ML19210B124
| ML19210B124 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 04/06/1977 |
| From: | Reid R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19210B123 | List: |
| References | |
| NUDOCS 7911040059 | |
| Download: ML19210B124 (11) | |
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UNITED STATES l'.,
5 NUCLEAR REGULATORY COMM!sstON f i.".
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METROPOLITAN EDISON COMPANY JERSEY CENTRAL POWER AND LIGHT C0"PANY PENNSYLVANIA ELECTRIC COMPANY DOCKET NO. 50-289 THREE MILE ISLAND NUCLEAR STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 28 License No. DPR-50 1.
The Nue' ear Regulatory Coninission (the Comission) has found that:
A.
The application for amendment by Metropolitan Edison Company, Jersey Central ?cwer & Light Company, and Pennsylvania Electric Compa?y (the licensees) dated October 8,1976, as supplemented Oc<cber 21, 1976, and February 3,1977, 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 confonnity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance (1) that the actlyities 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 satisfled.
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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-50 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 28. 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 hWh b
- Robert W. Reid, Chief Operating Reactors Branch #4 Division of Operating Reactors
Attachment:
Changes to the Technical Specifications Date of Issuance:
April 6,1977 1556 219
ATTACHMENT TO LICENSE AMENDMENT NO. 28 FACILITY OPERATING LICENSE NO. DPR-50 DOCKET NO. 50-289 Remove Pages Insert Paces 2 2-7 2 2-7 2-9 2-9 Figure 2.3-1 Figure 2.3-1 3-1 & 3-2 3-1 & 3-2 The changed areas on the revised pages are shown by marginal lines.
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2.2 SAFE'"Y LIMITS - PEAC*CR SYSTD! PUESSUEE Aeolicability Applies to the limit en reactor coolant system pressure.
Objective To maintain the integrity of the reactor coolant system and to prevent the release of significant amounts of fission product activity.
Srecification 2.2.1 The reactor coolant system pressure shall not exceed 2750 psig when there are fuel assemblies in the reactor vessel.
Bases _
The reactor coolant system (1) serves as a barrier to prevent radionuclides in the reactor coolant from reaching the atmosphere.
In the event of a fuel cladding failure, the reactor coolant system is a barrier against the release of fission products. Establishing a system pressure limit helps to assure the integrity of the reactor coolant system. The maximum transient pressure allevable in the reactor coolant system pressure vessel under the ASME Code,Section III, is 110%
of design pressure.(2)
The maximum transient pressure allowable in the reactor coolant system piping, valves, and fittings under ANSI Section B31.7 is 110% of design pressure. Thus, the safety limit of 2750 psig (110% of the 2500 psig design pressure) has been established.(2) The maxi:mim settings for the reactor high pressure trip (2h05 psig) and the pressurizer code safety valves (2500 psig)
(3) have been established for Cycle 3 in accordance with ASME Boiler and Pressure Vessel Code,Section III, Article 9, Winter, 1968 to assure that the reactor coolant system pressure safety limit is not exceeded. The initial hydrostatic test was conducted at 3125 psig (125%
of design pressure) to verify the integrity of the reactor coolant system. Additional assurance that the reactor coolant system pressure does not exceed the safety limit is provided by setting the pressurizer electromatic relief valve at 2255 psig. (4)
Beferences (1)
FSAR, Section h (2)
FSAR, Section h.3.10.1 (3)
FSAR, Section h.2.h (h)
FSAR, Table h-1 n'}
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) J., a D L-2h Amendment No.
, 28
2.3 LIMITINO SAFETY CY"Gt CFTI"T, TECTECTICN INC'~m7NTATION Annlicability Applica to instruments monitoring reactor power, reactor power imbalance, reactor coolant system precoure, reactor ecolant cutlet te=perature, flow, number of pumps in operation, and high reactor building preccure.
Obiectiva To provide automatic protecticn action to pre's..t any combination of procecb variables from exceeding a cafety limit.
Specifientien 2.31 The reactor protection cycten trip cetting licits and the perniccible bypacces for the instrument channels chall be as ctated in Table 2.3-1 and Figure 2 3-2.
