ML19210A515

Tech Spec Change Request 30 Supporting Licensee Request to Change DPR-50,App a Re Cycle 2 Operation.Certification of Svc Encl

ML19210A515
Person / Time
Site:	Three Mile Island
Issue date:	01/13/1976
From:	Arnold R METROPOLITAN EDISON CO.
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ML19210A510	List: ML19210A509 ML19210A515 ML19210A518
References
NUDOCS 7910300554
Download: ML19210A515 (32)

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PENNSYL E ;IA ELECTRIC COMPANY THREE MILE ISLAND NUCLEAR STATION LWIT 1 Operating License No. DPR-50 Docket No. 50-289 Technical Specification Change Request No.30 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 Nuclear Station Unit 1. As a part of this request, proposed replacement pages for Appendix A are also included.

FETROPOLITAN EDISON COMPANY By Vice President-Generation Sworn and subscribed to me this day of ,1976 Notary Public

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. s UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION IN THE MATTER OF DOCKET NO. 50-289 OPERATING LICENSE NO. DPR-50 METROPOLITAN EDISON COMPANY -

This is to certify that a copy of Technical Specification Change Request No. 30 to Appendix A of the Operating License for Three Mile Island Nuclear Station, Unit 1, dated January 13, 1976 and filed with the U.S. Nuclear Regulatory Commission January 13, 1976, has this 13th day January, 1976, been served on the chief executives of Londonderry Township, Dauphin County, Pennsylvania, and of Dauphin County, Pennsylvania, by deposit in the United States Mail, addressed as follows:

Mr. Weldon B. Arehart, Chairraan Mr. Charles P. Hoy, Chairman Board of Supervisors of Board of County Commissioners of Londonderry Township Dauphin County R.D. #1, Geyers Church Road Dauphin County Courthouse Middletown, Pennsylvania 17057 Harrisburg, Pennsylvania 17120 METROPOLITAN EDISON COMPANY n

By Vice President-Generation 1487 196

o s Metropolitan Edison Co. (Met-Ed)

Three Mile Island Nuclear Station Unit 1 (TMI-1)

Operating License No. DPR-50 Docket No. 50-289 TECHNICAL SPECIFICATION CHANGE REQUEST NO. (Proposed Change 47)

The licensee requests that the attached changed pages replace pages 2-1, 2-2, 2-3, 2-5, 2-6, 2-7, 2-9, 3-16, 3-34, 3-35, 3-36, figures 2.1-1, 2 & 3, figures 2.3-1 & 2, and figures 3.5-2A thru F of the existing Technical Specifications.

REASON FOR PROPOSED CHAPGE These changes to technical specifications are necessary to ensure safe operation of TMI-l at rated power of 2535 MWt for the duration of Cycle 2 and are based on a Cycle 1 burnup of 440 + 10 EFPD.

Changes to the present technical specifications are necessary as a result of: the effects of introducing 56 fresh batch 4 fuel assemblies combined with relocation of once burned Batch 2 and 3 fuel assemblies; use of the B&W-2 CHF correlation with a 95/95 confidence level, and extended pressure application to 1750 psi; use of a RC flow equal to 106.5% of Cycle 1 design flow; and ECCS Final Acceptance Criteria (FAC).

The 56 batch 4 fuel assemblies are not in general the technical specifi-cation limiting assemblies. Their presence combined with relocation of the once burned batch 2 and 3 assemblies produces a redistribution of fuel and assemblies which results in changed core physics and thermal-hydraulic calculations. Further burnup and the cycle 2 locations of the batch 2 and 3 assemblies results in these assemblies being the limiting assemblies thermally and mechanically. Other factors that were considered in the derivation of the Cycle 2 specification limits are the slight dif ferences between the new and once burned fuel assemblies. These minor differences are reduced active length, slightly higher pellet density, and improved flow characteristics for the new assemblies compared to the burned assemblies.

In addition to Fuel changes, the use of the BaW-2 CHF correlation combined with the assumed minimum Flow of 106.5% have had an influence on these proposed specifications. Use of this correlation and flow more realistically predict core performance but still provide conservative technical specifica-tion limits.

