ML19210A428

Tech Spec Change Request 45 Supporting Licensee Request to Change App a of License DPR-50 Re Changes to Ensure Safe Operation of TMI-1 at Rated Core Power of 2535 Mwt During Cycle 3 Per Cycle 2 Burnup Rate.Certificate of Svc Encl

ML19210A428
Person / Time
Site:	Three Mile Island
Issue date:	01/26/1977
From:	Arnold R METROPOLITAN EDISON CO.
To:
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ML19210A417	List: ML19210A416 ML19210A428
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NUDOCS 7910290647
Download: ML19210A428 (42)

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% ee METROPOLITAN EDISON COMPA?iY JERSEY CENTRAL PO*4ER & LIGHT CCMPA:lY AIID PE:iNSYLVANIA ELECTRIC COMPA!If THREE MILE ISLAIID NUCLEAR STATION UNIT 1 Operating License No. DPR-50 Docket ?!c. 50-269 Technical Snecification Chance Request No. L5 This Technical Specification Change Request is submitted in support of Licensee's request to change Appendix A to Operating License No. DPR-50 for Three Mile Island IIuclear Station Unit 1. As a part of this request, proposed replacenent pages for Appendix A are also included.

METROPOLITAN EDISON COMPANY By /

Vice Pr"quh ent-Generation d

Sworn and subscribed to me this b day of , 1977.

G /

f Notar;gtblic

\h69 201

'910200 697

% ee U'lITED STATES OF M.! ERICA IIUCLEAR REGULATORY COICtISSIO!I III THE MATTER OF DOCKET ?!O. 50-289 LICE ISE I:0. DPR-50 METROPOLITAII EDICOIl COMPA!IY This is to certify that a copy of Technical Cpecification ' hange Request I!o . h5 to Appendix A of the Operating Licence for Three Mile Island Iluelear Station Unit 1, has, on the date given below, been filed with the U. S. Iluelear Regulatory Corniccion and been served on the chief executives of Londonderry Township, Dauphin County, Pennsylvania and Dauphin County, Pennsylvania by deposit in the United Statec mail, addrenced ac followc:

Mr. Weldon B. Archart Mr. Harry B. Reece, Jr.

Board of Cupervicora of Board of County Cormiccionerc Londonderry Township of Dauphin County R. D. #1, Geyers Church Road Dauphin County Court Houce Middletown, Pennsylvania 17057 Harricburg, Penncylvania 17120 METROFOLITA'I EDI30:1 COMFA iY i

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'/ '

By Vice F1451 dent-Generation Dated: January 26, 1977

\h

Metropolitan Edison Co. (Met-Ed)

Three Mile Island liuclear Station, Unit 1 ('INI-1)

Operating Licence No. DPR50 Docket No. 50-289 Technten1 Gnociff eation Change Request No. 'v; The Licencee requests that the attached changed pages replace pagea y, vi, vii, 1-5, 2-3, 2-5, 2-6, 2-7, 3-1, 3-2, 3-3h, 3-3ha, 3-35, 3-T;a, 3-36, h-59 figures 2.1-1, 2.1-2, 2.1-3, 2.3-1, 2.3-2, 3.5-2A through 3 5-2J and table 2.3-1 of the existing technical specifications.

Reason for Proroced Change These changen are necessary to ensure safe operation of 'INI-1 at a rated core power of 2535 !Gt for the duration of Cycle 3 and are baced on a Cycle 2 burnup of 253 10 EFPD.

• Changes to technical specifications are necessary as a result of the changed distribution of fuel by insertion of h8 tesh batch 5 accemblies and 13 once burned batch la assemblies, discharge of batch 2 fuel, and relocation of the remaining twice burned batch 3 and once burned batch h fuel. Other changes to the technical specification are responsive to recent NRC concerns relative to postulated fuel rod bowing and correcting a previouc error in densification penalty as discussed in our Change Requent 36 (July 7, 1976). In addition, revised quadrant tilt limits and new limits for the Axial Power Shaping Rods (APCR's) are specified. Furthermore our change request to revise the high precsure trip and pressurizer code cafety valve cettings (Change Requect #39, October 8, 1976 as supplemented by oc letter c; October 21, 1976) is hereby extended to apply to Cycle 3, therefo;e approval of these propoced technical specifications effectively fulfills our Chang" Request #39 Safety Evaluation Justifying Change Cycle 3, like the reference Cycle 2, was evaluated using the B&W-2 CHF correlation using a 95/95 confidence level and a RC flow equal to 106.5% of Cycle 1 design flow. The B&W 2 CHF correlation and the 106.5% RC flow value have been reviewed and approved by the NRC. In addition, the accured flow of 106.5% for Cycle 3 is even more conservative than for Cycle 2 in that Cycle 3 inecrporates L8 additional Mark Bk assemblies with improved flow characteistics, for a total of 10h of 177 accemblies with improved flow relative to tne Mark B decign of batchen 1, 2 and 3.

The recent NRC concernc relative to rod boving have been considered by including a DNBR penalty of 11.2% imposed by the NEC. It should be noted that the former error made in the application of dencification penalty has nico been corrected (reference CQL-0821, June 5,1976).

)hY

% ee Revised quadrant tilt limits are proposed to clarify how the tilt limits should be applied based on the method used in determining tilt. The revised tilt limits are more restrictive than formerly used resulting in additional operating flexibility for the rod insertion limits while at the same time not reducing the degree of conservatism that would exist had the insertion limits been based on a larger tilt value.

