ML19345E774
| ML19345E774 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 01/15/1981 |
| From: | Varga S Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19345E775 | List: |
| References | |
| NUDOCS 8102060089 | |
| Download: ML19345E774 (32) | |
Text
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o UNITED STATES
!' ) y ' j, NUCLEAR REGULATORY COMMISSION r
W ASHINGTON. D. C. 20555
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POWER AUTHORITY OF THE STATE OF NEW YORK DOCKET NO. 50-286 INDIAN P0 INT NUCLEAR GENERATING UNIT NO. 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 34 License No. DPR-64 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The applications for amendment by Power Authority of the State of New York (the licensee) dated September 29, 1980 and November 7,1980, and two applications dated October 31, 1980, comply 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 applications, 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 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 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 Commission's regulations and all applicable requirements i
have been satisfied.
l Bicuocoog
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t o.
L 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, by changing paragraphs 2.C(2) and 2.J to read as follows:
2.C(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 34, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
2.J The plant shall be brought to the cold shutdown condition within twelve equivalent months of operation from achieving criticality after the Cycle 3 mid-cycle outage, but in any event, no later than February 1,1982.
For the purpose of this requirement, equivalent operation is defined as operation with reactor coolant temperature greater than 350'F. An inspection of all four steam generators shall be performed and Nuclear Regulatory Commission approval shall be obtained before bringing the reactor critical following this inspection.
3.
This license amendment is effective as of the date of its issuance.
FCR THE NUCLEAR R GULATORY COMMISSION
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' Operating Reactors Br n h #1 Division of 1.icensing
Attachment:
Changes to the Technical Specifications Date of Issuance:
January 15, 1981
ATTACHMENT TO LICENSE AMENDMENT NO. 34 FACILITY OPERATING LICENSE NO. DPR-64 DOCKET NO. 50-286 Revise Appendix A as follows:
Remove Pages Insert Pages 1-4 1-4 1-5 1-5 3.1-12 3.1-12 3.1-13 3.1-13 3.1-21 3.1-21 3.1-25
.3.1-25 3.3-2 3.3-2 3.3-3 3.3-3 3.3-5 3.3-5 3.3-6 3.3-6 3.3-7 3.3-7 3.3-10 3.3-10 3.3-13 3.3-13 3.3-15 3.3-15 3.6-1 3.6-1 3.7-2 3.7-2 3.7-3 3.7-3 3.7-3a 3.8-2 3.8-2 3.8-3 3.8-3 3.8-4 3.8-4 3.10-4 3.10-4 3.10-7 3.10-7 3.10-14 3.10-14 3.10 3.10-15 3.10 3.10-16.
3.11-1 3.11-1 3.12-1 3.'12-1 4.4-3 4.4-3 5.3-2 5.3-2 5.4-1 5.4-11 r
I i
.. - ~
1.9.2 Instrument Channel Functional Test Injection of a simulated signal into the channel to verify that it is operable, including alarm and/or trip initiating action.
1.9.3 Instrument Channel Calibration Adjustment of channel output such that it responds, with acceptable range and accuracy, to know values of the para-meter which the channel measures. Calibration shall encom-pass the entire channel, including alarm or trip, and shall be deemed to include the channel functional test.
1.9.4 Logic Channel Functional Test The operation of relays or switch contacts, in all the combinations required, to produce the required output.
1.10 CONTAINMENT INTEGRITY Containment integrity is defined to exist when:
1.10.1 All non-automatic containment isolation valves which are not required to be open during accident conditions, except those required to be open for normal plant operation or testing as identified in Table 3.6-1, are closed and blind flanges are installed where required.
1.10.2 The equipment door is properly closed.
1.10.3 Both doors in each personnel air lock are properly closed unless being used for entry, egress or maintenance, at which time at least one air lock door shall be closed.
1.10.4 All automatic containment isolation valves are either operable or in the closed position, or isolated by a closed manual valve or flange that meets the same design criteria as the isolation valve.
Amendment No. 34 14
1.11 QUADRANT POWER TILT RATIO The quadrant power tilt ratio shall be the ratio of the maximum upper excore dector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is With one excore detector inoperable, the remaining three detectors greater.
shall be used for computing the tverage.
1.12 SUijVEILLANCE INTERVAL with the exception of those with shift awl Each Surveilli.nce Requirement, daily frequencies, shall be performed within the simci t ied tirne interval with:
A maximum a110wat>1e extension not to exceed 2 'it of the surveillance interval; and A total naximum combined interval time for any three (3) b.
consecutive surveillance interval not to exceed 3.25 times the specified surveillance interval.
1.13 OPERATION IN A DEGRADED MODE The plant is said to be operating in a degraded mode when it is operating with one or more systems listed herein inoperable as permitted by the Technical If inoperable components or systems are subsequently made Specifica; ions.
operable, the action statements requiring plant shutdown.no longer apply.
1.14 E-AVERAGE DISINTEGRATION ENERGY Noble gas E shall be the average (weighted in proportion to the concentration of of the sum of each radionuclide in the reactor coolant at the time of sampling)
(in MeV) for isotopes the average beta and gamma energies per disintegration with half lives greater than 10 minutes, making up at least 95% of the total activity in the coolant.
1.15 DOSE EQUIVALENT I-131_
l DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcurie / gram) whi:;h alone would produce the same thyroid dose as the quantity and isotopic mixture of The thyroid dose con-I-131, I-132, I-133, I-134, and I-135 actually present.
version factors used for this calculation shall be those listed in Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites."
l Amendment No.
