ML19312C383

Amends 45,45 & 42 to Licenses DPR-38,DPR-47 & DPR-55, Respectively,Changing Tech Spec Operating Limits for Unit 2, Cycle 3 Operating & Requirements for Testing Reactor Core Internal Vent Valves

ML19312C383
Person / Time
Site:	Oconee
Issue date:	07/29/1977
From:	Schwencer A Office of Nuclear Reactor Regulation
To:
Shared Package
ML19312C379	List: ML19312C378 ML19312C383 ML19312C388
References
NUDOCS 7912130910
Download: ML19312C383 (36)

Text

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UNITED STATES y"

NUCLEAR REGULATORY COMMISSION

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.....s DUKE POWER COMPANY DOCKET NO. 50-269 OCONEE NUCLEAR STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.45 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 6, 1977, as supplemented June 21 and July 11, 1977, and application dated March 1,1977, as supplemented May 5,1977, comply 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 Chaptar I; B.

The facility will operate in confonnity 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 l

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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2.

Accordingly, the license is amended by changes to the Technical Soecifications as indicated in the attachment to this license amendment and paragraph 3.8 of Facility License No. DPR-38 is hereby amended to read as follows:

(2) Technical Specifications _.

The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 45, 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 its date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION 1

if h A. Schwencer, Chief Operating Reactors Branch #1 Division of Operating Reactors

Attachment:

Changes to the Technical Specifications Date of Issuance:

July 29, 1977 i

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• g UNITED STATES

-/gp' *j NisCLEAR REGULATORY COMMISSION y4 3

WASHINGTON, D. C. 20666 DUKE POWER COMPANY DOCKET NO. 50-270 OCONEE NUCLEAR STATION, UNIT NO. 2 AMENDMENT TO FACILITY OPEPATING LICENSE Amendment No.45 License No. DPR-47 1.

The Nuclear Regulatory Cor.iission (the Comission) has found that:

A.

The application for amendment by Duke Power Company (the licensee) dated May 6, 1977, as supplemented June 21 and July 11, 1977, and application dated March 1,1977, as supplemented May 5,1977, comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), ar.d the Comission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in confonnity with the application, the provisions of the Act, and the rules and regulations of the Comissien; 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.

, 2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 3.8 of Facility License No. OPR-47 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A and 3, as revised through Amendment No. 45, are hereby incorporated in the license. The licensee shal1 operata the facility in accordance with the Technical Specifications.

3.

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

FOR THE NUCLEAR REGULATORY COMMISSION l

4WAYO A. Schwencer, Chief Operating Reactors Branch #1 Division of Operating Reactors

Attachment:

Changes to the Technical Specifications Date of Issuance:

July 29, 1977 l

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Q UNITED STATES NUCt. EAR REGULATORY COMMISSION g

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DUKE POWER COMPANY DOCKET NO. 50-287 OCONEE NUCLEAR STATION, UNIT NO. 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 42 License No. OPR-55 1.

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

A.

The application for amendment by Duke Power Company (the licensee) dated May 6, 1977, as supplemented June 21 and July 11, 1977, and application dated March 1, 1977, as supplemented May 5,1977, comply with the standards and requir.ements of the Atomic Energy Act of 1954, as amended a ct), and the Comission's rules and regulations set in 10 CFR Chapter I; B.

The facility will operate in confonnity 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 au?.horized by this amendment can be conducted without i

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.

i-2.

Accordingly, the license is amended by changes to the Tecnnical Specifications as indicated in the attachment to this license amendment and paragraph 3.8 of Facility License No. DPR-55 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appencices A and B, as revised through Amendment No. 42, 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 its date of issuance.

FOR THE NUCLEAR REGULATORY COMNISSION AWAYD A. Schwencer, Chief Operating Reactors Branch #1 Division of Operating Rer,toi,

Attachment:

Changes to the Technical Specifications Date of Issuance:

July 29, 1977 i

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ATTACRMENT TO LICENSE AMENDMENTS AMENDMENT NO. 45 TO DPR-38 AMENDMENT NO. 45 TO DPR-47 AMENDMENT NO. 42 TO DPR-55 DOCKETS NOS. 50-269, 50-270 AND 50-287 Revise Appendix A as follows:

Remove the following pages and insert revised identically numbered pages.

