ML20151Z554
| ML20151Z554 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 08/19/1988 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20151Z546 | List: |
| References | |
| NUDOCS 8808290355 | |
| Download: ML20151Z554 (6) | |
Text
-
- ga. sea,#'o u
'.INITED STATES
['
k NUCLEAR REGULATORY COMMISSION g
j WASHINGTON, D. C. 20666
\\...../
NORTHEAST NUCLEAR ENERGY COMPANY 00CXET NO. 50-245 MILLSTONE NUCLEAR _ POWER STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 21 License No. DPR-21 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for ame0dment by Northeast Nuclear Energy Company, et al. (the licensee), cated Decunber 18, 1987, 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 rea:onable assurance (1) 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, t
esosa9o355ggQ$$45 PDS ADOCK PNV P
l l
]
, 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. DPR-21 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 21, 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 issuance, to be implemented within 30 days of issuance.
F THE NUCLEAR REGULA ORY COMMISSION
\\
r b
J n F. Stolz, Director oj ct Directorate I-4 iv sion of Reactor Projects I/II ice of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
August 19, 1988
]
ATTACHMENT TO LICEfiSE AMEN 0HENT NO. 21 FACILITY OPERATING LICENSE NO. OPR-21 DOCKET fl0. 50-245 Replace the followin the enclosed pages. g pages of the Appendix "A" Technical Specifications with and contain vertfeal lines indicating the areas of change.The revised pag Remove Insert 3/4 7-4 3/4 7-4 83/4 7-1 83/4 7-1 B3/4 5-4 83/4 5-4 l
1 e
i
LIMITING CONDITION FOR OPERATION 37 CONTAINENT SYSTEMS 3 7.A.3 Primary containment integrity, as defined in section 1.0, shall be maintained at all times when the reactor is critical or when the reactor water temperature is above 212'F and fuel is in the reactor vessel.
SURVEILLANCE REQUIREENT 4.7.A.3 The primary containment integrity shall be demonstrated as follows:
Integrated Primary Containment Leak Test (IPCLT) a.
The containment leake rates shall be demonstrated at the following test schedule and shall be determined in conformance with the criteria specified in Appendix J of 10 CFR 50 using the methods and provisions of ANSI N45.4-1972 and BN-TOP-1 1.
Three Type A Overall Integrated Containment Leakage Rate tests shall be conducted at 40 0 10 month intervals during shutdown at P (43 psig) during each ten-year service period.
TheIbirdtestofeachsetshallbeconducted during the shutdown for the ten-year plant insarvice inspection.
2.
If any periodie Type A test fails to meet 0.75 L, the test schedule for subsequent Type A tests shall be re$iewed and l
approved by the Commission.
If two consecutive Type A tests fail to meet 0.75 L, a Type A test shall be performed at least every T8 months until two consecutive Type A tests meet 0.75 L, at which time the above schedule l
may be resumed.
a 3
The accuracy of each Type A test shall be verified by a i
supplemental test which:
l l
a.
Confirms the accuracy of the test by verifying that the difference between the supplemental data and the Type A test data is within 0.25 L,.
l b.
Has duration sufficient to establish accurately the l
change in leakage rate betwen the Type A test and the supplemental tast.
l l
c.
Requires t ie quantity of gas injected into containment i
or bled from containment during the supplemental test to be equivalent to at least 25 percent of the total measured leakage at P,.
Millstone Unit 1 3/4 7-4 l
3.5 CORE AND CONTAD#1ENT COOLING SYSTEMS BASES gg-The function of the Isolation Condenser during a small besak accident is to assist the autcoatic pressure relief system in depressurizing the reactor as & backup to the FWCI system.
The two effects of isolation condenser depressurization are:
(1) the minimization of reactor inventory loss which normally occurs during APR blowdown; this reduces the time of core uncovery prior to reflooding; and (2) earlier onset of low. pressure core spray. cooling.
Analysis performed by General Electric in March 1976, in support of extended operation of Millstone while the isolation cdndenser was be.ing retubed indicated that from 40% rated power, over 30 minutes is available to initiate operator action to mitigate the consequences of a loss of all feedwater.
This is based upon manual depressurization with APR and coolant supplied by all LPCI and core spray subsystems.
The FWCI was assumed lost as part of the nn-'aechanistic assumption of loss of feedwater, i
The successful mitigation of tiis postulated event was no uncovering of the fuel.
4 F.
Eneroency Coolino Availability The purpose of Specification F is to assure a sinimum of core cooling equipment is available at all times.
If,.for example one core spray system were out of service and the emergency power sou,rce which powered the opposite core spray system were out of service, only two LPCI pumps would be available. 'hikewise, if two LPCI pumps were out of service and two emergency service water pumps on the opposite side were also out of service, no containment cooling would be available.
It is during refueling outages 1: hat major saintenance is performed and during such time that low pressure core cooling systans may be cut of service depending on the activities being performed.
Specification F allows removal of one CRD mechanism or fuel removal and replacement while the torus is in a drained condition without comproeising core cooling capability.
The specification establishes the minimum operable low pressure core cooling systems, water inventories, electrical power supplies and other additional requirements that must exist to allow such activities as CRD mechanisa maintenance or fuel removal and replaconent to be performed in parallel with other major activities.
The available core cooling capability for a potential draint of the Mactor vessel while this work is performed is based on an est sated drain rate and the j
maintained minimum water level in the refueling cavity to be supplied to the reactor vessel.
In addition, the available low pressure core cooling systems are lined up to the condensate storage tank which supplements the reactor cavity water with an additional 450,000 gallons of water.
- Thus, with the torus drained, a volume of approxistely 800,000 gallons of water will be maintained available to be supplied to, the reactor ves'el.
s Millstone Unit 1 3 3/4 5-4
37 CONTAIN)ENT SYSTEMS BASES A.
1.
Ptimary Containment The integrity of the primary containmen* and operation of the emergency core cooling system in combination, limit the off-site doses to values less than those specified in 10 CFR 100 in the event of a break in the primary system piping. Thus, coiltainment 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.
2.
Suppression Chamber The pressure suppression pool water provides the heat
' *t for the reactor primary system energy release followin a pot led rupture of the system or for releases through the safety relie r talves. The pressure suppression chamber water volume must absorb the associated decay and structural sensible heat released during primary system blowdown from 1035 psig.
Since all of the gases in the drywell are considered purged into the pressure suppression chamber air space during a loss of coolant accident, the pressure resultirg from isothermal compression plus the vapor pressure of the liquid must not exceed 62 psig, the suppression chamber design 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 maximum water volumes given in the specification, containment pressure during the design basis accident isapproximately42psigwhichisgelowthedesignof62psig.
Maximum water volume of 100,400 ft results in a downconer submergence of 3 33 feet and, the minimum volume 98,000 ft3 results in a submergence of 3 0 feet. The majority of the Bodega tests were run with a submerged length of four feet and with complete condensation.
Additional condensation Hillstone Unit 1 B 3/4 7-1 Amendrent No. 21
. - -. - -.... -