ML20205C866
| ML20205C866 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 03/20/1987 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Metropolitan Edison Co, Jersey Central Power & Light Co, Pennsylvania Electric Co, GPU Nuclear Corp |
| Shared Package | |
| ML20205C872 | List: |
| References | |
| DPR-50-A-126 NUDOCS 8703300282 | |
| Download: ML20205C866 (35) | |
Text
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- ga ae go, UNITED STATES 4
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NUCLEAR REGULATORY COMMISSION i
WASHINGTON, D C. 20555 l'ETPOP01! TAN E0! SON COMPANY JERSEY CENTRtf POWEP Af!D LIGHT COMPAf:Y l
I PENNSYLVANI A ELECTF IC COPF Af'Y GPU f.lfCLEAP CORP 0FATION DOCV.ET NO. 50-289 THREE MILE IS!AND NUCLEAP STATIOF. UNIT NO. 1 AMENDFEP:T TO FACILITY OPEPATING LICENSE Amendment f;o.126 License No. DPP-Fr 1.
The Nuclear Reculatory Commission (the Commission) has found that:
A.
The applications for amendment by GPU Nuclear Corporation, et el.
(the licensees) dated July 16, 1986, and November 3, 1986, comply with the standards and requirements of the Atomic Energy Act of 1954, as amerded (the Act), and the Commission's rules and regulations set forth in 10 CFR Charter I; y
B.
The facility will operate in conformity with the applicatiers, the provisions of the Act, and the rules and regulations of tha Commission; C.
There is reasonable assurance (1) that the activities authori7ed 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 reculations; D.
The issuance of this amendment will not be inimical to the commor defense and security or to the health and safety of the public:
and E.
Tha issuance of this amendrent is in accordance with 10 CFP Part 51 of the Commission's regulations and all applicable recuirerents have been satisfied.
8703300282 070320 PDR ADOCK 05000289 PDR p
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2.
Accordirely, the licerse is amended bv changes to the Techrical Specificstions as irdicated in the attachmert to this license arendrert, arid paragraph ?.c.(?) of Facility Operrtina 1icense No. OPP-50 is hereby amer.deo to read as follows:
I Technical Specifications 4
i The Technical Specifications contained in Apperdix A, as revised through Arendment ho.126, are j
herehy incorporated in the license.
GPU Nuclear i
Corporation shall operate thF facility in aCCordance j
with the Technicel Specifications.
3.
This license amendment is effective as of its date of issuance.
FOR THE NUCI EAP PEGllLITGRY. C0FF'T.? ION
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/oIN. Stolz,'DirectqrT I-lPWR Project Directorate #6
\\ Division of PWP Licensing-P
Attachment:
Changes to the Technical u
i Specifications Date of Issuance:
March 20,1987 t
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6 MTAC4t4Et i TO licit'SE AFENn'E' f.0.l?6.
FACILITY bptrATING LICEl'SF NO.' DPFy'g.
00CVET l'O. EC-280
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Pepl a ct. the followirn paces of the Appendi>
"A" Techricci pecifications with the att acher' races, lhe_ revised parcs are idertified by Arerdrent rurber and certain vertictl i'ines ir:dicatino tFr. area cf charice.
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b Pemove insort v 'i i vii viii viii s
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?-3
?-3
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?-5 2-6 7-6 2-7
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?-9 249 3-34 3-34 3-34a 3-34a 3-35 3 35 3-35a 3-3Ea f
3-36 3-36 o
3-36a 3-3fe 5-3 5-3*
5f 5-4 Ficures:
7.1-?
?.1-?
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?.1-3
?.1-3 7.3-1
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?.3-?
?.3-?
3.5-PA 3.5-FA 3.5-?B 3.5-?P
'J 3.5-?C 3.5-?C
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- 3. 5-F r, 3.5-?D 3.5-?E 3.5-?E 3.5-?F 3.5-?F 3.5-?G
- 3. 5-Fr, 3.5-?H 3.5 2H 3.5-?!
3.5-21
- 3. 5-N <
3.5-?V 3.5-?L
- 0verleaf pace provided for docwent completeness 4
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-s LIST OF FIGURES s
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Figure Title
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- 2. P.1 TMI-l Core Protection Safety Limit 2.1-2 TMI-l Core Protection Safety Limits
!A 2.1-3 TMI-1 Core ^ Protection Safety Bases s
i 2.3-1; TMI-l Protection System Maximum Allowable Set Points 3
'c 2.3-2 Protection System Maximum Allowable Set Points for Reactor Power Imbalance, TMI-l
, ; 3.1 -1 Reactor Coolant System Heatup/Cooldown Limitations ( Applicable to 5 EFPY) l' 3.1-2 Reactor dooiant System, Inservice Leak and Hydrostatic Test 3
Limitat' ions -( Applicable to 5 EFPY)
~
3.1-3 Limiting Pressure vs. Temperature Curve for 100 STD cc/ Liter HO j
2 0.5-2A Rod Position Limits for 4 Pump Operation from 0 to 30+10/-0
/
EFPD, TMI-l
'3.5-2B
' Rod Position f.imits for 4 Pump Operation from 30+10/-0 to 250:10 t
EFPD, THI-l 3.5-2C,
Rod Position Limits for 4 Pump Operation af ter 250:10 EFPD, TMI-l
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3.5-2D Rod Position Limits for 3 Pump Operation from 0 to 30+10/-0 EFPD, TMI-1 3.5-2E Rod Position Limits for 3 Pump Operation from 30+10/-0 to 250:10 EFPD, TMI-l
,k5-2F Rhd P:sition. Limits for 3 Pump Operation af ter 250:10 EFPD, TMI-l C, F.od, Posi tion' Limits for 2 Pump Operation f rom 0 to 30+10/-0
'3.5-?G i
EFPD, TMI-I 3.5-2H
-Rod Position Limits for 2 Pump Operation from 30+10/-0 to 250:10 EFPD, THI-l 3.5-?!