Limits for the APSR's have been established which provide additional control of power peaking through an improved definition of core power distribution.

This change provides additional operating flexibility while at the same time not reducing the degree of conservatism relative to the limits that vould be appropriate if no APSR limits were specified.

The increases in the high pressure trip and pressuriser code safety valve settings are justified by Change Request #39 (October 8,1976) as supplemented October 21, 1976. Since the parameters relevant to the supporting analyses of Change Request #39 remain unchanged for Cycle 3, except for the nominal moderator and doppler coefficients used, and since these coefficients are a little different for Cycle 3 and are bounded by the moderator and doppler sensitivity studies, the justification given in Change Request #39 is applicable to Cycle 3.

The FSAR accidents which are dependent on core design or fuel loading have been reviewed for Cycle 3 The results presented in the FSAR bound the Cycle 3 results.

The LOCA analyses were reviewed and approved by the NRC with raendment 17 to the TMI-l operating license. Since that time an error has b en identified in the CRAFT 2 code which comprises part of the ECCS evaluatit - model required by Appendix K to 10CFR50. The errors to this code have been corrected and selected analysis have been performed to show that the code errors resulted in much less than a 20 F increase in the peak cladding temperatures previously predicted which the=selves were well below the 2200 0 F limit specified by regulations.

~

The NRC evaluated these additional analyses and concluded that no changes in Operating reactor technical specifications were necessary. In spite of the known error in the Appendix K model, the model has been shown to be conservative.

Consequently, B&W 10103 Rev. 1 is still considered a valid reference to support approval of these proposed technical specifications while revised analyses are being conducted. Additional and more detailed information supportive of these proposed technical specifications is provided in the attached Three Mile Island Unit 1, Cycle 3, Reload Report.

For all of the attached proposed technical specifications not previously requested the limits have beccme more conservative or are as conservative as their equivalent for Cycle 2. The above is true since these limits were developed for both Cycle 2 and Cycle 3 to meet the same criter .. The rod insertion and power imbalance limits =ay be different for various fuel distributions frca cycle to cycle, however, the revised limits should not t* considered a change that involves significant hazards since the degree of conservatism has not been reduced. The revised high pressure trip and code safety valve settings and the elimination of the reactor vent valve flow penalty (Change Requests 39 and 36 respectively) have been previously noticed to the public. As explained above, no other significant hazards are involved, therefere, public notice should not be required.

1469 210 (2)

% eo Based on the above, it is concluded that this change does not increase the probability of occurrence or increase the severity of an accident, nor does it create the possibility of cccurrence of an accident not previously considered.

Therefore, this change doc.s not represent undue risk to the health and safety of the public and contir aed operation of TMI-l at a rated core power of 2535 MWt is justified.

1469 211 (3)

% ee LIST OF TABLES Table Title Page 2.3-1 Reactor Protection System Trip Jetting Limito 2-9 2.3-2 High Precoure Reactor Trip and Pressurizer Code 2-10 39 Safety Valves Setting Faira 3 5-1 Instrumento operating conditiono 3-29 L.1-1 Instrument Surveillance Requiremento h-3 h.1-2 Inctrument Surveillance Requiremento h-3 h.1-3 Minimum Sampling Frequency h-8 h.2-1 Inctrument Surveillance Program h-lh h.h-1 Selected Tendonc and Correcronding Instection h - 3';a Periodo h.h-2 Tendona Selected for Tendon Physical Condition Test h-36 h.h-3 Ring Girder Surveillance h-36g h.15-1 Radioactive Liquid Wacte Sampling and Analycia L-59 h.15-2 Radioactive Gaceous Waste Sampling and Analycio h-63 6.12-1 Protection Factors for Reapiratora 6-23 9

)k ) L

% ee LIST OF FIGURES Ficuve Title 2.1-1 Core Protection Safety Limit 2.1-2 Core Protection Safety Limits 2.1-3 Core Protection Safety Basis 2.3-1 Protection System Maximum 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 5-1 Incore Instrumentation Specification Axial Imbalance Indication 3.5-2 Incore Instrumentation Specification Radial Flux Tilt Indication 3 5-2A Rod Position Limits for h Pump Operation Applicable During the Period from 0 to 100 10 EFPD; Cycle 3 3.5-2B Rod Position Limits for h Pump Operation Applicable During the Period from 100 10 E7PD; to 2h6 10 EFFD; Cycle 3 3 5-2C Rod Position Limits for h Pump Operation Applicable During the Period after 2h6 i 10 EFPD; Cycle 3 3 5-2D Rod Position Limits for 2 and 3 Pump Operation Applicable During the Period from 0 to 100 10 EFPD; Cycle 3

~

3.5-2E Red Position Limits for 2 and 3 Pump Operation Applicable During the Period from 100 1 10 to 2h6 10 EFPD; Cycle 3 3 5-2F Rod Position Limits for 2 and 3 Pu=p Operation Applicable During the Period After 2h6 10 EFPD; Cycle 3 3.5-2G Operational Power Imbalance Envelope Applicable to Operation from 0 to 100 10 EFPU; Cycle'3 vi )kb

