ML20065U439
| ML20065U439 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 12/19/1990 |
| From: | Marsh L Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20065U440 | List: |
| References | |
| NUDOCS 9101020338 | |
| Download: ML20065U439 (8) | |
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UNITED STATES -
g NUCLE AR REGULATORY COMMISSION L
t WASHINoToN. D. C. 20656 '
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NORTHERN STATES POWER COMPANY DOCKET NO. 50-263 MONTICELLO NUCLEAR GENERATING PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 76 License No. DPR-22 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application'for amendment by Northern. States Power Company (the= licensee) dated October 4, 1990, 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 1;;
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 (.onducted without endangering the health;and safety of the public, and (ii) that such activities
-will;be conducted in compliance with the Commission's regulations;
'D.
- The issuance of this amendment'will not be inimical-to the common-defense andisecurity-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 Commission's regulations and all applicable requirements!have
- been satisfied.
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-Accordingly,-the license is-amended by changes to the Technical Specifica-tions as-indi_cated in the attachment to this licenst amendment, and para-graph 2.C.2 of: Facility Operating License-No.-OPR-22-is hereby amended =to read as.follows:
o 9101020338 901219 ADOCKOSOOg3.
DR
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Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 76, 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 NUCLEAR REGULATORY COMMISSION 7
L. B. Marsh, Director Project Directorate III-1 Division of Reactor Projects III/IV/V Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
December 19, 1990 l
ATTACHMENT TO LICENSE AMEN 0 MENT NO. 76 FACILITY OPERATING LICENSE NO. OPR-22 DOCKET NO. 50-263 Revise Appendix A Technical $pecifications by removing the pages identified below and inserting the attached pages.
The revised pages are identified by amendment number and contain marginal lines indicating the area of change.
Remove Insert 39-39 127 127 151 151 169 169 189 189 198b 198b l
y Bases Continued:
3;l The IRMs are calibrated by the heat balance method such that 120/125 of full scale on the highest IRM range is below 20% ofi rated neutron. flux '(see ' Specification 2.3. A.2).
The
' requirement that the IRM detectors be inserted in the core assures that the heat balance calibration is not invalidated by the withdrawal of the detector.
Although the operator 'will set th' j set" points within the trip settings specified on Table 3.1.1, e
the actual values of the various set points can differ appreciably from the value the operator is attempting to set.
The. deviations could be caused by inherent instrument error, operator setting error, drift of.the set point,'etc.'
Therefore, such deviations have.been accounted for in the various-transient analyses and the actual trip settings may vary.by.the following amounts:
. Trio Function
. Deviation' Trio Function Deviation
- 3. High Flux IRM
+2/125 of scale ' *7. Reactor Low Water
-6 inches.
Level S. High Reactor Pressure
+10 psi
+1 gallon High Level
- 6. High Drywell Pressure
+1 psi
- 9. Turbine Condenser Low
-1/2 in. Hg vacuum
- This indication is reactor coolant temperature sensitive. The
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i calibration is thus'made for rated conditions. The level error at ' low pressures and temperatures is bounded by the safety analysis j
which reflects the weight-of-coolant above the lower. tap, and not the indicated level.
A violation of this specification is assumed to occur only when a device.is knowingly set outside of the limiting trip setting, or a sufficient number of devices have been affected by any means such that the automatic function is incapable of operating within the allowable deviation while in a re-37 actor mode in which the specified function must be operable, or the actions specified in 3.1.B.2 are
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not initiated as specified.
3 If an unsafe failure is detected during surveillance testing, it is desirable to' determine as soon 5
as possible if other failures of a similar type have occurred and whether the particular function in-volved is still operable or capable:of meeting the single failure criterion. To meet the require-z l
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ments of Table 3.1.1, it is necessary that all instrument channels in cae trip system be operable E
39 gj 3.1 BASES
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3.0 LIMITING. CONDITIONS FOR OPERATION
-4.0 ' SURVEILIANCE REQUIREMENTS E.
Safety / Relief Valves E.. Safety / Relief Valves 1.
During-power operating conditions 1.
a.
A minimum of c2ven safety / relief and whenever reactor coolant pressure, valves shall be bench checked or:
is greater than 110 psig arid--
replaced with a bench checked temperature is greater than 345*F:,
valve each refueling outage.
The nominal self-actuation setpoints.are.specified in a.
The safety valve function (self- '
- actuation) of seven safety /
Section 2.4.B.
relief valves shall be operable.
- b.. At Icast two of the safety / relief -
b.
The solenoid activated _ relief '
valves shall be disassembled and function (Automatic-Pressure inspected each refueling' outage.
. Relief) shall be operable as required by Specification 3.5.E.
c.
The integrity of the safety / relief
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valve bellows'shall-.be continuously c.
The Low-Low' Set function.for three monitored.
non-Automatic Pressure Relief; Valves shall be operable'as required by d.
The operability of_the bellows.
Specifica tion - 3.2.II.
monitoring system shall be demon-strated at least once every three 2.
If Specification 3.6.E.1.a is not met, months.
Initiate an orderly shutdownLand have reactor coolant pressure and temperature 2.