Rod Position Limits for 2 Pump Operation af ter 250:10 EFPD, TMI-l 3.5-?J Power Imbalance Envelope for Operation from 0 to 30+10/-0 EFPD, 1NI-l o
vii Amendment Nos. JJ, A/, 29, 39, AS, 50, 59, 72, JOE, 309, N,126
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LIST OF FIGURES s.
Figure Title t
3.5-2K Power Imbalance Envelope for Operation af ter 30-10/-0 EFPD. 'M!-l 1'
3.5-2L LOCA Limited Maximun Allowable Linear Heat Rate
,r-i;
- 3. 5-1 Incore Instrumentation Specification Axial Imbalance Indication, TMI-1 3.5-2 Incore Instrumentation Specification Radial Flux Tilt j
Indication, TMI-l 3.5-3 Incore-Instrumentation Specification 3.11-1 Transfer Path to and from Cask Loading Pit 4.17-1 Snubber Functional Test - Sample Plan 2 5-1 Extended Plot Plan TMI 5-2 Site Topography 5 Mile Radius 5-3 Site Boundary for Gaseous Effluents 5-4 Site Boun'dary for Liquid Effluents 6-1 GPU Nuclear Corporation Organization Chart 6-2
.TMI-l Onsite Organization q
viii 3
Amendment' Nos. 77.,7(126
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a conservative margin to DNB for all operating conditions.
The difference between the actual core outlet pressure and the indicated reactor coolant system pressure has been considered in determining the core protection safety limits. The dif ference in these two pressures is nominally 45 psi; however, only a 30 psi drop was assumed in reducing the pressure trip set points to correspond to the elevated location where the pressure is actually measured.
The curve presented in Figure 2.1-1 represents the conditions at which a DNBR of 1.3 or greater is predicted for the maximum possible thermal power (112 l
percent) when the reactor coolant flow is 139.8 x 10+6 lbs/h, which is less than the actual flow rate for four operating reactor coolant pumps. This curve is based on the following nuclear power peaking f actors (2) with potential fuel densification and fuel rod bowing effects; N
N N
F
= 2.82, F
= 1. 71 ; F
= 1.65 q
AH z
The 1.65 axial f eaking f actor associated with the cosine flux shape provides a l
lesser margin to a DNBR of 1.3 than the 1.7 axial peaking f actor associated with a lower core flux distribution. For this reason the cosine flux shape and the associated Fy = 1.65 is more limiting and thus the more l
conservative assumption.
The 1.65 cosine axial flux shape in conjunction with F AH = 1.71 define the l
reference design peaking condition in the core for operation at the maximum overpower.
Once the reference peaking condition and the associated thermal-hydraulic situation has been established for the hot channel, then all other combinations of axial flux shapes and their accompanying radials must result in a condition which will not violate the previously established design criteria on DNBR. The flux shapes examined include a wide range of positive and negative off set for steady state and transient conditions.
These design limit power peaking f actors are the most restrictive calculated at full power for the range from all control rods fully withdrawn to maximum allowable control rod insertion, and form the core DHBR design basis.
The curves of Figure 2.1-2 are based on the more restrictive of two thermal limits and include the effects of potential fuel densification and fuel rod bowi ng:
Thg 1.3 DNBR limit produced by a nuclear power peaking f actor of a.
F q= 2.82 of the combination of the radial peak, axial peak, and l
position of the axial peak that yields no less than 1.3 DNBR.
b.
The combination of radial and axial peak that prevents central fuel melting at the hot spot. The limit is 20.50 kW/f t.
l Power peaking is not a directly observable quantity and therefore limits have been established on the basis,of the reactor power imbalance produced by the power peaki ng.
the expected minimum flow rates with fou,r pumps,of Figure 2.1-2 correspond tothree pu The specified flow rates for curves 1 2 and 3 each loop, respectively.
Amendment No. 77, M, H,126 2-2
-The curve of Figure ?.1-1 is the most restrictive of all possible reactor coolant pump-maximum thermal power combinations shown in Figure 2.1-2.
The curves of Figure 2.1-3 represent the conditions at which a DNBR of 1.3 or greater is predicted at the maximum possible thermal power for the number of reactor coolant pumps in operation or the local quality at the point of minimum DNBR is equal to 22 percent, (3) whichever condition is more restrictive.
The maximum thermal power for three pump operation is 89.3 percent due to a l
power level trip produced by the flux-flow ratio (74.7 percent flow x 1.03 =
80.6 percent power) plus: the maximum calibration and instrumentation error.
l The maximum thermal power for other reactor coolant pump conditions is produced in a similar manner.
Using a local quality limit of 22 percent at the point of minimum DNBR as a basis for curve 3 of Figure 2.1-3 is a conservative criterion even though the quality at the exit is higher than the quality at the point of minimun DNBR.
The DNBR as calculated by the B&W-2 correlation continually increases from the point of minimum DNBR, so that the exit DNBR is always higher and is a function of the pressure.
For each curve of Figure 2.1-3, a pressure-temperature point above and to the lef t of the curve would result in a DNBR greater than 1.3 or a local quality at the point of minimum DNBR less than 22 percent for the particular reactor coolant pump situation. Curve 1 is more restrictive than any other reactor coolant pump situation because any pressure / temperature point above and to the lef t of this curve will be above and to the lef t of the other curves.
