ML19312C331

Amends 63,63 & 60 to Licenses DPR-38,DPR-47 & DPR-55, Respectively,Changing Tech Specs Re Unit 3,Cycle 4 Operation at Full Power After Reload & Removal of Orifice Rod Assemblies from Core

ML19312C331
Person / Time
Site:	Oconee
Issue date:	07/06/1978
From:	Reid R Office of Nuclear Reactor Regulation
To:
Shared Package
ML19312C330	List: ML19312C329 ML19312C331 ML19312C342
References
NUDOCS 7912130861
Download: ML19312C331 (29)

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UNITED STATES

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W ASHINGTON, D. C. 20555 s, ~. f DUKE POWER COMPANY DOCKET N0. 50-269 OCONEE NUCLEAR STATION, UNIT !!0.1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.

63 License No. OPR-38 1.

The Nuclear Regulatory Comission (the Comission) has found that:

A.

The application for amendment by Duke Power Company (the licensee) dated May 30, 1978, as supplemented June 14, 23 and 28,1978, complies with the standards and requirements of the Atcmic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in confomity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.

There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations;.

D.

The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied, t

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Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Facility Operating License No. DPR-38 is hereby amended to read as follows:

(2) Technical Specifications The Technical Scecifications contained in Appendices A and B, as revised through Amendment No. 63, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULA MMISSION k'V-Robert W. Reid, Chief Operating Reactors Branch #4 Division of Operating Reactors

Attachment:

Changes to the Technical Specifications Date of Issuance:

July 6, 1978 i

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UNITED STATES NUCLEAR REGULATORY COMMisslON y*(

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gv.....f DUKE POWER COMPAtlY DOCKET NO. 50-270 OCONEE NUCLEAR STATION, UNIT NO. 2 AMEN 0 MENT TO FACILITY OPERATING LICENSE Amendment No. 63 License No. DPR-47 1.

The Nuclear Regulatory Comission (the Comission) has found that:

A.

The application for amendment by Duke Power Company (the licensee) dated fiay 30, 1978, as supplemented June 14, 23 and 28,1978, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in confomity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.

There is reasorable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.

The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part l

51 of the Comission's regulations and all applicable requirements have been satisfied.

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Accordingly, the license is amended by changes 'o the Technical t

2.

Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Facility Operating License No. DPR-47 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 61 are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATO r0MMISSION g

Robert W. Reid, Chief Operating Reactors Branch #4 Division of Operating Reactors Attachnent:

Changes to the Technical Specifications Date of Issuance: July 6,1978

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DUKE POWER COMPANY DOCKET N0. 50-287 OC0 flee tlUCLEAR STATI0ti, UtlIT ?!0. 3 AMENDMEllT TO FACILITY OPEPATItiG LICENSE Amendment No. 60 License tio. DPR-55 1.

The Nuclear Regulatory Comission (the Comission) has f ound that:

A.

The application for amendment by Duke Power Company (the licensee) dated fiay 30, 1978, as supplemented June 14, 23 and 28,1978, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in confomity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in. compliance with the Comission's regulations; D.

The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

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Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Facility Operating License No. DPR-55 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 60, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of itt.

issuance.

FOR THE NUCLEAR REGULATORY COMMISSION s

Mb/e Robert W. Reid, Chief Operating Reactors Branch #4 Division of Operating Reactors

Attachment:

Changes to the Technical Specifications Date of Issuance:

July 6, 1978

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t ATTACHitENT TO LICENSE AltENDfiENTS A!!ENDMENT N0. 63 TO DPR-38 AMENDMENT NO. 63 TO DPR-47 AliENDMENT NO. 60 TO DPR-55 00CKETS NOS. 50-269, 50-270 AND 50-287 L

Revise Appendix A as follows:

Remove the following pages and insert revised identically numbered pages:

2.1-3c & 2.1-3d 2.3-3 3.2.1 & 3.2-2 3.5 3.5-11a*

3.5-16 & 3.5-16a i

3.5-17 3.5 3.5-20b 3.5 3.5-23b 3.5-23i - 3.5-23k Delete the following page:

3.2-la

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Changes on the revised pages are indicated by marginal lines.

• New page

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, Bases - Unit 3 The safety limits presented for Oconee Unit 3 have been generated using 3AW-2 critical heat flux correlation (l) and :he Reactor Coolant System flow rate of 6 lbs/hr for four-pump operation).

106.5 percent of the design flow (131.32 x 10 The flev rate utiliced is conservative cecpared :o the actual =easured flow rate.(2)

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 transf er, wherein the hea: transfer coefficient is large enough so that the clad surface te=perature is only slightly greater than the coolant te=perature.