Banen The reactor protection cystem concists of four instrument channels to moniter each of several celected plant cenditions which vill cause a reacter trip if any one of these conditions deviates frcm a pre-selected operating range to the degree that a safety limit may be reached.
The trip cetting limits for protection system instrumentation are listed in Table 2.3-1.
These trip setpoints are setting limita-en the setroint side of the protection system tictable ecmparators. The cafety analysis has been based upon these protection system inctrumentation trip set points plus calibratien and instrumentation errors.
Luclear Overrever A reactor trip at high power level (neutron flux) is provided to prevent damage to the feel cladding from reactivity excursiens too rapid to be detected by preccure and teaperature =cacurements.
During normal plant operation with all reactor coolant pumps operating, reactor trip is initiated when the reacter pcVer level reachec 105.5% of rated power. Adding to this the possible variatien in trip cet points due to calibration and instrument errors, the maximum actual power at which a trip would be actuated could be 1125, which is the value used in the safety analysis (1).
a.
OverpcVer trip based en flow and imbalance The power level trip cet point produced by the reactor coolant system flow is based en a power-to-flow ratio which has been establiched to accenmodate the moct severe thermal transient censidered in the design, the less-of-cociant flew accident frem high power. Analysis has de=enstrated that the specified power to f1cv ratio is adequate to prevent a DUER of less than 1.3 chould a lov flow condition exist due to any malfunction.
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The power level trip set point produced by the power-to-flow ratio provides both high power level and Icv flow protection in the event the reactor pcVer level increaces or the reactor coolant flow rate decreases. The pcver level trip set point produced by the power to flev ratio provides cverpower DN3 protection for all modes of pump operation. For every flow rate there is a maximum permissible pcVer level, and for every pcVer level there is a minimum permissible lov flow rate.
Typical power level and icv flev rate combinations for the punp situations of Table 2 3-1 are as follevs:
1.
Trip would occur when four reactor coolant punps are operating if power is 103 percent and reactor flev rate is 100 percent, or flow rate is 92.6 percent and pcver level is 100 percent.
2.
Trip would cecur when three reactor coolant pu=ps are operating if power is 60 7 percent and reacter flow rate is Th.7 percent or flev rate is 69.2 percent and pcVer level is 75 percent.
3 Trip would occur when ene reactor coolant pump is cperating in each loop (total of two pumps operating) if the ;over is 52 9 percent and reactor flev rate is L9 2 percent or ficv rate is h5 4 percent and the power level is h9 percent.
The flux / flow ratios account for the maximum calibration and instrumentation errors and the maxinun variation frem the average value of the FC flev signal in such a =1nner that the reactor protective cystem receives a concervative indication of the RO flow.
No penalty in reactor ecclant flow through the core was taken for an cpen cere vent valve tecause of the core vent valve surveillance program during each refueling outage.
For safety analysis calculations the maximum calibration and in trumentation errors for the power level were used.
The pcVer-imbalance boundaries are established in order to prevent reactor thermal limits from being exceeded. These thermal licits are either power peaking kW/ft limita or DNER limits. The reactor power i= baler.ce (pcVer in the top half of the core minus pcVer in t'c.c bottom half of core) reduces the power level trip produced by the power-to-flow ratic :o that the boundariec of Figure 2.3-2 are produced.
The pover-to-flev ratic reduces the pcuer level trip and acucciated reacter pover/ reactor power-i= balance boundaries by 1.0S percent for a cne percent flev reducticn.
b.
Pump menitors The redundant pump conitors prevent the minimum core DNLR frem decreasing below 13 by tripping the reacter due to the loss of reactor coolant punp(s).
The pun; =cniterc also restrict the power level for the nu=ber of pu=ps in operation.
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c.
Reactor coolant uystem proccare During a ctartup accident frca low power or a clow red withdrawal frca high power, the cyctem hich preocure trip cet point is reached before the nuclear overpower trip set point. The trip setting limit chovn in Figure 2.3-1 for high reacter coolant cystem pressure has been catabliched to maintain the syatem precsure below the cafety limit (2750 poig) for any decign trancient. Due to calibration and inctrument errors, the safety a.talycis assumed a 30 psi pressure error in the high reactor coolant cyctem precsure trip setting.