1kt-Ed submitted revised technical specifications based on FAC guidelines in our Technical Specification Change Request 17 (August 8,1975). Additional ECCS supporting information was provided in our letters of April 19, 1975, July 9,1975, July 15,1975, and October 23, 1975. The attached changed pages for TMI-l Cycle 2 operation include changes that were requested in Change Request 17 which apply to Cycle 2 operation. The number 17 beside the marginal bars indicates those changes that were requested in Change Request 17 and all other marginal bars indicate Cycle 2 changes. All appropriate cycle 2 Technical Specifications were developed based on FAC guidelines.

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e e SAFETY EVALUATION JUSTIFYING CHANGE The influence of the minor design changes of the batch 4 fuel assemblies (i.e. increased pellet density and reduced active length) compared to the batch 2 & 3 assemblies have been considered for Cycle 2 operation.

The effect on core flow distribution as a result of batch 4 assembly end fittings, and the absence of orifice rods in the core's periphery have also been considered. These proposed specifications were developed conservatively accounting for the above considerations for all applicable transient and steady state conditions.

In general the governing hydraulic, mechanical, thermal and nuclear parameters have been develo;ed using NRC accepted practices, models, and correlations.

Note, however, that the applicability of the B&W-2 correlation has been extended downward to 1750 psia. This extention is justified since this correlation produces conservative predictions of data in this range and use of this correlation provides conservative but realistic results.

The FSAR accidents wherein core design or fuel loading are important were considered for Cycle 2 operation. In all instances except for the LOCA, which is based on the FAC guidelines of 10 CFR 50.46 and 10 CFR 50 appendix K, the parameters for the Cycle 2 core are bounded by the FSAR analysis; consequently, the FSAR analyses are valid for Cycle 2. All other FSAR accidents are independent of core physics parameters and therefore also remain valid. The LOCA analysis has been reported in BAW-10103 Rev.1.

With our Change Request 17 and its supporting correspondence we have shown that the TMI-l case is at least as conservative as the analyses done in BAW-10103 Rev. 1.

Additional and more detailed information supporting the above safety evaluation and providing more detailed design information is provided in the attached Babcock and Wilcox Cycle 2 Reload Report.

Based on the above, it is concluded that this change does not increase the probability of occurrence or the severity of an accident. This change does not create the possibility of occurrence of an accident not previously analyzed. Therefore, this change does not represent undue risk to the health and safety of the public and continued operation of TMI . at rated power of 2535 MRt is justified.

1487 198

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2. SAFETY LIMITS AND LIMITMG SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS, REACTOR CORE Acplicability Applies to reactor thermal power, 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 combination of the reacter system pressure and coolant temperature shall not exceed the safety lLnit as defined by the locus of points established in Figure 2.1-1. If the actual pressure / temperature point is below and to the right of the line, the safety limit is exceeded.

2.1.2 The combination of reactor thermal power 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 power) shall not exceed the safety limit as defined by the locus of points (solid line) for the specified flow set forth in Figure 2.1-2. If the actual-reactor-thermal-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 fission product release, it is necessary to prevent overheating of the cladding under normal operating conditions. This is accomplished by operating within the nucleate boiling regime of heat transfer, wherein the heat transfer coefficient is large enough se that the clad surface temperature is only slightly greater than the coolant temperature. The upper boundary of the nucleate boiling regime is termed , departure f* a nucleate boiling (DNB) . At this point there is a sharp reduction of tne heat transfer coefficient, which would result in high cladding temperatures and the possibility of cladding failure.

Although DNB is not an observable parameter during reactor operation, the observab).e parameters of neutron power , reactor coolant flow, temperature, and pressure can be related to DNB through the use of the B&W-2 correlation. (1) the B&W-2 correlation has been developed to predict DNB and the location of l DUB for axially uniform and non-unifor= heat flux distributions. The local DNB ratio (DNBR), defined as the ratio of thc heat flux that would cause DNB at a particular core location to the actual heat flux, is indicative of the nargin to DNB. The minimum value of the DN BR, during steady-state operation, normal operational transients, and anticipated transients is limited to 1 3 A DNBR of 13 corresponds to a 95 percent probability at a 95 percent confidence level that DNB vill not occur; this is ccncidered 1487 19o9 2-1

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a conservative =argin to EUB 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 limit s . The difference in these two precsures 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 minimumDNBRof13ispredictedforthemaximumpossiblegher=alpower (112 percent) when the reactor coolant flov is 139.8 x 10+ 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 factors (2) with poten-tial fuel densification effects; N N N F = 2.67; F = 1 78; F = 1 50 q M z The 1.5 axial peaking factor associated with the cosine flux shape provides a lesser margin to a DNBR of 13 than the 17 axial peaking factor associated with a lover core flux distribution. For this reason the cosine flux shape and the associated F3 = 1.50 is more limiting and thus the more conservative assumption.