The upper boundary of the nucleate boiling regime is ter=ed " departure fr:m nuclea:e boiling" (DN3).

At this point, there is a sharp reduction of the hea: ::ansfer coef ficient, which would result in high cladding :a=peratures and the possibill:7 of cladding f ailure.

Although DN3 is not an observable parameter during reactor operation, the observable parameters of neutron power, reactor coolan: flow, te=perature, and pressure can be related :o DN3 through the use of the 3AW-2 correlation (l).

The 3AW-2 correlatica has been developed to predic: DN3 and the location of DN3 for axially uniform and non-unifors heat flux distributions.

The local DNB ra:io (DNBR), defined as :he ratio of the heat flux that wculd cause DN3 at a particular core location to the ac:ual heat flux, is indicative of the margin to DN3.

'"he sini=um value of the DN3R, during steady-state operation, normal operational transients, and anticipated transien:s is li=1ced to 1.30.

A DN3R of 1.30 correspends to a 95 percent probabili:y at a 95 percent confi-dance level that DN3 will not occur; this is considered a conservative =argin to DN3 for all operating conditions.

The difference between the actual core outlet pressure and the indicated reactor coolant system pressure has been considered in deter =ining the core protection safety li=1:s.

The difference in chese two pressuras is nenisally 45 psi; hcvever, only a 30 psi drop was assumed in reducing the pressure trip se:poin:s to correspond to the elevated location where the pressure is ac:ually neasured.

The curve presented in Figure 2.1-lc represents the conditions at which a DN31 of 1.30 is predicted for the =aximum possible :her=al power sw w (112 percent) when four reactor coolant pumps are cperacing (21=i=us reactor coolant f1:v is 129.36x 106 lbs/hr.).

This curve is based on the following nuclear power peaking fac: ors vi:h poten:ial fuel densifica 1ca and fuel rod O

N

= 2.265; e

= 1. 71 ); 5

= 3.50.

The design peaking I

bowing eff ec:s: _N N

I q

aH z

combination results in a sore ccuserrative DNBR than any other power shape that exists during nor=al operation.

The curves of Figure 2.1-2C are based en the more restrie:ive of evo thermal 11=1:s t=d include :he eff ects af potential fuel densification and fuel rod bowing.

2:565'or l

1.

The 1.30 DN3R 11=1: produced by a nuclear peaking factor of Fg

=

the cembinatics of the radial peak, axial peak and posi:1cn or the axial peak that yields no less than a 1.30 DN3R.

Amendrents Mos. 63, 63, and 60 2.1-3c 4

2.

The combination of radial and axial psak that causas etntral fuel =2iting at the hot spot.

The limit is 20.15 kw/ f t for Uni: 3.

6 Power peaking is not a directly observable quanti:y, and, therefore, ILsits have been established on the bases of the reactor power 1.-balance produced by the power peaking.

The specified flow rates for Curves 1, 2 and 3 of yigure 2.1-2C correspond to the expected minimum flow rates wi:5 four pc=ps, three pumps and one pu=p in each loop. respec:1vely.

The max 1=um thermal power for :hree-pu=p opera:1on is 35. 3 percent due :o a power level trip produced by :he flux-flow ratio 74.7 percen: flow x l.055=

error.

The 78.8 percent power plus the =ax1=um calibration and instru=ent saximum ther=al power f or other coolant pump conditions are produced in a similar =anner.

For each curve of Figure 2.1-3C a pressure-temperature point above and to the lef t of the curve would result in a DN3R grea:er than 1.30 or a local quality at the point of mini =um DN3R less than 22 percent for that par:1cular reac:or coolant pump situation. The rurve of Figure 2.1-1C is the most restrictive of all possible reae:or coolant pu=p-maxinun thermal power combinations shnen in Figure 2.1-3C.

References (1)

Correlation of Cri:1 cal Hea: Flux in a Sundle Cooled by Pressuri:ed Water, 3AW-10000, March 1970.

(2)

Oconee 3, Cycle 3 - Reload Repor: - 3 AW-14 53, August, 1977.

(3)

A=end=en: 1 - Oconee 3, Cycle 4 - Reload Report - 3AW-1436, June 12, 1973.

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emendments Nos. $3,63,and 60 2.1-3d 1

Icval' trip cod arsociated rarctor power /rcactor povar-imbalanca boundaries by 1.035% for 1: flow reduction.

Puco Monitors The pump onitors prevent the =inimum core DN3R from decreasing belov 1.3 by tripping the reactor due to the loss of reactor coolant pu=p(s).

