ML19317D262

Proposed Revisions to Tech Specs 2.1,2.3,3.2 & 3.5 Re Core Protection Safety Limits & Protective Sys Max Allowable Setpoints

ML19317D262
Person / Time
Site:	Oconee
Issue date:	09/18/1978
From:	DUKE POWER CO.
To:
Shared Package
ML19317D258	List: ML19317D257 ML19317D262
References
NUDOCS 7911190640
Download: ML19317D262 (26)

Text

-

Attach::2ent 1 Proposed Technical Specification Rerision Pages 7911190[fC

Besus - Unit 2 The safety limits presented fo{ggconee Unit 2 have been generated using BAW-2 critical heat flux correlation and the Reactor Coolant System flow rate of 106.5 percent of the design flow (design flow is 350,000 gpm for four pump operation). The f measuredflowrate{gyrateutilizedisconservativecomparedtotheactual 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 so that the clad surface temperature is only slightly greater than the coolant temperature. The upper boundary of the nucleate boiling regime is termed " departure from nucleate boiling" (DNB) . At this point, there is a sharp reduction of the heat transfer coefficient, which would result in high cladding temperatures and the possibility of cladding fail-ure. Although DNB is not an observable parameter during reactor operation, the observable parameters of neutron power, reactor coulant flow, temperature, and pressure can be related to DNB through the use of the BAW-2 correlation (1). The BAW-2 correlation has been developed to predict DNB and the loca-tion of DNB for axially uniform and non-uniform heat flux distributions. The local DNB ratio (DNBR), defined as the ratio of the heat flux that would cause DNB at a particular core location to the actual heat flux, is indicative of the margin to DNB. The minimum value of the DNBR, during steady-state operation, normal operational transients, and anticipated transients is limited to 1.30.

A DNBR of 1.30 corresponds to a 95 percent probability at a 95 percent confi-dence level that DNB will not occur; this is considered a conservative margin to DNB for all operating conditions. The difference between the actual core outlet pressure and the indicated reactor coolant system pressure has been considered in determining the core protection safety limits. The difference in these two pressures is nominally 45 psi; however, only a 30 psi drop was assumed in reducing the pressure trip setpoints to correspond to the elevated location where the pressure is actually measured.

The curve presented in Figure 2.1-13 represents the conditions at which a mini-mum DNBR of 1.30 is predicteed for the maximum possible thermal power (112 percent) when four reactor coolant pumps are operating (minimum reactor coolant flow is 374,380 gpm). This curve is based on the following nuclear power peak-ing factors with potential fuel densification and fudl rod bowing effects:

F{ = 2.565; FaH *

• I = 1.50 z

l The design peaking combination results' in a more conservative DNBR than any other power shape that exists during normal operation.

The curves of Figure 2.1-23 are based on the more restrictive of two thermal limits and include the effects of potential fuel densification and fuel rod l bowing:

l l . -

2.1-3a

1. Tha 1.30 DNBR linit producsd by the conbination of the radial psek, exial l prak and position of the axial peak that yields no less than a 1.30 DNBR.

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 flow rates for Curves 1, 2, and 3 of Figure 2.1-2B :orrespond to the expected minimum flow rates with four pumps, three pumps, and one pump in each loop, respectively.

The curve of Figure 2.1-1B 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-pamp operation is 85.3 percent due to a power level trip produced by the flux-flow ratio 74.7 percent flow x 1.055 =

78.3 percent power plus the maximum calibration and instrument error. The maximum thermal power for other coolant pump conditions are produced in a similar marner.

For each curve of Figure 2.1-3B, a pressure-temperature point above and to the lef t of the carve would result in a DNER greater than 1.30 or a local quality at the point of minimum DNBR less than 22 percent for that particu-lar 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 Bundle Cooled by Pres.arized Water, BAW-10000, March 1970.