The 150 cosine axial flux shape in conjunction with FAH = 178 define the reference design peaking condition in tne 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 vill not violate the previously established design criteria on DNBR. The flux shapes examined include a vide range of positive and negative offset for steady state and transient conditions.

These design limit power peaking factors are the most restrictive calculated at full power for the range from all control rods fully withdrawn to maximum allovable control rod insertion, and form the core DUBR design basis.

The curves of Figure 2.1-2 are based on the more restrictive of two thermal li=its and include the effects of potential fuel densification:

a. The 13 DNBR limit produced by a nuclear power peaking factor of FN = 2.67 of the combination of the radial peak, axial peak, and p8sition of the axial peak that yields no less than a 13 DNBR.

b. The combination of radial and axial peak that prevents central fuel melting at the hot spot. The limit is 19.6 kW/ft.

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 peaking.

The specified flow rates for curves 1, 2, and 3 of Figure 2.1-2 correspond to the expected minimum flow rates with four pumps, three pumps, and one pump in each loop, respectively.

2-2

The curve of Figure 2.1-1 is the most restrictive of all possible ree.ctor coolant pump-maximum ther=al power ecmbinations shcvn in Figure 2.1-3 The curves of Figure 2.1-3 represent the conditicns at which a minimum DNBR of 13 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. l 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 minimum ENBR.

The DNBR at calculated by the B&W-2 correlation continually increases from the point of m:nimum DNBR, so that the exit DNBR is always higher and is a function c,f the pressure.

The maximum thermal pcVer for three pump operation la 86.7 percent due to a pcuer level trip produced by the flux-flow ratio (74.7 percent flov x 1.08 =

80.7 percent power) plus the maximum calibration and instrumentation errcr.

The maximum thermal power for other reactor coolant pump conditions is pro-duced in a similar manner.

For each curve of Figure 2.1-3, a pressure-temperature point above and to the left of the curve vould result in a DNBR greater than 1.3 or a local quality at the point of minimum DNBR less than 22 percent for that particular reactor coolant pu=p eituation. The 1.3 DNBR curve for four pump operation is more restrictive than any other reactor coolant pump situation because any pressure / temperature point above and to tne left of the four pump curve vill be above and to the left of the other curves.

REFERENCES (1) FSAR, Section 3.2.3 1.1 (2) FSAR, Section 3 2 3 1.1.c (3) FSAR, Section 3 2 3 1.1.k 01 2-3

8 t 2.3 LIMIfING SAFETY SYSTEM SETTINGS, PROTECTION INSTRUIENTATION Applicability Applies to instru=ents monitcring reacter pcVer, reactor pcVer inbalance, reactor coolant system pressure , reacter coolant outlet te=perature, flow, nu=ber of pu=ps in operation, and high reactor building pressure.

Objective To provide automatic protection action to prevent any cc=bination of process variables frem 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.

Eases The reacter protection system censists of four instrument channels to =enitor each of several selected plant ecnditienc which vill ecuse a reactor trip if any one of these conditicns deviates from a pre-selected operating range to the degree that a safety limit may be reac ed.

The trip setting limits for protection system instru=entatien are listed in Table 2 3-1. The safety analysis has been based upon these protection system instru=entation trip set points plus calibration and instrumentation errors.

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

During normal plant operation with all reactor coolant pu=ps cperating, 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 calibratien and instrument errors, the maxi =um actual power at which a trip would be actuated could be 112%, which is the value used in the safety analysis (1).

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a. Overpower trip based on flew and imbalance The pcVer level trip set point produced by the reacter coolant system flow is based on a power-to-ficw ratio which has been established to acconmodate the =ost severe thermal transient considered in the design, the loss-of-coolant flow accident from high pcver. Analysis has de=cnstrated that the specified pcVer to flev ratio is adequate to prevent a DNER of less than 13 should a lov flev conditien exist due to any =alfunction.

1487 202 2-5

The power level trip set point produced by the power-to-flow ratio provides both high power level and lov 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 p; er 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 Icv 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 in 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 74.7 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 49.2 percent or flow rate is h5.h percent and the power level is 49 percent.