REFERENCES (1) FSAR, Section 3.2.3.1.1 (2) FSAR, Section 3.2.3.1.1.3 l
(3) FSAR, Section 3.2.3.1.1.11 Amendment No. 77, 29, 39, $$, 720,126 2-3
2.3 LIMITING SAFETY SYSTEM SETTINGS, PROTECTION INSTRUMENTATION Applicability Applies to instruments monitoring reactor power, reactor power imbalance, reactor coolant system pressure, reactor coolant outlet temperature, flow, number of pumps in operation, and high reactor building pressure.
Objective To provide automatic protection action to prevent any combination of process variables f rom exceeding a saf ety limit.
Specification 2.3.1 The reactor protection system trip setting limits and the permissible bypasses for the instrument channels shall be as stated in Table 2.3-1 and Figure 2.3-2.
Bases The reactor protection system consists of four instrument channels to monitor each of several selected plant conditions which will cause a reactor trip if any one of these conditions deviates from a pre-selected operating range to the degree that a safety limit may be reached.
The trip setting limits for protection system instrumentation are listed in Table 2.3-1.
These trip setpoints are setting limits on the setpoint side of the protection system bistable comparators. The safety analysis has been based upon these protection system instrumentation trip set points plus i
calibration and instrumentation errors.
Nuclear Overpower A reactor trip at high power level (neutron flux) is provided to. prevent damage to the fuel cladding f rom reactivity excursions too rapid to be detected by pressure and temperature measurements.
During normal plant operations with all reactor coolant pumps operating, reactor trip is initiated when the reactor power level reaches 105.1% of rated l
Adding to this the possible variation in trip set points due to power.
calibration and instrument errors, the maximum actual power at which a trip would be actuated could be 112%, which is the value used in the safety analysis (1).
a.
Overpower trip based on flow and imbalance The power level trip set point produced by the reactor coolant
)
system flow is based on a power-to-flow ratio which has been established to accommodate the most severe thermal transient
]
considered in the-design, the loss-of-coolant flow accident from high power. Analysis has demonstrated that the specified power to
. flow ratio is adequate to prevent a DNBR of less than 1.3 should a low flow condition exist due to any malfunction.
Amendment No. D, U, 4,126 2-5
The power level trip set point produced by the power-to-flow ratio provides both high power level and low flow protection 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 modes of pump operation.
For every flow rate there is a maximum permissible power level, and for every power level there is a mininu, permissible low flow rate.
Typical power level and low flow rate combinatices for the pump situations of Table 2.3-1 are as follows:
1.
Trip would occur when four reactor coolant pumps are operating if power is 108 percent and reactor flow rate is 100 percent, cr. fl ow rate is 92.5 percent and power level is 100 percent.
2.
Trip would occur when three reactor coolant pumps are operating if power is 80.6 percent and reactor flow rate is 74.7 percent or flow l
rate is 69.4 percent and power level is 75 percent.
3.
Trip would occur when one reactor coolant pump is operating in each loop (total of two pumps operating) if the power is 53.1 percent and reactor flow rate is 49.2 percent or flow rate is 45.3 percent and the power level is 49 percent.
The flux / flow ratios account for the maximum calibration and instrumentation errors and the maximum variation from the average value of the RC flow signal in such a manner that the reactor protective system receives a conservative indication of the RC flow.
No penalty in reactor coolant flow through the core was taken for an open core vent valve because of the core vent valve surveillance program during each refueling outage.
For safety analysis calculations the maximum calibration and instrumentation errors for the power level were used.
The power-imbalance boundaries are established in order to prevent reactor thermal limits from being exceeded. These thermal limits are either power peaking Kw/f t limits or DNBR limits. The reactor power imbalance (power in the top half of the core minus power in the bottom half of core) reduces the power level trip produced by the power-to-flow ratio so that the boundaries of Figure 2.3-2 are produced. The power-to-flow ratio reduces the power level trip and associated reactor power / reactor power-imbalance boundaries by 1.08 percent for a one percent flow reduction.
b.
Pump Monitors The redundant pump monitors prevent the minimum core DNBR f rom decreasing below 1.3 by tripping the reactor due to the loss of reactor coolant pump (s).
The pump monitors also restrict the power level for the number of pu'mps in operation.
i l
Amendment No. 72, 77, RE, RS, JP, ER.126 2-6 i
- _. _ _. _ _ _ _ _ _. _ _. - _ _.- - - - _ - _ ~
e c.
Reactor coolant system pressure During a startup accident from low power or a slow rod withdrawal from high power, the system high pressure trip set point is reached before the nuclear overpower trip setpoint.
The trip setting limit shown in Figure 2.3-1 for high reactor coolant systen pressure his been established to maintain the system pressure below the safety limit (2750 psig) for any design transient (6).
Due to calibration and instrument errors, the safety an. ysis assumed a 45 psi pressure error in the high reactor coolant system pressure trip setting.
The high pressure trip setpoint was subsequently lowered from 2390 psig to 2300 psig. The lowering of the high pressure trip setpoint and raising of the setpoint for the Power Operated Relief Valve (PORV), from 2255 psig to 2450 psig, has the effect of reducing the challenge rate to the PORV while maintaining ASME Code Safety Valve capabil ity, i
The low pressure (1800 psig) and variable low pressure (11.75 TngB 5103) trip setpoint were initially established to maintain the D9 ratio greater than or equal to 1.3 for those design accidents that result in a pressure reduction (3,4).
The B&W generic ECCS analysis, however, assumed a low pressure trip of 1900 psig and, to establish conformity with this analysis, the low pressure trip setpoint has been raised to the more conservative 1900 psig.