% ee Firure Title 3.5-2H operational Power Imbalance Envelope Applicable to Operation from 100 10 to 2h6 10 EFPD; Cycle 3 3.5-2I Operational Power Imbalance Envelope Applicable to Operation after 2h61 10 EFPD; Cycle 3 3.5-2J LOCA Limited Maximum A11ovable Linear Heat Rate 3 5-2K APSR Position Limits for Operation from 0 to 100 10 EFPD; Cycle 3 3 5-2L APSR Position Limits for Operation from 100t 10 to 2h6t 10EFFD; Cycle 3 3.5-2M APSR Position Limits for Operation after 2h6t 10 EFPD; Cycle 3 3 5-3 Incore Instrumentation Specification h.2-1 Equipment and Piping Requiring Inservice Inspection in Accordance with Section XI of the ASME Code h.h-1 Ring Girder Surveillance L.4-2 Ring Girder Surveillance Crack Pattern Chart h.h-3 Ring Girder Surveillance Crack Pattern C'. art k.h b Ring Girder Surveillance Crack Pattern Chart h.h-5 Ring Girder Surveillance Crack Pattern Chart 6-1 Organization Chart 1469 214

% es 1.6 POWER DISTRIBUTIO::

1.6.1 QUADRAI;T POWER TILT Quadrant power tilt is cefined by the following equation and is expressed in percent. __

100 Power in any core cuadrant ~y Average power of all quadrants The quadrant tilt limits are stated in Specification 3 5 2.h.

1.6.2 REACTOR PCWER IMBALANCE Reactor power i= balance is the power in the top half of the core minus the power in the bottom half of the core expressed as a percentage of rated power. Imbalance is menitored continuously by the RPS using input from the power range channels. Imbalance limits are defined in Specification 2.1 and imbalance setpoints are defined in Specification 2.3 17 CONTAII.7ENT' INTEGRITY Containment integrity exists when the following conditions are satisfied:

a. The equipment hatch is closed and sealed and both doors of the personnel hatch and emergency hatch are closed and scaled except as in "b" below,

b. At least one docr on each of the personnel hatch and emergency hatch is closed and sealed during refueling or personnel passage through these hatches.

c. All non-autcmatic containment isolation valves and blind flanges are closed as required by the " Containment Integrity Check List" attached to the operating procedure " Containment Integrity and Access Limits."

d. All automatic containment isolation valves are operable os locked closed.

e. The containnent leakage determined at the last testing interval satisfies Specification h.h l.

1469 215 1-

% eo The curve of Figure 2.1-1 is the most restrictive of all possible reactor coolant pump-maximum thermal power combinations shown in Figure 2.1-3 The curves of Figure 2.1-3 represent the conditions at which a =ini=u=

DNBR of 1.3 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 minimus DNER is equal to 22 percent, (3) whichever condition is more restrictive.

The maximum thermal power for three pump operation is 86.7 percent due ta a power level trip produced by the flux-flow ratio (Th.7 percent flov x 1.08 = 80 7 percent power) plus the =aximum calibration and instrumentation error. 'Jhe maxi =um 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 DNER 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 minimum DNER.

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 left of the curve vould result in a DNER greater than 1.3 or a local quality at the point of minimum DNBR less than 22 percent for that 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 left of this curve vill be above and to the left of the other curves.

REFERENCES (1) FSAR, Section 3.2.3.1.1 (2) FSAR, Sectica 3.2.3.1.1.c (3) FSAR, Section 3.2.3.1.1.k t

i 2-3 )&(30 L\N

% ee 2.3 LEtITING 3AFETY SYSTDt 3ETTINGS, PROTECTION INSTRGENTATION Apulicability_

Applies to instru=ents monitoring reactor power, reactor power imbalance, reactor coolant syste= pressure, reactor coolant outlet te=perature, flow, number of pumps in operation, and high reactor building pressure.

Objective To provide automatic protection action to prevent any combination of process variables from exceeding a safety 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 proLection system consists of four instrument channels to monitor each of seve: C selected plant conditions which vill cause a reactor trip if any one ef these condations 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 "9 2

side of the protection system bistable comparators. The safety analysis has been based upon these protection system instrumentation trip set points plus calibration and instrumentation errors.

Nuclear Overpower A reactor trip at high power level (neutron flux) is provided to prevent damage to the fuel cladding from reactivity excursions too rapid to be detected by pressure and temperature measurements.

During normal plant operation with all reactor coolant pumps operating, reactor trf.p is initiated when the reactor power level reaches 105 5% of rated powe: . Adding to this the possible variation in trip set 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 uced in the safety analysis (1).

a. Overpower trip based on flow and i= balance j469 2}[

The power level trip set point produced by the reactor coolant system flow is baced on s pcwer-to-flow ratio which has been established to acconmodate the most severe thermal transient censidered in the design, the loss-of-coolant flow accident from high power. Analysis has demonstrated that the specified pcVer to flow ratio is adequate to prevcct a OUER of less than 1.3 should a low flow condition exist due to any malfunction.

2-

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 increaces 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 minimum permissible lov flow rate. Typical power level and lov flow rate combinations 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, or flow rate is 92.6 percent and power level is 100 percent.

2. Trip would occur when three reactor coolant pumps are operating if power is 80 7 percent and reactor flow rate is Th.7 percent or flow rate is 69.2 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 52 9 percent and reactor flow rate is h9 2 percent or flow rate is h5.h percent and the power level is 49 percent.

The flux / flow ratios account for the maximum calibration w.d instrumentation errors and the maximum variation from the average value of the RC flow signal 39 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 fcr an open core vent valve because of the core vent valve surveillance program during each 36 refueling outage.

For safety analysis calculations the maxi =um calibration and instrumentation errors for the power level were used.