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/ft limits or DNBR limits. The reactor power bubalance (power in the top half of 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 pover/ reactor power-imbalance boundaries by 1.08 percent for a one percent flow reduction. l

b. Pu=p monitors The redundant pump monitors prevent the minimum core DNBR 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 number of pumps in operation.

c. Reactor coolant system pressure Durir4 a startup accident from low pcver or a slow rod withdrawal from high power, the system high pressure trip set point is reached before the nuclear overpower trip set point. The trip setting limit shown in Figure 2 3-1 for high reactor coolant system pressure (2355 rsig) has been established to maintain the system pressure below the safety limit (2750 peig) for any design transient.

87 203 2-6

The icv pressure (1800 psig) and variable lov pressure (1175 Tout - 5103) trip setpoint shown in Figure 2 3-1 have been established to maintain the RTB ratio greater than cr equal to 1.3 for those design accidents that result in a pressure reduction (3,4).

Due to the calibration and instrumentation errers, the safety analysis used a variable low reactor coolant syste= pressure trip value of (11.75 Tout - 5103).

d. Coolant outlet temperature The high reacter coolant cutlet temperature trip setting li=1t (619 F) shcvn 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 RFS is 520 to 600 F. The trip setpoint of the channel is 619 F. Under the vorst case environment , pcVer cupply perturbaticns, and drift, the accuracy of the trip string is+1F. This accuracy was arrived at by su==ing the worst case accuracies of each =cdule. This is a conservative method of errer 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 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 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 operaticn within the calibrated range. Additional testing has demonstrated that in fact, the temperature channel is fully operational approximately 10% above the calibrated range.

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Since it has been established that the channel vill trip at a value of RC outlet temperature no higher than 620F even in the worst case, and since the channel is fully operational approximately 107, above the calibrated range and exhibits no hysteresis or foldover characteristics, it is concluded that the instrument design is accept able.

e. Reacter building pressure The high reactor building pressure trip setting limit (h psig) provides positive assurance that a reactor trip will occur in the unlikely event of a steam line failure in the reacter building or a loss-of-coolant accident, even in the absence of a low reacter coolant syste= pressure trip.

1487 204 2-7

TABLE 2.3-1 -

REACTOR PROTECTION SYSTEM TRIP SL'2 TING LIMITS One Reactor Coolant Pump Four Reactor Coolant Pumps Three Reactor Coolant Pumps Operating in Each Loop Shutdown Operating (Hominal Operating (Nominal (Nominal Operating Bypass Operating Power - 100%) Operating Power - 75%) Power 49%)

1. Nuclear power, Max.

% of rated power 105 5 105.5 105.5 5.0(3)

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) power

3. Nuclear power based (5) NA NA 91% Bypassed on pump monitors, max.

5 of rated power y h. High reactor coolant 2355 2355 2355 1720(h) do system pressure, psig, max.

5. Iov 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)(1) Bypassed coolant system

-* pressure, psig, min.

4 C33 7 Reactor coolant temp. 619 619 619 619

'%J F., Max.

N h 4 g 8. High Reactor Building h h m pressure , psig, max.

(1) Tout is in degrees Fahrenheit (F)

(2) Reactor coolant system flow, 5 (3) Administrative 1y controlled reduction set only during reactor shutdown (h) Automatically set when other segments of the RPS (as specified) are bypassed (5) The pump nonitors also produce a trip on: (a) loss of two reactor coolant pumps in one reactor coolant loop, and (b) loss of one or two reactor coolant pumps during two-pump operation.

3.1.7 MODERATOR TEMPERATURE COEFFICIEIIT OF REACTIVII?

Apolicability Applies to maximum positive moderator temperature coefficient of reactivity at full pcVer conditions.

Objective To assure that the moderator temperature coefficient stays within the limits calculated for safe operation of the reactor.

Specification 3.1.7.1 The moderator temperature coefficient shall not be positive at power levels above 95% of rated power.

Bases A non-positive moderator coefficient at power levels above 95% of rated pcVer is specified such that the maximum clad temperatures vill not exceed the Final 17 Acceptance Criteria based on LOCA analyses. Belov 95% of rated power the Final Acceptance Criteria vill ot be exceeded with a positive moderator temperature coefficient of +0.5 x 10- AK/K/F. All other accident analyses as reported in the FSAR have been performed for a range of moderator temperature coefficients including +0 5 x 10-h 6K/K/F.

