ML19210B286
| ML19210B286 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 05/18/1976 |
| From: | Goller K Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19210B285 | List: |
| References | |
| NUDOCS 7911060528 | |
| Download: ML19210B286 (37) | |
Text
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4 UNITED STATES d
NUCLEAR REGULATORY COMMISSION y-j j
WASHINGTON. D. c. 2o555 e
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METROPOLITAN EDISON COMPANY JERSEY CENTRAL Pot:ER AND LICHT COMPANY PENNSYLVANIA ELECTRIC COMPANY DOCKET NO.
50-299 THREE MILE ISLAND NI* CLEAR STATION, C;IT NO.1 AMENDMENT TO FACILITY OPE?ATI';C LICE';SE A=end=ent No. 17 License No. DPR-50 1.
The Nuclear Regulatory Cc==1ssion (the Co==ission) has found that:
A.
The applications for a=end=ent by Metropolitan Edison Co=pany, Jersey Central Power and Light Cc=pany, and Pennsylvania Electric Co=pany (the licensees) dated August 8, 1975, as supported by filings dated July 9 and 15, 1975, and October 23, 1975; and January 13, 1976, as amended February 11, 1976, and April 2, 1976, and supported by filings dated January 23, 1976, April 5 and'8, 1976, comply with the standards and require =ents of the Atecic Energy Act of 1954, as a= ended (the Act) and the Coc=ission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in confor=1ty with the applications, the provisions of the Act, and the rules and regulations of the Co==ission;
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C.
There is reasonable assurance.(1) that the activities authorized by this a=end=ent can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in co=pliance with the Coc=ission's regulations; D.
The issuance of this amend =ent will not be inimical to the co==on defense and security or to the health and safety of the public; and E.
After weighing the environ = ental aspects involved, the issuaace of this a=and=ent is in accordance with 10 CFR Part 51 of the Co==ission's regulations and all appidpttb requ.._ents have been satisfied.
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Accordingly, the license is a ended by a change to the Technical Specificatiens as indice.ted i= the attachment to this license a:end:ent.
3.
This license amend:ent is effective as of the date of its issuance.
FOR T*.II Ni.'.CLFAR REGL7 ATORY CC:CESSICli
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/)(w: ;\\n:, T M..'L r Karl R. Geller, Assistant Director for Operating Reactors Divisica of Operating Reactors
Attachment:
Changes to the Technical S,pecifications Date of Issuance: IMY l 8 ;g73 o
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ATTAC102NT TO LICENSE AME: !ENT NO. 17 FACILITY OPERATING LICENSE NO. DPR-50 DOCKET No. 50-289 Revise Appendix A as follows:
Remove Pages Insert Pages vi vi, vii 2-1 2-1 2-2 2-2 2-3 2-3
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2-5 2-5 2-Sa 2-6 2-6 2-7 2-7 2-8 2-8 2-9 2-9 3-1 3-1 3-2 3-2 3-15 3-15 3-16 3-16 3-33 3-33 3-34 3-34 3-35 3-35 3-35a 3-35a 3-36 3-36 Remove Figures Insert Figures 2.1-1 2.1-1 2.1-2 2.1-2 2.1-3 2.1-3 2.3-1 2.3-1 2.3-2 2.3-2 3.5-2A - 2F 3.5-2A - 2J Changes on the revised pages are indicated by marginal lines. Pages 2-8, 3-15, and 3-33 are unchanged and are included for convenience only.
1585 149
LIST OF FICt'RIS r
3 Figure Title 2.1-1 Core Protection Safety Limit 2.1-2 Core Protection Safety Litics 1
2.1-3 Core Protection Safety Basic 2.3-1 Protection System Maxi =u: Allowable Set Points 2.3-2 Protection System Maximum Allowable Set Points 3.1-1 Reactor Coolant System Heatup Limitations 3.1-2 Reactor Coolant System Cooldown Limitations 3.1-3 Limiting Pressure vs. Temperature Curve for 100 STD cc/ Liter h 0 2
3.5-1 Incore Instrunentation Specification Axial Imbalance Indication 3.5-2 Incore Instrumentation Specification Radial Flux Tilt Indication 3.5-2A Rod Position Limits for 4 Pump Operation Applicable During the Period from 0 to 152 i 10 EFPD; Cycle 2 3.5-2B Rod Position Limits for 4 Pump Operation Applicable During the Period frca 152 1 10 EFPD; to 265 1 10 EFPD; Cycle 2 3.5-2C Rod Position Limits for 4 Pump Operation Applicable During the Period after 265 i 10 EEPD; Cycle 2 3.5-2D Rod Position L1=its for 2 and 3 Pu=p Operation Applicable During the Period from 0 to 152 1 10 EEPD; Cycle 2 3.5-2E Rod Position Limits for 2 and 3 Pump Operaticn Applicable During the Period from 152 i 10 to 265 1 10 EFFD; Cycle 2 3.5-2F Rod Position Limits for 2 and 3 Pucp Operation Applicable During the Period After 265 1 10 EFPD; Cycle 2 3.5-2G Operational Power Imbalance Envelope Applicable to Operation from 0 to 152 1 10 EFPD; Cycle 2 Amendment No.
, 17 vi 1585 150
Vigure Title 3.5-2H Operaticnal Power I balance Envelope Applicable to Operation frem 152110, to 265 + 10 EFPD; Cycle 2 3.5-21 Operational Power Imbalance Envelope Applicable to Operation after 265 1 10 EFPD; Cycle 2 3.5-2J LOCALicited$aximumAllowableLinearHeatRate 3.5-3 Incore Instru=entation Specification 4.2-1 Equipment and Piping Requiring Inservice Inspection in Accordance with Section XI of the ASMI Code 4.4-1 Ring Girder Surveillance 4.4-2 Ring Girder Surveillance Crack Pattern Chart 4.4-3 Ring Girder Surveillance Crack Pattern Chart 4.4-4 Ring Girder Surveillance Crack Pattern Chart 4.4-5 Ring Girder Surveillance Crack Pattern Chart 6-1 Organization Chart e
vii 4
AmendmentNo.j}[,17 3
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2.
SAFETY LIMITS A'D LIMITING SAFETY SYSTEM SETTINGS
'2.1 SAFE *Y LIMITS, REACTOR CCRE Applicability Applies to reactor themal pcver, reactor power imbalance, reactor coolant system pressure, coolant te=perature, and coolant flow during power operation of the plant.