The power-imbalance boundaries are established in order to prevent reactor thermal limits frem being exceeded. These thermal limits are either power peaking kW/ft 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 r7dundant pump monitors prevent the minimum core DNER from decreasing below 13 by tripping the reactor due to the loss of reactor coolant pump (s). The pump monitors also restrict the power level for the nu=ber of pumps in operation.

1469 2IB 2-6

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 nucleer overpower trip set point. The trip setting limit shown in Figure 2.3-1 for high reactor coolant system pressure has been established to maintain the system pressure below the safety limit (2750 psig) for any design transient. Due to calibration and 39 instrument errors, the safety analysis assumed a 30 psi pressure error in the high reactor coolant system pressure trip setting.

The low pressure (1800 psig) and variable low pressure (11 75 Tout - 5103) trip setpoint shown in Figure 2.3-1 have been establishod to maintain the DNB ratio greater than or equal to 1 3 for those design accidents that. result in a pressure reduction (3, b).

Due to the calibration and instrumentation errors, the safety analysis used a variable low reactor coolant system pressure trip value of (11.75 Tout -

Slh3) and a low pressure trip value of 1770 psig. l39

d. 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 excessive core coolant temperatures 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 619 F. Under the vorst case environment, power supply perturbations, and drift, the accuracy of the trip string is IF. This accuracy was arrived at by summing the vorst 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 vorst 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 drift, linearity, repeatability, etc. This does not imply that the equipment is restricted to operation within the calibrated range. Additional testing has demonstrated that in fact, the temperature channe? is fully operational approximately 10% above the calibrated range.

Since it has been established that the channel vill trip at a value of RC outlet temperature no higher than 620 F even in the worst case, and since the channel is fully operational approximately 10% above the calibrated range and exhibits no hysteresis or foldover character-istics, it is concluded that the instrument design is acceptable.

e. Reactor building pressure }4hh 2l9 The high reactor building pressure trip setting limit (L psig) provides positive assurance that a reactor trip will 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 reactor coolant system pressure trip.

TABLE 2 3-1(6)

REACTOR PROTECTION SYSTEM TRIP SETTING LIMITS Four Reactor Coolant Three Reactor Coolant One Reactor Coolant Pumpn Operating Pumps Operating Pump Operating in (Nominal Operating (Nominal Operating Each Loop (Nominal Shutdown Power - 100%) Power - 75%) Operating Power 19%)

4 Bypass

1. Huclear power, Max. 105 5 105 5 105.5 5 0 (3)

% of ated power P 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 max. of rated power imbalance (s) imbalance (r) imbalance (s)

3. Nu lear power based NA NA 91% Bypassed (5 on pump monitors, max. % of rated power 14 High reactor coolant See Table 2.3-2 II) See Table 2.3-2 (7) See Table 2.3-2 (7) 1720 U') 3 y system pressure, psig, O 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 coolant system pressure psig, min.

7 React.or cooiant temp. 619 619 619 619 F., Max.

8. High Reactor Building la la 14 14 pressure, psig, max.

Tout is in degr es Fahrenheit, (F) e Reactor coolant system flow, %

(3) Administratively controlled reduction set only during reactor shutdown N (1 ) Automatically set when other segments of the RPS (as specified) are bypassed 4

N (5) The pump monitors also produce a trip on: (a) loss of two reactor coolant pumps in one reactor coolant loop, and O (b) loss of one or two reactor coolant pumps during two-pump operation

6) Trip settings limits are setting limits on the set point side or the protection system bistable coraparators (7) Settinga are specified in Table 2.3-2 corresponding to the pressurizer code safety valve settings. 3.

TABLE 2 3-2 HIGH PRESSURE REACTOR TRIP M.'D PRESSURIZER CODE SAFETY VALVES SETTING PAIRS Alternate Fairs High Pressure Pressurizer Code of Settings Reactor Trip Setting Safety Valve Setting 2375 psig 2h35 psig A[/

2 2h05 psig B/ 2500 psig 39 Notes: .

1/ The pair of settings denoted as A vill be utilized until the pressurizer code safety valve setting can be increased from 2h35 psis to 2500 psig.

2/ The pair of settings denoted as 3 vill be utilized on a Ier=anent basis following resetting of the pressurizer code safety valves.

1469 22I 2-10

2600 2400

.5 E

ACCEPTABLE OPERATION /

[

fy ,

2200 f

E 2000

/ UNACCEPTABLE OPERATION 1800

/

1600 560 580 600 620 640 660 Reactor Outlet Temperature, F THI 1, UNIT 1, CYCLE 3 CORE PROTECTION SAFETY LIMIT Figure 2.1 1 1469 222

Tnermal Power Level, 5

-- 120 (112)

( 33,112) (+35,112)

_g ACCEPTABLE Kw/ft Limit - - 100 4 MP Kw/ft limit OPERATION

( 56,90) (-33,86.7) -- 90 (86.7) (+35,86.7) '

h ,, gg ACCEPTABLE 3 & 4 PUMP OPERATION

-- 70

( ds,64.7) ( 33,59.1) (+35,59.1)

__ 00 (59.1)

W ACCEPTABLE 5

2,3 & 4 PUMP (+47,48)

( 46,47)

OPERATION

-- 40

-- 30

-- 20

-- 10 I I I I i 60 -40 -20 0 20 40 60 Reactor Power imoalance, 5 CURVE REACTOR COOLANT FLOW (lo/nr) 1 139.8 x 106 2 104.5 x 106 3 68.8 x 106 TMl 1, UNIT 1, CYCLE 3 CORE' PROTECTION SAFETY LIMITS Figure 2.1 2 1469 223