Figure 2.3-1 shows the high pressure, low pressure, and variable low pressure trips.
d.
Coolant outlet temperature The high reactor coolant outlet temperature trip setting limit (618.8F) shown in Figure 2.3-1 has been established to prevent l
excessive core coolant temperature in the operating range.
The calibrated range of the temperature channels of the RPS is 520*
to 620*F. The trip setpoint of the channel is 618.8F. Under the l
worst case environment, power supply perturbations, and drif t, the accuracy of the trip string is 1.?*F.
This accuracy was arrived at l
by summing the worst case accuracies of each module.
This is a conservative method of error analysis since the normal procedure is to use the root mean square method.
Therefore, it is assured that a trip will occur at a value no higher than 620*F even under worst case conditions.
The safety analysis used a high temperature trip set point of 620*F.
The calibrated range of the channel is that portion of the span of indication which has been qualified with regard to drif t, linearity, repeatability, etc. This does not imply that the equipment is restricted to operation within the calibrated range.
Additional testing has demonstrated that in f act, the temperature channel is fully operational approximately 10% above the calibrated ra nge.
Amendment No. 77, 4, D,
y, /g,126 2-7
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Tab 1'J 2.3-1 p
REACIOR twiu;nON SYS'IDi 'IRIP SETTDK; LIMITS (6)
Ibur Reactor Coolant
'Ihree Reactor Coolant One Reactor (bolant m
5 Punps Operating Pmps Operating Pmp OperatinJ in (Naninal Operating (Naninal Operating Each Icop (Hanninal Shutdown 2
?
Power - 1001)
Ibwer - 751)
Operating Power - 49%)
Bypnse C0
- 1. Nuclear power, Max.
105.1 105.1 105.1 5.0(3) 1 m -
g
% of rated power
- 2. Nuclear power based cn 1.08 times flow 1.08 times flow 1.08 times flow minus Bypassed flow (2) and inbalance minus reduction due minus reductica due reduction due to
~
y max. of rated power to inbalance to imbalance inhalance D
3.
Nuclear power based NA NA 55%
BymW
~
(5) on punp ronitors, y
Max. % of rated PUW8f ww
~
4.
High reactor coolant 2300 2300 2300 1720(4)
)
system pressure, poig w4 box.
1 5.
Iow reactor coolant 1900 1900 1900 Bypassed systern pressum, poig min.
6.
Variable low reactor (11.75 Tout-5103)(1)
(11.75 Tout-5103)(1)
(11.75 Tout-5103)(1)
Byp.5 coolant system pressure psig, min.
7.
Reactor coolant tenp.
618.8 618.8 618.8 618.8 F., Max.
8.
High Reactor Dailding 4
4 4
4 pressure, poig, max.
(1) 'Ibut is in degrees Fahrenheit (F)
(2) Reactor coolant system flow, %
(3) Ahninistratively controlled reduction set only during reactor shutdown.
(4) Autcmatically set when other segments of the RTS (as specified) are bypassal.
(5) '1he punp monitors also produce a trip on: (a) loss of two reactor coolant punps in one reactor coolant loop, and (b) loss of one or two reactor coolant numps during two-purp operation.
(6) Trip setting limits are cetting limits on the cetpoint side of the protection systan histable connectors.
f.
If a control rod in the regulating or axial power shaping groups is declared inoperable per Specification 4.7.1.2., operation nay continue provided the rods in the group are positioned such that the rod that was declared inoperable is maintained within allowable group average position limits of Speci#ication 4.7.1.?.
g.
If the inoperable rod in Paragraph "e" above is in groups 5, 6, 7, or 8, the other rods in the group may be trimmed to the same position.
Normal operation of 100 percent of the thermal power allowable for the reactor coolant pump combination may then continue provided that the rod that was declared inoperable is maintained within allowable group average position limits in 3.5.2.5.
3.5.2.3 The worth of single inserted control rods during criticality is limited by the restriction of Specification 3.1.3.5 and the Control Rod Position Limits defined in Specification 3.5.2.5.
3.5.2.4 Ouadrant Tilt:
a.
Except for physics tests the quadrant tilt shall not exceed
+4.12% as determined using the full incore detector system.
l b.
When the full incore detector system is not available and except for physics tests quadrant tilt shall not exceed +1.96%
as determined using the power range channels displayed on the console for each quadrant (out of core detection system),
c.
When neither detector system above is available and, except for physics tests, quadrant tilt shall not exceed +1.901 as determined using the minimum incore detector system.
d.
Except for physics tests if quadrant tilt exceeds the tilt limit power shall be reduced 2 percent for each 1 percent tilt in excess of the tilt limit. For less than four pump operation, thermal power shall be reduced 2 percent of the thermal power allowable for the reactor coolant pump combination for each 1 percent tilt in excess of the tilt li mi t.
e.
Within a period of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, the quadrant power tilt shall be reduced to less than the tilt limit except for physics tests,-
or the following adjustments in setpoints and limits shall be made:
1.
The protection system reactor power / imbalance envelope trip setpoints shall be reduced 2 percent in power for each 1 percent tilt, in excess of the tilt limit.
l Amendment No. 77, 79, 79, $$, 50, PE, 126 3-34
2.
The control rod gioup withdrawal limits (Figures 3.5-2A to 3.5-21) shall be reduced 2 percent in power for each I percent tilt in excess of the tilt limit.
3.
The operational imbalance limits (Figures 3.5-2J and 3.5-2K) 4 shall be reduced 2 percent in power for each 1 percent tilt in excess of the tilt limit.
f.
