the level
)
to assure adequate cooling.
I D.
Bg1ticle Control Rod Remon1 1
These specifications ensure maintenance r,r repair of control rods or rod drives will be performed under conditior.s that limit the probability of inadvertent criticality.
The requirement that the fuel assemblies in the cell controlled by the control rod be removed from the reactor core before the interlock can be bypassed ensures 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 the control rod, Thus, removal of an entire cell (fuel assemblies plus control rod) results in a lower reactivity potential of the core.
4.10 DASES:
A.
Refuelina Interlocks Complete functional testing of all refueling interlocks before any refueling outage will provide positive indication that the interlocks operate in the situations for which they were designed.
By loading each hoist with a weight equal to ine fuel astembly, positioning the refueling platform, and withdrawing control rods, the interlocks can be subjected to valid operational tests. Hhere redundancy is provided in the logic circuitry, tests can be performed to assure that each redundant logic element can independently perform its functions.
B.
Core Monitorina Requiring the SRH's to be functionally tested prior to any core alteration ensures the SRM's will be operable at the start of that alteration.
The daily response check of the SRM's ensures their continued operability.
Amendment No.
AT: 137
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4 6.9.A Routine Reports (Continued) 3.
Occuoational_(xposure Tabulation A tabulation of the number of station, utility and other personnel (including contractors) receiving exposures greater than 100 mrem /yr and their associated man-rem exposure according to work and job functions, e.g. reactor operations and surveillance inservice inspection, routine maintenance, special maintenance (including a description), waste processing, and refueling shall be submitted on an annual basis.
This tabulation supplements the requirements of 20.407 of 10 CFR 20.
The dose assignment to various duty functions may be estimates based on pocket dosimeter, TLD, or film badge measurements.
Small exposures totalling less than 20% of the individual total dose need not be accounted for.
In the aggregate, at least 80% of the total whole body dose received from external sources shall be assigned to specific major work functions.
4.
Core Ooeratina Limits Reoort a) Core operating limits shall be established and documented in the v0RE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle, b) The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC in NEDE-240ll-P-A, " General Electric Standard Application for Reactor Fuel " (the approved version at the time the reload analyset are performed shall be identified in the CORE OPERATING LIMITS REPORT),
and in NEDC-31852P, " Pilgrim Nuclear Power Station SAFER /GESTR-LOCA Loss of Coolant Accident Analysis", dated September,1990 (the approved version at the time the reload analyses are performed shall be identified in the CORE OPERATING LIMITS REPORT).
c) The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) are met, d) The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.
6.9.B Deleted Amendment No. 122, 137 217a