Objective To maintain the integrity of the fuel cladding.
Specification 2.1.1 The cc=bination of the reacter system pressure and coolant te=perature shall not exceed the safety limit as defined by the locus of points established in Figure 2.1-1.
If the actual pressure /te.perature point is below and to the right of the line, the safety limit is exceeded.
2.1.2 The ec=bination of reacter therm 1 pcver and reactor power imbalance (power in the top half of core minus the power in the bottom half of the core expressed 'as a percentage of the rated pover) shall not exceed the safety limit as defined by the locus of points (solid line) for the specified ficv set forth in Figure 2.1-2.
If the actual-reactor-ther=al-power reactor-power-imbalance point is above the line for the specified flow, the safety limit is exceeded.
Bases To maintain the integrity of the fuel cladding and to prevent fissica product release, it is necessary to prevent overheating of the cladding under normal operating conditions. This is accerplished by operating within the nucleate boiling regi=e of heat transfer, wherein the heat transfer coefficient is large enough so that the clad surface te=perature is only slightly greater than the coolant te=perature. The upper boundar/ of the cucleate boiling regime is ter=ed, departure fres nucleate boiling (D:G). At this point there is a sharp reduction of the heat transfer coefficient, which vonld result in high cladding te=peratures and the possibility of cladding failure.
Although D:iB is not an cbservable para =eter during reactor operation, the observable parameters of neutren power, reacter coolant flev, temperature, and pressure can be related to DNB through the use of the BW-2 correlatien.
(1) the BW-2 correlation has been developed to predict D:iB and the locatica cf DUB for axially uniform and non-uniform heat flux distributions. The local DNB ratio (DNBR), defined as the ratio of the heat flux that vould cause DNB at a particular core location to the actual heat flux, is indicative of the cargin to DNB. The mini =u= value of the D!I.BR, during steady-state operation, nomal cperational transients, and anticipated transients is limited to 1.3 A D2R of 1.3 corresponds to a 95 percent probability at a 95 percent confidence level that D:i3 vill not occur; this is censidered Amendnent No. 17 2-1
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i The difference a e,onservative mar 6 n to L. 3 for all operating conditient.
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between the actual core outlet pressure and the indicated reactor coolant systes pressure has been considered in deterM"ing the core protectica safety li=its.
The difference in these two pressures is nocinally h5 Tsi; however, only a 30 psi drop vas assu=ed in reducing the pressure trip set points to correspcnd to the elevated locatica where the pressure is actually measured.
The curve presented in F16ure 2.1-1 represents the conditiens at which a minimum DIGR of 1.3 is predicted for the mhum possible t,hernal power (112 percent) when the react'or ecolant flow is 139 8 x 10+0 lbs/h, which is less than the actual flow rate for four operating reactor ecolant pumps. This curve is based on the following nuclear power peaking factors (2) with poten-tial fuel densification and fuel rod bovin6 effects; 5
5 5
F
= 2.67; F
= 178; F
= 1 50 q
M z
The 15 axial peaking factor associated with the cosine flux shape provides a lesser margin to a DIER of 13 than the 17 axial peMeg tactor asscefated with a lover core flux distributica. Fcr this reason the cosine flux shape artd the. associated.F,=-1,10 is more -limiting, and thus the.mcre conservative
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assu=ption.
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The 150 cosine axiar flux shape in conjunction vith FAH = 176 define the reference design peMrs condition in the core for operatien at the maximus overpower. Once the reference peaking condition and the associated therdal-hydraulic situatien has been established for the hot channel, then all other combinations of axial flux shapes and their accccpanying radials must result in a conditica which vill not violate the previcusly established design criteria on D:3R. The flux shapes exe=ined include a vide range of positive and negative offset for steady state and transient conditions.
These design li=it pcVer peakin5 factors are the most restrictive calculated at full power for the range fres all centrol reds fully vithdrawn to =aximu=
allovable control red insertics, and for= the core D:aR design basis.
The curves of Figure 2.1-2 are based en the more restrictive of two ther=al and fuel red li=its and include the effects of potential fuel densification boving; The 13 DNER li=it prcduced by a nuclear pcver pea'dng facter of a.
FN = 2.67 of the ec=bination of the radial peak, axial peak, and p8sitien of the axial peak that yields no less than a 13 D;GR.
b.
The co=bination of radial and axial peak that prevents central fuel celting at the hot spot. The limit is 19.6 kW/ft.
Pcver peaking is not a directly observable quantity and therefore limits have been established en the basis of the reactor power i= balance produced by the pcuer peaking.
The specified flow rates for curves 1, 2, and 3 of Figure 2.1-2 ccrrespond to the expected =ini=us flow rates with four pu=ps, three punps, and one pu=p in each icop, respectively.
Acendnent No. 17 22 s
8'he curve of Figure 2.1-1 is the =ost restrictive of all possible reacter I
The coole.a.: pu=p-caxi=u= ther=al power cc=binations shown in Figure 2.1-3 curves of Figure 2.1-3 represent the conditions at which a mini =u= DIGR of 13 is predicted at the =axi== possible ther=t1 pcVer fer the nu=ber of reacter coolant pu=ps in operatica or the local quality at the point of mini =us D:GR is equal to 22 percent, (3) whichever conditica is more restrictive.
l Using a local quality li=it of 22 percent at the point of =ini=u= D:GR as a basis for curve 3 of Figure 2.1-3 is a censervative criterien even though the quality at the exit is higher than the quality at the point of mini =u= DIGR.
Tce DIGR as calculated by the BW-2 correlatica continually increases fres *.he point of =ini=us DIGR, so that the exit DIGR is always higher and is a function of the pressure.
The =aximum ther=al power for three pu=p operatien is 86.7 percent due to a power level trip produced by the flux-flev ratio (74.7 percent flev x 1.C8 =
80 7 percent pcver) plus the =axi=u= calibration and instru=entation errer.
The =aximum t* u=al power for other reactor coolant pu=p conditions is pro-duced in a similar =anner.
For each curve of Figure 2.1-3, a pressure-te=perature point above and to the left of the curve would result in a DIGR greater than 1.3 cr a local quality at the point of minimu= DIGR less than 22 percent for that particular reacter coolant pu=p situation. The 1.3 DIGR curve for four pu_p operation is core restrictive than any other reacter ecolant pu=p situatien because any pressure /te=perature point above and to the left of the four pu=p curve vill be above and to the left of the other curves.