2500 P = As specified in i able 2.3,2 a 2300 -..

" ACCEPT!.2:

OPERATION T = C13 F 5

m

[ 2100 5 9

= s a se 1900 4/ UNACCEPTABLE g

3 P = 1800 psig OPERATION f

1700 1500 560 580 600 620 640 540 Reactor Outlet Temperature, F TMl-1, UNIT 1, CYCLE 3 PROTECTION SYSTEM MAJ' MUM ALLOWABLE SET POINTS Figure 2.3-1

2400 1

\

5 2200

• {

2 5

e

/

2000

% ._ 3 5

E 1800 y <

t 1600 620 640 660 560 580 600 Reactor Outlet Temperature, *F REACTOR COOLANT FLOW POWER PUMPS OPERATING (TYPE OF LIMIT)

CURVE (LBS/HR) 139.8 x 106 (100",)* 112% Four Pumps (DNBR Limit) 1 86.7% Three Pumpt (DNBR Limit) 2 104.5 x 106 (74.7%)

59.1% One Pump in Eacn 1. cop (Quality Limit) 3 68.8 x 106 (49.2%)

• 106.5% of Cycle 1 Design Flow TMl-1, UNIT I, CYCLE 3 CORE PROTECTION SAFETY Figure 2.1 -3 1469 225

Power Level, %

- - 120

- 110 (108) _ (+21,108)

(-20,108) -

$+

g% -- 100 o

x- ACCEPTABLE 4

1. 4 PUNP -- go

+' OPERATION

- (80.7) (48,80)

(-46,80) 80 1 -- 70 ACCEPTABLE 3 & 4 PUNP OPERATION 60 (53.I) (+4 8,52. 7 )

(-46,52.7) ~~

50 I I 40 ACCEPTA8LE 2,3 & 4 PUMP

-- 30 OPERATION

(+48,25.I)

(-46,25.I)

-- 20 l

I -- io la  ?

,l ,l [ I ,, ,,

i- i i i i I r, l , y

-10 10 20 30 40 50

-50 -40 -30 -20 0 Power Imbalance, %

THI-l, UNIT I, CYCLE 3 PROTECTION SYSTEN MAtlMUM ALLOWASLE SET PolNTS Figure 2.3-2 1469 226

3 LIMITING CONDITIONS FOR OPERATION 3.1 REACTCR C00LAliT SYSTEM 3 1.1. OPERATIONAL CCMPONE1TS Atelicability  !

Applies to the operating status of reactor coolant system components.

Objective To specify those limiting conditions for operation of reactor coolant system components which must be met to ensure safe reactor operations.

Scecifica'. ion 3 1.1.1 Reactor Coolant Pumps

a. Pump conbinations permissible for given power levels shall be as shown in Specification Table 2.3 1.

b. Fover cperation 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 combination at the end of the 2k 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 generator shall be operable whenever the reactor coolant average temperature is above 2500F.

3.1.1.^ Pressurizer Safety Valves

a. The reactor shall not remain critical unless both pressurizer code safety valves are operable with one of the lift 39 settings specified in Table 2.3-2 tl% allowance for error.

b. 'a' hen the reactor is 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.

)hY 3-1

Bases The limitation on power cperation with one idle RC pump in each loop has been imposed since the ECCS cooling performance has nct 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 pump (s) and to return the reactor to an acceptable ecmbination of operating RC pumps. The 2h hours 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 pump or decay heat removal pump is required to be in operation before the boron concentration is reduced by dilution with makeup water.

Either pump vill provide mixing which vill prevent sudden positive reactivity changes caused by dilute coolant reaching the reactor. One decay heat removal pump vill 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 pressurizer heaters, and reactor decay heat.(h) sources which are puap Both pressurizer codeenergy, safety valves are required to be in serv.' .e prior to criticality to conform to the system design relief capabilit for a rod withdrawal accident. N . The The code safety valves pressurizer prevent code safety overpressure valve lift set point shall be set at one of the settings specified in Table 2 3-2 t1 percent allevance for error and each valve shall be capable of relieving l_9 0

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 k-T (2) FSAR, Sections h.2.5 1 and 9 5 2.3 (3) FSAR, Section h.2 5.k (h) FSAR, Sections h.3.10.h and h.2.h 4gg 228 (5) FSAR, Section h.3.T 3-2

O

f. If a control rod in the regulating or axial power shaping groups is declared inoperable per Specification h.T.1.2.,

operation may 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 Specification h.7 1.2.

g. If the inoperable rod in Paragraph "e' above is in groups 5, 6, T, or 8, the other rods in the group shall 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 rol that was declared inoperable is maintained within allowable group average position limits in 3 5 2 5 3 5.-.3 The worth of single inserted control rods during criticality are limited by the restrictions of Specification 3 1.3.5 and the Control Rod Position Li=its defined in Specification 3.5.2 5 3.5.2.h Quadrant tilt:

a. Except for physics tests the quadrant tilt shall not exceed

+2.66% as determined using the full incorc detector system.

b. When the full incore detector system is not available and except for physics tests quadrant tilt shall not exceed +1.k7% as determined using the minimum incore detector system.

c. When neither incore detector system above is available and except for physics tests quadrant tilt shall not exceed +0.81%

as determined using the power range channels dispicyed on the console for each quadrant (out of core detector system).

d. Except for physics tests if quadrant tilt exceeds the tilt limit power shall be reduced immediately to below the power level cutoff (see Figures 3 5-2A, 3.5-23 and 3.5-2C). Moreover, the power level cutoff value 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 limit.