34 1-5 L
C.
MINIMUM CONDITIONS FOR CRITICALITY 1.
Except during low power physics test, the reactor shall not be made critical at any temperature above which the moderator temperature coefficient is positive.
2.
The reactor shall not be brought to a critical condition until the pressure temperature state is to the right of the criticality limit lina shown in Figure 3.1-1, 3.
At all times during critical operation, T,yg should be no lower than 450 F.
4.
The reactor shall be maintained suberitical by at least lti k
-T until normal water level is established in the pressurizer.
Basis:
During the early part of the initial fuel cycle, the moderator temperature coefficient is calculated to be slightly positive at coolant temperatures below the power operating range.
The moderator coefficient at low temperatures will be most positive at the beginning of life of the fuel cycle, when the boron concentration in the coolant is the greatest. Later in the life of the fuel cycle, the boron concentration in the coolant will be lower and the moderatior coefficient will be either less positive or will be negstive. At all times, the moderator coefficient is negative in the power operating range.
Suitable physics measure.ients of moderator coefficient of reactivity will be made l
l as part of the startup program to verify analytic predictions.
l The requirement that the reactor is not to be made critical when the moderator l
i coefficient is positive has been imposed to prevent any unexpected power excursion l
This l
during normal operations as a result of an increase in moderator temperature.
requirement is waived during low power physics tests to permit measurement of reactor I
, Amendment No. "34 -
3.1-12
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-r s
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moderator coefficient and other physics design parameters of interest. During physics tests, special operating precautions will be taken.
The requirement that the reactor is not to be made critical except in accordance with Figure 3.1-1 provides increased assurance that the proper relationship between reactor coolant pressure and temperature will be maintained during system heatup and pressurization whenever the reactor vessel is in the nil-ductility temperature range. Heatup to this temperature will be accomplished by operating the reactor coolant pumps.
The requirement for bubble formation in the pressurizer w'.en the reactor has passed the threshold of 14 suberiticality will assure that the Reactor Coolant not be solid when criticality is achieved.
References:
L.
FSAR Table 3.2.1-1 2.
FSAR Figure 3.2.1-9 f
Amendment No. 34 3.1-13
i F.
LEAKAGE OF REACTOR COOLANT Specification 1.
If leakage of reactor coolant is indicated by the means available such as water inventory balance, monitoring equipment or direct observation a follow-up evaluation of the safety implications shall be initiated as practicable but no later than within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Any indicated leak shall be considered to be a real leak until it is determined that the indicated leak cannot be substantiated by direct observation or other indication.
2.
If the leakage rate, excluding controlled leakage sources such as the Reactor Coolant Pump Controlled Leakage Seals and Leakage into Closed Systems, exceeds 1 gpm and the source of leakage is not identified, reduce the leakage rate to within limits within four hours or be in hot shutdown within the next six hours er,d in cold shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
3.
If the sources of leakage are identified and the results of the evaluation are that continued operation is safe, operation of the reactor with a total leakage, other than from controlled sources or into closed systems, not exceeding 10 gpm shall be permitted except as specified in 3.1.F.4 below.
Amendment No.
34 3.1-21 4
Maccuremtnt of the lockcga rcto to the containment atmogphare in alro possible
- However, through humidity detection and condensation collection and measurement.
it is expected that the containment activity method will give the initial indi-cation of coolant leakage. The other methods will be employed primarily to confirm that leakage exists, to indicate the location of the leakage sources, and to measure the leakage rate.
As described above, the four reactor coolant leak detection systems are based on three different principles, i.e., activity, humidity and condensate flow measure-Two systems of different principles provide, therefore, diversified ways ments.
of detecting leakage to the containment.
Total reactor coolant leakage can be determined by means of periodic water inventory balances.
If leakage is into another closed system, it will be detected by the plant radiation monitors and/or inventory control.
Four hours is allowed from the time of leakage detection to identify the leakage source and to measure the leakage rate. This time period is required since identification and quantification of leakage sources of less than ten gallons per minute require a careful gathering and evaluation of data and/or a visual inspec-tion of the reactor coolant system.
The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those limits found to result in negligible corrosion of the steam generator tubes.
If stress corrosion cracking occurs, the extent cf cracking during plant operation would be limited by the limitation of steam gene-rator leakage between the primary coolant system and the secondary coolant system.
Cracks having a primary-to-secondary leakage less than 500 gallons per day during operation will have an Amendment No.
34 3.1-25
?-*v ee
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One residual heat removal pump and heat exchanger together c.
with the associated piping and valves operable.
One recirculation pump together with its associated piping d.
and valves operable.
2.
If the Safety Injection and Residual Heat Removal Systems are 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> not restored to meet the requirements of 3.3.A.1 within the reactor shall be in the cold shutdown condition within the next 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />.
shall not exceed 350 F unless 3.
The reactor coolant system T,yg the following requirements are mets The refueling water storage tank contains a minimum of 346,870 a.
gallons of water at a boron concentration of at least 2000 ppm.
The boron injection tank contains 900 gallons of a boric acid b.
solution of 11-1/2% to 13% by weight (20,000 ppm to 22,500 ppm of boron) at a temperature of at least 145 F.
Two channels 0
of heat tracing shall be operable for that portion of the flow l
path bounded by the boron injection tank inlet and outlet motor l
operated valves and the recirculation flow path to and from the l
i boric acid tanks.
The four accumulators are pressurized between 600 and 700 psig I
c.