REFERE' ICES (1)
FSAR, Section 3.2 3 1.1 (2)
FSAR, Section 3.2.3 1.1.c (3)
FSAR, Sectica 3.2 3 1.1.k 1585 154 Amendment No. 17 2-3
- ^.
,23 LIMITING SAFETY SYSTEM SETTINGS, PROTECTION INSTELHiliTATION Applicability Applies to instru=ents monitoring reacter power, reactor pcVer imbalance, reactor coolant syste= pressure, reacter coolant outlet te=perature, flov, nu=ber of pu=ps in operation, and high reactor building pressure.
Objective To provide autc=atic protection action to prevent any cc=bination of process variables frc= exceeding a safety li=it.
Specification 231 The reactor protection syste= trip setting li=its and the permissible bypasses for the instru=ent channels shall be as stated in Table 2 3-1 and Figure 2 3-2.
Bases The reacter protection syste= consists of four instru=ent channels to =cnitor each of several selected plant conditions which vill cause a reactor trip if -
any one of these conditiens deviates frc= a pre-selected operating range to the degree that a safety li=it =ay be reached.
The trip setting li=its for protection syste= instrumentatica are listed in Table 2.3-1.
The safety analysis has been based upon these protection syste:
instru=entatica trip set points plus calibratica and instru=entatica errors.
Nuclear Overeever A reactor trip at high power level (neutron flux) is provided to prevent da= age to the fuel cladding frc= reactivity excursicas too rapid to be detected by pressure and te=perature =easure=ents.
During nor=21 plant operatica vith all reactor coolant pu=ps operating, reacter trip is initiated when the reacter power level reaches 105 5% of rated power.
Adding to this the possible variation in trip set points due to calibratic and instru=ent errers, the =axi=u= actual power at which a trip would be actuated could be 1125, which is the value used in the safety analysis (1).
Ov'rpower trip based en flow and imbalance a.
e The power lesei trip set point produced by the reacter coolant syste=
flow is based en a pcver-to-flow ratio which has been established to accon=odate the most severe ther=al transient considered in the design, the loss-of-coolant flev accident frc= high power. Analysis has de=enstrated that the specified power to flow ratio is adequate to prevent a DNER of less than 13 should a low flev condition exist due to any =alfunctica.
1585 155 A=end=ent No. y[, 17 s
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The pcver level trip set point produced by the power-to-flev ratio provides both high power level and icv flow.protecticn in the event the reactor power level increases or the reactor coolant flow rate decreases. The power level trip set point produced by the power to flow ratio provides overpower DNB protection for all =cdes of pu=p operatien. For every flow rate there is a maxi =u= per=issible power level, and for every power level there is a =ini=u= permissible lov flow rate. Typical power level and lov flow rate ec=binations for the pu=p situations of Table 2 3-1 are as follevs:
1.
Trip would occur when four reacter coolant pumps are operating if power is 108 percent and reactor flow rate: is 100 percent, or flow rate is 92.6 percent and power level is 100 percent.
2.
Trip would occur when three reactor coolant pu=ps are opersting if power is 80 7 percent and reacter flev rate is 74 7 percent or flow rate is 69 2 percent and pcver level is 75 percent.
3 Trip would occur when one reactor coolant pu=p is operating in each loop (total of tvo pumps operating) if the power is 52 9 percent and reacter flev rate is h9 2 percent er flev rate is h5.h percent and the pcver level is k9 percent.
For safety analysis calculations the -av4 calibration and instru=entation errcrs for the pcVer level were used.
The power-i= balance boundgries are established in order to prevent reacter ther=al limits frc= being exceeded. These thermal limits are either power peaking kW/ft li=its er D:GR li=its. The reacter power i= balance (pcver in the top half of core minus pcVer in the bottc= half of core) reduces the power level trip produced by the power-to-flev ratio so that the boundaries of Figure 2 3-2 are produced. The power-to-flow ratio reduces the pcver level trip and associated reactor pcver/ reactor pcver-i= balance boundaries by 1.08 l
percent for a one percent flev reduction.
b.
Pump =onitors The redundant pu=p =ccitcrs prevent the =ini=u= core D:GR frc=
decreasing belev 13 by tripping the reacter due to the loss of reacter coolant pu=p(s). The pump =eniters also restrict the power level for the nu=ber of pu=ps in operation.
c.
Reactor coolant syste= pressure During a startup accident frc= lov pcVer or a slev rod withdraval from high power, the systen high pressure trip set point is reached before the nuclear overpever trip set point. The trip setting limit shown in Figure 2 3-1 for high reacter ecolant syste= pressure (2355 psig) has been established to =aintain the syste= pressurt 585 15 6 below the safety li=it (2750 psig) for any design transient.
AmendmentNo.[,17 2-6
e The low pressure (1600 psig) and variable icv pressure (11 75 Tout - 5103) trip setpoint shown in Figure 2 3-1 have been established to maintain the DtTB ratio greater than or equal to 13 for those design accidents that result in a pressure reduction (3,4).
Due to the calibration and instru=entation errcrs, the safety analysis used a variable low reactor coolant system pressure trip value of (11 75 Tout - Slh3).
d.
Coolant outlet temperature The high reactor coolant outlet te=perature trip setting li=it (619 F) shown in Figure 2.3-1 has been established to prevent excessive core coolant te=peratures in the operating range.
The calibrated range of the te=perature channels of the RPS is 520 to 620 F.
The trip setpoint of the channel is 619 F.
Under the verst case enviren=ent, power supply perturbations, and drift, the accuracy of the trip string is+lF.
This accuracy was arrived at by su==ing the vorst case accuracies of each =cdule. This is a ecuservative method of errer analysis since the nor=al procedure is to use the root =ean square method.
Therefore, it is assured that a trip vill occur at a value no higher than 620F even under vorst case conditions. The safety analysis used a high temperature trip set point of 620F.
The calibrated range of the channel is that pcrtion of the span of indication which has been qualified with regard to drift, linearity, repeatability, etc.
This does not i= ply that the equip =ent is restricted to operation within the calibrated range. Additional testics has de= castrated that in fact, the te=perature channel is fully operational approxi=ately 10% above the calibrated range.
