ML20116M200
| ML20116M200 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 04/22/1985 |
| From: | Vassallo D Office of Nuclear Reactor Regulation |
| To: | Boston Edison Co |
| Shared Package | |
| ML20116M202 | List: |
| References | |
| DPR-35-A-087 NUDOCS 8505060143 | |
| Download: ML20116M200 (6) | |
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o UNITED STATES NUCLEAR REGULATORY COMMISSION 5
- j WASHINGTON, D. C. 20555 i.,...../
April 22, 1985 BOSTON EDISON COMPANY DOCKET NO. 50-293 PILGRIM !'JCLEAR POWER STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.87 License No. DPR-35 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Boston Edison Company (the licensee) dated December 6, 1984, complies with the standards and l
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 (1) that the activities authorized by this amendment can be conducted without endangering the health i
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 requirements have been satisfied.
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 Facility Operating License No. DPR-35 is hereby amended to read as follows:
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B.
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 87, 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 the date of its issuance.
FOR THE NUC EAR REGULATORY COMMISSION Domenic B. Vassallo, Chief Operating Reactors Branch #2 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of issuance: April 22, 1985 I
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ATTACHMENT TO LICENSE AMENDMENT NO. 87 FACILITY OPERATING LICENSE NO. DPR-35 DOCKET NO. 50-293 Replace the following pages of the Technical Specifications with the enclosed pages. The revised pages are identified by amendment number and contain a vertical line indicating the areas of change.
Remove Insert 157 157 157a 157a 167 167 i
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7 Primary Containment 4.7 Primary Containment b.
Any drywell-suppression chamber b.
During each refueling outage:
vacuum breaker may be non-fully closed as determined by the position (1) Each vacuum breaker shall be switches provided that the drywell tested to determine that the to suppression chamber differential disc opens freely to the touch decay rate is demonstrated to be not and returns to the closed greater than 25% of the differential position by gravity with no pressure decay rate for the maximum indication of binding.
allowable bypass area of 0.2ft".
c.
Reactor operation may continue pro-(2) Vacuum breaker position vided that no more than 2 of the dry-switches and installed alarm well-pressure suppression chamber systems shall be calibrated and vacuum breakers are determined to be functionally tested.
inoperable provided that they are secured or known to be in the (3) At least 25% of the vacuum closed position.
breakers shall be visually inspected such that all vacuum d.
If a failure of one of the two breakers shall have been installed position alarm systems inspected following every fourth occurs for one or more vacuum refueling outage.
If deficien-breakers, reactor operation may cles are found, all vacuum continue provided that a breakers shall be visually differential pressure decay rate inspected and deficiencies test is initiated immediately and corrected.
performed every 15 days thereafter until the failure is corrected.
(4) A drywell to suppression chamber The test shall meet the leak rate test shall demonstrate requirements of that the differential pressure Specification 3.7.A.4.b.
decay rate does not exceed the rate which would occur through a 1 inch orifice without the addition of air or nitrogen.
5.
Oxygen Concentration 5.
Oxygen Concentration a.
The primary containment atmosphere The primary containment oxygen shall be reduced to less than concentration shall be measured 4% oxygen by volume with nitrogen and recorded at least twice gas during reactor power weekly.
operation with reactor coolant pressure above 100 psig, except as specified in 3.7.A.5.b.
Amendment No. g 87 i
157 l
LIMITING CONDITION FOR OPERATION SURVEILLANCE RE00XREMENTS 3-7 Primary Containm@nt 4.7 Primary Containment b.
Within the 24-hour period subsequent to placing the reactor in the Run mode following a shutdown, the containment atmosphere oxygen concentration shall be reduced to less than l
4% by volume and maintained in this condition. De-inerting may commence 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to a shutdown.
6.
If the specifications of 3.7.A cannot be met, an orderly shutdown shall be initiated and the reactor shall be in a Cold Shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Amenament No. 87 157a
BASES:
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3.7. A & 4.7. A Primary Containment In addition to the limits on temperature of the suppression chamber pool water.. operating procedures define the action to be taken in the event a relief valve inadvertently opens or sticks open.
This action would include:
(1) use of all available means to close the valve, (2) initiate suppression pool water cooling heat exchangers, (3) initiate reactor shutdown, and (4) if other relief valves are used to depressurize the reactor, their discharge shall be separated from that of the stuck-open relief valve to assure mixing and uniformity of energy insertion to the pool.
Because of the large volume and thermal capacity of the suppression pool, the volume and temperature normally changes very slowly and monitoring these parameters daily is sufficient to establish any temperature trends.
By requiring the suppression pool temperature to be continually monitored and frequently logged during periods of significant heat addition, the temperature trends will be closely followed so that appropriate action can be taken.
The requirement for an external visual examination following any event where potentially high loadings could occur provides assurance that no significant damage was encountered.
Particular attention should be focused on structural discontinuities in the vicinity of the relief valve discharge since these are expected to be the points of highest stress.
If a loss-of-coolant accident were to occur when the reactor water temperature is below approximately 330*F, the containment pressure will not exceed the 62 psig code permissible pressure, even if no condensation were to occur.
The a
maximum allowable pool temperature, whenever the reactor is above 212*F, shall be governed by this specification.
Thus, specifying water volume-temperature requirements applicable for reactor-water temperature above 212*F provides additional margin above that available at 330*F.
Inerting The relatively small containment volume inherent in the GE-BWR pressure suppression containment and the large amount of zirconium in the core are such that the occurrence of a very limited (a percent or so) reaction of the zirconium and steam during a loss-of-coolant accident could lead to the liberation of hydrogen combined with an air atmosphere to result in a flammable concentration in the containment.
If a sufficient amount of hydrogen is generated and oxygen is available in stoichiometric quantities, the subsequent ignition of the hydrogen in rapid recombination rate could lead to failure of the containment to maintain a low leakage integrity.
The 47.
l oxygen concentration minimizes the possibility of hydrogen combustion following a loss-of-coolant.
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l Amendment No. 87 167 1
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