(2) Oconee 2, Cycle 3 - Reloao Report -

BAW-1452, April, 1977. l (3) Oconee 2, Cycle 4 - Reload Report -

3AW-1491, Augus t, 1978. l 4

2.1-3b l

G OF RATED THERMAL POWER DNBR LIMIT -- 120 -

i 112.0 @

( 33.60.112) (10.64.112.0) 110 Ks/FT LIMIT ACCEPTABLE 4 PUMP l FT I I

LIMIT l OPERATION -

- 100 g (31.0,100.0)

( 52.0.95.0) l l g l --

90 l l . e5.31 l a) l ,

I l

ACCEPTABLE l I

- 80 l 3 & 4 PUMP l l lGPERATION g 173.31 68 31

- 70 g I

I i 58.20 60

@li ,

1 l

l l l l ACCEPTABLE l

2,3 14 PUMP - 50 l l

~

OPERATION l

1. I 20 1 -

40 l l I I I I - 3 I I i l I I l l - - 20 l l

' I l l I ~ 10 j i i l ,

I: '

I: ' -

I , l ' , .

50 -50 40 -30 -20 10 0 10 20 30 40 50 50 AXi31 Power Imaalance, ",

CURVE REACTOR COOLANT FLOW (GPM) 1 374.380 2 280.035 3 133.690 CORE PROTECTION SAFETY L y.ITS g UTAT 2 icanoast OCONEE NUCLEAR STATION M Figure 2.1-23 2.1-10

THERNAL POWER LEVEL. 4 UNACCEPTABLE OPERATION 2.463

( 19.0.105.5) i0s.5 T '7, $

ACCEPTABL E.-- 100 l4PUNP k

s@ l0PERATION l (20.0,95.0)

,N' ~ 90 4%, l I I

( 42.0,80) 78.81 -- m ACCEPTABLE I

3 f. 4 PUMP -~

lg OPERATION (20.0,68.31) i I _ _g

( -42.0.53.31' . 51 70 -

l ACCEPTABLg"~ 50 N l 2,3, f. 4 l PUNP l

4

\ (20.0,41.20) 1 l0PERATION I

I -m

( 4 2.0.25.20)  :

I

-m 1

i l -

- 10 l I I

:  :  : :I  :  :  :  :  :  :

-50 -50 -40 -30 20 -10 0 10 20 30 40 50 50 Power imcalance, 5 PROTECTIVE SYSTEM MAXIMUM ALLCWA3LE SETP0ISTS g

pu Maa; UNII 2 DY' OCONEE Figure NUCLEAR 2.3-23 STATIO l

2.3-9

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

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

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

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

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

This source will be the concentrated boric acid storage tank contain-ing at least the equivalent of 995 ft3 of 8700 ppm boron as borie l acid solution with a temperature at least 10 F above the crystalliza-

tion 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 acid storage tank. At least one channel of heat tracing capable of meeting the above temperature requirement shall be in operation. One associated boric acid pump shall be operable.

If the concentrated boric acid storage tank with its associated flow-path is unavailable, but the borated water storage tank is available and operable, the concentrated boric acid storage tank shall be re-stored 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 l in a hot shutdown condition and be borated to a shutdown margin l equivalent to 1% ak/k at 200 F within the next twelve hours; if the t

concentrated boric acid storage tank has not been restored to opera-bib ty within the next 7 days the reactor shall be placed in a cold l the down condition within an additional 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

! If the concentrated boric acid storage tank is available but the l borated water storage tank is neither available nor operable, the 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 addition-al 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

l t

l 1

r 3.2-1 l

l l . . . .. -- .--

Bases The high pressure injection system and chemical addition system provide con-trol 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 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 th reactor coolant systen to a 1% ak/k suberitical margin at cold conditions (70*F) with the maximum worth stuck rod and no credit for zenon at the worst time in core life. The current cycles for each unit, Oconee 1 Cycle 5, Oconee 2, Cycle 4, l 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.

ft3 of 8,700 ppm boric acid in the concentrated boric acid A minimum of 995 storage tank, or l a minimum of 350,000 gallongs of 1800 ppm boric acid in the borated water storage tank (3) will satisfy the requirements. The volume requirements in-clude a lot, margin and in addition allow for a deviation of 10 EFPD in the cycle lens h. The specifiistion 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 free 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 injected in less than six hours using only one of the makeup pumps.