Since it has been established that the channel vill trip at a value of RC outlet te=perature no higher than 620F even in the verst case, and since the channel is fully operational approxi=ately 10% above the calibrated range and exhibits no hysteresis er foldever characteristics, it is concluded that the instru=ent design is acceptable, e.
Reacter building pressure The high reactor building pressure trip setting li=it (h psig) provides positive assurance that a reactor trip vill occur in the unlikely event of a steam line failure in the reactor building or a loss-of-coolant accident, even in the absence of a low reacter coolant system pressure trip.
1585 157 Amendnent No. 17 2-I ga,
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Shutdovn bypass In order to provide for control red drive tests, :ero power physics testing, and startup procedures, there is provision for bypassing certain segnents of the resetor protection system. The reactor protection systes segnents which can be bypassed are shown in Table 2 3-1.
?io conditions are imposed when the bypass is used:
1.
By ad=inistrative control the nuclear overpower trip set point s:ust be reduced to a value 15 0 percent of rated power during reactor shutdown.
2.
A high reactor coolant systes pressure trip set point of 1720 psig is auto =stically i= posed.
The purpose of the 1720 psig high pressure trip set point is to prevent normal operation with part of the reactor protection syste=
bypassed. This high pressure trip set point is lower than the nor=al low pressure trip set point so that the reactor =ust be tripped before the bypass is initiated. The overpower trip set point of 15 0 percent prevents any significant reactor power from being produced when perfor=ing the physics tests. Sufficient natural circulation (5) would be available to re=cve 5 0 percent of rated power if none of the reactor coolant pu=ps were operating.
REFERENCES (1)
FSAR, Section ik.1.2.3 (2)
M AR, Section 1k.1.2.2 (3)
FSAR, Section ik.1.2 7 (k)
FSAR, Section 1h.l.2.8 (5)
E AR, Section 14.1.2.6 1585 158 2-8 3
TABLE 2.3-1 i
REACTOR PROTECTION SYSTEM TRIP SETTING LIMITS
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One Reactor Coolant Pump "n
0 Four Reactor Coolant Pumps Three Reactor Coolant Pumps Operating in Each Ioop Shutdown operating (Nominal Operating (Nominal (Nominal Operating Bypass E.
Operating Power - 100%)
Operating Power - 75%)
Power - 49%)
F I I.
Nuclear power, Max.
5 0(3)
% of rated power 105.5 105.5 105.5 2.
Nuclear Power based 1.08 times flow minus 1.08 times flow minus 1.08 times flow minus Bypassed on flow (2) and imbal-reduction due to reduction due to reduction due to ance, max. of rated imbalance (s) imbalance (s) imbalance (s) i power Bypassed NA NA 91%
3.
Nuclear power based (5) i on pump monitors, max.
% of rated power h.
High reactor coolant 2355 2355 2355 1720(h)
,m system pressure, psig, max.
5 Low reactor coolant 1800 180G 1800 Bypassed systen pressure, psig, min.
6.
Variable low reactor (11.75 Tout - 5103)(1)
(11.75 Tout - 5103)(1)
(11.75 Tout - 5103)(1) Bypasses coolant system pressure, peig, min.
7 Reactor coolant temp.
619 619 619 619 r., Max.
8.
High Reactor Building h
h h
h w
y
- J1 pressure, psig, max.
(1) Tout is in degrees Fahrenheit (F)
W (2) Henctor coolant system flow, %
(3) Administratively controlled reduction set only during reactor shutdown (4) Automatically set when other segments of the RPS (as specified) are bypassed (5) The pump monitors also produce a trip on:
(a) loss of two reactor coolant pumps in one renet.or coolant loop, and (b) loss of one or two reactor coolant pumps during two-pump operation.
, 3.
LIMITING C05sfTIONS FOR OPERATION 3.1 REACTOR COOLANT SYSTEM 3.1.1 OPERATIONAL COMPONENTS Applicability Applies to the operating status of reacter coolant system cocponents.
Objective To specify those limiting conditicas for operation of reactor coolant system components which cust be met to ensure safe reactor operations.
Specification 3.1.1.1 Reactor Coolant Pu=ps a.
Pump combinations per=issible for given power levels shall be as shown in Specification Table 2.3.1.
b.
Power operation with one idle reactor coolant pu=p in each loop shall be restricted to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
If the reactor is not returned to an acceptable RC punp operating conbination at the end of the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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 unitss at least one reactor coolant pu=p or one decay heat removal pu=p is circulating reactor coolant.
3.1.1.2 Steam Generator a.
One steam generator shall be operable whenever the reactor coolant average temperature is above 250*F.
3.1.1.3 Pressurizer Safety Valves a.
The reactor shall not recain critical unless both pressurizer code safety valves are operable with a lift setting of 2435 PSIG : 1%.
b.
When the reactor i's suberitical, at least one pressurizer code safety valve shall be operable if all reactor coolant system openings are closed, except for hydrostatic tests in accordance with ASME Boiler and Pressure Vessel Code,Section III.
,1585 16 0 Amendment No. }/[, 17 3-1
Bases The limitation on power operation with one idle RC pu=p in each loop has been imposed since the ECCS cooling performance has not been calculated in accordance with the Final Acceptance Criteria requirements specifically for this mode of reactor operation. A time period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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 a'ceptable 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 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period is considered very remote.
A reactor coolant pu=p or decay heat removal pu=p is required to be in operation before the boron concentration is reduced by dilution with makeup water.
Either pump will provide mixing which will prevent sudden positive reactivity changes caused by dilute coolant reaching the reactor. One decay heat re= oval 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 pressurizer code safety valve is capable of preventing overpressurization when the reactor is not critical since its relieving capacity is greater than that required by the sum of the available heat sources which are pump energy, pressurizer heaters, and reactor decay heat.(4) 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 overpres-sure for a rod withdrawal accident.(5)
The pressurizer code safety valve lift set point shall be set at 2435 psig 1 percent 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 4-7 (2) FSAR, Sections 4.2.5.1 and 9.5.2.3 (3) FSAR, Section 4.2.5.4 (4) FSAR, Sections 4.3.10.4 and 4.2.4 (5) FSAR, Section 4.3.7 1585 16i Amendment No. 17 3-2 m.-
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f)
If reactor coolant leakage is to the contain=ent, it =ay be identified by one t
or more of the folleving =ethods:
o a.
The contain=ent air particulate =onitor is sensitive to lov leak rates.
The rate of leakage to which the instru=ent is sensitive is 0.05h gp=
vithin sixty =inutes, assu=ing the presence of corresion product activity.
b.