The circuitry monitoring pump operational status provides redundant trip protection for DN3 by tripping the reactor on a signal diverse from that of the power-to-flow ratio.

The pump sanitors also restrict the power level for the number of pumps in operation.

The reactor trip upon loss of one pump during 4-pump oper-ation above SO: FP is specified for Unit 1 in order to provide a minimum of 11.2% DNBR =argin in the flux / flow trip setpoint to accommodate the possible reduction in thermal =argia due to rod bowing.

For unit 2, loss of one pu=p l

trip is not required because of thermal credits from excess RC flow, i.e., by saintaining a ainimum RC flow of 109.5%.

For unit 3, the required DNER =argin for rod bowing is included in the analysis of the flux / flow trip setpoint.

Reactor Coolant Svstem Pressure During a startup accident from low power or a slow rod withdrawal from high power, the system high pressure set point is reached before the nuclear over-power trip set point.

The trip setting limit shown in Figure 2.3-LA - Unit 1 2.3 Unit 2 2.3-1C - Unit 3 for high reactor coolant system pressure (2355 psig) has been established to maintain the system pressure below the safety limit (2750 psig) for any design transient.

(1)

The low pressure (1800) psig and variable low pressure (11.14 T(11.14 T

".-4 706) trip

' 4706)

(1800) psig (1800) psig (11.14 T "*-4706) 0 out setpoints shown in Figure 2.3-lA have been established to maintain the DNB 2.3-13 2.3-1C ratio greater than or equal to 1.3 for those design accidents that result in a pressure reduction.

(2,3)

Due to the calibration aid instrumentation errors the safety analysis used a variable low reactor coolant ' system pressure trip value of (11.14 T

- 4746)

(11.14 I " - 47464 (11.14 I "* - 4746) out Coolant Outlet Temnerature The high reactor coolant outlet temperature trip setting limit (619 F) shown in Figure 2.3-lA has been established to prevent excessive core coolant 2.3-13 2.3-1C temperatures in the operating range.

Due to calibration and instru=entation errors, the safety analysis used a trip setpoint of 620 F.

l Reactor Buildine Pressure 1

The high reactor building pressure trip setting limit (4 psig) prevides positive l

assurance that a reactor trip will occur in the unlikely event of a loss-of-coolant accident, even in the absence of a low reactor coolant system pressure trip.

2.3-3 Amendments Nos. 63, 63, and 60

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HIGH PRESSURE INJECTION AND CHEMICAL ADDITION SYSTEMS 3.2 Applicability Applies to the high pressure injection and the chemical addition syste.ms.

Objective To provide for adequate boration under all operating conditions to assure ability to bring the reactor to a cold shutdown condition.

Soecification The reactor shall not be critical unless the following conditions are met:

Two high pressure injection pumps per unit are operable except 3.2.1 as specified in 3.3.

One source per unit of concentrated soluble boric acid in addi-3.2.2 tion to the borated water storage tank is available and operable.

This source will be the concentrated boric acid storage tank 3

containing at least the equivalent of 980 ft of 8700, ppm boron as boric acid solution with a temperature at least 10 F above the crystallization temperature.

System piping and valves necessary to establish a flow path from the tank to the high pressure injection system shall be operable and shall have the same temperature requirement as the concentrated boric c,cid storage tank.

At least one channel of heat tracing capable of shall be in operation.

meeting the above temperature requirement One associated boric acid pump shall be operable.

If the concentrated boric acid storage tank with its associated flowpath is unavailable, but the borated water storage tank is available and operable, the concentrated boric acid storage tank shall be restored to operability within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or the reactor shall be placed in a hot shutdown condition and be borated to a shutdown margin equivalent to 17; ak/k at 200 F.within the next twelve hours; if the concentrated boric acid storage tank has not been restored to operability within the next 7 days the reactor shall be placed in a cold shutdown condition within an additional 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

the If the concentrated boric acid storage tank is available but the borated water storage tank is neither available nor operable, borated water storage tank shall be restored to operability within one hour or the reactor shall be placed in a hot shutdown condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in a cold shutdown condition within an additional 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

FD 'T h h' D"D (

A k!nks AmendmentsNos. 63, 63, and 60 g

3.2-1

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Bases The high pressure injection system and chemical additien system provide control of the reactor coolant system boron concentration. (1)

This is normally accomplished by using any of the three high pressure injection pumps in series with a boric acid pump associated with either the boric acid mix tank or the concentrated boric acid storage tank.

An alternate I

method of boration will be the use of the high pressure injection pumps taking suction directly from the borated water storage tank. (2)

The quantity of boric acid in storage in the concentrated boric acid storage tank or the borated water storage tank is sufficient to borate the reactor coolant system to a 1% Sk/k suberitical margin at cold conditions (70 F) with the maximum worth stuck rod and no credit for xenon at the worst time in core life.