The experkental value of the moderator coefficient will be corrected to obtain the hot full power moderator coefficient. The correction factor vill be verified during startup testing en earlier B&W reactors.

The Final Acceptance Criteria states that post-LOCA clad te=perature vill not 17 exceed 2200 F.

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REFEREITCES (1) FSAR, Section lb (2) FSAR, Section 3 1487 206 3-16

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

to the sa:e position. Normal operation of 100 percent of tee ther=al power allowable for the reactor coolant pu=p co=-

bination cay then continue provided that the rod that was declared inoperable is =aintained within allovable group average position limits in 3.5.2 5 3.5.2.3 The worth of single inserted control rods during criticality are limited 17 by the restrictions of Specification 3.1.3 5 and the Control Rod Position Limits defined in Specification 3.5.2.5 3.5 2.h quadrant tilt:

a. Except for physics tests if quadrant tilt exceeds h percent, power shall be reduced i==eidately 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 h percent tilt. For less than four pu=p operation, ther=al power shall be reduced 2 percent of the thermal power allovable for the reactor coolant purp combination for each 1 percent tilt in excess of h percent,

b. Within a period of k hours, the quadrant power tilt shall be reduced to less than h percent except for physics tests, or the following adjustments in setpoints and limits shall be made: ,

l. The protection systes reactor pover/ imbalance envelope trip setpoints shall be reduced 2 percent in power for each 1 percent tilt.

2. The control rod group withdrawal limits (Figures 3.5-2A 3.5-23, 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 h percent.

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 h percent.

1487 207 3-3h

c. If quadrant tilt is in excess of 25 percent, except for physics tests or diagnostic testing, the reactor vill be pliced in the hot shutdown condition. Diagnostic testing during power operation with a quadrant power tilt is permitted provided the ther=al power allevable for the reactor coolant pump combinations is restricted as stated in 3.5 2.h.a above.

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

148;7 208 3-3ha

3 5.2 5 control rod positions:

a. Operating red 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 /vithdrawal limits are specified on Figures 3 5-2A,3 5-2B, and 3 5-2C for four pu=p operation and Figures 3 5-2D, 3 5-2E, and 3 5-2F, for three or two pu=p operation. If the control rod position limits are exceeded, corrective measures shall be taken inmediately 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 th_e xenon reactivity is within 10 percent of the equilibrium value foroperation at rated power and asy=ptotically approaching stability.

d. Core imbalance shan be monitored on a tinicus frequency of once every two hours during power operation above 40 percent of rated power. Except for physics tests, corrective measures (reduction llT of libalance by ApSR movements and/or reduction in reactor pover) shall be taken to maintain operation within the envelope defined by Figures 3 5-2G, 3 5-2H and 3 5-2I. If the Dnbalance 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 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 sball be locked at all times with limited access to be authorized by the superintendent.

3 5 2.7 A power map shall be taken to verify the expected power distribution at periodic intervals of approximately 10 full power days using the incore instrunentation detection system.

Bases The pover-inbalance envelope defined in Figures 3.5-2G, 3 5-2H, and 3.5-2I is based on LOCA analyses which have defined the maximus linear heat rate (see Figure 3.5-2J) such that the r.axi=us clad tenperature vill not exceed the Final Acceptance Criteria (2200?) . Operation outside of the power idbalance envelope alone does not constitute a situation that vould cause~ the Final Acceptance Criteria to be exceeded should a LOCA occur. The power iibalance envelope represents the boundary of operation 3-35 1487 209

li=ited by the Final Acceptance Criteria only if the control rods are at the withdrawal / insertion li=its 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 4 percent quadrant power tilt exists. Additional conservatis= is introducted by application of:

a. Nuclear uncertainty factors.

b. Thermal calibration uncertainty.

c. Fuel densification effects,

d. Hot rod =anufacturing tolerance factors.

The Rod index versus Allowable Power curves of Figures 3 5-2A, 3 5-2B, 3.5-2C, 3 5-2D, 3 5-2E and 3 5-2F, describe three regions. Thece three regions are:

1. Permissible operating Region

2. Restricted Regions 3 Prohibited Region (Operation in this region is not allowed) 17 Note: Inadvertent operation within the Restricted Region for a period of k hours is not considersd_a violation of a lir.iting condition for operation. The limiting criteria within the Restricted Region are potential ejected rod voith and ECCS power peaking and since the probability of these accicents is very lov especially in a h hour time frate, inadvertant opt ration within the Restricted Region for a period of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is allowed.