The contain=ent radicactive gas =cnitor is less sensitive but can be used as a backup to the air particulate =onitor. The sensitivity ran6e of the instru=ent is approxi=ately 2 gp= to greater than 10 gp=.
c.
A leakage detection syste= vhich deter =ines leakage losses frc= vater and stea= syste=s within the contain=ent. This syste= collects and measures =oisture ecndensed from the contain=ent at=cephere by c Oling coils of the =ain recirculation units. This system provi:les a depend-able and accurate =eans of =easuring total leakage, including leaks frc= the cooling coils the=selves which are part of the contain=ent boundary.
d.
Indication of leakage frc= the above scurces shall be cause to recuire a contain=ent entr/ and li=ited inspection at power of the reactor coolant syste=. Visual inspection =eans, i.e., looking for ste2=, ficer vetness, or boric acid crjstalline for=ations, vill be used.
Pericdie inspections for indications of leaka6e within the contain=ent vill be conducted to enhance early detection of proble=s and to assure bast on-line reliability.
If reactor coolant leakage is to the auxiliary building, it may be identified by one or = ore of the following methods:
a.
The auxiliary and fuel handling building vent radicactive gas =enitor is sensitive to very lov activity levels and would shev an increase in activity level shortly after a reactor coolant leak developed within the auxiliar/ building.
b.
Water invent ries around the auxiliar/ building su=p.
c.
Periodic equip =ent inspections, d.
In the event of gross leakage, in excess of 13 + 2 gp=, the individual cubicle leak detectors in the =akeup and decay heat pu=p cubicles, vill al.ar= in the control rec = to backup "a",
"b", and "c" above.
When the source and location of leakage has been identified, the situation can be evaluated to determine if operation can safely continue. *his evaluation vill be perfor=ed by the Three Mile Island Operations Group according to routine established in Section 12.1.1 of the FSAR. Unde. these conditions, an allevable leakage rate of 30 gpn has been established.
1585 162 3-15
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8 3.1 7 MODERATOR TE4PERATURE COEFFICIENT OF REA m y m Applicability Applies to =axim:= positive =oderator tenperature coefficient of reactivity at full power conditions.
Objective To assure that the moderator te=perature coefficient stays within the limits calculated for safe operation of the reactor.
Specificatien 3.1.7.1 The moderator te=perature coefficient shall not be positive at pcver levels above 95% of rated power.
Bases A non-positive moderator coefficient at power levels above 95% of rated power is specified such that the - M m clad te=peratures vill not exceed the Final Acceptance Criteria based en LOCA analyses. Belov 95% of rated pcVer the Final Acceptance Criteria vill got be exceeded with a positive moderator te=perature coefficient of +0 5 x 10- 4K/K/F. All other accident analyses as reported in the FSAR have been performed for a range of moderater te=perature coefficients including +0.5 x 10 k 4K/K/F.'
The experi= ental value of the moderator coefficient vill be corrected to e
obtain the hot full power moderator coefficient. The correction factor vill be verified during startup testing on earlier B&W reactors.
The Final Acceptance Criteria states that post-LOCA clad te=perature vill not exceed 2200 F.
REFERENCES (1)
FSAR, Section 1h (2)
FSAR, Secticn 3 1585 163 Amendment No. 17 3-16 w,
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3 5.2 corrInoL acD cacvP AnD Po'nza DISTRIS'J: Ion LI c S dbu}.k CD cn en en Applicability as es This specification applies to power distribution and cperation of control rods during pcVer operation.
objective To assure an acceptable core power distribution during pcVer Operation, to set a li=1t en potential reactivity insertion frc= a hypothetical centrol rod ejection, and to assure core suberiticality after a reactor trip.
Speci fic ation 3.5.2.1 The available shutdcun nargin shall not be less than one per-cent aK/k with the highest verth control rod fully ' ithdrawn.
3 5 2.2 operation with inoperable rods:
a.
Operation with =cre than ene inoperable rod as defined in Specification h.7.1 and h.7.2.3 in the safety cr regulating rod banks shall not be permitted.
b.
If a centrol rod in the regulating and/or safety red banks is declared inoperable in the withdrawn position as defined in Specification paragraph h.7.1.1 and k.7.1.3, an evalustlun shall be initiated i=nediately to verify the existence cf ene percen Ak/k hot shutdevn nargin. 3cration =ay be initiated te in rease the available red verth either to ec=penscte for the vorth of the inoperable rod er until the regulating banks are full / vith-dravn, whichever occurs first.
Si=ulta.eously a progra= cf exercising the rensining regulating and safety rods shall be initiated to verify operability.
c.
If vithin one hour of deter =ination of an inoperable red as defined in Specification h.7.1, it is not dete--i- * ' a a
one percent ik/h hot shutdevn =argin exists cenbining the worth of the inoperable rod with each of the other reds, the reattor shall be brought to the het standby cor.dition until this =argin is established.
d.
Folleving the deter =dnation of an inoperable red as defined in Specification h.7.1, all reds shall be exercised within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and exercised veekly until the red proble= is solved.
If a control red in the regulating or safety red groups is e.
declared inoperable per L.7.1.2, pcuer shall be reducei to 60% of the ther-a' pcver allevable for the reactor cecis-t pump ec=hination.
2-33 1585 164
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f.
If a control rod in the regulating or axial power shaping groups is declared inoperable per Specification 4.7.1.2.,
operation =sy continue provided the rods in the group are positioned such that the red that was declared inoperable is maintained within allovable gro g average position 1 hits of Specification k.71.2.
g.
If the inoperable rod in Paragraph "E" above is in groups 5, 6, 7, or 8, the other rods in the group shall be tri=ed to the sa=e position. Ucr=al operation of 100 percent of the ther=al power allowable for the reactor coolant pu=p co=,
bination nay then continue provided that the red that was declared inoperable is =sintained vithin allevable group average position li=its in 3 5 2.5 3 5 2.3 The verth of single inserted control rods during criticality are li=1ted by the restricticas of Specification 3.1.3.5 and the control Rod Positien Lir.its defined in Specification 3 5 2 5 3 5 2.h Quadrant tilt:
Except for physics tests 'if ouadrant tilt exceeds k percent, a.
power shall be reduced i==ediately to below the power level cutoff {see Figures 3 5-2A, 3 5-2B and 3 5-2c). Moreover, the power level cutoff value shall be reduced 2 percent for each 1 percent tilt in excess of 4 percent tilt. For less than four pu=p operation, ther al power shall be reduced 2 percent of the thez..a1 pover allevable for the reactor coolant pt. p co=bination for each 1 percent tilt in excess of 4 percent.
b.