Except for physics or diagnostic testing, if quadrant tilt is in excess of +16.80* determined using the full incore detector system (FIT), or +14.2% determined using the out of core detector system (OCT) if the FIT is not available, or +9.5% using the minimum incore detector system (MIT) when neither the FIT nor DCT are available, the reactor will be placed in the hot shutdown condition.
Diagnostic testing during power operation with a quadrant tilt is permitted provided that the thermal power allowable is restricted as stated in 3.5.2.4.d above.
g.
Quadrant tilt shall be monitored on a minimum frequency of once every two hours during power operat. ion above 15 percent of rated
- power, i
i 1
3-34a i
Amendment No. 29. 28, 39,/0,AE, E0,lyf126
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3.5.2.5 Control Rod Positions a.
Operating rod group overlap shall not exceed 25 percent +5 percent, between two sequential groups except for physics tests. _
b.
Position limits are specified for regulating control rods.
Except l
for physics tests or exercising control rods, the regulating control rod insertion / withdrawal limits are specified on rigures 3.5-2A, 3.5-2B, and 3.5-2C for four pump operation and Figures 3.5-2D, 3.5-2E, and 3.5-2F for three pump operation. Two pump operation is specified on Figures 3.5-2G, 3.5-2H, and 3.5-21.
If any of these control rod position limits are exceeded, corrective measures shall be taken immediately to achieve an acceptable control rod position.
Acceptable control rod positions shall be attained within four hours, c.
Deleted d.
Core imbalance shall be monitored on a minimum frequency of once every two hours during power operation above 40 percent of rated power.
Except for physics tests, corrective measures (reduction of imbalance by APSR movements and/or reduction in reactor power) shall be taken to maintain operation within the envelope defined by Figures 3.5-2J and 3.5-2K.
If the imbalance is not within the envelopes defined by Figures 3.5-2J or 3.5-2K at the appropriate time in cycle, corrective measures shall be taken to achieve an acceptable imbalance.
If an acceptable imbalance is not achieved within four hours, reactor power shall be reduced until imbalance limits are met, e.
Safety rod limits are given in 3.1.3.5.
3.5.2.6 The control rod drive patch panels shall be locked at all tims with limited access to be authorized by the superintendent.
3.5.2.7 A power map shall be taken at intervals not to exceed 30 effective full power days using the incore instrumentation detection system to verify the power distribution is within the limits shown in Figure 3.5-2L.
l Arendment No. 70, 77,77, yg, 33, gg, 3-35 17@.126 i
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ls 1
l Bases f
The power-imbalance envelope defined in Figures 3.5-2J and 3.5-2K is based on l
LOCA analyses which have defined the maximum linear heat rate (see Figure 3.5-2L) such that the maximum clad temperature will not exceed the Final l
Acceptance Criteria (2200*F).
Operation outside of the power imbalan:e envelope alone does not constitute a situation that would cause the Final Acceptance Criteria to be exceeded should a LOCA occur.
The power inbalance envelope represents the boundary of operation limited by the Final Acceptance Criteria only if the control rods are at the withdrawal /# nsertion limits as defined by Figures 3.5-2A, 3.5-2B, 3.5-2C, 3.5-20, 3.5-2E, 3.5-2F, 3.5-2G, 3.5-2H, 3.5-21, and if quadrant tilt is at the limit.
The effects of the gray APSRs are also included.
Additional conservatism is introduced by application of:
4 a.
Nuclear uncertainty factors b.
Thermal calibration uncertainty c.
Fuel densification effects A
d.
Hot rod manuf acturing tolerance f actors l
e.
Postulated fuel rod bow effects f.
Peaking limits based on initial condition for loss of Coolant Flow tra nsients.
The Rod index versus Allowable Power curves of Figures 3.5-2A, 3.5-2B, 3.5-2C, 3.5-2D, 3.5-2E, 3.5-2F, 3.5-2G, 3.5-24, and 3.5-21 describe three regions.
These three regions are:
1.
Permissible operating Region 2.
Restricted Regions 3.
Prohibited Region (Operation in this region is not allowed)
NOTE:
Inadvertent operation within the Restricted Region for a period of four hours is not considered a violation of a l
I J
i, Amendment No. 17, 79, 3$. 39, 50, 179 3-35a 126
The 25+5 percent overlap between successive control rod groups is allowed since the worth of a rod is lower at the upper and lower part of the stroke.
Control rods are arranged in groups or banks defined as follows:
Group Function 1
Safetf 2
Safety 3
Safety 4
Safety 5
Regula ti ng 6
Regulating 7
Regulating (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 normal position at power is for group 7 to be partially inserted.
The rod position limits are based on the most limiting of the following three criteria:
ECCS power peaking, shutdown margin, and potential ejected rod worth. As discussed above, compliance with the ECCS power peaking criterion is ensured by the rod position limits.
The minimum available rod worth,
- onsistent with the rod position limits, provides for achieving hot shutdown by reactor trip at any time, assuming the highest worth control rod that is withdrawn remains in the full out position (1).
The rod position limits also ensure that inserted rod groups will not contain single rod worths greater than: 0.65% ak/k at rated power.
These values have been shown to be safe by the safety analysis (2) of the hypothetical rod ejection accident.
A maximum single inserted control rod worth of 1.0% Lk/k is allowed by the rod position limits at hot zero power.
A single inserted control rod worth 1.0% Ak/k at beginning of life, hot, zero power would result in a lower transient peak thermal power and, therefore, less severe environmental consequences than 0.65% Ak/k ejected rod worth at rated power.