The current cycles for each unit, Oconee 1 Cycle 4, Oconee 2 Cycle 3, and Oconee 3 Cycle 4 were analyzed with the most limiting case selected as the basis for all three units.

Since only the present cycles were analyzed, the specifications will be re-evaluated with each reload.

A minimum of 980 ft3 of 8,700 ppm boric acid in the concentrated boric acid storage tank, or a minimum of 350,000 gallons of 1800 ppm boric acid in the borated water storage tank (3) will satisfy the requirements. The volume requirements include a 10% margin and in addition allow for a deviation of 10 EFPD in the cycle length. The specification assures that two supplies are available whenever the reactor is critical so that a single failure will not prevent boration to a cold condition.

The required amount of boric acid can be added in several ways.

Using only one 10 gpm boric acid pump taking suction from the concentrated boric acid storage tank would require approximately 12.25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> to inject the required boron.

An alternate method of addition is to inject boric acid from the borated water storage tank using the makeup pumps.

The required boric acid can be in-jected in less than six hours using only one of the makeup pumps.

The concentration of boron in the concentrated boric acid secrrge tank may be higher than the concentration which would crystallize at ambic.: conditions.

For this reason and to assure a flow of boric acid is available when needed, these tanks and their associated piping will be kept at least 10 F above the crystallization temperature for the concentration present.

The boric acid concen-tration of 8,700 ppm in the concentrated boric acid storage tank corresponds to a crystallization temperature of 77 F and therefore a temperature require-ment of 37 F.

Once in the high pressure injection system, the concentrate is sufficiently well mixed and diluted so that normal system temperatures assure boric acid solubility.

REFERENCES (1)

FSA.1, Section 9.1; 9.2 D*yD

• ]D ~3h ev Ju. eh adbu 22 (2)

FSAR, Figure 6.2 g; e5 (3)

Technical Specification 3.3 Amendments Nos. 63, 63, & 60 3.2-2

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If within onn (1) hour of dstermination of an inoperable rod, g.

it is not determined that a 1%3k/k hot shutdown margin exists combining the worth of the inoperable red with each of the other rods, the reactor shall be brought to the hot standby condition f

until this =argin is established.

h.

Following the determination of an inoperable rod, all rods shall be exercised within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and exercised weekly until the rod problem is solved.

1.

If a control rod in the regulating or safety rod groups is declared

.i inoperable, power shall be reduced to 60 percent of the thermal power allowable for the reactor coolant pump combination.

l j.

If a control rod in the regulating or axial power shaping groups is declared inoperable, operation above 60 percent of rated power may continue provided the rods in the group are positioned such that the rod that was declared inoperable is maintained with-in allowable group average position limits of Specification 3.5.2.2.a and the withdrawal limits of Specification 3.5.2.5.c.

3.5.2.3 The worths of single inserted control rods during criticality are li=ited by the restrictions of Specification 3.1.3.5 and the control rod position limits defined in Specification 3.5.2.5.

3.5.2.4 Quadrant Power Tilt Except for physics tests, the maximum positive quadrant power a.

tilt shall not exceed the Steady State Limit of Table 3.5-1 during power operation above 15% Iull power.

4 b.

If the maximum positive quadrant power tilt exceeds the Steady State Limit but is less than or equal to the Transient Limit of Table 3.5-1, then:

1.

Either the quadrant power tilt shall be reduced within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to within its Steady State Limit or 2.

The reactor thermal power shall be reduced below the power level cutoff (as specified in Specification 3.5.2.5) and further reduced 2% thermal power for each 1% of quadrant power tilt in excess of the Steady State Limit, and the Nuclear Overpower Trip Setpoints, based on flux and flux /

flow inbalance, shall be reduced within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by 2%

thermal power for each 1% tilt in excess of the Steady State Limit.

If less than four reactor coolant pumps are in operation, the allowable thermal power for the reactor coolant pump combination shall be reduced by 2% for each 1% excess tilt.

Am endments Nos. 63, 63, & 60 3,5-7

Qundrtnt.awar tilt shall ba reduced with. 24 hsurs to with-c.

in its Steady Stata Limit or, 1.

The reactor thermal power shall be reduced within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to less than 60% of the allowable pcwer for the reactor coolant pump ccmbination and the Nuclear Overpower Trip Setpoints, based on flux and flux / flow imbalance, shall be reduced within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to 65.5% of the thermal power value allowable for the reactor coolant pump combina-tion.

d.