Within a period of k hours, the quadrant power tilt shall be reduced to less than 4 percent except for physics tests, or the following adjust =ents in setpoints and limits shall be. =ade:
.l.
The protection syste= reactor pover/i= balance envelope trip setpoints shall be reduced 2 percent in power for each 1 percent tilt.
2.
The control red group vithdrawal li=its (Figures 3 5-2A 3 5-2B, 3 5-2C, 3 5-2D, 3 5-2E, and 3 5-2F) shall be reduced 2 percent in power for each 1 percent tilt in excess of k percent.
3 The operational i= balance li=its (Figure 3 5-20, 3 5-2H and 3 5-2I) shall be reduced 2 percent in pover for each 1 percent tilt in excess of 4 percent.
3-3' 1585 165 Amendment No. 17
3 5'.2 5 control ror esttions:
N s.
Operating rod group overlap shall not exceed 25 percent
- 5 percent, between two sequential groups except for physics tests.
b.
Except for physics tests or exercising control rods, the control rod insertion /vithdraval limits are specified on Figures 3 5-2A,3 5-23, and 3 5-2c for four pu=p operation and Figures 3 5-2D, 3 5-2E, and 3 5-2F, for three or tvo pt'. p operation.
If the control rod position li=its are exceeded, corrective =easures shall be taken ie=ediately to achieve an acceptable control rod position. Acceptable control rod positions shall be attained within four hours.
c.
Except for physics tests, power shall not be increased above the pover level cutoff (See Figures 3.5-2A, 3 5-23 and 3 5-2c) Mess the xenon reactivity is within 10 percent of the equilibrius value for peration at rated power and asymptotically approaching stability.
d.
Core i= balance shall be =enitored on a ef 4-frequency of once every two hours during power operation above 40 percent of rated power.
Except for physics tests, corrective =easures (reduction l
of i= balance by APSR r.ove=ents and/or reduction in reactor pover) shall be taken to =aintain operation vithin the envelope defined by Figures 3 5-20, 3 5-2H and 3 5-2I.
If the i=talance is not within the envelope defined by Figures 3 5.-20, 3 5-2H and 3 5-2I corrective measures shall be taken to achieve an acceptable i= balance. If an acceptable i= balance is not achieved within four hours, reactor power shall be reduced until imbalance limits are =et.
e.
Safety rod limits are given in 3.1.3 5 3 5 2.6 The centrol rod drive patch panels shall te locked at ell'ti=es vith limited access to be authorized by the superintendent.
3527 A power =ap shall be taken to verify the expected power distribution at periodic intervals of approximately 10 full power days using the incere instrumentation detectio,n system.
Esses The pover-i= balance envelope der.ned in Figures 3 5-2G, 3 5-2H, and 3 5-2I is based on LOCA analyses which have defined the maximu= linear heat rate (see Figure 3.5-2J) such that the =axi=us clad temperature vill not exceed the Final icceptance Criteria (2200F). Operation outside of the power imbalance envelope alone does not constitute a situation that vould cause'the Final Acceptance Criteria to be exceeded should a LOCA occur, The power imbalance envelope represents the boundary of operation 3-35@$ }hb Amend =ent No. 17 hg6em M q pe W W g*' q
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o limited by the Final Acceptance Criteria enly if the centrol rods are at the withdrawal / insertion limits as defined by Figures 3 5-2A, 3.5-23, 3 5-2C, 3 5-2D, 3.5-2E and 3.5-2F and if a h percent quadrant pcVer tilt exists. Additional conservatis: is introducted by application of: a*. Nuclear uncertainty factors. b. Ther=al calibratica uncertainty. c. Fuel densification effects. d. Hot rod manufacturing tolerance facters. The Rod index versus A11cvable Power curves of Figures 3.5-2A, 3 5-23, 3 5-2C, 3 5-2D, 3 5-2I and 3 5-2F, describe three regicas. These three regions are: 1. Pezuissible operating Regica 2. Restricted Regions 3 Prohibited Regica (Operatica in this regica is not allowed) Note: Inadvertent operaticn within the Restricted Region for a period of k hours is not considersd a violation of a li=iting condition for operatica. The li=iting criteria within the Restricted Regicn are potential ejected rod verth and ECCS power peaking and since the i probability of these accidents is very lov especially in a k hour time fra=e, inadvertant operation within the Restricted Region for a period of k hours is alleved. Amendment No. 17 1585 167 3-35a