2.1-3a 2.1-3b 2.1-5 2.1-8 2.1-11 2.3-3 2.3-4 2.3-6 2.3-9 2.3-12 3.5-8 3.5-9 3.5-10 3.5-11 3.5-14 3.5-14a 3.5-15 3.5-19 3.5-19a 3.5-19b 3.5-22 3.5-22a 3.5-22b 3.5-23f 3.5-239 3.5-23h 4.1-9 4.2-3 4.20-1

Bases - Unit 2 ThesafetylimitspresentedforOconeey* nit 2 have been generated using g

BAW-2 critical heat flux correlation and the Reae:or Coolan: System flow rate of 136.5 percent of the desian ficw (design ficw is 352,000 gpm for four-pump operation).

The flew rate utilized is conservative compared to the ac:ual measured flow ra:e(2).

To maintain the integrity of the fuel cladding and to prevent fission product release, it is necessary :o prevent overheating of the cladding under normal operating conditions.

This is accomplished by operating within the nucleate boiling regime of hea: transfer, wherein the heat transfer coefficient is large enough so that the clad surface temperature is only slightly greater :han the coolan: temperature.

The upper boundary of the nuclea:e boiling regime is termed " departure frem nucleate boiling" (DNB). At this point, there is a sharp redue:1on of the heat transfer coefficient, which would result in high cladding :emperatures and the possibility of cladding failure.

Although DN3 is not an observable parameter during reactor operation, the observable parame:ers of neutron power, reactor coolant flow, temperature, and pressure can be related to DN3 through :he use of the BAW-2 correlation (1).

The BAW-2 correlation has been developed to predic: DN3 and the location of DNS for axially uniform and non-unifor= heat flux dis:ributions.

The local DN3 ratio (DNBR), defined as the ra:io of the heat flux that would cause DN3 at a particular core location to the actual heat flux, is indicative of the margin to DNS.

The =inimum value of the DNBR, during steady-state operation, normal operational transients, and anticipated transients is limited to 1.30.

A DN3R of 1.30 corresponds to a 95 percent probability at a 95 percent confidence level that DNB will not occur; this is considered a conservative =argin to DN3 for all operating conditiens.

The difference between the actual core outlet pressure and the indicated reacter coolant sys:em pressure has been considered in de:ermining the core protection saf ety limits.

The difference in these two pressures is ncminally 45 psi; however, only a 30 ps' drop was assumed in reducing the pressure trip se: points to correspond to the elevated loca:1on where the pressure is actually measured.

The curve presented in Figure 2.1-13 represen:s the conditions at which a minimum DN3R of 1.30 is predicted for the maximum possible thermal power (112 percent) when four reactor coolant pumps are operating (minimum reactor coolant flow is 374,830 gpm).

This curve is based on the following nuclear power peaking factors with potential fuel densitication and fuel N

N N

2.67; F

= 1.78; F

= 1. 50.

The rod bowing effects:

F

=

q aH z

design peaking combination results in a more conservative DN3R than any other power shape that exists during normal operation.

i The curves of Figure 2.1-23 are based on the acre restrictive of two thermal limits and include the effects of potential fuel densification

)

and fuel rod bcwing:

4 l

2.1-3a i

Amendments 45, 45 & 42 l

~T N

The 1.30 DNBR limit produced by a nuclear peaking factor of F

= 2.67 1.

or the combination of the radiol peak, axial peak and positioE of the axial peak that yields no less than a 1.30 DN3R.

2.

The combination of radial and axial peak that causes central fuel melting at the hot spot.

The limit is 19.8 kw/ft for Unit 2.

Power peaking is not a directly observable quantity, and, therefore, limits have been established on the bases of the reactor power imbalance produced by the power peaking.

The specified flew rates for Curves 1, 2, and 3 of Figure 2.1-23 correspond to the expected minimum flow rates with four pumps, three pu=ps, and one pump in each loop, respectively.

The curve of Figure 2.1-13 is the most restrictive of all possible reactor coolant pump-maximum thermal power combinations shown in Figure 2.1-33.

The maximum thermal power for three-pump operation is 85.3 percent due to a power level trip produced by the flux-flow ratio 74.7 percent flow x 1.055= 73.8 percent power plus the maximum calibration and instrument error.

The maxi =um thermal power for other coolant pu=p conditicns are produced in a similar manner.

For each curve of Figure 2.1-33, a pressure-temperature point above and to the left of the curve would result in a DNBR greater than 1.30 or a local quality at the point of minimum DNBR less than 22 percent for that particular reactor coolant pump situation. The 1.30 DNBR curve for four-pump operation is more restrictive than any other reactor coolant pump situation because any pressure / temperature point above and to the left of the four-pump curve will be above and to the left of the other curves.