3
[
3 and a maximum of 815 ft l
and each contains a minimum of 800 ft Accumu-of water at a boron concentration of at least 2000 ppm.
i lator isolation valves 894A, B, C, and D shall be open and their power supplies de-energized whenever the reactor coolant system pressure is above 1000 psig.
34 Amendment No.
3.3-2
's.
11 t he Safet y In je. t ion and Ronidual lle.it Ren. val Synt em ;
are not restored to meet the requirements of 3.3.A.3 within the time periods specified in 3.3.A.4; then:
a.
If the reactor is critical, it shall be in the hot shutdown condition within four hours and the cold shutdown condition within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
b.
If the reactor is subcritical, the reactor coolant system temperature and pressure shall not be increased more than 25 F and 100 psi, respectively, over existing values.
If the requirements of 3.3. A.3 are not satisfied within an additional 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the reactor shall be brought to the cold shutdown procedures.
The shutdown shall start no later than the end of the 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> period.
B.
Containment Cooling and Iodine Removal Syst ems 1.
The reactor shall not-be brought above the cold shutdwon condition unless the following requirements are met:
a.
The spray additive tank contains a minimum of 4000 gallons of solution with a sodium hydroxide concentration of not less than 30% by weight.
b.
The five fan coolcr-charcoal filter units and the two spray pumps, with their associated valves and pipino, are ora 7rable.
2.
The rc<pii rements of 1.1.H.1 may be modified to allow any one of the following cominnents to be inoperabic at one time:
Amendment No.
34 3.3-5
d f
a.
Van co-ter unit 12. 34, or 35 or the flow path -for fan coolant unit 32, 34, or 35 may be out'os service for a.
period not to exceed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, provided both containment spray pumps-are demonstrated to be operable.
i OR Fan cooler unit 37. or 33, or the flow path for fan cooler t
unit 31 or 33 may be cut of service for a' period not.to exceed 7 days provided both containment spray pumps are denonstrated daily to be operable.
b.
One containment spray pump may be out of service for a
~
period not to exceed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, provided the five fan cooler units are operable and the remaining containment spray pump is demonstrated to be operable.
c.
Any valves required for the functioning of the system during and following accident conditions may'be inoperable provided it is restored.to an operable status within 24'
].
hours and all' valves in the system that provide'the. dup-
.licated function are demonstrated to be operable.
i 3.
If the Containment Cooling and Iodine Removal ' arc not restored to meet the requi rewnt s of 1.3.11.1 within. the t imo; period. specified
.e in 3.3.H.2, ther..
a.
I f t he rtactor is critical, it nhall-be in-tlie hot'ahutdown
-condition within.four hours and in the cold shutdown condition'
~
l within' the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, T
b.
If the reactor _i's suberitical,~the reactor coolant system.
temperature and-pressure shall not be increased moresthanj250F and 100 psi,~ respective, over existiniJ values. If.the requirements
~
~
of 3.3. A.3 are not satis fied within an additional 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, tho '
reactor shall' he broughti tlo ;the cold, shutdown condition. utilizing,
' normal-olerating pro <sedures. The shutdownishalle start no :later. thari I
i t.he ; cnd of f t he.4H -hour period..
Rtc19 %D9NR N R
. T. W.
C.
Isolation Valve Seal Water System (VSWS) 1.
The reactor shall not be brought above cold shutdown un.1.ess the following requirements are met:
a.
The IVSWS shall be operable.
b.
The IVSW tank shall be maintained at a minimum pressure of 45 psig and contain a minimum of 144 gallons of water.
2.
The requirements of 3.3.C.1 may be modified to allow any one of the fallowing components to be inoperable at any one time:
a.
Any one header of the IVSWS nw r be inoperable for a period not to exceed 4 consecutive days.
b.
Any valve required for the functioning of the system during and following accident conditions provided it is restored to an operable status within 4 days and all valves in the system that provide a duplicate function are demonstrated to be operable.
3.
If the IVSW System is not restored to an operable status within the time period specified, then:
i a.
If the reactor is critical, it shall be brought to the I
hot shutdown condition utilizing normal operating procedures.
The shutdown shall start no later than at the end of the specified time period.
l Amendment No. 34 3.3-7
1 IT the component C< nal i niq ;iynt em i t, n<>t restot ett t o ineet t.he requirements of 3.3.E.1 within the tirne periodo specified in 3.3.E.2, thent a.
If the reactor is critical, it shall be in the hot shutdown condition within four hours and in the cold shutdown conditien within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
b.
If the reactor is subscritical, the reactor coolant system temperature and pressure shall not be increased more than 25"F and 100 psi, respectively, over existing values.
If the requirements of 3.3.A.3 are not satisfied within an additional 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the reactor shall be ifrought to the cold shutdown condition utilizing normal operating proceduren. The shutdown shall start no later than the end of the 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> period.
F.
Service Wat er Syntem 1.
The reactor shall not be brought above cold shutdown unless three service water pumps on the designated essential header and a minimum of two service water pumps on the designated non-essential header together with their associated piping and valven are operabic.
2.
When the reactor in above cohl shut down and i t' the requirements of 1.1.r.1 cannot 1,e met within twelve hours, the r. actor shall be brought to the cold shiit down condition, starting no lat cr than the end of the twelve hour period, utilizing normal operating procedures.
Amendment No. 34 3.3-10
2.
Thw rcquirements of 3.3.H.1 may be modificd c3 followas The control room ventilation system may be inoperable a.
At the for a period not to exceed seventy-two hours.
end of this period if the mal-condition in the control room ventilation system has not been corrected, the reactor shall be placed in the hot shutdown condition utilizing normal operating procedures.