1487 210 3-35a

The 25t5 percent overlap between successive control rod groups is allowed since the vorth of a rod is lover at the upper and lover part of the stroke. Control 17 rods are arranged in groups or banks defined as follows:

Groun Function 1 Safety 2 Safety 3 Safety b Garety 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 normal position at power is for groups 6 and 7 to be partially inserted.

The rod position limits are based on the cost liciting of the following three criteria: ECOS power peaking, shutdown margin, and potential ejected rod vorth.

As discussel above, compliance with the ECCS power peaking criterion is ensured 17 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 tite, assuming the highest verth control rod that is withdrawn re=ains in the full out position (1). The rod position limits also ensure that inserted rod groups vill 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 hypot hetical rod ejection accident.

A taxitus single inserted control rod worth of 1.0% Ak/k is allowed by the rod position limits at hot zero power. A single insert.d .

control rod worth 1.0% Ak/k at beginning of life, ho , zero power vould result in a lover transient peak thermal power and, therefore, less severe environmental consequences than 0.65% Ak/k ejected rod vorth at rated power.

The plant co=puter vill scan for tilt and imbalance and vill satisfy the technical specification requirements. If the computer is out of service, than canual calculation for tilt above J 5 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.- have been established within the thermal analysis design base using the definition of quadrant power tilt given in Technical Specifications, Section 1.6.

During the physics testing program, the high flux trip setpoints are ad=inistratively set as follows to assure an additional safety targin is provided:

Test Power Trin Eetroint 0 <5%

15 50%

h0 50%

5

>75 6l 1487 211 105.5%

REFEFE CES (1) FSAR, Section 3.2.2.1.2 (2) FSAR, Section Ih.2.2.2 3-36

2600 2400 ACCEPTABLE u,

OPERATION 2200 5  !

5 0

e U 2000

=

UNACCEPTABLE a OPERATION 1800 7

1600 560 580 600 620 640 660 Reactor Outlet Temperature, F CORE PROTECTION SAFETY LIMIT F i gu r e 2.1 -1 1487 212

Thermal Power Level, 'l UNACCEPTABLE OPERATION 120 --

(-19,i12) (112) (+18,i12)

ACCEPTABLE Kw/ft Limit 4 PUNP Kw/ft Limit 100 -- 0PERATl0N

(+4 5, 9 3 )

(-45.91)

(gg,7)

(-45.86.7) ( *

• ACCEPTABLE 2 80 -- 3 & 4 PUMP OPERATION

(-us.se.i) 60 __ (59.1)

(,4,,gg,,)

ACCEPTABLE 2,3 & 4 PUMP @

OPERATION 40 --

20 - -

l 1 I i

-60 -40 -20 0 +20 +40 +60 Reactor Power Imualance, %

CURVE REACTOR COOLANT FLOW (Ib/hr) 1 139.8 x 106 2 104.5 x 106 3 68.8 x 106 l4g7 pl}

CORE PROTECTION SAFETY LIMITS Figure 2.1-2

2600 2400 i

2h Y

.?

E. 2200 _

/

5 U

E

~

2000 b

8 f l l e

1800

///

1600 560 580 600 620 640 660 Reactor Outlet Temperature, F REACTOR COOLANT FLOW CURVE (LBS/HR) POWER PUMPS OPERATING (TYPE OF LIMIT) 1 139.8 x 106 (100%)* I I2P, Four Pumps (DNBR Limit) 2 104.5 x 106 (7k.7%) 86.75 Three Pumps (DNBR Limit) 3 68.8 x 106 i L9 2%) 59.15 One Pump in Each Loop (Quality Limit)

• 106.5% of Cycle 1 Design Flov.