References (1)

Correlation of Critical Heat Flux in a Sundle Cooled by Pressurized Water, BAW-10000, March 1970.

I (2) Oconee 2, Cycle

- Reload Report - RAW-1452, April, 1977.

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2.1-3b Amendments 45, 45 & 42 J

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2000 UNACCEPTABLE OPERATION

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1800 1600 l

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CURVE RE1:! R 000Last FLO, < GPS,

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s0 035 COP.E PP.0TECTION SAFETY LIMITS 3

123 530

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ocaten. OCONEE NUCLEAR STATION Figure 2.1-25

?.1-3 Amendments 45, 45 & 42

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level trip and associated reactor power / reactor power-imbalance boundaries by 1.07% -Unit 1 for a 17. flow reduction.

1.055%-Unit 2 1.072-Unit 3 For Unit 2, the power-to-flow reduction ratio is 0.949, and for Units 1 and 3 l the power-to-flow reduction factor is 0.961 during single loop operation.

Pump Monitors The pump monitors prevent the minimum core DNBR from decreasing below 1.3 by tripping the reactor due to the loss of reactor coolant pump (s).

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

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

Reactor Coolant Systen 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 2.3-1A - Unit 1 2.3 Unit 2 2.3-1C - Unic 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

-4706) trip out (1800) psig (11.14 Tout-4706) l (1800) psig (10.79 Tout -4539) setpoints shown in Figure 2.3-1A have been established to maintain the DNB 2.3-13 2.3-lC ratio greater than or equal to 1.3 for those design accidents that result in a pressure reduction. (2,3)

Due to the calibration and instrumentation errors the safety analysis used a variable low reactor coolant system pressure trip value of (11.14 T

-4746) g (11.14 Tout -4746) l (10.79 Tout -4579)

Ceolant Outlet Temperature The high reactor coolant outlet temperature trip setting limit (619 F) shown in Figure 2.3-1A has been established to prevent excessive core coolant 2.3-13 2.3-1C temperatures in the operating range.

Due to calibration and instrumenta'. ion errors, the safety analysis used a trip set point of 620 F.

Reactor Building Pressure The high reactor building pressure trip metting limit (4 psig) provides positive assurance that a reactor trip will occur in the unlikely event of j

a loss-of-coolant accident, even in the absence of a low reactor coolant system pressure trip.

2.3-3 Amemdments 45, 45 & 42

m D*9D Ml-3_ f A

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Shutdown Bvpass_

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In order to provide for con:rol red drive tests. zero power physics testing, and s:artup procedures, there is provision for bypassing certain segnents of The reactor protec: ion system segments which the reactor protection system.

Two conditions are impesed when can be bypassed are shewn in Table 2.3-1A.

2.3-15 2.3-1C the bypass is used:

point must be By administra:ive centrol the nuclear overpower trip se:

1.

reduced to a value < 5.0". of ra:ed power during react:r shutdown.

2.

A high reac::: coolan: system pressure trip setpoin: of 17:0 psig is automatically imposed.

The purpose of the 1720 psig high pressure : rip se: point is :: preven: normal of the reac:or pre:ec: ion system bypassed.

This high opera: ion wi:h part than the normal low pressure trip se: point pressure trip set poin: is lowet The over be tripped before :he bypass is ini:iated.

so that the reae:ct must power trip se: point of 1 5.0% preven:s any significan: reac:or power from Sufficien: na: ural being produced when performing the physics tests. remove 5.0* cf ra:ed pcwer if none of circulation (S) would be available ::

the reAc c. coolant pumps were operating.

Two pumo Operation A.

Two Leop Operation Operation with one pump in each loep will be allowed only following reacter shutcown.

After shutdown has occurred, rese: the pu.rp centact monitor po.er level trip setpcint :o 55.0*.

B.

Single Loop Opera: ion Single loop operation is permi::ed only af:er the reac:cr has been After the pump contact monitor trip has occurred, the following tripped.

actions will permit single loop operation:

to 55.0%.

the pump contact moni:or pcwer level trip se: point 1.

Rese:

Trip one of the two protective channels receiving outlet temperature 2.

information frem sensors in the Idle Loep.

3.