If the quadrant power tilt exceeds the Transient Limit but is less than the Maximum Limit of Table 3.5-1 and if there is a simultaneous indication of a misaligned control rod then:

1.

Reactor thermal power shall be reduced within 30 minutes at least 2% for each 1" of the quadrant power tilt in ex-cess of the Steady State Limit.

2.

Either quadrant power tilt shall be reduced within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to within its Transient Limit or, 3.

The reactor thermal power shall be reduced within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to less than 60% of the allowable power for the reactor coolant pump combination and the Nuclear Overpower Trip Setpoints, based on flux and flux / flow imbalance, shall be reduced within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to 65.5% of the thermal power value allowable for the reactor coolant pump combination.

If the quadrant power tilt exceeds the Transient Limit but is e.

less than the Maximum Limit of Table 3.5-1, due to causes other than simultaneous indication of a misaligned control rod then:

1.

Reactor ther=al power shall be reduced within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to less than 60% of the allowable power for the reactor coolant pump combination and the Nuclear Overpower Trip Setpoints, based on flux and flux / flow imbalance, shall be reduced within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to 65.5% of the thermal power value allowable for the reactor coolant pump com-

bination, f.

If the maximum positive quadrant power tilt exceeds the Maximum Limit of Table 3.5-1, the reactor shril be shut down within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Subsequent reactor operatica is permitted for the purpose of measurement, testing, and corrective action provided the thermal power and the Nuclear Overpower Trip Setpoints allowable for the reactor coolant pump combination are restricted by a reduction of 2% of ther=al power for each 1% tilt for the max-tmum tilt observed prior to shutdown.

g.

Quadrant power tilt shall be monitored on a mini =um frequency of once every 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> during power operation above 15% full power.

Amendments Nos. 63, 63, s 60 3.5-8

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3.5.2.'

. Rod Positions

hnical Specification 3.1.3.5 does not prohibit the exercising-individual safety rods as required. by Table 4.1-2 or apply to-operable safety rod limits in Technical Specification 3.5.2.2.

cept for physics tests, operating red group overlap shall be 25%

5% between two sequential groups.

If this limit is exceeded, arrective measures shall be taken i=nediately to achieve an accept-aole overlap.

Acceptable overlap shall be attained within two hours or the reactor shall be placed in a hot shutdown condition within an additional 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, Position limits are specified for regulating and axial power shap-c.

ing control reds.

Except for physics tests or exercising control rods, the regulating control rod insertion / withdrawal limits are specified on figures 3.5.2-1A1 and 3.5.2-1A2 (Unit 1); 3.5.2-131, 3.5.2-132 and 3.5.2-133 (Unit 2); 3.5.2-1C1, 3.5.2-1C2 and 3.5.2-1C3 (Unit 3) for four pump operation,'and on figures 3.5.2-2A1 and 3.5.2-2A2 (Unit 1); 3.5.3-231, 3.5.2-232 and 3.5.2-233 (Unit 2);

3.5.2-2C1, 3.5.2-2C2 and 3.5.2-2C3 (Unit 3) fo two or three pump operation.

Also, excepting physics tests or exercising control rods, the axial power shaping control rod insertion / withdrawal limits are specified on figures 3.5.2-4A1, and 3.5.2-4A2 (Unit 1); 3.5.2-431, 3.5.2-4B2, and 3.5.2-433 (Unit 2); 3.5.2-4C1, 3.5.2-4C2, and 3.5.2-4C3 (Unit 3).

If the control rod position limits are exceeded, corrective measures shall be taken immediately to achieve an acceptable control rod position.

within two hours.An acceptable control rod position shall then be attained cation 3.5.2.1 shall be maintained at all times,The minimum shutdown d.

Except for physics tests, power shall not be increased above the power level cutoff as shown on Figures 3.5.2-1A1, and 3.5.2-1A2 (Unit 3.5.2-1C2, 3.5.2-1C3 (Unit 3), unless the following req are met:

(1)

The xenon reactivity shall be within 10 percent of the value for operation at steady-state rated power.

(2)

The xenon reactivity worth has passed its final maximum or minimum peak during its approach to its equilibrium valve for i

operation at the power level cutoff.

3.5.2.6 Reactor power imbalance shall be monitored on a frequency not to exceed two hours during power operation above 40 percent rated power for physics tests, imbalance shall be maintained within the envelope Execpt defined by Figures 3.5.2-3A1, 3.5.2-331, 3.5.2-332, 3.5.2-333, 3.5.2-3C1 3.5.2-3C2, and 3.5.2-3C3.

If the imbalance is not within the envelope defined by these figuras, corrective measures shall be taken to achieve an acceptable imbalance.