CORE PROTECTION SAFETY BASES F i gu r e 2.1 -3 1487 214

2500 2300 P = 2355 PSIG T = 619 F ACCEPTABLE ea OPERATION 2100 i

G <e

\*

E 1900

=

ss E + UNACCEPTABLE O P = 1800 PSIG

• OPERATION a

t 3 1700 1500 540 560 580 600 620 640 Reactor Outlet Temperature, F PROTECTION SYSTEM MAXIMUM ALLOWABLE SET POINTS F i gu re 2.3-1 1487 215

Power Level, $

UNACCEPTABl.E 120--

OPERATION (108,0)

@ l10E #g

.,+,

1 ACCEPTABLE (80.7) 4 PUMP 8f OPERATION t

60- - ACCEPTABLE (53.1) 3 & 4 PUMP OPERATION 40- -

ACCEPTABLE 2,3 & 4 PUMP OPERATION

, 20-- , o ,

k Y 13 11 Il  !!

m m =

i i 1 lm i i i 40 -20 0 +20 +40 +60 Power Imualance, %

PROTECTION SYSTEM MAXIMUM ALLOWABLE SET POINTS Figure 2.3 2 1487 216

100 iss.9.102 196.4.102 POWER LEVEL 196.4.92 -CUT 0FF 168.9.85 207.85

RESTRICTED REGION RESTRICTED REGION me o 70 -

Q 126.7.67 N

60 -

Q 50 300.44.5 f 40 -

PERMISSIBLE 222.3.44.5 OPERATING 30 -

REGION 20 -

10 -

/

'O e e i e i e i e i e i i i 0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 Rod Index. 5 Withdrawn G 25 50 75 100 0 25 50 75 100 1 1 i t i i t t Group 5 Group 7 0 25 50 75 100 t t I t i

, Group 6 s

RCD POSITION LIMITS FOR 4 PUMP OPERATION APPLICABLE DURING THE PERIOD FROM 0 TO 152 ! 10 EFPO;

. CYCLE 2 Figure 3.5-2A 1487 217

OPERATION IN THIS REGION 175.7.102 202.I,102 IS NOT ALLOWED POWER LEVEL 90 -

202.i.92 CUT 0FF 80

- RESTRICTE0 SHUTOOWN MARGIN 826.s,73 REGION

- 70 -

E LIMIT e

60 36.50 300,ss

,. RESTRICTED PERMISSIBLE

" I"

! 40 -

OPERATING 222.3,us REGION 30 -

20 -

0.15 10 -

0 i i e i i i e i , i i i , ,

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 t t t t t t , g g Group 5 Group 7 0 25 50 75 100 t 1 1 1 1 Group 6 R00 POSITION LIMITS FOR 4 PUMP OPERATION APPLICABLE CURING THE PERIOD FROM 152 1 10 TO 275 1 10 EFPD; CYCLE 2 Figure 3.5-28 1487 218

is ,i 2 OPERATION IN THIS REGION 253,4,102 IS NOT ALLOWED POWhR LEV t 90 -

CUT 0FF 80 b 70 -

SHUTOOWN MARGIN g LIMIT

~

~ 60 o

50 -

sg,9,,7 E '

" 222.3.s7 40 -

PERMISSIBLE 30 OPERATING REGION 20 -

is. t. is 10 _

I I l f f I f f f I t t I t 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 I I t 1 I f I e e Group 5 Group 7 0 25 50 75 100 t t t l l Group 6 R00 POSITION LIMITS FOR 4 PUMP OPERATION APPLICABLE CURINri THE PERICO AFTER 275 1 10 EFP0; CYCLE 2 Figure 3.5-2C 1487 219

104,102 158.102 100 -

209.102 218.102 RESTRICTED REGION RESTRICTED RESTRICTED REGION 90 - FOR 2 AND 3 PUMP REGION FOR 2 AND 3 OPERATION R3 p p p p

12s.6.s5 s oo. a 2. 5 g 80 -

OPERATIO

222.3.a2.5 g

g 70 -

RESTRICTED REGION

$ FOR 3 PUMP 60 - OPERATION

$ PERMISSIBLE l 50 OPERATING 222.3,56.5 300.56.5 REGION m"

3 40 -

E.

3 30 -

[ 20 -

j 10 -

a.

[o. o i i i i i i i e e i i i 0 20 40 60 80 100 120 140 160 ISO 200 220 240 260 280 300 Rod Index, % Withdrawn 0 25 50 75 100 0 25 g g t I 50 75 100 t

f f f 1 Group 5 Group 7 0, 25

, 50, 7,5 100 Group 6 R00 POSITION LIMITS FOR 2 AND 3 PUMP OPERATION APPLICABLE DURING THE PERICO FROM 0 TO 152 1 10 EFP0; CYCLE 2 Figure 3.5-20 1487 220

v OPERATION IN THis REGION 100 -

i7s.io2 , 207.io2 2ia 'o2 15 NOT ALL0nEO RESTRICTED REGION 90 -

FOR 2 AND 3 PUWP OPERATION 3co.s5 222.e5 5

0 70 -

SHuiOOWN MARGIN ' RESTRICTED REGION g LIMIT FOR 3 PUMP PERYlS$1BLE

[ 60 -

OPERATING OPERATION I '" '"