The concentration of boron in the concentrated boric acid storage tank may be higher than the concentration which would crystallize at ambient 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 concentration of 3,700 ppm in the concentrated boric acid storage tank cor-responds to a crystallization temperature of 77'F and therefore a temperature requirement of 37*F. Once in the high pressure injection system, the concen-trate is sufficiently well mixed and diluted so that normal system tempera-tures assure boric acid solubility.

REFERENCES (1) FSAR, Section 9.1; 9.2 (2) FSAR, Figure 7.2 (3) Technical Specification 3.3 3.2-2

3.5.2.5 Control Rod Positions

a. Technical Specification 3.1.3.5 does not prohibit the ex*ercising of individual safety rods as required by Table 4.1-2 or apply to in-operable safety rod limits in Technical Specification 3.5.2.2.

b. Except for physics test, operating rod group overlap shall be 25% !

5% between two sequential groups. If this limit is exceeded, cor-rective measures shall be taken immediately 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 2.5.2-1A1 and 3.5.2-1A2 (Unit 1); 3.5.2-131, 3.5.2-132 and 3.5.2-1B3 (Unit 2); 3.5.2-1C1, 3.5.2-1C2 and 3.5.2-IC3 (Unit 3) for four pump operation, and on figures 3.5.2-2A1 and 3.5.2-2A2 (Unit 1); 3.5.3-231, 3.5.2-2B2 and 3.5.2-233 (Unit 2); 3.5.2-2C1, 3.5.2-2C2 and 3.5.2-2C3 (Unit 3) for 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); 3.5.2-4B1, 3.5.2-4B2, and 3.5.2-4B3 (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,posi-tion. An acceptable control rod position shall then be attained with-in two hours. T'ne minimum shutdown margin required by Specification 3.5.2.1 shall be maintained at all times.

3.5.2.6 Xenon Reactivity Except for physics tests, reactor power shall not be increased above the power-level-cutoff shown in Figures 3.5.2-1A1, and 3.5.2-1A2 for Unit 1; Figures 3.5.2-131, 3.5.2-182, and 3.5.2-133 for Unit 2; and Figures 3.5.2-1C1, 3.5.2-1C2, and 3.5.2-1C3 for Unit 3 unles: one of the following conditions is satisfied:

1. Xenon reactivity did not deviate more than 10 percent from rie equi-librium value for operation at steady state power.

2. Xenon reactivity deviated more than 10 percent but is now within 10 percent of the equilibrium v'alue for operation at steady state rated power and has passed its final maximum or minimum peak during is ap-proach to its equilibrium value for operation at the power level cut-off.

3. Except for xenon free startup (when 2. applies), the reactor has oper-ated within a range of 37 to 92 per ent of rated thermal pcwer for a period exceeding 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. l

. s 3.5-9

- __ __ - _ , _. ~_ _ _ _ _.

d 3.5.2.7 Reactor power imbalance shall be monitored on a frequency not to exceed two hours during power operation above 40 percent ratet: power. Except for physics tests, imbalance shall be maintained within the envelope defined by Figures 3.5.2-3A1, 3.5.2-3A2, 3.5.2-3B1, 3.5.2-392, 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 figures, 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.

3.5.2.3 The control rod drive patch panels shall be locked at all times with limited access to be authorized by the manager or his designated alternate.

. .s 3.5-10

t Bases The power-imbalance envelope defined in Figures 3.5.2-3A1, 3.5.2'3A2, 3.5.2-3B1, j j 3.5.2-382, 3.5.2-3B3, 3.5.2-3C1, 3.5.2-3C2 and 3.5.2-3Ci is based on LOCA analyses  ;

which have defined the maximum linear heat rate (see Figure 3.5. 2-5) such that i the maximum clad temperature will not exceed the Final Acceptance Criteria.