If an acceptable imbalance is not achieved within two hours, reactor power shall be reduced until imbalance limits are met.

3.5-9 Amtndments Nos. 63, 63, & S0 I-

~

s O

/

is allowed by ths rod position limits at hot zsro pow 2r.

A single inserted in control rod worth of 1.0%ak/k at beginning-of-life, hot zero pow r would result a lower transient peak. thermal power and, therefore, less severe environmental consequences than a 0.65%Ak/k ejected rod worth at rated pcwer.

Control red groups are withdrawn in sequence beginning with Group 1.

Groups 5,6, The normal position at power is for Groups 6 and 7 are overlapped 25 percent.

and 7 to be partially inserted.

The quadrant power tilt limits set forth in Specification 3.5.2.4 have been established to prevent the linear heat rate peaking increase associated with a positive quadrant power tilt during normal power operation from exceeding 5.10% f or Unit 1.

The limits shown in Specification 3.5.2.4 5.10% for Unit 2 l

7.50% for Unit 3 are measurement system independent. The actual operating limits, with the for each appropriate allowance for observability and instrumentation errors, measurement system are defined in the station operating procedures.

tilt and axial imbalance monitoring in Specification 3.5.2.4 and The quadrant The 3.5.2.6, respectively, nor= ally will be performed in the process computer.

two-hour frequency for monitoring these quantities will provide adequate surveillance when the computer is out of service.

Allowance is provided for withdrawal limits and reactor power imbalance limits to be exceeded for a period of two hours without specification violation.

Acceptable red positions and imbalance must be achieved within the two-hour time period or appropriate action such as a reduction of power taken.

Operating restrictions are included in Technical Specification 3.5.2.5d to prevent excessive power peaking by transient xenon.

The xenon reactivity must be beyond the its final maximum or mini =um peak and approaching its equilibrium value at power level cutoff.

REFERENCES 1 FSAR, Section 3.2.2.1. 2 FSAR, Section 14.2.2.2 FSAR, SUPPLEMENT 9 B&W FUEL DENSIFICATION REPORT BAW-1409 (UNIT 1) 3AW-1396 (UNIT 2)

BAW-1400 (UNIT 3)

'Cconee 1, Cycle 4 - Reload Report - 3AW-1447, March 1977, Section 7.11.

s 3.5-11 Amendments Nos. 63, 63,& 60

m n

.)

p TABLE 3.5-1

.Ouadrant Power Tilt Limits Steady State Transient Maximum I

Limit Limit Limit i-Unit 1 3.41 9.44 20.0 Unit 2 3.41 9.44 20.0 I

Unit 3 5.00 9.44 20.0 i

4 t

4 i

t i

5 A

Amendments Nos. 63, 63, & 60 3.5-11a

~

e

,/

110 (82.1.102; (174.1,102) o

,,(212.S,102) 100 RESTRICTED GFERAll0NS REGION sc NOT (174.1. 00)

( 212. G. 9 0 )

REtiRICTED ALLCHE0 REGION POWER (251.4,80) 80 (161.2,80)

LEVEL CUTOFF SHUT 00KN 70 CRGIN (151.4,70)

=

(300,70)

=R LIMIT 60 o

s.

3 50 2

(23,50)

PERMisSISLE

~

OPERAilNG REG 10N 30 20 w

0,15) 10 0

('N i

t t

0 20 40 60 80 100 120 140 160 100 200 220 240 260 250 300 Re: Index, 5 Witnarawn r

I f

e 0

25 50 75 100 0

25 50 75 100 Grcup 5 Grcup 7 f

e i

0 25 50 75 100 Group S ROD POSITION LIMITS FOR FOUR-PUMP OPERATION FROM 0 TO 100 + 10 EFPD OCONEE 3 a,ositreatt OCONEE NUCLEAR STATION Y'

Amendments flos. 63, 63, & 60 Figure 3.5.2-1C1 l

3.5-16

~'

'110 (164.9,102)

(174.1.102)

(232,102) 0 0

RESTRICTED 100 REGION OPERATIGN IN THl$

(174,j,gg)

(232,gg) 90 REGION IS NOT ALLONE3 POWER LEVEL CUTOFF 80

[*

(151.2.50) g 70 8

(

SHUT 00hN PARGIN 50 llHis

~

50 (90,50)

PERMISSIBLE CPERATING REGION a

40 e

30 20

-( 0.13)

(30,15) 10 RESTRICTED REGION (I '

i t

i e

i i

t 0

0 20 40 60 80 100 120 140 ISO ISO 200 220 240 250 290 300 Rea Inder, 5 Witnerawn i

