ML20080J720
| ML20080J720 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 02/16/1995 |
| From: | Butler W Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20080J727 | List: |
| References | |
| NUDOCS 9502280070 | |
| Download: ML20080J720 (5) | |
Text
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3 UNITED STATES g*
j NUCLEAR REGULATORY COMMISSION WASHINGTON. D.C. 2066H001
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BOSTON EDISON COMPANY-DOCKET NO. 50-293 PILGRIM NUCLEAR POWER STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.158 License No. DPR-35 1.
The Nuclear Regulatory Comission (the Comission or the NRC) has found that:
A.
The application for amendment filed by the Boston Edison Company (the licensee) dated September 6, 1994, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),
and the Comission's rules and regulations; 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 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.
9502200070 950216 PDR ADOCK 05000293 P
'i 3.
This license amendment is effective as of its date of issuance and shall be implemented prior to startup from refueling outage #10.
FOR THE NUCLEAR REGULATORY COMMISSION 0.
W YOR Walter R. Butler, 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:
February 16, 1995 l
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ATTACHMENT TO LICENSE AMENDMENT NO.158 FACILITY OPERATING 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 3/4.7-8 3/4.7-8 B3/4.7-7 83/4.7-7
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N LIMITING CONDITION FOR OPERATION SURVEILIANCE REQUIREMENTS 3.7 CONTAINMENT SYSTEMS (Cont) 4.7 CONTAINMENT SYSTEMS (Cont) l A.
Primary Containment (Cont)
A.
Primary Containment (Cont)
- 1. The valve is demonstrated
- b. During each refueling to open with the applied interval:
force of the installed test actuator as
- 1. Each vacuum breaker shall-indicated by the position be tested to determine switches and remote that the disc opens position indicating freely to the touch and
- lights, returns to the closed position by gravity with
- 2. The valve shall return by no indication of binding.
gravity when released after being opened by
- 2. Vacuum breaker position remote or manual means, switches and installed to within 3/32" of the alarm systems shall be fully closed position.
calibrated and functionally tested.
- 3. Neither of the two position alarm sys tems,
- 3. At least 25% of the which annunciate in the vacuum breakers shall be Control Room when any visually inspected such' vacuum breaker opening that all vacuum breakers exceeds 3/32", are in shall have been inspected alarm.
following every fourth refueling interval.
If
- b. Any drywell-suppression deficiencies are found, chamber vacuum breaker may all vacuum breakers shall be non-fully closed as be visually inspected and determined by the position deficiencies corrected.
switches provided that the drywell to suppression
- 4. A drywell to suppression chamber differential decay chamber leak rate test rate is demonstrated to be shall demonstrate that not greater than 25% of the the differential pressure differential pressure decay decay rate does not rate for the maximum exceed the rate which allowable bypass area of would' occur through a 1 2
0.2f t.
inch orifice without the addition of air or
- c. Reactor operation may nitrogen.
continue provided that no more than 2 of the drywell-pressure suppression chamber vacuum breakers are determined to be inoperable provided that they are secured or known to be in the closed position.
J Amendment No. 68r-87,-L49, 158 3/4.7-8
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13/4.7 CONTAINMENT SYSTEMS (Cont)
- A.
Primary Containment (Cont)
.Each drywell suppression chamber vacuum breaker is equipped with three switches. One switch provides full open indication only. Another switch provides closed indication and an alarm should any vacuum breaker come off l
its. closed seat by greater than 3/32".
The third switch provides a' separate
[
and redundant' alarm should any vacuum breaker come off its closed seat by-l
. greater than.3/32". The two alarms above are those referred to in Section 3.7.A.4.a.3 and 3.7.A.4.d.
The water in the suppression chamber is used only for cooling in the event of.
an' accident; i.e., it is not used for normal operation; therefore, a daily H
check of the temperature and volume is adequate to assure that adequate heat removal capability is present.
l i
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 j
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, i
1 the subsequent ignition of the hydrogen in rapid recombination rate could lead to failure of the containment to maintain a low leakage integrity. The 4%.
oxygen concentration minimizes the possibility of hydro 5en combustion
'l following a loss-of-coolant.
The occurrence of primary system leakage following a major refueling outage or other scheduled shutdown is much more probable than the occurrence of the-loss of-coolant accident upon which the specified oxygen concentration limit is based.
Permitting access to the drywell for leak inspections during a startup is judged prudent in terms of the added plant safety. offered without significantly reducing the margin of safety. Thus, to preclude the possibility of starting the reactor and operating for extended periods of time.
with significant leaks in the primary system, leak inspections are scheduled-during startup periods, when the primary system is at or.near rated operating temperature and pressure. The 24-hour period to provide inerting is judged to i
be sufficient to perform the leak inspection and establish the required oxygen.
concentration.
The primary containment is normally slightly pressurized during periods of reactor operation. Nitrogen used for inerting could leak out of the containment but air could not leak in to increase oxygen concentration. Once the containment is filled with nitrogen to the required concentration, no monitoring of oxygen concentration is necessary. However, at least twice a week the oxygen concentration will be. determined as added assurance.
Mark'I Containment Long Term Program testing showed that maintaining a drywell to k
i 311-53 -55,-113, 158 B3/4 7-7 Amendment No.
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