Reset flux-flow setpoint to 0.961 (Unit 1) 0.949 (Unit 2) 0.961 (Unit 3)

REFERENCES (1) FSAR, Section 14.1.2.2 (4) FSAR, Section 14.1.2.3 (2) FSAR, Section 14.1.2.7 (5) FSAR, Section 14.1.2.6 (3) FSAR, Section 14.1.2.8 2.3-4 Amendments 45, 45 & 42

n 2400 T = 619'F P = 2355 pstg 2300 y

2200 ACCEPTABLE OPERATION S

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UNACCEPTABLE 3

2000 2

OPERATION cc ll A

1900 P = 1800 psig 1800 T = 584F I

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i Reactor Outlet Temperature, *F PROTECTIVE SYSTEM MAXIvuM l

ALLOWABLE SETPOINTS, U'4IT

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PROTECTIVE SYSTEM VAXIMUt' ALLOWASLE SETPOI?!TS UNIT 2 lM'; OCONEE NUCLEAR STATION Figure 2.3-2B Amendments 45, 45 & 42 2.3-9

(3) Except as provided in specification 3.5.2.4.b, the reactor shall be brought to the hot shutdown condition within four hours if the quadrant power tilt is not reduced to less than 3.41% Unit I within 2' hours.

3.41% Unit 2 3.41 Unit 3 b.

If the quadrant tilt exceeds +3.41: Unit I and there is simultaneous 3.41% Unit 2 3.41?. Unit 3 indication of a misaligned control red per Specification 3.5.2.2, reactor operation =ay continue provided power is reduced to 60%

of the ther=al power allowable for the reactor coolant pump combination.

c.

Except for physics test, if quadrant tilt exceeds 9.44% Unit 1, 9.44% Unit 2 9.44 Unit 3 a controlled shutdown shall be initiated im=ediately, and the reactor shall be brought to the hot shutdown condition within four hours.

d.

Whenever the reactor is brought to het shutdown pursuant to 3.5.2.4.a(3) or 3.5.2.4.c above, subsequent reactor operation is permitted for the purpose of measurement, testing, and corrective action provided the thermal power and the power range high flux setpoint allowable for the reactor coolant pump combination are restricted by a reduction of 2 p,ercent of full power for each 1 percent tilt for the maximum tilt observed prior to shutdown.

e.

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

3.5.2.5 Control Rod Positions Technical Specification 3.1.3.5 does not prohibit the exercising a.

of individual safety rods as required by Table 6.1-2 or apply to inoperable safety rod limits in Technical Specification 3.5.2.2.

b.

Except for physics tests, operating rod group overlap shall be 25 + 5%

between two sequential groups.

If this limit is exceeded, corrective measures shall be taken imediately to achieve an acceptable 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 />.

c.

Position limits are specified for regulating and axial power shaping control rods.

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 ('Jnit 2); 3.5.2-1C1, 3.5.2-1C2 and 3.5.2-1C3 CUnit 3) for four pump operation, and on figures 3.5.2-2Al and 3.5.2-2A2 (Unit 1); 3.5.2-231 3.5.2-232 and 3.5.2-233 (Unit 2); 3.5.2-2C1, 3.5.2-2C2 and 3.5.2-2C3 (Unic 3) for two or three 3.5-8 Amendments 45, 45 & 242

pump opera. on.

Also, cxcepting physics t,ts or exnrcising con-trol rods, tha exial power shaping control rod inser: ion /withd:cw-al limits are specified on figures 3.5.2-431, 3.5.2-432. and

3. 5. 2-433 ('.'ni: 2).

If the control rod position limits are ex-ceeded, corrective measures shall be taken i==ediately to achieve an acceptable control rad position.

An acceptable control red posi:1on shall then be.2::ained within two hours.

The =inimum shutdown margin required by Specifica:1on 3.5.2.1 shall be main-

ained at all times.

d.

Except for physics tests, power shall not be increased above the power level cutoff as shown on Figures 3.5.2-1A1, 3.5.2-1A2 (Uni: 1), 3.5.2-131, 3.5.2-132, and 3.5.2-1B3 (Unit 2), and 3.5.2-1C1, 3.5.2-1C2, 3.5.2-1C3 (Unit 3), unless the following requirements are met.

(1)

The xenon reactivity shall be wi:hin 10 percent of the value for operation at steady-sta:e rated power.

(2) The xenon reactivity worth has psssed its final =aximum or mini =um peak during 1:s approach to its equilibrium value fer 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.

Except for physics tests, imbalance shall be maintained within the envelope defined by Figures 3.5.2-3A1, 3.5. 2-3A2, 3.5.3-331, l

3.5.2-3B2, 3.5.2-333, 3.5.2-3C1. 3.5.2-3C2, and 3.5.2-3C3.