I I

t i

f f

e f

0 25 50 75 100 0

25 50 75 100 Group 5 Group 7 i

i i

0 25 50 75 100 Group G ROD POSITION LIMITS FOR FOUR-PUMP OPERATION FROM 100 + 10 TO 235 + 10 EFPD OCONEE 3 -

b\\

bat nau, OCONEE NUCLEAR STATION D'

Amendments flos. 63, 63, & 60 Figure 3.5.2-1C2 3.5-16a

r 110 (191.2,102)

(251.4,102) 100 OPERATION IN THIS RE0!ON NOT ALLONE0 RESTRICTED

'(251.4,90)

SG REGION POWER LEVEL BD CUTOFF (241.8,80)

SHuiOONN MARGIN EO

'o (110,50) 50 PERMISSISLE OPERATINO j

43 REGION 30 20 (0,7) 10 (0,0) 0 0

20 40 60 80 100 120 140 160 180 200 220 240 260 230 300 Roc Index, ; Witncrann I

f I

f f

f f

I t

0 25 50 75 100 0

25 50 15 100 Group 5 Group 7 i

t i

0 25 50 75 100 Grcup 6 i

R00 POSITION LIMITS FOR FCUR-PUMP OPERATION l

AFTER 235 3,10 EFPD OCONEE 3

\\

,opi nan; OCONEE NUCLEAR STATION

'M Figure 3.5.2-Ic3 A=endments Nos. 63, 63, & 60 3.5-17

m 110.

~

(

}

(

OPERATION IN THIS RE310N NOT (209.0,102) 100 ALLONEO RESir.lCTED RESTRICTED FOR 3 PU!'?S HEM ON FOR 5

SO (151.4.03.7)

(300,93.7) 3 PU','P SHUTOCNN OPER.

b 60

- ERGIN

• g limit u

{

70

- (50,70)

[

PERulSSIBLE OPERATING

[

60 REGION (20,50)

.g 50 7

2 40 O

30 RESTRICTED FOR 3 PUEP OPERATION 2.

20 (0,15) 10 (0,0) 0 t

0 20 40 00 80 100 120 140 150 180 200 220 240 250 280 300 Rcd inaet, 5 witnerann i

f f

f e

e t

I t

0 25 50 75 100 0

25 50 75 100 Grcup 5 Group 7 f

?

9 I

f 0

25 50 75 100 Grcup 6 ROD POSITION LIMITS FOR TVO & THREE-PUMP OPERATION FROM 0 TO 100 + 10 EFPD

~~

OCONEE 3 lsuusan; OCONEE NUCLEAR STATION i

l W~

l Figure 3.5.2-2Cl Amendments Nos. 63, 63, 3 60 3.5-20

^

110 (202.5.102)

(164,5,102)

RESTRICTED FOR 3

~

OPER*il0N IN THis REGICN PUYP OPERifl0H E

IS t.0T ALLONEC (300,S3.7)

C 90 2

a

!! 80 u

?

O ShUIO0%N MARGIN Ll'!!T -

PERMISSIBLE OPERAT!N3 REGION 3 50 2

3 50 (30,50) 2 40 E

,- 30 E

20

-( 0,13 )

(30,15)

RESTRICTED FOR 2L3 PUMP QPERATION 10

- - =

(0,0)

O e

i e

i i

0 20 40 50 80 100 120 140 150 180 200 220 240 250 250 300 R:3 lacer, 5 #itnerann v

e

+

i i

0 25 50 75 100 0

25 50 75 100.

Group 5 Group 7 i

i i

0 25 50 75 100 Group $

R00 POSITION LIMITS FOR TWO & THREE-PUMP OPERATION FROM 100 + 10 to 235 + 10 EFPD OCONEE 3 b\\

,FUcata; OCONEE NUCLEAR STATIOf1 D

Figure 3.5.2-2c2 Amendments Nos. 63, 63, & 60 3.5-20a

e m,

! !,3 REST!;tCTE0 FC2 2 E3 PU!.'P CFEttAil0N

g _

(151.2.102) /

(213.5.102)

(130.5.93.7)

[o $3 2

f50 CPERAT10!; !N TH!S REGICM IS NOT ALLG#EC 2s 10 c

M 0

~

FERM105 TELE j

OPERATING o

j 0

gE;;g#

(110,50) 3 40 C

30 Y

E 20 (48,15)

(C 71 10 RESTRitiED FOR 2 L 3

~

(0,0)