The low preccure (1600 psig) and variable low prescure (1175 Tout - 5103) trip cetpcint shown in Figure 2 3-1 have been established to maintain the DNB ratio greater than or equal to 1.3 for those design accidents that recult in a precoure reduction (3, h).
Due to the calibraticn and inctrumentation errors, the safety analycia used a variable low reactor coolant system preccure trip value of (11.75 Tout -
51h3) and a low precsure trip value of 1770 psig.
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Coolant outlet temperature The high reactor coolant outlet temperature trip setting limit (619 F) shown in Figure 2.3-1 has been established to prevent exceccive core coolant temperatures in the cperating range.
The calibrated range of the temperature channels of the RFS is 520 to 620 F.
The trip cetpoint of the channel is 619 F.
Under the vorst come environment, pnver enpply perturbations, and drift, the accuracy of the trip string is tlF.
This accuracy vac arrived at by summing the verat case accuracies of each module. This is a concervative method of error analysis since the ncrual procedure is to use the root mean square method.
Therefore, it is accured that a trip util occur at a value no higher than 620 F even under verst cace conditions.
The safety analycis used a high temperature trip cet point of 620 F.
The calibrated rance of the channel is that portion of the span of indicatien which has been qualified with regard to drift, linearity, repeatability, etc.
Tnis does not imply that the equipment is restricted to cperation within the calibrated r1nge. Additional testing has demonctrated that in fact, the te=perature channel is fully operational approximately 10% above the calibrated range.
Since it has been establiched that the channel vill trip at a value of RC outlet temperature no higher than 620 F even in the worst cace, and since the channel is fully cperational approximately 105 acove the calibrated range and exhibito no hysteresis or foldover character-istica, it is ccncluded that the instrument design is acceptable.
e.
Reactor building prescure The high reactor building precoure trip Jetting limit (h pcig) providec pocitive accuracce that a reactor trip vill occur in the unlikely event of a cteam line failure in the reactor building cr a lecc-of-coolant accident, even in the absence of a low reacter ecolant syctem proccure trip.
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TABLE 2 3-1(6)
REACTOR PROTECTION GYSTDf TRIP SETTING LIMITS m
Four Reactor Coolant Three Reactor Coolant One Reactor Ccolant E.
Fumps Operating Pumpn Operating Pump Operating in S
(Nominal Operating (Nominal Operating Each Loop (Nominal Shutdown S
Power - 100%)
Power - 75%)
Operating Power - 49%)
Bypass z
1.
Nuclear power, Max.
105.5 105 5 105.5 5.0 (3)
% of rated power 2.
Nuclear Power based on 1.08 times flow minus 1.08 times flow minus 108 times flow minus Bypassed flow (2) and imbalance reduction due to reduction due to reduction due to h
max. of rated power imbalance (s) imbalance (s) imbalance (s) y co 3.
Uuq] ear power based NA NA 91%
Bypassed (51 on pump monitors,
% of rated power max.
)
k.
High reactor coolant 2405 (7) 1720 2405 (7)
= 2405 (7) y syatem pressure, psig, m
max.
5 Low reactor coolant 1800 1800 1800 Bypassed system pressure, psig min.
6.
Variable low reactor (11.75 Tout-5103) (1)
(11.75 Tout-5103) (1)
(11.75 Tout-5103) III Bypassed coolont system pressure puig, min.
7 Reactor coolant temp.
619 619 619 619 F., Max.
8.
Ilich Reactor Building h
h h
4
(.n pressure, psig, max.
CO CIs (1) Tout is in degrees Fahrenheit (E)
(2) Reactor coolant system flow, %
N (3) A:1r.inistratively controlled reduction set only during reactor shutdown N
( 1. ) Aut.or.intically set when o'her segments of the RPS (as specified) are bypassed (5) The 1 unip monitora nloo produce a trip on:
(a) loss of two reactor coolant pumps in one reactor coolant loop, and (1.) loss of one or two reactor coolant pumps during two--pump operation (6) Trip at tire l i mi t :, n re u. tt i to-limitr on the cet point cide of the protection system bistable comparators (7) These limits applicable for operation in Cycle 3 only.