If the im-balance is not within the envelope defined by these figures, corrective measures shall be taken to achieve an acceptable i= balance.

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

3,5.2.7 The control rod drive patch panels shall be locked at all times w1:h limited access to be authorized by the manager or his designated alternate.

3.5-9 Amendments 45, 45 & 42

d 1

Bases The power-imbalance envelope defined in Figures 3.5.2-3A1, 3.5.2-3A2, 3.5.2-3B1, 3.5.2-332, 3.5.2-333, 3.5.2-3C1, 3.5.2-3C2 and 3.5.2-3C3 is based on LOCA analyses which have defined the maximum linear heat rate (see Figure 3.5.2-5) such : hat the maxi =um clad temperature will not exceed the Final Acceptance Criteria.

Corrective measures will be taken i= mediately should the indica:ed quadrant tilt, rod position, or imbalance be cu: side their specified boundary.

Operation in a situation that would cause the Final Acceptance Criteria to be apprcached should a LOCA occur is highly improbable because all cf the pcwer distribution parameters (quadrant tilt, red positien, and imbalance) must be at their limits while simultaneously all other engineering and uncertainty fac: ors are also at their limi:s.**

Conservatism is introduced by application of:

a.

Nuclear uncertainty factors b.

Ther=al calibration I

c.

Fuel densification effects d.

Hot red =anufacturing tolerance factors e.

Fuel red bewing effects The 25 - 57, overlap between successive con:rol red gr. ups is allowed since the worth of a red is lower at the upper and icwer par: cf :he streke.

Control rods are arranged in g cups or banks defined as follows:

Graue Function 1

Safety 2

Safety 3

Safety 4

Safety 5

Regulating 6

Regulating 7

Xenen transient override 8

AFSR (axial power shaping bank)

The red posi: ion limits are based on the =os: limiting of the following three criteria:

ECCS power peaking, shutdown margin, and potential ejected red worth.

Therefore, ce=pliance with the ECCS power peaking criterion is ensured by the rod position limits.

The minimum available rod worth, concistent with the red position limits, provides for achieving hot shutdown by reactor trip at any time, assuming the highest worth control red 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 0.5% ak/k (Unit 1) or 0.65%

ak/k (Units 2 and 3) at rated power.

These values have been shown to be safe by the safety analysis (2, 3, 4, 5) of the hypothetical rod ejection accident.

A max 1mus 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 of 1.0% ak/k at beginning-of-life, hot zero power would result in a lower transient peak thermal power and, therefore, less severe environmental consequences than a 0.5% ak/k (Unit 1) or 0.65% ak/k (Units 2 and 3) ejected rod worth at rated pcwer.

l

• • Actual operating limits depend on whether or not incore or excore detectors are used and their respective instrument calibration errors.

The method used to define the operating limits is defined in plant operating procedures.

3.5-10 Amendments 45, 45 & 42

.?

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

Groups 5, 6, and 7 are everlapped 25 percent.

The normal position at power is for Groups 6 and 7 to be partially inserted.

The quadrant power tilt limi:s set forth in Specificatien 3.5.2.4 have been established with consideration of potential effects of rcj bowing and fuel densification to prevent.the linear heat rate peaking increase i

associated with a positive quadrant pcwer til: during normal power cperation i

from exceeding 5.10% for Uni: 1.

The limits shown in Specification 3.5.2.4 5.10% for Uni: 2 5.10% for Unit 3 are measurement system independent.

The ac:ual operating limits, with the appropriate allowance for observabili:y and instru=entatic.. errors, for each measurement system are defined in :he station operating procedures.

The quadrant til: and axial imbalance moni:oring in Specifica:icn 3.5.2.4 and 3.5.2.6, respectively, nor= ally vill be performed in the pr: cess computer.

The two-hour frequency f : =eni:cring these quanti:ies vill provide adequa:e surveillance when the c:=pu:er is out of service.

Allevance is provided for wi:hdrawal limi:s and reac

poyer i=balan:e limits to be exceeded for a period of :ve hours without specification violation.

Acceptable red positions and i= balance must be achieved within the :vo-hour time period er appropria:e ac:1cn such as a reduction of power taken.

Operating restrictions are included in Technical Specifica:icn 3.5.2.5d to preven excessive power peaking by transient xenon.

The xenon reactivity must be beyond its final maximum or =ini=um peak and approaching its equili-brium value at the power level cuteff.