Corrective measures will be taken immediately should the indicateo quadrant tilt, rod position, or imbalance be outside their specified boundary. Opera-tion in a situation that would cause the Final Acceptance Criteria to be a,-

proached should a LOCA occur is highly improbable because all of the power distribution parameters (quadrant tilt, rod position, and imbalance) must be at their limits while simultaneously all other engineering and uncertainty factors are also at their limits.** Conservatism is introduced by application 4

of:

a. Nuclear uncertainty factors

! b. Thermal calibration

c. Fuel densification power spike factors (Unit I and 2 only)

d. Hot rod manufacturing tolerance factors

e. Fuel rod bowing power spike factors J

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

Group Function .

1 Safety 2 Safety 3 Safety 4 Safety 5 Regulating 6 Regulating 7 Xenon transient override 3 APSR (axial power shaping bank)

I x* 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.

l 3.5-11

The rod position limits are based on the most limiting of the following three criteria: ECCS power peaking, shutdown margin, and potential ejected rod worth.

Therefore, co=pliance with the ECCS power peaking criterion is ensured by the rod position linits. The minimum available rod worth, consistent with the rod position limits, provides for achieving hot shutdown by reactor trip at any time, assuming the highest worth control rod that is withdrawn remains in the full out position (l). The rod position limits also ensure that in-serted rod groups will not contain single rod worths greater than 0.65% ak/k at rated power. These values have been shown to be safe by the safety analysis (2,3,4,5) of hypothetical rod ejection accident. A maximum single inserted control rod worth of 1.0% ak/k is allowed by the rod position limits at hot l

zero power. A single inserted control rod worth of 1.0% a k/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 0.65% ak/k ejected rod worth at rated power.

Control rod groups are withdrawn in sequence beginning with Group 1. Groups 5, 6, and 7 are overlapped 25 percent. The nor=al position at power is for Groups 6 and 7 to be partially inserted.

The quadrant power til: linits set forth in Specification 3.5.2.4 have been established to prevent the linear hes rate peaking increase associated with a positive quadrant power til: during nor=al power operation from exceeding 7.50% for Uni: 1. The limits shown in Specification 3.5.2.4 7.50% for Uni: 2 l 7.50% for Uni: 3 are measurement system independent. The actual operating li=its, with the appropriate allowance for observability and instru=entation errors, for each measurenent system are defined in the station operating procedures.

L The quadrant tilt and axial imbalanc e moni:oring in Specification 3.5.2.4 and 3.5.2.6, respectively, nor= ally will be performed in the process computer.

The two-hour frequency for monitorir.g these quantities will provide adequa'te surveillance when the ec=puter is oat of service.

Allowance is provided for withdrawal limits and reactor power imbalance limi:s to be exceeded for a period of two hours withou: specification violation.

Acceptable rod positions and inbalance must be achieved within the two-hour

i=e period or appropria:e action such as a reduction of power taken.

Operating res:rictions are included in Technical Specification 3.5.2.? to prevent excessive power peaking by transient xenon. For Uni: 1, a 5%

peaking increase is applied to calculated peaks at equilibrium conditions for powers above the power level cutoff. For Units 2 and 3, an 3% peaking increase is applied. These values conservatively bound the peaking effects of ::ansien:

xenon once the applicable requirement of 3.5.2.6 has been satisfied.

l l

l 3.5-11a I

- , _ . , , . - , , - - .-,- --. ._ - . _, . ~ , . . _ , _ - _ , _ . - . ._ . _ _ _ . _ _ _ . , _ _ - - _ _ . . , _ _ _ _ , _ , . _ . _ _ _ _ _ _ _

REFERENCES FSAR, Section 3.2.2.1.2 FSAR, Section 14.2.2.2 FSAR, SUPPLEMENT 9 4

B&W FUEL DENSIFICATION REPORT BAW-1409 (UNIT 1)

BAW-1396 (UNIT 2)

BAW-1400 (UNIT 3)

Oconee 1, Cycle 4 - Reload Report - BAW 1447, March 1977, Section 7.11 l

l l

.o 3.5- 11b l

. 't

/

TABLE 3.5-1 Quadrant Power Tilt Limits .

