ML20041B621
| ML20041B621 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 02/05/1982 |
| From: | Vassallo D Office of Nuclear Reactor Regulation |
| To: | Boston Edison Co |
| Shared Package | |
| ML20041B622 | List: |
| References | |
| DPR-35-A-053 NUDOCS 8202240371 | |
| Download: ML20041B621 (6) | |
Text
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UNITED STATES 0
'h NUCLEAR REGULATORY COMMISSION h
j WASHINGTON, D. C. 20555 x
BOSTON EDISON COMPANY DOCKET NO. 50-293 PILGRIM NUCLEAR POWER STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 53 License No. DPR-35 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Boston Edison Company (the licensee) dated January 15,1982,. complies with the standards and requirements of the Atomic Energy Act of 1954, 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 wit.hout endangering the health and safety of the public, and (ii) that such activities will be l
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 l
l 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.
2.
Accordingly, the license is amended by changes to the Technical Spec-ifications as indicated in the attachment to this license amendment and paragraph 3.B of Facility Operating License No'. DPR-35 is hereby amended to read as follows:
B.
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 53, are hereby incorporated h
in the license. The licensee shall operate the facility in accordance with the Technical _ Specifications.
i 8202240371 820205 PDR ADOCK 05000293 P
2 3.
This license amendment.is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Domenic B. Vassallo, Chief Operating Reactors Branch #2 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: February 5,1982 t
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ATTACHMENT TO LICENSE AMENDMENT NO. 53 FACILITY OPERATING LICENSE N0. DPR-35 DOCKET NO. 50-293 Replace the following pages of the Appendix "A" Technical Specifications with the enc.losed pages.
The revised pages are identified by Amendment Number and centain a vertical line indicating the area of change.
Remove Reolace 152A 152A 166 166 1 71 1 71
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMEhTS
(
3.7 COh7AIhMEh7 SYSTEMS (Cont'd) 4.7 CONTAINMENT SYSTEMS (Cont'd)
The pressure differential h.
During reactor isolation e.
conditions, the reactor pressure between the.drywell and vessel shall be depressurized suppression chamber shall be to less than 200 psig at normal recorded at least once.each
. cool down rates if the pool shift when the differential temperature reaches 120 F.
pressure is required.
0 1.
Differential pressure between the f.
Suppression chamber water drywell and suppression chamber level shall be recorded at shall be maintained at equal to or least once each shift when-greater than 1.17 psid, except as the differential pressure specified in j and k.
is required.
j.
The differential pressure shall be established within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of placing the reactor in the run mode following a shutdown. The differential pressure may be reduced to less than 1.17 psid 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to a scheduled shutdown.
k.
The differen:ial pressure may be reduced to less than 1,17 psid for f
a maximum of four (4) hours fo'r maintenance activities on the differential pressure control system and during required operability testing of the EPCI system, the relief valves, the RCIC system and the dryaell-suppre.=sion chanber vacuum breakers.
1.
If the specifications of item 1, above, cannot be met, and the dif ferential pressure cannot be restored within the subsequent six (6) hour period, an orderly shutdown shall be initiated and the reactor shall be in a cold i
shutdown condition in twenty-four l
(24) hours.
i m.
Suppression cha=ber water. level shall be maintained between -6 to -3 inches on torus level instrument which corresponds to a downcomer submergence of 3.00 and 3.25 feet respectively.
Amsudment No. 31, 53 151A
I
, BASES:
3.7.A & 4.7.A Primr.r1 Containm:nt The integrity of the primary containment and operation of the core standby cooling systein in combinatio. liLit the off-site doses to values less than thos,e suggested in 10 CFR 100 in the event of a break in the. primary system piping. Thus, containment integrity is specified whenever the potential for violation of the primary reactor system integrity exists. Concern about such a violation exists whenever the reactor is critical and above atmospheric pressure. An exception is made to this requirement during initial core load-ing and while the low power test program is being conducted and ready access to the reactor vessel is required. There vill be no pressure on the system at this time, thus greatly reducing the chances of a pipe break. The reactor may be taken critical during this period; however, restrictive operating procedures will be in effect again to minimize the probability of an accident occurring. Procedures and t'..e Rod Worth Minimizer would limit control worth such that a rod drop would not result in any fuel damage. In addition, in the unlikely event that an excursion did occur, the reactor building and standby gas treatment system, which shall be operational during this time, offer a sufficient barrier to keep off-site doses well below 10 CFR 100 limits.
The pressure suppression pool vater provides the heat sink for the reactor primary system energy release following a postulated rupture f the system.
The pressure suppression chamber water volume must absorb the associated decay and structural sensible heat released during primary"r.yste= blowdown from 1035 psig.
Since all of the gases in the drywell are purged into the pressure supression chamber air space during a loss-of-coolant accident, the pressure resulting from isother=al co=pression plus the vapor pressure of the liquid must not exceed 62 psig, the suppression chamber maxi =um pressure. The design volume of the suppression chamber (water and air) was obtained by considering that the total volume of reactor coolant to be condensed is discharged to the suppression chamber and that the drywell, volume is purged to the suppression chamber.
Using the minimum or maxi =um vater volu=es given in the specification, containment pressure during the design basis accident is approximately 45 psig 3
which is below the maximum _of 62 psig. Maximum water volume of 94.000 ft 3
l results in a downcomer submergency of 4'-0" and the mininum volume of 84,000 f t results in a submergence approximately 12-inches less. Mark I Containment Long Term Program Quarter Scale Test Facility (QSTF) testing at a downcomer submerger.cy of 3.25 feet and 1.17 psi wetwell to drywell pressure differential shows a signifi-
)
cant suppression chamber load reduction and Long Term Program analysis and modifica-l tions are based on the above submergence and AP.
Should it be necessary to drain the suppression chamber, provision vill be made to caintain those requirements as described in Section 3.5.F BMES of this l
Technical Specification.
l Experimental data indicates that excessive steam condensing loads can be avoided if the peak temperature of the pressure suppression pool is maintained l
.below 1600F during any period of relief-valve operation with senic conditions et the discharge exit.
Specifications have been placed on the envelope of reactor operating conditions so that the reactor can be deprescurized in a timely manner to avoid the regime of potentially high pressure suppression chamber loadings.
166 Amendment No. 39, 53
BASES:
3.7.A & 4.7.A Primary Containment (Cont'd)
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 concentra-tion, no monitoring of oxygen concentration is necessary.
However, at least 1:iice a week the oxygen concentration will be determined as added Mark I Containment Long Term Program testing showed that assurance.
maintaining a drywell to wetwell pressure differential to keep the sup-pression chamber downcomer legs clear of water significantly reduced suppression chamber post LOAC hydrodynamic loads. A pressure of 1.17 psid is required to sufficiently clear the water legs of the downcomers without. bubbling nitrogen into the suppression chamber at the 3.00 ft.
downcomer submergence which corresponds to approx. 84,000 ft.3 of water.
Maximum downcomer submergence is 3.25 ft. at operating suppression chamber water level. The above pressure differential and submergence number will be used in the Pilgrim I Plant Unique Analysis to be submitted to the NRC.
1 Amendment No. 3J, 53 171
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