.R.EFERENCES 1TSAA, Section 3.2.2.1.2 2PSAR, See:1on 14.2.2.2 3FSAR, SUPPLDiENT 9 4

BW MJEL DENSITICATION RIPORT 3AW-1409 (UNIT 1)

SAW-1396 (UNIT 2; 1AW-1400 (UNIT 3) l l

3.5-11 Amendments 45, 45 & 42

(174 302)

,,(225 8 102)

OPf 4Afl04 IN I'll!

b!f REGION IS 407 RESTRic7E0 Att03E0 ci74 2 323 (22 8 92)i REGION PueER LEVEL (160.2.80) g.gg g ggy 10 3 HUT 30th EAR 31% Llulf (146 2 70)

(253 I 70) to (132 2.60)

(257.3 60)

.C (45.50)

(119 2.506 PERul:31 Bt!

(251. B 50)

OPERail%G REGION 4g (300.37) 2 20 (0.15)

(73 15, t

i i

e 0

O 50 100 150 200 250 300 Rec inces, i sitneraen 0

25 50 75 100 a

25 50 75 100 t

t f

f I

Group 5 Group 7 0

25 50 75 100 i

G r s u,,

6 ROD POSITION LIMITS FOR 4 PUMP OPERATION FRCM 0-100 + 10 EFFD UNIT 2-1 um t CCONEE NUCLEAR STATIOfJ 5

Figure 3.5.2-181 4

3.5-14 Amendments 45, 45 & 42 i

j l

l 1

(181 102)

.,(225 8.102) 100 OPERATION 14 THIS RESTRICTIO REGION REGION is wCT ALL0t!O (225 8 9D Pat!R I,fvEL CUTOFF

$Hufacew v1RGry 83 LluiT 15J 2.!O)

(239 8.80)

(14 2.70) 253 8.70)

RESTRICTED 50 REGION (132 2.50)

(257.8.50)

I 3

}

(105.50)

(118 2.50)

(;g, 3 3g3 so (300 37)

PERMl!!!!LE

~

OPERATIN' REGION g

(56,15i (70.15) 0 0

50 100 150 200 250 30 0 Rc: Incer. % fitneraen 0

25 50 75 100 0

25 50 15 100 i

Groug 5 Group 7 0

25 50 75 100 i

t Group 5 ROD POSITICN LIMITS FOR 4 PUMP GPERATION FROM 100 + 10 'O 250 + 10 EFPD UNIT 2

'; OCONEE NUCt. EAR STATION Figure 3.5.2-182 3.5-14a Amendments 45, 45 & 42 i

i I

~

( $' 8 '

CPERATION IN THl3 L

REGl0N 15 NOT ALLQ8E0 CUTOFF gg (237 6.50)

$ HUT 00th

'223 I IO' uAR3lN -

g Lluii RESTRIOTED 60 RISION t203 $ $0) i:

(121.50)

(135 6 50; j

40 E

PERutSSIBLE OPERATING (63,15)

(146.15) 0 O

50 100 150 200

,250 300 43C fflCet, S flt50faen 0

25 50 75 100 0

25 50 75 100 I

f f

f t

j Group 5 Groua 7 0

25 50 15 100 1

t t

e I

Group 6 R00 POSITION LIMITS FOR 4 PUMP OPERATION AFTER 250 + 10 EFPD UNIT 2 OCONEE NUCLEAR STATION Figure 3.5.2-1B3 3.5-15 Amendments 45, 45 & 42

(90 t02)

(132.2.102) 100 Ulg (160 2.102)

(239 8.102)

OPERAfl04 IN THl!

E3 RESTRICTED REGION 15 NOT FJ ! L 3 P'JP gggggy gg, 3 E

ALL0sEO fiTH 2 OR 3 PJF LuTim (148.2,89)

(253 1,89) p,p 3

RESTRICTE0 GPERAfl04 3

30 IN Thf3 3

(132.2,76)

(III 8.78)

REGl>

t

[

IIN!WUM SHUT 00sN m

=

80

- Llulf

< 45,50 :

PERulS lBLE (300 47)

OPERATING

~

RE310N

!2 20 (0.15)

O t

t e

i O

50 100 150 200 250 300 Ras Innes, i fitnarasq 0

25 50 75 100 0

25 50 75 100 t

t i

f f

I

?

f Groap 5 Groua 7 0

25 50 75 100 t

a y

f Groua 6 R00 POSITION LIMITS FOR T'WO AND THREE PUMP OPERATION FRCM 0 70 100 EFPD UNIT 2 mt OCONEE NUCLEAR STATION Figure 3.5.2-2B1 3.5-19 Amendments 45, 45 & 42