If after an additional 48' hours the mal-condition still exists, the reactor shall be placed in the cold shutdown condition utilizing normal operating procedures.
Basis to heat the rer ar The normal procedure for starting the reactor is, first, coolant to near operating temperature, by running the reactor coolant pumps.
The reactor is then made critical by withdrawing control rods and/or diluting boron in the coolant. III With this mode of startup, the energy stored in the reactor coolant during the approach to criticality is substantially equal to that during power operation, and, therefore, the minimum regnired engineered safeguards and auxiliary cooling systems are required to be operable.
The probability of sustaining both a major accident and a simultaneous failure of a safeguards component to operate as designed is necessarily very small.
(
Thus, operation with the reactor above the cold shutdown condition with minimum safeguards operable for a limited period does not significantly increase the probability of an accident having consequences which are more severe than the Design Basis Accident.
The operable status of the various systems and components is demonstrated by periodic tests defined by Specification 4.5.
A large fraction of 1
3.3-13 Amendment No. 34
Assuming the reactor has been operating at full rated power, the magnitud2 Thus, the require-of the decay heat decreases after initiating hot shutdown.
ment for core cooling in case of a postulated loss-of-coolant accident while in the hot shutdown condition is significantly reduced below the requirements Putting for a postulated loss-of-coolant accident during power operation.
tial the reactor in the hot shutdown condition significantly reduces the poten and also allows more free access consequences of a loss-of-coolant accident,to some of the engineered safeg hour of going to the hot shutdown j
Failure to complete repairs within 1 condition is considered indicative of a requirement for major maintenance and, therefore, in such a case the reactor is to be put into the cold shutdown condition.
The limits for the Boron Injection Tank, Refueling Water Storage Tank, and the accumulators insure the required amount of water with the proper boron concentration for injection into the reactor coolant system following a loss-of-These limits are based on values used in the coolant accident is available.
(13) accident analysis.I9)
The specified quantities of water for the RWST include unavailable waterin the alarm getpcints in the tank bottom, inaccuracies (1406 gals)
(3) and recir-(4687 gals) and minimum quantities required during injection (246,000 gals)346,870 gals) provides (80,000 gals).I4) The minimum RWST (e.g.,
culation phases approximately 13,370 gallons margin.
are maintained in the The four accumulator isolation valves (894 A. B, C, D) open position when the reactor coolant pressure is above 1000 psig to assure flow passage from the accumulators will be available during the injection of a loss-of-coolant accident.
panel, should any of these valves not be in the full open position even with the The 1000 psig limit is derived from the minimum valve operator de-energized.
pressure requirements of the accumulators combined with instrument error and a operational band and is based upon avoiding inadvertent injection int reactor coolant system.
energized to prevent an extremely Amendment No. 34 3.3-15
3.6 CONTAINMENT SYSTEM Applicability Applies to the integrity of reactor containment.
objective To define the operating status of the reactor containment for plant operation, specification A.
Containment Integrity 1.
The containment integrity (as defined in 1.10) shall not be violated unless the reactor is in the cold shutdown condition. However, those non-automatic valves listed in Table 3.6-1, may be opened if necessary for plant operation and only as long as necessary to perform the intended function.
2.
The containment integrity shall not be violated when the reactor vessel head is removed unless the boron concen-tration is sufficient to maintain the shutdown margin
> 10% &k,
k 3.
If the containment integrity requirements are not met when the reactor is above cold shutdown, containment integrity shall be restored within one hour or the reactor shall be in the hot shutdown condition within six hours and in cold shutdown condition within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
B.
Internal Pressure If the internal pressure exceeds 2.5 psig or the internal vacuum exceeds 2.0 psig, the condition shall be corrected or the reactor shutdown.
Amendment No. 34-3.6-1
and is in addition to the fuel requirements for other nuclear units on the site.
6.
Three batteries plus three chargers and the D. C. distribution systems operable.
7.
No more than one 120 volt A. C. Instrument Bus in the backup lighting supply.
The requirements of 3.7.A may be modified to allow any one of the following B.
power supplies to be inoperable at any one time.
1.
One diesel or any diesel fuel oil system or a diesel and its associated fuel oil system may be inoperable for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> f
provided the 138 KV and the 13.8 KV sources of offsite power are available and the remaining diesel generators are tested daily to ensure operability and the engineered safety features associated with these diesel generator buses are operable.
2.
The 138 KV or the 138 KV sources of power may be inoperable for This 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> provided the three diesel generators are operable.
operation may be extended beyond 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> provided the failure is reported to the NRC within the 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> period with an outline of the plans for restoration of offsite power and NRC approval is granted.
l l
3.
One battery may be inoperable for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> provided the other batteries and the three battery chargers remain operable with one battery charger carrying the D. C. load of the failed batterar i
supply system.
l Amendment No.34 3.7-2
C.
If the electrical distribution system is not restored to meet the requirements of 3.7.A within the time periode specified in 3.7.B, ti.en :
1.
If the reactor is critical, it shall be in the hot shutdown condition within six hours and in the cold shutdown condition within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
2.
If the reactor is suberitical, the reactor coolant system temperature and pressure shall not be increased more than 25 F and 100 psi, respectively, over existing values.
D.
The requirements of Specification 3.7.A.1 may be modified during an emergency system-wide blackout condition as follows:
Two of the three 13.8 KV feeders (13W92,13W93 and/or 13W94) to the Buchanan Substation 138 KV buses operable with at least 37 MV power 0
from any combination of gas turbines (nameplate rating at 80 F) at the Buchanan Substation and onsite and onsite available for exclusive use on Indian Point Unit No. 3.