6
h 2500 P = 2405 psig 2300 E
ACCE?TA5LE d
OPERAil0N T = 619 F 5
[
2100 N
5 E
s a
se
- c. !l s%
1900 j
g UNACCEPTABLE P = 1800 psig OPERAil0N 1700 1500 540 550
'580 600 620 040 Reactor Outlet Temperature, F TP.1-1, UNIT 1, CYCLE 3 PROTECTION SYSTE!J MAXIMU3 ALLOWABLE SET P0ll!TS Figure 2.3-1
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Amendment No.
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LIMITI:;G CGTTCITIC:!3 FCR OPERATIC:I 3.1 REACTOR COOLA::T SYSTEM
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3.1.1 OPEATIO:IAL CCMPO:IE:;TS Applicability Applies to the operating status of reactor coolant system components.
3 Objective To specify those limiting conditions for operation of reactor coolant system compcnents which must te met to ensure safe reactor operations.
Snecification 3.1.1.1 Reactor Coolant Pumps a.
Pump ecmbinations permissible for given power levels shall be as shown in Specification Table 2.3.1.
b.
Power operation with one idle reactor coolant pump in each loop shall be restricted to 2h hours.
If the reactor is not returned to an acceptable RC pump operating ccmbinaticn at the end of the 2h hour period, the reactor shall be in a hot shutdown condition within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, c.
The boron concentration in the reactor coolant system shall not be reduced unless at least one reactor coolant pump or one decay heat removal pump is circulating reactor coolant.
3.1.1.2 Steam Generator a.
One steam generatcr shall be operable whenever the reactor coolant avera6e temperature is above 2500F.
3.1.1.3 Pressurizer Safety Valves a.
The reactor shall not remain critical unless both pressurizer code safety valves are operable with a lift setting of 2500 psig
- 1%.
b.
When the reactor is suberitical, at least one precsurizer code safety valve shall be operable if all reactor coolant sycten openings are closed, except for hydrostatic tests in accordance with ASME Boiler and Pressure Vessel Code,Section III.
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Bases Tne limitation on power operatien with one idle RC pump in each loop has been imposed since the ECCS cco2.ing performance has not been calculated in accordance with the Final Acceptance Criteria requirements specifically for this mode of reactor operation. A time period of 2h hours is allowed for operation with one idle RC pump in each loop to effect repairs of the idle pu=p(s) and to return the reactor to an acceptable combination of operating RC pumps. The 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for this mode of operation is acceptable since this mode is expected to have considerable margin for the peak cladding temperature limit and since the likelihood of a LOCA within the 2h hour period is considered very remote.
A reactbr coolant pump or decay heat removal pump is required to be in operation before the boron concentraticn is reduced by dilution with makeup water.
Either pump will provide mixing which will prevent sudden positive reactivity changes caused by dilute coclant reaching the reactor. Cne decay heat rencve.1 pump will circulate the equivalent of the reactor coolant system volume in one half hour or less.
The decay heat removal system suction piping is designed for 300 F and 370 psig; thus, the system can remove decay heat when the reactor coolant system is below this temperature.
(2, 3)
One precourizer code safety valve is capable of preventing overpressurizaticn when the reactor is not critical since its relieving espacity is greater than that required by the sum of the available heat) sources which are pump energy, pressurizer heaters, and reactor decay heat.(k Both pressurizer code safety valves are required to be in service prior to criticality to conform to the system design relief capabilities. The code safety valves prevent overpressure for rod withdrawal or feedvater l'ine break accidents. (5)(6)~ The pressurizer code safety valve lift setpoint shall be set at 2500 psig i 1% allowance for error and each valve shall be capable of relieving 311,700 lb/h of saturated steam at a pressure not greater than three percent above the set pressure.
REFERENCES-(1) FSAR, Tables 9-10 and 4-3 through h-7 (2) FSAk, Sections 4.2.5.1and9.52.3 (3) FSAR, Section 4.2.5.h (h) FSAR, Sections k.3.10.4 and 4.2.4 (5) FSAR, Section L.3.7 (6) Met Ed letter GQL-1410 of October 8, 1976.
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