The plant computer will scan for tilt and imbalance and will satisfy the tecnnical specification requirements.
If the computer is out of service, then manual calculation for tilt above 15 percent power and imbalance above 40 percent power must be performed at least every two hours until the computer is returned to service.
The quadrant power tilt limits set forth in Specification 3.5.2.4 have been established within the thermal analysis design base using an actual core tilt of +4.92% which is equivalent to a +4.12% tilt measured with the full incore instrumentation with statistically combined measurement uncertainties included. The maximum allowable quadrant power tilt setpoint of +16.8% tilt measured with the full incore detector system represents a +20% actual core tilt and includes bounding measurement uncertainty allowances.
1 i
Amendment No. J7, 79, 3$, 90, 50,126 3-36 i
. _ _ _ _ _ _ _ _, _ _.. - _ _ _.., _ _. ~. _.
o During the physics testing program, the high flux trip setpoints are administratively set as follows to assure an additional safety margin is provided:
Test Power Test Setocint 0
'5%
15 50*
40
- 509, 50 60?:
75 85%
>75 105.1%
l REFERENCES (1 ) FSAR, Section 3.2.2.1.2 (2) FSAR, Section 14.2.2.2 l
Amendment No. 79,126 3-36a
=
O O
r k
1 The principal design basis for the structure is that u be capatia j
cf withstandinc the internal pres:ure resulting frc 1,less of i
c clant accident, as defined in Sectier. Ih, with ne less of integrity.
In this event the total energy ::ntained in the water j
of the reacter coolant system is accused to be released int: the reactor building through a break in the rea:tcr cc:lant piping.
)
Sutcequent pressure behavior is deter:f ned by the buildine volume,
engineered safeguards, and the contined influence of energy source: and heat sinks.
5 2.2 ps;;;;p 33:; ping isola 7 Ion sys;gg i
Leakage through all fluid penetrations not serving a:cident-consequence-limiting systems is minimized by a double barrier i
so that no single, credible failure er calfuncticr. of an active component can result in loss-of-isolation er intolerable leakage.
The installed double barriers take the fer of closed piping systems, both inside and ou+(*s de the reactor building and various i
t types of isolation valvet.
2 REFERENCES (1) FSAR Section 5.1 j
(2) PSAR Section 5 3.1 j
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1 4
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I 5-3 4
l
_ - - - -.=
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5.3 REACTOR Applicability i
Applies-to the design features of the reactor core and reactor coolant systen.
Objective To define the significant design features of the reactor core and reactor I
j coolant system.
l t
Specification 5.3.1 REACTOR CORE i
5.3.1.1 The reactor core contains approximately 93.1 metric tons of slightly enriched uranium dioxide pellets. The pellets are encapsulated in Zircaloy-4 tubing to form fuel rods. The reactor core is made up of rods.tpd23agsemblies. Each fuel assembly contains 208 fuel 177 fy l
1 l
\\
+
l 5.3.1.2 The reactor core shall approximate a right circular cylinder with an equivalent diameter of 128.9 inches and an active height of 142.25 l
t i nches.12 7 i
- 5. 3.1. 3 The average initial enrichment o the current core for Unit 1 is a nominal 2.86 weight percent of g 35 The highest enrichment is less than 3.5 weight percent U?
5.3.1.4 There are 61 full-length control rod assemblies (CRA) and 8 axial power shaping rod assemblies (APSRA) distributed in the reactor core l
as shown in FSAR Figure 3.2-1.
The full-length CRA contain a 134 l
inch length of silver-indium-cadmium alloy clad with stainless steel.(3) The gray APSRA contain a 63 inch length of Inconel.
5.3.1.5 The core will have 68 burnable poison spider assemblies with similar l
l dimensions as the full-length control rods.
The cladding will be l
zircaloy-4 filled with alumina-boron.
l l
5.3.1.6 Reload fuel assemblies and rods shall conform to design and j
evaluation descrggd in FSAR and shall not exceed an enrichment of l
i l.
3.5 percent of U 3 i
1 5.3.2 REACTOR COOLANT SYSTEM 5.3.2.1 The reactor coolant system shall be d accordance with code requirements.(4)esigned and constructed in 5.3.2.2 The reactor coolant system and any connected auxiliary systems exposed to the reactor coolant conditions of temperature and r
pressure, shall be designed for a pressure of 2,500 psig and a i
temperature of 650 F.
The pressurizer and prgSqurizer surge line shall be designed for a temperature of 670 F lDJ I
i Amendment No. 126 5-4 I
Thennal Power Level, %
. 120
_ (-39.4.112)
(38.1,112) 1 ACCEPTABLE
~110 l
~
4 PUMP
- 100 OPERATION
~2
_(-39.4.89.3) 00 (38.1,89.31 1
ACCEPTABLE
(-61.3,80.4)
- 80 (44.2,80.4) 3 & 4 PUMP OPERATION
- 70
_ (-39.4.62.6)
(38.1.62.61 3
(-61.3,57.8)
ACCEPTABLE
-- 60 L (44.2.57.8) 2,3, & 4 PUMP 50 OPERATION
- 40
(-61.3.30.4) 30 I (44.2,30.4) 20 10 t
i i
f f
f f
f f
I I
l f
f I 70 50 30 10 0 10 20 30 40 50 60 70 80 Reactor Power Imbalance, %
Curve Reactor Coolant Flow (lb/hr) 6 1
139.8 x 10 6
2 104.5 x 10 6
3 68,8 x 10 TMI-1 CORE PROTECTION SAFETY LIMITS l
Figure 2,1-2 Amendment No. 17, 17. 77, $$. $$, 17),
126
4
[
l 2400 4
1
\\
~
f 1200 E
[
i 2
4 Ti i
E 2000
/
i
/
T
/
4 J
1800 j
1 1
1600 560 580 800 820 640 660
)
Reactor Outlet Temperature, *F REACTOR COOLANT FLOT CURVE (LBs/HR)
POWER PUMPS OPERATI,NG (TYPE OF LIMIT) 1 139.I : 106 (1005)*
112; Four Puv.s (DNBR Limit)
I 2
104.5 x 106 (74.75) 89.3; inree Pumps (DN8R Limit)
I 3
88.8 x 106 (49.25) 62.6t; One Pump in Eacn Loop (Quality Limit)
- 106.55 of Cycle I Gesign Flow l
TMI l i
CORE PROTECTION SAFETT BASES Amendment No. 59.126 Figura 2.1-3
-. - ~, - - - -, -, -
.-r,
o 2500 g
F = 2300 psig M;CIFTABLE DFDAn0N T = 618.8 F l
g
.#)
y 2100 5/
E
.y
,I,,
F = 1900 psig g
1900 J
UNACCIFTABLE 0? dan 0N
?