E.
Whenever the reactor critical, the circuit breaker on the electrical feeder to emergency lighting panel 318 inside containment shall be locked open except when containment access is required.
As a minimum, under all conditions including cold shutdown, the following F.
A.C. electrical power sources shall be operable:
One transmission circuit to Buchanan Substation, except for testing.
1.
2.
Either 6.9 KV buses 5 or 6 energized from the 138 KV feeder 95331 a.
or 95332, or 13.8 KV feeder 13W92 or 13W93 and its associated 13.8/6.9 KV b.
transformer available to supply 6.9 power, 3.
Two of the four 4BO-volt buses 2A, 3A, 5A and 6A energized.
Amendment No.34
.3.7-3
4.
Two operable diesel generators together with tots 1 underground storage containing a minimum of 5676 gallons of fuel.
Basis ne electrical system equipment is arranged so that no single contingency can inactivate enough safeguards equipment to jeopardize the plant safety. We 480-volt equipment is arranged on 4 buses. The 6900-velt equipment is supplied from 6 buses.
The Buchanan Substation has both 345 KV and 138 KV transmission circuits which are capable of supplying startup, normal operation, shutdcwn and/or engineered safeguards loads.
The 138 KV supplies or the gas turbines are capable of providing sufficient power for plant startup. Power via the station auxiliary transformer can supply all the required plant auxiliaries during normal operation, if required.
In addition to the unit transformer, four separate sources supply station service power to the plant.(1)
Amendment No.34 3.7-3a
7.
The containment vent and purge system, including the radiation monitors which initiate isolaticn, shall be tested and verified to be operable within 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> prior te refueling operations.
3.
No movement of irradiated fuel in the reactor shall be t
made until the reactor has been subcritical for at least 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br />.
In addition, movemen t of fuel in the reactor before the reactor has been subcritical for equal to or greater than 365 hours0.00422 days <br />0.101 hours <br />6.035053e-4 weeks <br />1.388825e-4 months <br /> will necessitate, operation of the Containment Building Vent and Purge System through the
!! EPA filters and charcoal adsorbers.
For this case operability of the Containment Duilding ' lent and Purge,
System shall be estab ished in accordance with Section 4.13 Of the Technical Specifications.
In the event that more than one region of fuel (72 assemblies) is to be-dischargeed from the reactor, those assemblies in excess of one recion shall not be discharged-before an interval of 400 hours0.00463 days <br />0.111 hours <br />6.613757e-4 weeks <br />1.522e-4 months <br /> has elapsed after shutdown.
3.
Whenever movement of irradiated fuel is being~ made, the minimum water level in the area of movement shall be maintained 23 feet oter tha top of the reactor pressure
-l I
vessel flange.
10.
Iloists or cranes utilized in handling irradiated fuel shall' be dead-load tested before movement begins.
The-load assumed by the hoists or cranes for this test must be equal to or greater than maximum load to be assumed by the. hoists or crancs during the refueling operation.
A thorough visual inspection of the hoists or cranes shall.be made after the idead-load test and prior to fuel. handling.
A test of inter-Iccks shall also be performed, 11.
The fuel. storage building emergency ventilation system shall be cperable whenever irradiated tuel is'being-handled within the fuel storage building.
The emergency ventilation system may be inoperable when irradiated fuel is-in'the fuel storage building,-provided. irradiated fuel ~is'not being handled and neither the spentifuel cask-nor the-cask crane are moved over the spent fuel pit during the period of inoperability.
Amendment ::c, '4 34 3.3-2
.m.
If any of the specified limiting condition for refueling are not met, B.
refueling shall cease until the specified limits are met, and no operations which may increase the reactivity of the core shall be made.
During fuel handling and storage operations, the following conditions shall C.
be met.
1.
Radiation levels in the spent fuel storage area shall be monitored continuously whenever there is irradiated fuel stored therin.
If the monitor is inoperable, a portable monitor may be used.
2.
The spent fuel cask shall not be moved over any region of the spent fuel pit which contains irradiated fuel. Additionally, if the spent fuel pit contains irradiated fuel, no loads in excess of 2,000 pounds shall be moved over any region of the spent fuel pit.
3.
During periods of spent fuel cask or fuel storage building cask crane movement over the spent fuel pit, or during periods of spent fuel movement in the spent fuel pit, when the pit contains irradiated fuel, the pit shall be filled with borated water at s concentration of
>>1000 ppm.
4.
Whenever movement of irradiated fuel in the spent fuel pit is being made, the minimum water level in the area of movement shall be main-tained 23 feet over the top of irradiated fuel assemblies seated in the storage rack.
5.
Hoists or cranes utilized in handling irradiated fuel shall be dead-load tested before fuel movement begins. The load assumed by the hoists or cranes for this test must be equal to or greater than the maximum load to be assumed by the hoists or cranes during the fuel handling operation. A thorough visual inspection of the hoists or cranes shall be made after the dead-load test prior to fuel handling.
Inspection of fuel handling equipment is to include testing of overload cutoff devices on the manipulator.
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Amendment No. M4 3.8-3 L_..__________________
5.
The fuel storage building emergency ventilation system shall be operable whenever irradiated fuel is being handled within the fuel stcrage building.
The emergency ventilation system may be inoperable when irradiated fuel is in the fuel storage building, provided irradiated fuel is not being handled and neither the spent fuel cask nor the cask crane are moved over the spent fuel pit during the periods o f inoperability.