REGION S _ 3s.\o 2

a

~

g 40 -

2 30 -

d. 20 -

f 0.l5 10 -

t i I I I I ' I I I I I ' '

0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 Rod Index, 5 Witnarann 0 25 50 15 100 0 25 50 75 100 t i f I I I '

Group 5 Group 7

,0 2,5 5,0 75 100 i i Group 6 R00 POSITION LIMITS FOR 2 AND 3 PUMP OPERATION APPLICABLE DURING THE PERIOD FROM 152 1 10 TO 275 1 10 EFP0; CYCLE 2 Figure 3.5-2E 1487 221

OPERATION IN THIS REGION 192.102 224.102 233.102 100 ~

I 15 NOT ALLOWED

+-- - -RESTRICTED c

o 90 - 159 87 REGION FOR 222.3.17 2 AND 3 PUMP S

g 80 -

/

[ OPERAil0N g RESTRICTED REGION a FOR 3 PUMP

~

$ OPERATION o SHL'TOOWN MARGIN 60 - Liggi

[o 98.60 222.3.60 3 50 -

j 68.9.47 PERMISSIBLE 3 40 -

OPERATING

~

o REGION e 30 -

N j 20 -

is.o.is 10 1 1 I I I I I I I I I I I I O 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 Rod index. 5 Witridrawn 0 25 50 75 100 0 25 50 75 100 I I t t I f 9 f Group 5 Group 7 0 25 50 75 100 t t f 1 9 Group 6 R00 .- .ITION LIMITS FOR 2 AND 3 PUMP OPERATION APPLICABLE DURING THE PER130 AFTER 275 1 10 EFPO; CYCLE 2 Figure 3.5-2F 1487 222

Power, % of 2535 MWt RESTRICTED REGION

-11.22,102 t1.22,102

-11.04,92 l 10.12,92

-14.45,85 13.6,85

- 80

-18.75,75 18.75,75

- 70

-- 60

-- 50 PERMISSIBLE OPERATING

-20.03,44.5 REGION - 40

- 30

-- 2C

-- 10 i i I i l I I I I 50 40 30 -20 -10 0 10 20 30 40 50 Axial P0wer Imbalance, %

OPERATIONAL POWER IMBALANCE ENVELOPE APPLICABLE TO OPERATION FROM 0 TO 152 1 10 EFPD CYCLE 2 Figure 3.5-2G 1487 223

Power, % of 2535 MWt RESTRICTED REGION

-16.32,102 7 11.22,102

-15.64,92 l 10.12,92 90

-15.3,85

-- 80

#5'#'

10 60 i

f

-- 50

-20.7,4s PERMISSIBLE OPERATING REGION - 40 30

-- 20

.. 10 l l I I i i I 3 50 -40 -30 20 -10 0 10 29 30 40 i0 Axial Power imbalance, %

OPERATIONAL POWER IMBALANCE ENVELOPE APPLICABLE TO OPERATION FROM 152 1 10 TO 275 10 EFPD, CYCLE 2 Figure 3.5-2H 1487 224

Power, 4 of 2535 MWt RESTRICTED REGION

-- 100

-22.54,92 13.8,92

-- 90 80

-- 70 15.7!,,63

-- 60 50

-22.56,47 PERMISSIBLE OPERATING --

40 REGION 30 20 10 1 I I I I I I I I I

-50 -40 -30 -20 -10 0 10 20 30 40 50 Axial Power Imbalance, %

OPERATIONAL POWER IMBALANCE ENVELOPE APPLICABLE TO OPERATION AFTER 275 1 10 EFPD, CYCLE 2 Figure 3.5-21 1487 225

21 20 19 -- - - - - -- ---

= l 18 I S

E 17 /

[\

e N__. i

= N f

16 l

, [-/ ,

-- V i

=

is L s 14

_5 j 13 12 0 2 4 6 8 10 12 Axial Location of Peak Power From Bottom of Core. ft LOCA LIMITED MAXIMUM ALLOWABLE LINEAR HEAT RATE Figure 3.5-2 J 1487 226