3 1700 1300 540 560 580 600 620 640 Amactor Outlet Tamperature, F DC-1 FROTICn0N ST5TD'. F.AIOC.H ALL3iA3LI SET F01lCS Tigure 2.3-1 Amendment No. 13. 77, gg, 37, p 78, 126
o Thennal Power Level, t
-- 120
(-30,108) 110 (25.108) i ACCEPTABLE i
g4 PUMP
- 100 g
y = +2.405 l OPERATION m
m 2 = -5.277 l
90 l
!(-30,80.6)
(25.80.6)!
- ACCEPTABLE 80
(-45.8,70) 3 & 4 PUMP 70 1
(32.2,70) l OPERATION l
1 60 I
!(-30.53.1)
(25.53.1)!
iACCEPTABLE 50
(-45.8,42.6) l2,3, & 4 PUMP l
(32.2,42.6) iOPERATION 40 i
l l
l 30 l
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l
, (32.2,15.1) l 20 ce
(-45.8.15.1) 0 RI 4
e 10 T/ I 7
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' 70 50 30 10 0
10 20 30 40 50 60 70 80 Reactor Power Imbalance, *.
PROTECTION SYSTEM ttAXIMUM ALLOWABLE SETPOINTS FOR REACTOR POWER IMBALANCE l
TMI-1 Figure 2.3-2 Amendment tio. 17, 77. 37, $0. 75. S@. Ji@.126 l
l 1
(300,102)
(76,102)
(268,102) -
100 - NOT ALLOWED (264,92) 90 RESTRIGED (200,80) 80 a
E 4
70 A
t 60 (32,50)
(125,50) i 50 b
h 40 PERMISSIBLE 30 (0,23) 20 10 (0,5) i 0
O 25 50 75 100 125 150 175 200 225 250 275 300 Rod Index, % Withdrawn0 25 50 75 100 a
t t
t Group 7 0
25 50 75 100 a
f f
f I
Group 6 0
25 50 75 100 l
I t
f I
I Group 5 i
i l
ROD POSITI0f4 LIMITS FOR 4 PUMP OPERAT!0ft FROM 0 to 30+10/-0 EFPD TMI-1 l
Figure 3.5-2A Amendment No. 17, 19, 37, 50, 126 r
l
(300,102)
(96,102)
(272,102)
NOT 100 (264,92)
(248,80) 80 a
iE RESTRICTED W
70
~
t 60 (46,50)
(232,50) w 50 I
2 40 30 PERMISSIBLE 20 '(0,15.5) 10 0
25 50 75 100 125 150 175 200 225 250 275 300 0
Rod Index, % Withdrawn
,0 25 50 7p 1,00 Group 7 0
25 50 75 100 t
i f
f e
Group 6 0
25 50 75 100 a
t t
f 3
Group 5 R00 POSITION Lift!TS FOR 4 PUMP OPERATION FROM 30+10/-0 TO 250+10 EFPD THI-1 Figure 3.5-25 Amendment No. 10. 17 27,37, H, 50,126
(300,102)
(171,102)
(272.102) -
100 NOT ALLOWED (264,92) 90
,0) 80 70 M
RESTRICTED 0
60 U
50 (94,50)
(232,50) f 40 E
30 20 PERMISSIBLE (42,15) 10 0
i i
i i
i O
25 50 75 100 125 150 175 200 225 250 275 300 Rod Index, *. Withdrawn 0
25 Sp 7p 1,00 Group 7 0,
25 50 75 100 Group 6 0
25 50 75 100 t
i f
I i
Group 5 R00 POSITION LIMITS FOR 4 PUMP OPERATION AFTER 250!10 EFPD THI-1 Figure 3.5-2C Amendment No. J7, 27, 37, 59,126
100 90 NOT (300,77) 80 a
( 6,77)
(268,77)
E ALLOWED (264,69) 70 RESTRICTED Po 60 (200,60) c 50 I
a.