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.

3-3L \

2. The control rod group withdrawal limits (Figures 3.5-2A, 3.5-2B, 3.5-2C, 3.5-2D, 3 5-2E, 3 5-2F, 3 5-2K, 3.5-2L, and 3 5-2M) shall be reduced 2 percent in power for each 1 percent tilt in excess of the tilt limit.

3. The operational imbalance limits (Figure 3.5-2G, 3.5-2H and 3.5-2I) 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 +25.72% determined using the full incore detector system (FIT), or +2h.09% determined using the minimum incore detector system (MIT) if the FIT is not =vailable, or +21 39%

determined using the out of core detector system (OCT) when neither the FIT nor MIT are available, the reactor vill be placed in the hot shutdown condition. Diagnostic testing during power operation with a quadrant tilt is permitted provided that the thermal power allovable is restricted as stated in 3 5 2.h.d above.

g. Quadrant tilt shall be monitored on a minimum frequency of once every two hours during power operation above 15 percent of rated power.

1469 230 3-3ha

3.5.2 5 control nod sition .

h

a. Operating rod group overlap shall not exceed 25 percent 1 5 percent, between two sequential groups except for physice tests.

b. Fosition limits are specified for regulating and axial power shaping control rods. Except for physics tests or exercising control rods, the regulating control rod insertion /vithdrawal limits are specified on Figures 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 or two pump operation. Also excepting physics tests or exercising control rods, the axial power shaping control rod insertion /vithdrawal limits are specified on Figures 3.5-2K, 3.5-2L, and 3 5-2M. If any of these control rod position limits are exceeded, corrective measures shall be taken i=nediately 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 power level cutoff (See Figures 3.5-2A, 3.5-23 and 3.5-2C) unless the xenon reactivity is within 10 percent of the equilibrium value for operation at rated power and asymptotically approaching stability.

d. Core imbalance shall be monitored on a minimum frequency of once every two hours during power operation above h0 percent of rated power. Except for physics tests, corrective measures (reduction of imbalance by AFSR movements and/or reduction a. eactor power) shall be taken to maintain operation within the envelope defined by Figures 3 5-2G, 3 5-2H and 3 5-2I. If the imbalance is not within the envelope defined by Figures 3.5-2G, 3 5-2H and 3.5-21 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 times with limited access to be authori::ed by the superintendent.

3.5.2.7 A power map shall be tsken at periodic intervals of 30 full power days using the incore instrumentation detection system, to verify that the power distribution is within the limits shown in Figure 3.5-2 J.

Bases The power-imbalance envelope defined in Figures 3.5-2G, 3.5-2H, and 3 5-2I is based on LOCA analyses which have defined the maximum linear heat rate (see Figure 3.5-2J) such that the maximum clad temperature vill not exceed the Final Acceptance 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 limited 'cy the Final Acceptance Criteria only if the control rods are at the withdrawal / insertion limits as defined by 3-35 1469 2M

Figures 3.5-2A, 3 5-2B, 3 5-20, 3.5-2D, 3 5-2E, 3 5-2F, 3 5-2K, 2.1-2L, and 3.5-2M and if quadrant tilt is at the limit. Additional conservatism is introduc by application of:

a. Nuclear uncertainty facters

b. Thermal calibratien uncertainty

c. Fuel densification effects

d. Hot rod manufacturing tolerance factors,

e. Postulated fuel rod bow effects 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, 3.5-2K, 3 5-2L and 3.5-2M describe three regions. These three regions are:

1. Permissible operating Region

2. Restricted Regions

3. Prohibited Region (Operatien 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 limiting condition for operation. The limiting criteria within the Restricted Region are potential ejected rod worth and ECCS power peaking and since the probability of these accidet.ts is very low especially in a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> time frame, inadvertant operation within the Restricted Region for a period of h hours is allowed.

1469 232 3-35a

The 25!5 percent overlap between successive control rod groups is allowed since the worth of a rod is lower at the upper and lover part of the stroke. Control rods are arranged in groups or banks defined as follows:

Group Function 1 Safety 2 Safety 3 Safety h Safety 5 Regulating 6 Regulating T

Regulating (Xenon transient override) l 8 APSR (axial power shaping bank)

Control rod groups are withdrawn in sequency beginning with group 1. Groups 5, 6 and T are overlapped 25 percent. The normal position at power is for groups 6 and T to be partially inserted.

The rod position limits are based on the most limiting of the folleving three criteria: ECCS power peaking, shutdown cargin, 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, consistent 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 e.c rated power. These values have been shown to be safe by the safety analysis (2) of the hypothetical rod ejection accident. A maximu=

single inserted control rod worth of 1.0% Ak/k is allowed by the rod position limits at hot zero power. A single inserted control rod worth 1.0%dk/k at beginning of life, hot, zero power vould result in a lower transient peak thermal power and, therefore, less severe environ = ental consequences than 0.65% Ak/k ejected rod vorth at rated power.

The plant computer vill scan for tilt and imbalance and will satisfy the technical specification requirements. If the computer is out of service, then manual calculation for tilt above 15 percent power and imbalance above h0 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.h have been established within the thermal enalysis design base using an actual core tilt of +3.bl% Which is equivalent to a +2.66% tilt measured with the full incore instrumentation with measurement uncertainties included.