Low-Iew. Set logic -surveillance shall reduced to 110 psig or less and 345*F or be performed in accordance with Table 4.2.1.
less within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
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! Bases Continued 3.6 and 4.6:
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1 The safety / relief valves ha~ e twol functions; i.eu power: relief or self-actuated by high pressure. -
l The solenoid actuated function"(Automatic Pressdre Pelief) in which external instrumentation signals of.
l v
j coincident' high drywelll pressure and low-low water level initiate opening of the' valves. This function is i
discussed.in"Specificationf3.5.E. 'In addition, the valves can be operated manually.
The safety l function.is performed by the'same: safety / relief valve with self-actuated integral.-
bellows and pilot valve causi,ng mainLyalve operation.. Article 9 of the ASME' Pressure Vessel Code Section:III r
Nuclear Vessels ; requires 'that. these bellows be monitored for ' failure since this would defeat the safety function of the safety /reliefzvalve.
Provision ' also has been.made to detect failure 'of, the bellows monitoring system.'
system quarterly provisions assurance of. bellows integrity.
Testing;of this When the. setpoint :is being bench checked, it. is prudent to disassemble one of the safety / relief valves to. examine for crud buildup, bending:of certain actuator members or other signs of possible l
dernriora tion Low-Low Se.+. Iogic has been provided on three non-Automatic Pressure Relief Jyste's valves.
I This logic is di.-~:ssed 'in detail in the Section 3.2 Bases. This. logic, through pressure rensing instrumentation... reduces the' opening setpoint and increases the blowdowa range of the three selected valves following a scram to eliminate the discharge line water leg clearing loads resulting.from multiple: valve openings.
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3.6/4.6 BASES 151 4
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3.0 LIM 1fING CONDITIONS FOR OPEP>. TION 4.0 SURVEILIANCE REQUIREMENTS C.
C.
Secondsry Containseent 1.
Except as specified in 3.7.C.2 and 1.
Secondary containment surveillance shall 3.7.C.3, Secondary Contai:nent Integrity be performed as indicated below:
shall be maintained during all modes of Secondary containment espability no plant operation.
a.
maintain at least a 1/4 inch of water 2.
Secondary Containment Integr'ty is not vacuum under enim vind 12 < u < 3 n.+)
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required when all of the following can-conditions with a filter train flov ditions are satisfied:
rate of <4,000 sefar, siall be dem-onstrated at each refueaing outage a.
The reactor is suberitical and prior to refueling. Verification Specification 3.3.A is met.
that each automatic damper actuates to its isolation position shall b.
The reactor water temperature is be performed at each refueling outage below 212*.
and after maintenance, repair or replace-ment work is perfornMd on the damper or c.
No activity is being performed which its associated. actuator, control circuit, can reduce the shutdown margin below or power circuit.
that specified in Specification 3.3.A d.
The foel cask er irradiated fuel is not being moved within the reactor building.
3.
With an inoperable secowlary contain-ment isolation damper, restore the inoperable damper to operable status or isolate th.
affected duct by use of a closed damper or 3g blind flange within eight hours.
4.
If Specifications 3.7.C.1 through 3.7.C.3 cannot be met, initiate a normal orderly shutdown and have the y
reactor in the Cold Shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Alterations of the
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3.7/4.7 169 wc~
de C
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Bases continued 1 D.
Primary Containment Isolation Valves Those large pipes comprising portion of the reactor coolant system whose failure could result in uncovering the reactor core are supplied with automatic isolation val.es (except those lines need-d
^
for emergency core cooling system operatien or containment cooling). The closure t mes sp+_cified in USAR Table 5.2-3b are adequate to prevent loss of more coolant from the circumferential rupture
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of any of these lines outside the containment than from a steem line rupture. Therefore, I
this isolation valve closure time is sufficient to prevent uncovering the core.
l I
The primary containment isolation valves are highly reliable, have low service requirement, and cost are normally closed. The initiating sensor and associated trip channels are also checked to demon-I strate the capability for automatic isolation. Reference Section 5.2.2.5.3 and Table 5-2-3b USAR.
The test intgrval of once per operating cycle for automatic initiation results in a failure probability of 1.1 x 10-that a line will not isolate. Store frequent testing for valve operability rest.lts in j
a more reifable system.
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R Ro IJ9 4.7 BASES z
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3.0 LIMITING CONDITIONS Fr;R OPERATION 4.0 SURVEILIANCE REQUIREMENTS 4
Offgas Treatment System 4
Offgas Treatment System F'llowing c ach isotopic analysis of a sample of The offgas treatment system shall be in a.
a.
operation whenever the main condenser air gases free the steam Jet air ejector required by ejector system is in operation. Components 4.8.B.S. verify that the maxista storage tank i
of the system shall be operated to previde activity limit specified in 3.8.B.4.e cannot be the maximum holdup time obtainable except exceeded using the method in the ODCM.
during periods of equipment maintenance.
b.
With gaseous vaste being discharged for more than 7 days with a holdup tima of less than 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br />, within 30 days submit to the Commission a special report which includes the following information:
1.
Identification of equipment or sub-systems not functional and the reason.
2 Actit,(s,\\ taken to restore equipment to funct!onal status.
3.
Summary description of action (s) taken to prevent a recurrence.
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3.8/4.8 w
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