B_a_ sis Th-o m:: men t and general : recedures to be utilized during refueling, f7el h,ndlin7, and starano are a.scussed in the ?sAR.
Detailed instructions, the ahcve spectfied precautiens, and the design of the fuel handling equipment incorporating built-in interlocks and safety features, provide assurance that no incident could occur during the refueling, fuel handling, reacter maintenance or storage operations that would result in a hazard to public health and safety. (1)
Whenever changes are not being made in core geometry, one flux monitor is sufficient.
This permits maintenance of the instrumentation.
Continuous monitoring of radiation levels and neutrcn flux provides immediate indication of an unsafe condition.
The residual heat removal pump is used to maintain a uniform baron concentration.
The shutdown margin indicated will keep the core subcritical, even if all control rods were withdrawn from the core.
During refueling the reactor refueling cavity is filled with approximately 342,000 gallons of water from the refueling water storage tank with a boren cencentration of 2000 ppm.
A shutdown margin of 10% AK/K in the cold condition with all rods inserted will also maintain the core sub--
c'ritical even if no control rods were inserted into the reactor.(~i Pericdic checks of refueling water boron concentratien and residual heat removal pump operation insure the proper shutd wn margin.
The requirement for direct communications allows the control recm operator to inform the manipulator operater of any imponding unsafe condition detected from the main control board indicators during fuel movement.
l In addition to the above safeguards, interlocks are utilized l
during re fueling to ensure safe handling.
An excess weight interlock is provided en the lifting hoist to prevent movement of more than one fuel asscmbly at a time.
The spent fuel transfer mechanism can accomodate only one fuel assembly at a time.
The 120-hour decay time following the suberitical condicion and the 23 feet of water above the top of the reactor pressure vessel flange is consistent with the assumptions used in the dose calculatien for the fuel-handling accident.
The waiting time of 400 hours0.00463 days <br />0.111 hours <br />6.613757e-4 weeks <br />1.522e-4 months <br /> required following plant shutdown before unloading more than one recion of fuel from the reactor assures that the maximum peol water cemperature will be within desian objectives as stated in the F5AR.
I AT.e nd me n t No. J4 34 3.3-4 L
Alarms are provided to indicate non-conformance with the flux 3.10.2.8, difference requirements of 3.10.2.5.1 and the flux difference-time requirements of 3.10.2.6.1.
If the alarms are temporarily out of service, conformance with the applicable limit shall be demonstrated by logging the flux difference at hourly intervals for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and half-hourly thereafter.
If the core is operating above 75% power with one excore nuclear 3.10.2.9 channel out of service, then core quadrant power balance shall be determined once a day using movable incore detectors (at least two thimbles per quadrant).
3.10.3 Quadrant Power Tilt Limits When ever the indicated quadrant power tilt ratio exceeds 1.02, 3.10.3.1 except for physics tests, within two hours the tilt condition shall be eliminated or the following actions shall be taken Restrict core power level and reset the power range high f
a) flux setpoint three percent of rated value for every percent of indicated power tilt ratio exceeding 1.0, and b)
If the tilt condition is not eliminated after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the power range nuclear instrumentation setpoint shall be rest to 55% of allowed power. Subsequent reactor operation is per-mitted up to 50% for the purpose of measurement, testing and corrective action.
Except for physics tests, if the indicated quadrant power tilt 3.10.3.2 ration exceeds 1.09 and there is simultaneous indication of a misaligned control rod, restrict core power level 3% of rated l
value for every percent of indicated power tilt ratio exceeding 1.0 If the rod is not realigned and realign the rod within two hours.
within two hours or if there is no simultaneous indication of a mis-alighned rod, the reactor shall be brought to the hot shutdown condition If the reactor is shut down, subsequent testing up to 50%
within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of rated power shall be permitted to determine the cause of the tilt.
Amendment No.34
-~ -
3.10-4
If a control rod having a rod position indicator chann21 out of 3.10.6.3 service, is found to be misaligned from 3.10.6.la. above, then Specification 3.10.5 will be applied.
3.10.7 Inoperable Rod Limitations An inoperable rod is a rod which does not trip or which is 3.10.7.1 declared inoperable under Specification 3.10.5 or fails to meet the requirements of 3.10.8.
Not more than one inoperable control rod shall be allowed any 3.10.7.2 time the reactor is critical except during physics tests Otherwise, the plant requiring intentional rod misalignment.
shall be brought to the hot shutdown condition.
If any rod has been declared inoperable, then the potential 3.10.7.3 ejected rod worth, associated transient power distribution factors and the accidents listed in Table 3.10-1 shall peaking be analyzed within 5 days, or the reactor brought to the hot shutdown ec ; 11 tion using normal operating procedures. The analysis shall include due allowance for non-uniform fuel depletion in the neigh-borhood of the inoperable rod.
If the analysis results in a more limiting hypothetical transient than the cases reported in the safety analysis, the plant power level shall be reduced to an analytically j
determined part power level which is consistent with the safety analysis.
I 3.10.8 Rod Drop Time At operating temperatu.te and full flow, the drop time to each control rod shall be no greater than 1.8 seconds from loss of stationary gripper coil voltage to dashpot entry.
34 3.10-7 Amendment No.
l
described below.
The radial power distribution within the core must satisfy the design values assumed for calculation of power capability. Radial power die ributions are measured as part of the startup physics testing and are periodically measured at a monthly or greater frequency. These measurements are taken to assure that the radial power distribution with any quarter core radial power asymmetry conditions are consistent with the assumptions used in power capability analyses.