During the physics testing program, the high flux trip setpoints are admini-stratively set as follows to assure an additional safety margin is provided:

Tric Setpoint Test Power 0 <5%

15 So5 5 0,- 1469 233

~0 50 50%

85%

75

>75 105 5%

3-36

REFERETCES (1) FSAR, SEction 3 2.2.1.2 (2) FSAR, Section 14.2.2.2 1469 234 3-36a a

h.16 REACTOR INTERNALS VENT VALVES SURVEILIANCE Apolicability Applies to Beactor Internals Vent Valves.

Objective '

To verify that no reactor internals vent valve is stuck in the open position and that each valve continues to exhibit freedom of movement.

Specification 36 h.16.1 At intervals not exceeding the refueling interval, each reactor internals vent valve will be tested to verify that no valve is stuck in the open position and that each valve continues to exhibit freedom of movement.

Bases Verifying vent valve freedom of movement insures that coolant flow does not bypass the core through reactor internals vent valves during operation and therefore insures the conservatism of Core Protection Safety limits as delineated in figures 2.1-1 and 2.1-3, and the flux / flow trip setpoint.

}khh .

h-59

174.102 tes.a.102 112.802 '

OPERATION IN THIS 100 -

REGION IS NOT ALLOWED 225,8.90 POWER LEVEL 174.2.90 ,

CUT 0FF 90 -

239.8.80 is0.2.s0 80 - RESTRICTED REGION SHUTOOWN 846.2.70 300.7o 239.s.70 70 MARGIN LIMIT 129.0.s0

60 -

m E ' PERMISSIBLE

~

OPERATING

• REGION

{ 40 -

E 30 -

I 20 _.

0.15 10 -

i i i I i f i I I I 250 275 300 150 175 200 225 75 100 125 O 25 50 R00 Index. 5 Withdrawn.

50 75 100 0 25 t 1 t t i

Group 7 75 100 25 50 0 t J I i 1 Group 6 75 100 0 25 50 i L # i i R00 POSITION LIMITS FOR 4 PUMP 10 EFFD Group 5 OPERATION FROM 0 TO 1001 T M I- 1, CYCLE 3 Figure 3.5-2A 1469 236

176.o,to2 225.s.102 r

100 _ OPERATION IN THIS REGION POWER LEVEL IS NOT ALLOWED 90 - .i7u.2.so '25.s.9 2n.s.so 80 - iso.2.so RESTRICTED REGION I - iss.2.70 239.s.70 300.70

=

N g 60 SHUT 00NN MARGIN 129.o.60 50 98.50 PERMISSIBLE OPERATING REGION 40 -

30 -

t 20 -

Ss.i5 70.l5 20 -

o.o I I I f I I I I i I I 175 200 225 250 275 300 50 75 100 125 150 G 25 Rod index. 5 Withdrawn 0 25 50 75 100 i f I I I Group 7 25 50 75 100 0

t f f 1 Group 6 25 50 75 100 0

t t i i i R00 POSITION LIMITS FOR 4 PUMP Group 5 OPERATION FROM 100 t 10 EFP0 TO 246110 EFP0 TMl-l. CYCLE 3 i Figure 3.5-2B

)hY

194.102 254.s.102 100 - OPERATION IN THIS OsE L L REGION IS NOT ALLOWED g0F 90 -

, RESTRICTED 237.s.s0 80 _ REGION

SHUT 00WN MARGlH

) 60 _

LIMIT 2

m 110,5 PERMISSIBLE f50 -

OPERATING

- REGION g

30 -

20 -

80.15 10 0.0 1 1 1 1 I I 1 1 1 l l 175 200 225 250 275 300 25 50 75 100 125 150 O

R0d index, ",Witndrawal 25 50 75 100 0

t I l I f Group 7 0 25 50 75 100 I ( l t 1 Group'6 0 25 50 75 100

' I t f t

Group 5 R00 POSITION LIMITS FOR 4 FUMP OPERAil0N AFTER 246 1 10 EFPD TMi-I CYCLE 3 Figure 3.5-2C 1469 238

160.102 240.102 I12.t02 119.102 RESTRICTED REGION 100 _ OPERATION IN THIS FOR 3 PUMP REGION IS NOT ALLOWED i gs, gg 300.89 90 - 240.89

.o 380 - SHUTOOWN

'E MARGIN '28 76 5 LIMIT Ti - &

E M 60 -

PERMISSIBLE OPERATING g REGION S6.50 I50 i

~

4 540 -

2 30 -

Restricted Region for 2 & 3 pu=p Operation j20 O. is 10 -

t I i i t _,

I I I 0*0 I i 1 I

250 275 300 i 150 175 200 225 75 100 125 0 25 50 R00 Index. 5 withdrawal 25 50 75 100 0 i i t i i

Group 7 75 100 0 25 50 J

1 I I 1 Group 6 75 100 0 25 50 i i t i Group 5 R00 POSITION LIMITS FOR 2 AND 3 PUMP OPERATION FROM 0 TO 100 t 10 EFPD TM l-1 CY CLE 3 Figure 3.5 20 1469 239

176.602 240.102 100 -

RESTRICTED REGIGN FOR 3 PUMP OPERATION IN THIS REGION ( IRATION 90 IS NOT ALLOWED 240.89 300.89 g 80 -

m 5 70 -

a p 60 -

PERMISSIBLE OPERATING g SHUT 001N MARGIN REGION o LIMIT a: 50 -

9a .50 o

w g 40 -

o 2 30 -

~

o

1. 20 -

, g' 58,15 j 10 -

0. 0 0 I f f I I f I I I f I O 25 50 75 100 125 150 175 200 225 250 275 300 R00 Index, 5 Withdrawal 0 25 50 75 100 t 1 I l 3roup 7 0 25 50 .75 100 I t i I Group 6 0 25 50 75 100 t I t 1 I Group 5 R00 POSITION LIMITS FOR 2 AND 3 PUMP OPERATION FROM 1001 10 EFP0 TO 2461 10 EFP0