(181 102)

( 239. B.102 )

0 GPERAfl0N IN THi!

g g

g afGION is NOT ALLOT 0 G53 8.89) 3 PUW CPERAilCN 5

I to 0

(267 8.76) seur ;eq usa;is I

n.

tisti W

60

=

(106.50)

(300,47) 1 40 Ptaul%318tf

]

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20 i55 15) i f

0 g

50 130 150 200 250 300 R:: Ir.ces. ', Eitnersen 0

25 53 75 100 0

25 50 75 100 t

e i

t t

f Groug 5 Grou3 7 0,

25 50 15 100 Group 6 P.00 POSITI0ti LIMITS FCP. It10 AMD THREE PUMP CPERATION FP0M 100 - 10 TO 250 + 10 EFPD "tJIT 2 Catme>

OCONEE NUCLEAR STATION Figure 3.5.2-232 1

3. 5 -19a hnendments 45, 45 & 42 l

100 OPERATION IN THl!

REGION 11 NOT ALL0910 IITH RESTRICTED RE;10N 3 Pur

=

m5W

=

2 OR 3 PuuPS FOR 2 & 3 P M mm OPERATION ag;1ti;;V 5 t

$o INTht$

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'89.15) 0 i

0 50 100 150 200 250 300 Roo inces. 5 fata:raen C

25 50 75 100 0

25 50 75 100 i

I I

t t

t t

g Groua 5 Group 1 0

25 50 75 100 t

i Group 6 R00 POSITION LIMITS FOR TWO AND THREE PUMP OPERATION AFTER 250 + 10 EFPD UNIT 2 OCONEE NUCLEAR STATION Figure 3.5.2-283 3.5-19b Amendments 45, 45 & 42

Power, 5 of 2568 utt RESTRICTED REGION

(.18.4,102)

(+14.2.102) 100 f

f(*I'#'02)

( 17.5,92) 90..

80 -

70..

60 50..

40 - -

30 -

20 - -

10 - -

t 0

l

-20 40 0

+10

+20 Core Imaalance, "<

OPERATIONAL POWER IMBALANCE '

ENYELOPE FOR OPERATION FRCM 0 TO 100 + 10 EFPD, UNIT 2

) OCONEE NUCLEAR STATION 3.5-22 Figure 3.5.2-381 Amndmnt.s 45. 45,. 47

Power, 4 of 2568 Mft RESTRICTED REGION

(-20.5.102)

( +14. 2,10 2 )

100

( 19.0,92)

(+13.4,92) 80 70 60 50 --

40 30 20 10 0

-30

-10

-0 0

+10

+20

+30 E

Core imoalance, 4 l

OPERATIONAL POWER IMSALANC ENVELOPE FOR OPERATION FRCM 100 + 10 TO 250 + 10 EFPO UNIT'2 OCONEE NUCLEAR STATION Figure 3.5.2-382 3.5-22a Amendments 45, 45 & 42

Poser. 5 of 2568 uit RESTRICTED REGION

( 23 3,102)

-(+11 1.102)

.. +10 7. 9 2 )

(.25 3.92) 90 30 -

70 -

60 -

50.

ID.

30 20 -

10 0

-30 20 10 0

+10

+20

+30 Core imoalance.

's OPERATIONAL POWER IMBALANC ENVELOPE FCR OPERATION AFT) 250 10 EFPD, UNIT 2

) OCONEE NUCLEAR STATION

~

ndments 45, 45 & 42 I

i (7.6,102)

(32.6.102) 100 RESTRICTED (5.2.92)

(35.0.92)

REGION 90 80

. ( 2. B. 80 )

(35.0.80)

RESTRICTE0 REGION 70 (0.70)

(42.70)

(86.60)

E 60 3

(100.60)

"_ 50 O

e f 40

[

PERMISSIBLE OPERATING 30 REGION 20 10 0

0 10 20 30 40 50 60 70 80 90 100 APSR 5 Wetnarawn APSR POSITICN LIMITS FOR CDERATION FROM 0 TO 100

• 10 EFPD, UNIT 2 Camtr OCONEE NUCLEAR STATION Figure 3.5.2-1B1 3.5-23f Amendments 45, 45 & 42

e RESTRICTED REGION f

100 RESTRICTED f

REGION (36.5.92)

(5.2 92) g 80

. (2.3.30)

(36.5 80)

(0.70)

43 5,70) 70 (57.5.60) 60 5

(100,60)

~

50 g

e 40 f.