It is not intended that reactor operation would continue with a power tilt condition which exceeds the radial power asymmetry considered in the power capability analysis.
The quadrant tilt power deviation alarm is used to indicate a sudden orThe unexpected change from the radial power distribution mentioned above.
two percent tilt alarm setpoint represents a minimum practical value consis-tent with instrumentation errors and operating procedures, This asymmetry level is su. icient to detect significant misalignment of control rods.
Misalignment of control rods is considered to be the most likely cause of radial power asymmetry. The requirement for verifying rod position once each shift is imposed to preclude rod misalignment which would cause a tilt condition less than the 2% alarm level.
The two hour time interval in this specification is considered ample to identify a dropped or misaligned rod and complete realignment procedures to eliminate the tilt.
In the event that the tilt condition cannot be eliminated within the two hour time allowance, additional time would be needed to investigate the casue of the tilt condition. The measurements would include For a tilt a full core physics map utilizing the moveable detector system.
condition < 1.09, an additional 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br /> time interval is authorized to accomplish these measurements.
However, to assure that the peak core power is maintained below limiting values, a reduction of reactor power of three per-l cent for each one percent of indicated tilt is required. Physics measurements have indicated that the core radial power peaking would not exceed a two to one relationship with the indicated tilt from the excore nuclear detec*or system for the worst rod misalignment.
In the event a tilt condition of ( l.09 cannot be eliminated after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the reactor power level will be reduced to the range required for low power physics testing. To avoid reset of a large number of portection setpoints, the power range nuclear instrumentation would be reset to cause an automatic reactor trip at 55% of allowed power. A reactor trip at this power has been selected to prevent, with margin, exceeding core safety limits even with a nine percent tilt condition.
If tilt ratio greater than 1.09 occurs which is not due to a misaligned rod, the
- However, reactor shall be brought to a hot shutdown condition for investigation.
if the tilt condition can be identified as due to rod misalignment, operation can l
continue at a reduced power (3% for each one percent the tilt ratio exceeds 1.0) for two hours to correct the rod misalignment.
Trip shutdown reactivity is provided consistent with plant safety analysis assumptions. One percent shutdown is adequate except for' steam break analysis, Figure 3.10-1 which requires more shutdown if the boron concentration is low.
is drawn accordingly.
M 3.10-14 Amendment No.
Rod insertion limits are used to assure adequate trip reactivity, to assure meeting power distribution limits, and to limit the consequence of a hypothetical rod ejection accident. The available control rod reactivity, or excess beyond needs, decreases with decreasing boron concentration because tne negative reactivity required to reduce the core power level from full power to zero power is largest when the boron concentration is low.
The intent of the test to measure control rod worth and shutdown margin (Specification 3.10.4) is to measure the worth of all rods less the worth of The the worst case for an assumed stuck rod, that is, the most reactive rod.
measurement would be anticipated as part of the initial startup program and infrequently over the life of the plant, to be associated primarily with deter-minations of special interest such as end of life cooldown, or startup of fuel cycles which deviate from normal equilibrium conditions in terms of fuel loading patterns and anticipated control bank worth. These measurements will augment the normal fuel cycle design calculations and place the knowledge of shutdown capability on a firm experimental as well as analytical basis.
Operation with abnormal rod configuration during low power and zero power testing is permitted because of the brief period of the test and because special pre-cautions are taken during these tests.
The rod position indicator channel is sufficiently accurate to detect a rod +7 inches away from its demand position. An indicated misalignment less than 12 steps does not exceed the power peaking factor limits.
If the rod position indicator channel is not operable, the operator will be fully aware of the inoperability of the channel, and special surveillance of core power tilt indications, using established procedures and relying on excore nuclear detectors, and/or movable incore detectors, will be used to verify power distribution These indirect measurements do not have the same resolution if the symmetry.
bank is near either end of the core, because a 12 step misalignment would have no effect on power distribution. Therefore, it is necessary to apply the indirect checks following signific&nt rod motion.
One inoperable control rod is acceptable provided that the power distribution limits are met, trip shutdown capability is available, and provided the potential hypothetical ejection of the inoperable rod is r.ot worse than the cases analyzed in the safety analysis report. The rod ejection accident for an isolated fully inserted rod will be worse if the residence time of the rod is long enough to cause significant non-uniform fuel depletion. The 5 day period is short compared.
l with the time interval required to achieve a significant, non-uniform fuel depletion.
The required drop time to dashpot entry is consistent with safety analysis.
REFERENCE WCAP-8576, " Augmented Startup and Cycle 1 Physics Programs, August 1975 1.
2.
FSAR Appendix 14C Amendment No. 34 MYME
TABLE 3.10-1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE EVENT OF AN INOPERABLE FULL LENGTH ROD Rod Cluster Control Assembly Insertion Characteristics Rod Cluster Control Assembly Misalignment Loss Of Reactor Coolant From Small Ruptured Pipes Or From Cracks In Large Pipes Which Actuates The Emergency Core Cooling System Single Rod Cluster Control Assembly Withdrawal At Full Power (Loss Of Coolant Major Reactor Coolant System Pipe Ruptures Accident)
Major Secondary System Pipe Rupture Rupture of a Control Rod Drive Mechanism Housing (Rod Cluster Control Assembly Ejection) l l
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i Amendment No. )&ik 34 3.10-16 g
f i
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3.11 MOVABLE IN-CORE INSTRUMENTATION Applicability Applies to the operability of the movable detector instrumentation system.