'MI-1, CYCLE 3 Figure 3.5-2E

)hh

194 602 238.102 100 -

OPERATION IN THIS REGION IS NOT ALLOWE0 RESTRICTED REGION FOR 90 -

3 PUMP OPERATION E

j 80 -

5 a

SHUT 00WN b 70 -

, MARGIN  :

S LIMIT PERMISSIBLE OPERATING

]60 cc REGION 2 50 - 180 50 2

340 2

G g 30 -

o f 20 -

g es.ls o.

, 10 -

0. 0 I l l l l I I l t f f 0 25 50 75 100 125 150 175 200 225 250 275 ~300 Rod Index, ",Witndrawn 0 25 50 75 100 i t t  ! I Group 7 0 25 50 75 100 l l t t t Grcup 6 0 25 'O 75 100 t I f I f Gr up 5 R00 POSITION LIMITS FOR 2 AND 3 PUMP OPERATION AFTER 246 t 10 EFPO TMI-1, CYCLE 3 Figure 3.5-2F

\h69

P0wer. 5 Of 2535 MWt RESTRICTED REGION 12.04.102

-17.54.102

-100

-18.00.90 -- 90 12.06.90 80 70 PERMISSIBLE OPERATING. 60 REGION

-- 50

-- 40 1

-- 30

-- 20

-- 10 t I i e I l i t g g 40 -30 -20 -10 0 10 20 30 40 50 Axial P0wer Imualance, 5 POWER IMBALANCE ENVFLOPE FOR OPERATION FROM 0 T0 100 I la EiPD TMl 1 CYCLE 3 Figure 3.5-2G 1469 242

P0wer, % pf 2535 MWt RESTRICTED REGION

,33,,9, ,

1

-- 100

-22.50.90 -- 90 9.49.90

-- 80

- 70 PERMISSIBl.E OPERATING

~~

REGION

-- 50

- 40

-- 30

-- 20

-- 10 t i I I I I I I I

-50 -40 -30 10 0 10 20 30 40 50 Axial P0wer Imualance, 5 POWER IMBALANCE ENVELOPE FCR OPERATION FROM 100 1 10.TO 246 1 10 EFP0 ,

TMl-1 CYCLE 3 Figure 3.5-2H 1469 243

Power, % of 2535 MWt RESTRICTED REGION

-31.21,102 100 '

-28.8,90 . 90 14.28,90

-- 80 PERMISSIBLE OPERATING

- - 70 REGION

_ 60

- 50

- 40

-- 30

- 20

- 10 I I I I I I I I t I

-50 -40 -30 -20 -10 0 10 20 30 40 50 Axial Power imbalance, 5 POWER IMBALANCE ENVELOPE FOR OPERATION AFTER 2461 10 EFPD TMI-l CYCLE 3 Figure 3.5-21 1469 244

21 ,

20 j 19 d

7 18 b

~

17 J

[ _\

S \

E 16

\

5

/

e 15 5

j 14 13 12 8 10 12 0 2 4 6 Axial location of Peak Power From Bottom of Core, f t LOCA LIMITED MAXIMUM ALLOWABLE tlNEAR HEAT RATE Figure 3.5-2J 1469 245

3

\

18.2.102 34.7.102 100 -

15.8.90 37.9.90 RESTRICTED 90 -

' REGION 37.9.80 80 -

6.5.80 45.2.70 70 s.s.70 0.70 C

60 -

45.2.so 100.60 2

N

50 _

" PERMISSIBLE

( f= 40 - OPERATING 2 REGION 30 -

20 -

10 -

i i t

, i e i i i

60 70 80 90 100 10 20 30 40 50 0

APSR 5 Witndrawn APSR POSITION LIMITS FOR OPERATION FROM 0 TO 100 1 10 EFP0 TMI-1 CYCLE 3 Figure 3.5 2K 1469 246

20.6.102 38.8.102 100 _

90 '**

• 8.9.90 RESTRICTED 80 - 6.5,80 44.6.60 REGION

$ 70 0,70 6.5,70 45.2.70 60 _

g 45.2,60 100.60 a

50 -

a 40 -

PERMISSIBLE 30 - OPERATING REGION 20 _

10 _

I t t i e  ! I i 0 10 20 30 40 50 60 70 80 90 100 APSR 5 Withdrawn APSR POSITION LIMITS FOR OPERATION FROM 100 1 10 TO 246 1 10 EFP0 TMI-I CYCLE 3

. Figure 3.5-2L

)h(30

6.5,102 49.0,102 100 -

- 2.5,90 RESTRICTED 90 REGION 80 2.5.sc 0.80 70 _

60 -

49.O,60 100.60 m

50 -

PERMISSIBLE o OPERATING 40 _ REGION J.

E e m 30 -

20 -

10 -

i i i i i r i i 0 10 20 30 40 50 60 70 80 90 100 APSR % Witnarawn APSR POSITION LIMITS FOR OPERATION AFTER 246 1 10 EFPD TMl-1 CYCLE 3 Figure 3.5-2M 1469 248