PUMP CPERAil0N 0

i t

e e

0 20 40 60 B0 100 120 140 150 180 20G 220 240 260 230 300 Rod incex, i Witncrann t

t t

I f

1 t

1 0

25 50 75 100 0

25 50 75 100 Group 5 Group 7 i

t t

9 i

O 25 50 75 100 Grcus 6 R00 POSITION t.lMITS FOR TWO & THREE-PUMP OPERATION AFTER 235 + 10 EFPD OCONEE 3 -

Or Wi min. OCONEE NUCLEAR STATION W'

Figure 3.5.2-2c3

(

Amendments Nos. 63, 63, & 60 3.5-20b

(

e Pp.er, a of 2558 Eat RESTRICIED REGION

~

-19,102

-- 100

-22.5,00

-- S0

?-20,90 80 o 20,60

-30,00 70

-- 60

-- 50

-- 40 PERMI3SIBLE

-- 30 OPERATING REGICN

-- 20

-- 10 i

i t

i 50

-40 30 20

-10 0

10 20 30 40 50 Arial Poner Imaalance, 5 OPERATIONAL PCWER IMSALANCE ENVELOPE FOR OPERATION FROM 0 TO 100 + 10 EFPC OCONEE 3 b

hatnate, OCONEE NUCLEAR STATION D E' Figure 3.5.2-3C1 Amendments Nos. 63, 63, & 60 3.5-23

em, s,

Pecer, 5 ef 2500 EN:

RESTRIOTED REGICN

-22,!02 15.102

- - l u c, 30,50

-- 90 20,90 35,30.

-- 80

-a.20.80

-- 70 60

-- 50 PE R4 l SS I E!.E OPE 3AT!NG

-- 40 RE C 10'4

-- 30

-- 20

-- 10 I

f f

8 t

i I

I

-50 40 30

-20 10 0

10 20 30 40 50 Axial Pener letalance, 5 OPERATIONAL POWER IMSALANCE ENVELOPE FCR OPERATION FROM 100 + 10 TO 235 + 10 EFPD OCONEE 3 pecem OCONEE NUCLEAR STATION

'M Figure 3.5.2-3C2 Amendtrents Nos. 63, 63, s 30 3.5-23a

m 1

Pe er, t :l 25SS MWt RESTRIGIED REGION m 20,102 26.102

,, i g,3

^

2.90

-39,90

.- 90 3

39,50 o

-- 60

-o 25,80

~~ 10

-- 60 PERMISSISLE

~

OPERAill;G

~~

REGION

-- 40 30

-- 20

-- 10 i

t

?

e 9

f f

f

-50 40

-30

-20

-10 0

10 20 30 40 50 Arial Power I-aalance, S OPERATIONAL POWER IMBALANCE ENVELOPE FOR OPERATION AFTER 235,1,10 EFPD OCONEE 3

[o:n nan; OCONEE NUCLEAR STATION Y'

Figure 3.5.2-3C3 5

Amendments-Nos. 63,63, & 60 3.5-23b

e i ICTED 100 REG'CN 19.1.90 90 26,80 gg 70 100,70 3

8

~

.C g

o O

PE R' ISSISLE 50 J

d OPERATING 40 REGION 30 20 10 0

e i

i i

e 0

10 20 30 40 50 60 70 80 90 100 APSR, 5 Witneraan

.i 4

i APSR POSITION LIMITS FOR OPERATION FROM 0 TO 100 + 10 EFPD OCONEE 3

.n u m te, OCONEE NUCLEAR STATION Y

Figure 3.5.2-4c1 Amendments Nos. 63, 63, & 60

J 23,102 100 RESTRICTED SO -

- 25.5,90 N

45.80 80 70 100,70 a"

X 60 O

E 50 o

40 E

PERMISSIELE ONRATING 30 REGICN 20 10 0

0 10 20 30 40 50 60 70 80 90 100 APSR, 5 litnataan APSR POSITION LIMITS FOR OPERATION FROM 100 1 10 to 235 + 10 EFPD OCONEE 3

, oat mais, OCONEE NUCLEAR 07% TION Y'

Figure 3.5.2-4C2 i

I Amendments Nos. 63, 63, 3_60 3.5-23J

m.

?

25.5.102 100 RESTRICTED REGICf4 i

33*C0 50 l

64.4.80 80 E

70 100,70 m5 60 a

C o

53 5

40 PERMISSIBLE OPERATING REG 10N 3g 20 10 l

1 0

1 I

1 0

10 20 20 40 50 60 70 60 90 100 i

APSR, 5 Witnarann APSR POSITION LIMITS FOR OPERATION AFTER 235 + 10 ETPD

~

I OCONEE 3 i

.w, OCONEE NUCt. EAR STATION Figure 3.5.2-4c3 I

f.

A,endme ns Nos. 63, 63, & 60 3.5-23k l