E PERMI SSIBl.E 30

.PERATING 0

REGION 20 10 i

1 0

0 10 20 30 40 50 60 70 80 90 100 APSR, * # tnaraan APSR POSITION LIMITS FOR OPERATION FROM 100 i 10 TO 250 i 10 EFPD, UNIT 2 OCONEE NUCLEAR STATION 3.5-23 Amendments 45, 45 & jgwe 3.5.2-4B2 F

)

RESTRICTED REGION 100 6.5.102)

(31.4,102)

RESTRICTED REGION 90 (4.1.92) q(33.8,92) 80 (1.7,80) o (33.8,80) 70 (0.70) a (33.8,70)

(34.8,60)

=*

S0 (100,60) m 50 - -

a f

40 PERMISSIBLE E

OPERATING REGION 30 20 10 0

+

0 10 20 30 40 50 60 70 80 90 100 APSR, 5 Witndrawn APSR POSITION LIMITS FOR OPERATION AFTER 250 + 10 EFPD UNIT 2 i

OCONEE NUCLEAR STATION Figure 3.5.2-48 3 s

3.5-2h Amendments 45, 45 & 42 l

l

4.2.10 The licensee shall submit a report or application for license amendment to the NRC within 90 days after the occurrence of the following: After March 13, 1978, any time that Crystal River Unit No. 3 fails to maintain a cumulative reactor utilitation factor of greater than 45'..

The report shall provide justification for continued operation of Oconee Nuclear Station Units 1, 2 and 3 with the reactor vessel surveillance program conducted at Crystal River Unit No. 3 or the application for license amendment shall propose an alternative program for conduct of the reactor vessel surveillance program.

4.2.11 During the first two refueling periods, two reactor coolant system piping elbows shall be ultrasonically inspected along their long-itudinal welds (4 inches beyond each side) for clad bonding and for cracks in both the clad and base metal.

The elbows to be inspected are identified in B5W Report 1364 dated December 1970.

Bases The surveillance pregram has been developed to comply with Section XI of the ASME Boiler and Pressure Vessel Code, Inservice Inspection of Nucitar Reacter Coolant Systems,1970, including 1970 winter addenda, edicion.

The progres places r.ajor emphasis on the area of highest stress concentrations and on areas where f ast neutron irradiation might be suf ficient to change material properties.

The number of reactor vessel specimens and the frequencios for recoving and testing these specimens are provided to assure compliance with the requirements of Appendix H to 10 CFR Part 50.

For the purpose of Technical Specification 4.2.10.

Cumulative reactor utilizatien factor is defined as: [(Cumulative thermal mejawatt hours since attainment of commerical operation at 100% power) x 100] +.(licensed thermal power) x (cumulative hours since attainment of commercial operation at 100% power)].

The definition of Regulatory Guide 1.16, Revision 4 (August 1975) applies for the term "commerical operation".

Early inspection of Reactor Coolant System piping elbows is considered desirable in order to reconfirm the integrity of the carbon steci base metal when explosively clad with sensitized stainless steel.

If no degradation is observed during the two annual inspections, surveillance requirements will revert to Section XI of the ASNE Boiler and Pressure Vessel Code.

4.2 3 Amendments.45, 45 & 42 L

4.20 REACTOR VESSEL INTERNALS '.'ENT VALVES i

Applicability Applies to reactor vessel internals vent valves used to prevent vapor lock in the reactor vessel following a postulated reactor coolant inlet pipe rupture.

Objective To verify that the reactor vessel internals vent valves operate as required.

Specification i

At least once each refueling cycle, each reactor vessel internals vent valve shall be demonstrated operable by:

Conductin6 a remote visual inspection of visually accessible surfaces a.

of the value body and disc sealing faces and evaluatin: any observed surface irregularities.

b.

verifying that the value is not stuck in an open position, and c.

Verifying that the valve can be fully opened with force equivalent to or less than 1.00 psid.

Bases The internals vent valves are provided to relieve the pressure generated by steaming in the core following a LOCA so that the core remains suffici-ently covered.

Inspection and manual actuation of the internals vent valves (1) assures-operability, (2) assures that the valves are not open during normal operation, and (3) demonstrates that the valves are fully open at the forces equivalent to the dif ferential pressures justifiable by the ECCS analysis.

4.20-1 Amendments 45, 45 & 42

.