Objective To specify functional requirements on the use of the in-core instrumentation system, for the recalibration of the excore axial off-set detection system.
Specification A minimum of 2 thimbles per quadrant and sufficient movable in-core A.
detectors shall be operable during re-calibration of the excore axial off-set detection system.
Power shall be limited to 90% of rated power for 4 loop or 65% of B.
rated power for 3 loop operation if re-calibration requirements for excore axial off-set detection system, identified in Table 4.1-1.
are not met.
During the incore/excore calibration procedure, all full core flux C.
maps will be made only when 75% of the movable detector guide thimbles are operable.
Basis In-core Instrumentation System (1) has six drives, six detectors, The Movable and 50 thimbles in the core. Each detector can be routed to sixteen or Consequently, the full system has a great deal more capability more thimbles.
than would be needed for the calibration of the ex-core detectors.
To calibrate the excore detectors system, it is only necessary that the Movable In-core System be used to determine the gross power distribution indicated by the power balance between the top and bottom in the core as halves of the core.
Amendment No. 34 3.11-1
3.12 RIVER LEVEL Applicability Applies to water elevation of the Hudson River as measured at the Indian Point Unit No. 3 intake structure.
Objective To specify the maximum water elevation of the Hudson River for safe operation of the reactor.
Specification When the Hudson River water elevation as measured at the Indian Point Unit 3 intake structure reaches 11'-0" above mean sea level, sandbagging Unit No.
the service water pumps will be initiated.
If the Hudson River water elevation reaches 12'-5" above mean sea level at the Indian Point Unit No. 3 intake structure, the reactor will be in the hot shutdown condition within six hours and in the cold shutdown condition within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
Basis Analyses have been performed wich indicate that the river water elevation would have to reach 15'-3" above mean sea level before it would seep into the lowest floor elevation of any of the buildings hoursing equipment vital for safe shutdown of the reactor.III Monitoring of the Hudson River water elevation will not be required until there is a flood warning notice dis-eminated by the New York City National Oceanographic and Atmosphere Administration (NOAA) office.
References:
(1)
FSAR, Section 2.5 3.12-1 amendment no. 34
b.
If repairs are not completed and conformance to the acceptance criterion is not demonstrated within 7 days, the reactor shall be shut down until repairs are effected and the continuous leakage meets the acceptance criterion.
C.
Sensitive Leakage Rate 1.
Test A sensitive leakage rate test shall be conducted with the containment penetrations, weld channels, and certain double gasketed seals and isolation valve interspaces at a minimum pressure of 41 psig and with the containment building at atmospheric pressure.
2.
Acceptance Criteria The test shall be considered satisfactory if the leak rate for the containment penetrations, weld channel and other pressurized zones is equal to or less than 0.2% of the containment free volume per day.
3.
Frequency A sensitive leakage rate test shall be performed at a frequency of at least every other refueling but in no case at intervals greater than 3 years.
l D.
Air Lock Tests 1.
The containment air locks shall be tested at a minimum pressure of 40.6 psig and at a frequency of every 6-months. The acceptance criteria is included in E.2a.
The equipment hatch is to be leak rate tested after every reinsertion prior to requiring containment I
(
integrity.
Whenever containment integrity is required, verification shall be 2.
f made of proper repressurization to at least 41 psig of the double-1 l
gasket air lock door seal'upon closing an air lock door.
t Amendment No. 34 4.4-3
The control 5.
There are 53 control rods in the reactor core.
rods contain 142 inch lengths of silver-indium-cadmium alloy clad with the stainless steel. (5)
B.
Reactor Coo 1&nt System The design of tha reactor coolant system compEes with the code 1.
(6)
Design values for system temperature and requirements.
pressure are 571.5 F and 2250 psig respectively.
All piping, components and supporting structures of the reactor 2.
coolant system are designed to class I requirements, and have been designed to withstand the maximum potential seismic ground acceleration, 0.15g, acting in the horizontal and 0.10g acting in the vertical planes simultaneously with no loss of function.
The total liquid volume.of the reactor coolant system, at rated 3.
operating conditior.s, is 11,522 cubic feet.
References (1) FSAR Section 3.2.2 (2)
FSAR Section 3.2.1 (3) FSAR Section 3.2.1 (4) FSAR Section 3.2.3 (5) FSAR Sections 3.2.1 & 3.2.3 (6) FSAR Table 4.1-10 t
1 Amendment No.
34 5.3-2
5.4 FUEL STORAGE Applicability Applies to the capacity and storage arrays of new and spent fuel.
Objactive To define those aspects of fuel storage relating to prevention of criticality in fuel storage areas.
Specification The spent fuel pit structure is designed to withstand the anticipated 1.
earthquake loadings as a Class I structure. The spent fuel pit has a stanless steel liner to insure against loss of water.
The new and spent fuel storage racks are designed so that it is impossible 2.
to insert assemblies in other than an array of vertical fuel assemblies with the sufficient center-to-center distance between assemblies to assure The nominal center-to-center spacing of the storage racks k,ff { 0.95.The capacity of the spent fuel pit is 840 assemblies and is 1 foot.
the new fuel area is 72 assemblies.
Whenever there is fuel in the pit (except in the initial core loading),
3.
the spent fuel storage pit is filled and borated to the concentration to match that used in the reactor cavity and refueling canal during refueling operations.
Fuel assemblies that contain more than 44.46 grams of uranium -235, or 4.
equivalent, per axial centimeter of fuel assembly shall not be stored in the spent fuel pit.
34 5.4-1 Amendment No.