The 25t5 percent overla: tveen successive control rod ' ups is allowed since the vorth of a rod is 1swer at the upper and lover part or! the stroke. Control rods are arranged in groups or banks defined as follows: Grouc Function 1 Safety 2 Safety 3 Safety 4 Safety 5 Regulating 6 Regulating 7 Xenon transient override 8 APSR (axial power shaping bank) Control rod groups are withdrawn in sequence beginning with group 1. Groups 5, 6 and 7 are overlapped 25 percent. The nornal cosition at power is for groups 6 and T to be partially inserted. The red position limits are based on the most limiting of the following three ~ criteria: ECCS power peaking, shutdown =argin, and potential ejected rod vorth. As discussed above, co=pliance with the ECCS power peaking criterion is ensured by the rod position li=its. The =ini=us available red verth, consistent with the rod position limits, provides for achieving het shutdown by reactor trip at any time, assu=ing the highest vorth control rod that,is withdrawn re=ains in the full out position (1). The rod position li=its also ensure that inserted rod groups vill not contain single rod vorths greater than: 0.65% Ak/k at rated power. Trese values have been shown to be safe by the safety analysis (2) of the hypothetical rod ejection accident. A m4mm single inserted control rod vorth of 1.0% Ak/k is allowed by the rod position li=its at hot zero power. A single inserted control red vorth 1.0% Ak/k at beginning of life, hot, zero power vould result in a lover transient peak themal power and, therefore, less severe enviren= ental consequences than O.65% Ak/k ejected rod vorth at rated power. The plant co=puter vill scan for tilt and i= balance and vill satisfy the technical specification require =ents. If the computer is out of service, than panual calculatica for tilt above 15 percent power and imbalance above 40 percent power must be perfor=ed at least every two hours until the cocputer is returned to service. The quadrant power tilt limits set forth in Specification 3 5 2.- have been established within the ther=si analysis design base using the definition of quadrant power tilt given in Technical Specifications, Section 1.6. During the physics testing progra=, the high flux trip setpoints are ad=inistratively set as follows to assure an additional safety =argin is provided: Test Pover Trin Setreint 0 <5% 15 50% ho-50% 50 60% 75 85% >T5 105.5% REFEPZiCES . (1) FSAR, Sectio. 3.2.2.1.2 1585 168 (2) FSAR, Section 14.2.2.2 Anendnene No. 17 3-36
2600 2400 ACCEPTABi.E { OPERAT10N 2200 l 0
- E -
5 2000 E o UNACCEPTABLE a OPERATION , 1800 p 1600 560 580 600 620 640 660 Raactor Outlet Temperature, F CORE PROTECTION SAFETY LIMIT Figure 2.1 1 1585 169 Amendment No. 17
t Thermal Power Le< DNACCEPTABLE OPERATION --120 (112) (+35.112) (-33.182) / - i0 ACCEPTABLE 4 PUNP -100 OPERAT10M g,fgg gg,;g Rw/ft Limit 90 ( 58.90) (-56.90) (.33,86.7) (86.7) (+35.86.7) ACCEPTitLE 80 3 & 4 PUHP OPERATION _ 70 (.33,59.1) __ 60 (59.I) (+35.59.I) ~ 4_CCEPTA8tE.. 50 2,3 & 4 PUHP -.- OPERATICM ' +47,48) (-46.47) 40 30 20' ~ 10 l I I I I 1 -60 -40 -20 0 20 to 60 teactor Power imbalance, 5 i. CURVE REACTOR COOLANT FLOW (Ib/hr) 0 I 139.8 x 10 6 2 104.5 x 10 6 3 68.8 x 10 UNIT I. CYCLE 2 CORE PROTECTION SAFETY LlHIT. Figiere 2.1-585 170 Amendment No. 17 g= guey = g - = = = - e===p ep e umummmer
l 1 2000 i . 2400 2e "f / e. E. 2200 I f R; t 2000 { -y o E l \\ u i 1800 /// i 1600 560 580 600 620 640 660 Reactor Outlet Temperature, F REACTOR COOLANT FLOW CURVE (LBS/HR) POWER PUMPS 0PERATING (TYPE OF LIMIT) 1 139.8 x 106 (1005)* 1125 Four Pumps (DNBR Limit) 2 104.5 x 106 (74.75) 86.75 Three Pumps (DNBR Limit) 3 68.8 x 106 (49.25) 59.15 One Pump in Eacn Loop (Quality Limit)
- 106.5% of Cycle 1 Design Flev CORE PROTECTION SAFETY BASES F i gu r e 2.1 -3 Amendment No. 17
.1585 171
2500 't = 2300 P = 2355 PSIG T = 619 F ACCEPTABLE w OPERATION l 2100 = '$s ac N4/ ug 1900 ~ E + UNACCEPTABLE ~ 0 P = 1800 PSIG 0PERATION ~g g 1700 1500 540 560 580 600 620 640 Reactor Outlet Temperature, F PROTECTION SYSTEM MAXIMUM ALLOWABLE SET POINTS ~ Figure 2.3-1 ' Amendment No. , 17 1585 172 ..__v.
Power Level '$ UN ACCEPT ABLE -.120 OPERATION (-20,108) 110 (108) (+21,108) 100 + f I l
- +
~~ ACCEPTA8tE + 4 PUHP 90 l 6 +N OPERATION g ) (80.7) (+48,80) (-46,80) [ 80 I l 70 l ' ACCEPTABLE t.A 4 P.u}lP 2 O P E R'AT I O,N.. 6,0 r ~ -(53.L) (+48. 52' 7 ) (-46,52.7) -- 50 ( .- 40 l 1 ACCEPTABLE 2,3 & 4 PUHP l (+ 8',2 5.1 ) (-46,25.1) OPERATION 20 E N o = T to + Y ul u u =l 1 I I i 1" -50 -40 -30 -20 -10 0 10 20 30 40 50 Power labalance, % UNIT I, CYCLE 2 PROTECTION SYSTEM MAXIHUM ALLOWABLE SET POINTS Amendment No. 17 ~ Figure 2.3-2 .mv 1585 173
las.9,lo2 200.5,802 100 Iss.s,92 200.5,92 POWER LEVEL 90 RESTRICTED CUTOFF REGION 20s.s,a5 80 RESTRICTED 10 s2s.s,ss REGION j 60 PERMISSIBLE b 50 OPERATINO REGION 3005 5 , 40 222.3,45.5 g 2 = E 30 20 10 p '0, 0 20 40 60 80 100 120 140 160 120 200 220 240 250 280 300 Rod index. 3 Vithdrawn 0 25 50 75 100 0 25 50 75 100 I I I I 3 I l t Group 5 Group 7 0 25 50 75 100 e t t t I Group 6 ROD POSITION LIMITS FCR 4 PLFP OPERATION APPLICABLE CURING ThE PERICO FRCM 0 TO 152 1 10 EFPO; CYCLE 2 Figure 3.5-2A 1585 174 Amendment No. 17 en- .g .,e,e. -g m