40 (32,38)
(125,37.5) 30 PERMISSIBLE 20, (0,17,7) 10 0
O 25 50 75 100 125 150 175 200 225 250 275 300 Rod Index, " Withdrawn0 25, 5,0 75 100 Group 7 0
25 50 75 100 a
i f
f f
Group 6 0
25 50 75 100 Group 5 ROD POSITION LIMITS FOR 3 PUMP OPERATION FROM 0 TO 30+10/-0 EFPD THI-1 Figure 3.5-20 Amendment No. 17, 79, 77, $5, 50,126
100 90 (300,77)
(96,77)
, 80 g
NOT (272,77)
ALLOWED W 70 (264,69) m 2 60 (248.60)
RESTRICTED
. 50 L
Ie 40 (46,38)
(232,37.5) 30 20 10 ' - (0.12.1)
PERMISSIBLE O'
0 25 50 75 100 125
.150 175 200 225 250 275 300 Rod Index, % Withdrawn 0 25 50 7p 1,00 Group 7
,0 25 50 75 100 Group 6 0
25 50 75 100 t
i I
t i
Group 5 ROD POSITION LIMITS FOR 3 PUMP OPERATION FR0'4 30+10/-0 TO 250+10 EFPD THI-1 Figure 3.5-2E Amendment No. 17, 17, 37, 70, 50, 170, 126 i
100 90 (300,77)
(171,77)
(272,77)-
80
[
NOT (264,69) m 70 ALLOWED G
(248,60) 60 t
50
- u f
40 (94,38)
(232,37.5) 30 PERMISSIBLE 20 (42,11.7) 10 0
0 25 50 75 100 125 150 175 200 225 250 275 300 Rod Index, t Withdrawn
,0 25 50 75 100 i
i Group 7 0
25 50 75 100 t
i I
I 8
Group 6 0
25 50 75 100 t
I i
i i
Group 5 R00 POSITION LIMITS FOR 3 PUMP OPERATION AFTER 250+10 EFPD TMI-1 Amendment No. 17. 19. 39. 95. 50. 77p,126 1
,.n--
-.n.
,y-,-
o 100 90 80 a
E 70 N
60 (300,52)
( 6,52)
M'W ^
50 NOT y
ALLOWED (264,46)
RESTRICTED E
40 (200,40)
~
(32,26)
(125,25) 20
( 'l '6)
PERMISSIBLE 10 O
O 25
-50 75 100 125 150 175 200 225 250 275 300 Rod Index, % Withdrawn 0
25 50 75 100 Group 7 0
25 50 75 100 t
i I
I B
Group 6 0
25 50 75 100 a
t i
I i
Group 5 R0D POSITION LIMITS FOR 2 PUMP OPERATION FROM 0 TO 30+10/-0 EFPD TMI-1 Figure 3.5-2G Amendment NO.17, 29, U, H, 90,126
o O
100 90 y
80 y
70 N
(96,52)
(272,52) f 50 NOT (264,46)
ALLOWED E
40 (248,40)
RESTRICTED 30 (46,26)
(232,25) 20 10 '
(0,8.7)
PERMISSIBLE 0,2,. 5 )
0 0
25 50 75 100 125 150 175 200 225 250 270 300
~
Rod Index, 7, Withdrawn 0
25 50 7h 100 1
i 4
i Group 7
,0 25 50 75 100 Group 6 0
25 50 75 100 i
i i
i Group 5 R00 POSITION LIMITS F,0R 2 PUMP OPERATION FROM 30+10/-0 TO 250 10 EFPD 1
iMI-1 Amendment No. 77, 77,#0, #5,Ef, 1/A,126
O O
.I
~v i. %
~,,
{
100 j
90 80 f
M 70 N
c'-
60 (300,52)
(171,52)
(272,52)-
50 NOT (264,46)
I ALLOWED (248,40) e 40 P
RESTRICTED 30 (94,26)
(232,25) co PERMISSIBLE 10 0i ( 0,,3. 5)42,8. 5 )
I 0
25 50 7.5 '
100 125 150 175 200 225 250 275 300 Rod Indet, % Withdrawn 0
2p 70 7p 1,00 Group 7 0
25 50 75 100 i
t i
I i
i Group 6 0
, 25 50 75 100 L,.
t i
I Group 5 R0D POSITION LIMITS FOR 2 PUMP OPERATION AFTER 250+10 EFPD TMI-1 s,
knendment sio. 110,126 Figure 3.5-21 9
(
l
Power, % of 2535 MWt l
RESTRICTED REGION - 110
(-20,102)
-100
(-20.7,92)
(26.7,92)
_ gg
(-24,80)<
80
>(32,80)
- 70 PERMISSIBLE
-- 60 OPERATING REGION
- 50
- 40
- 30
-- 20
-- 10 t
I e
t I
I l
e e
! 40 20 -10 0 10 20 30 40 50 Power Imbalance, %
POWER IMBALANCE ENVELOPE FOR OPERATION FROM 0 TO 30+10/-0 EFPD TMI-1 Amendment No.126 Figure 3.5-2J l
Power, % of 2535 MWt RESTRICTED REGION
--110
(-20,102)
(28.5,102) 100
(-22.1,92)<
(28.5,92)
,(32.8,80)
(-22.4,80)o
- 80
- 70 PERMISSIBLE OPERATING 60 REGION
- 50
- 40 30
- 20
- 10 i
i i
i i
i 40 20 -10 0
10 20 30 40 50 Power Imbalance, %
POWER IMBALANCE ENVELOPE FOR OPERATION AFTER 30+10/-0 EFPD TMI-1 Figure 3.5-2Kl Amendment No. 126
i i
i e
i 20
~
18 d1 a
.e lf
~
16 a
/
i W
/
=
/
/
e
/
2
/
14 f
/
~~
g 2
4 0-1000 Wd/mtU z
12
._1000-2600 Wd/mtU Af ter 2600 Wd/mtU 10 0
2 4
6 8
10 12 Axial Location of Peak Power From Bottom of Core, ft LOCA LIMITED MAXIMUM ALLOWABLE LINEAR HEAT RATE TMI-1 Figure 3.5-2L Amendment No.126
.