100 - OPERATION IN THIS REGION IS NOT ALLOWED iss.s,s2 2o=.7.92 POWER LEVEL 90 CUTOFF 80 - SHUTDOWN MARGIN 126.6,75 70 - LIMIT RESTRICTED REGION RESTRICTED a REGION b 60 a b I ~ PERMitSIBLE OPERATING 222.3 47 3 5 ~ REGION E 30 20 o,ss 10 0 0 20 40 60 80 100 120 140 160 180 200 220 240 260 230 300 Rod index. 5 Withdrawn 0 25 50 75 100 0 25 50 75 100 i I I I f f f f I j Group 5 Group 7 0 25 50 75 100 t t t I I Group 6 RCD POSITION LIMITS FCR 4 PLNP OPERATICN APPLICAELE CURING THE PERIOD FRCM 152 1 10 EFFO: TO 265 1 10 EFPD; CYCLE 2 Figure 3.5-28 1585 175 Amendment No. , 17
is2 :02 29s.2 02 3 3 100 -OPERATION IN THIS REGION is NOT ALL0 SED POTER LEVEL 235.s,s2 90 CUTOFF 80 SHUT 001N 70 MARGIN Llui-RESTRICTED REGION l 60 h 50 222.3,9s = so,ss A 40 5 ? 30 PERillSSIBLE 20 OPERATING 05 REGION 3 10 0 0 20 40 60 80 100 120 140 160 180 200 220 240 200 280 300 Rod index, 5 tithdrawn 0 25 50 75 100 0 25 50 75 100 t t t t i a t t Group 5 Group 7 0 25 50 75 100 r e t t t Group 6 ~ R00 POSITICN LIMITS FOR 4 PUMP CPERATICN APPLICABLE CURING THE PERICO AFTER 265 1 10 EFPO; CYCLE 2 Figure 3.5-2C neem no. p 1585 176 ~=,-- 6 v M O* $4#-W9% N=
RESTRICTED REGION 117,102 148,102 210,102 219,102 .g100 - FOR 2 AND 3 PUMP RESTRICTED OPERATION REGION FOR 2 & 3 5 90 80' 80 PUMP OPERAT10N E S 0 80 s [ 222,84, am,84 '8 g. p RESTRICTED E 70 REGION FOR 3 o PUMP OPERATION oc g 60 222,58 300,58 g 50 PER!alSSIBl.E a" 3 40 OPERATING REGION 30
== 20 u Eg 10 _ O' o 0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 Rod Index, 5 Withdrawn 0 . 25 50 75 100 0 25 50 75 100 e i i i 1 i Group 5 Group 7 i 0 25 50 75 100 - 9 f f f Group 6 ROD POSITION LIMITS FOR 2 AND 3 PUMP OPERATION APPLICABLE DURING THE PERICD FROM 0 TO 152 1 10 EFPD; CYCLE 2 Figure 3.5-2D 1585 177 Amendment No. 17
OPERATION IN THIS 117,102 I40.102 210,102 219.102 100 REGION IS NOT 1 RESTRICTED REGION ALLOWE0 g 127,96 FOR 2 & 3' PUMP g 90 OPERATION 3 74,80 g 222,87 300.87 80 g - SilVT00WN RESTRICTED REGION o MARGIN FOR 3 PUMP & 70 - LIMIT OPERATION E. o 60 g 222,60 300,60 t '7.50 ~ PERMISSIBLE ' OPERATING 40 REGION 3 30 [ w 0,15 d 10 E o.o a-0 i i e e i i 0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 Rod index, ", Withdrawn 0-25 50 75 100 0 25 50 75 100 Group 5 Group 7 0 25 50 75. 100 a t t t Group 6 ROD POSITION LIMITS FCR 2 AND 3 PUMP CPERATION APPLICABLE CURING THE PERICD FRCM 152 ! 10 TO 265 i 10 EFPD; CYCLE 2 Figure 3.5-2E 1585 178 Amendment No. 17 ~- -~w- = ,,,w.w, ,,m-
143,802 229..s2 232,102 100 OPERATION IN THIS REGION I f g IS NOT ALL0ffE0 RESTRICTED FOR 2 & 3 0 90 PUMP OPERATION 3 i17,89 222,89 E 80 0 RESTRICTED REGION FOR 70 3 PUMP OPERATION 1 E 60 62 si SHUTOOWN 222.si E 50 MARGIN LIMIT 90.50 = j 40 I E OPERATING E 30 REGION 'o w 20 .is ' RESTRICTED FOR 3 10 - 2 & 3 PUMP 0 i e f i 0 20 40 60 80 100 120 140 160 180 200 220 240 280 280 300 Rod index, ", Withdrawn 'O 25 50 75 100 0 25 50 7,5 100 Group 5 Group 7 0 25 50 75 100 .Ercup 6 RCD POSITION LIMITS FOR 2 AND 3 PUMP OPERATION APPLICABLE DURING THE PERICD AFTER 265 t 10 EFPD; CYCLE 2 Figure 3.5-2F Amendment No. 17'
Power, 5 of 2535 MWt 1ESTRICTED REGION h "ll.--e.v2 14.28.102 -11.04,92 12.88.92 90 4 -1s.sa,as 5, 73, "a 5 -19.25,77 19.25,77 s " 70 PERMISSIBl.E 60 OPERATING REGION 50 -20.48,45.s 40 30 20 4 10 I .50 -40 20 -10 0 10 20 30 40 50 Axial Power imoalance, 5 OPERATIONAL POWER IMBALANCE ENVELOPE APPLICABLE TO OPERATION FROM O TO 152 i 10 EFFD; CYCLE 2 . Figure 3.5-2G i 1585 180 Amendment No. 17
Power, 5 of 2535 p t RESTRICTED REGION -16.32,102 , g gg 14.79.102 -45.64,92 90 13.34.s2 -15.75,87 5 . 80 18.25,73 70 60 PERMISSIBLE OPERATING -11.15,47 50 REGION 40 30 ,20 10 -50 -40 20 -10 0 10 20 30 40 50 Axial Power imoalance, 5 3 OPERATIONAL PO?iER IMBALANCE ENVELCPE APPLICABLE TO CPERATICN FROM 152 t 10 265 1 10 EFPD; CYCLE 2 - s Figure 3.5-2H 1585 181 w dment No. 17
I Power, 5 of 2535 MWt RESTRICTED REGION t -18.87,102 1 13.26.102 100 -21.62,92 90 14.26,92 - 80 70 PERHISSIBLE 16.25,65 l OPERATING - 60 REGION 50 -23.04,48 40 . 30 20 10 e i ! 40 20 -10 0 10 20 30 40 50 t Axial Power Imaalance, 5 j 1 i l OPERATICNAL POWER IMBALANCE ENVELOPE APPLICABLE TO CPERATICN AFTEF. 265 t 10 EFPD; CYCLE 2 Figure 3.5-2I
- r i
I 4 Amendment No. 17 1585 182
21 20 19 a_ _. _ 18 ~ 5 [N E 17 / \\ i / \\ 16 ~ E 15 I .2 E 14 g j 13 It 0 2 4 6 8 10 12 Axial Location of Peak Power From Bottom of Core. ft LOCA LIMITED MAXIMLN ALLOWABLE LINEAR HEAT RATE Figure 3.5-2 J 1585 183 Amendment No. 17 ~~ ~ 89 ** -.'9_ m* g
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