Steady State Transient Maximum Limit Limit Limit Unit 1 5.00 9.44 20.0 Unit 2 5.00 9.44 20.0 l Unit 3 5.00 9.44 20.0 4

e 3.5- lic

110 (171.102) (20S.102) 100 OPERATION IN THIS REGION IS NOT (171 32) (206.92) \ POWER LEVEL 90 ALLOWED CUT 0FF S0 - (151.4,90) (225.5.90)

E 70 RESTRICTED RESTRICTED

= SHUT 00sN REGION REGION

= MARGIN

" ,0 L,3;7 50 -(37.50) (125.5.50) (251.4.50)

PERMISSISLE 40 ~

( 00.40, 5 OPERATING 30 (0.28) 20 -

10 10.0) , , . . . . , .

0 2 , , .

0 20 40 60 30 100 120 140 ISO ISO 200 220 240 250 230 300 Rac Incex, '- #0 t t I f 1 1 I l  ! l 0 25 50 75 100 0 2: :D 75 100 Group 5 .

Grouc 7

!  ! I f f 0 25 50 7: 100 Grcua 6 l

l l

l l

l l

l R0D ?0SITIO:i LIMITS FOR FOUR-PL72 0?ERATIO FRCM 0 TO 250 - EFPD

' g CCOiEE 2 OCONEE NUCLEAR STATION p@u ma; Figure 3.5.2-;31 3.5-14

_ . _ _ . - _ _ _ - _ _ . . _ - - _ _ = . - - _ _ _ , . __. . .

110 OPERATION IN THIS (133 102) (259.102) (300.102) 100 REGION IS NOT ALLOWED ,

(253 7.92)

(235.50) POWER 8

• 0 -

LEVEL SHUTDOWN MARGIN CUT 0FF LIMIT RESTRICTED

= 70 _

REGION g 60 -

~

PERMISSIBLE OPERATING 50 - (52.50) (ISE.50)

• REGION 40 S

s 0 r (0.27) 20 -

10 -

0 3 20 40 50 S0 100 120 140 ISO ISO 200 220 240 250 2S0 300 Roc Incex. 5 40 t t  ! ' t t  ? f f f 25 50 75 100 0 25 50 75 100 0

Grouu 5 Grouc 7 1  ! t i 1 0 25 50 75 100 Gr:up 'S l

I l

RCD FOSITION LIMITS FOR FOL7.-? CMP CPERATION AF~ER 250 - EEPD

[Mg CCONEE 2 ~

JutMars. OCONEE NUCLEAR STATION Figure 3.5.2-132 3.5-14a ,

Figure 3.5.2-133 Deleted During Oconee Unit 2, Cycle 4 Operation 3.5-15 l . _ . _ _ _ _ _ _ _ _ _ _ w - --. _

T'" - - . . _ _ ,_, __ _ _ _

OPERA 710N IN THIS 110 - REGION IS NOT ,

ALLOWED #1TH 2 OR (130.102) (151.4.102) (225.5.102) '

(246.5.102) 100 -

(108.102)/ -

(125.5.92) 4C*

30 - G, ( 251. 4. 92)

• C RESTRICTED

1. G4 REGION FOR y

80 .

[ @g h*g 21,3 PUYP

= 4 CO OPER IU ~

SHUTOOWN ,F S $## (300,74) 3 MARGIN g g (125.5.63) (251.4.53) 'S umT

- ,0 2

30 a 7.50) p,8, g 0 00.5i)

= ,,

l. .0 l. 6+@, PERMISSISLE OPERATING 30 H 0,2s>

REGION 20 -

10 t . I e t '  ! t '

f '  ? t ,  ?

}

3 20 40 60 30 100 120 140 160 ISO 200 220 240 250 230 300 320 i Rcc Incex. i 40 i  ! I t 1 t f f I t 100 0 25 50 75 100 0 25 50 75 Group 7 Groua 5 t 8 e f f 0 25 50 75 100 Group 3 ROD POSITION LIMITS FOR TWO A'O THREE PLT OPERATION FROM 0 TO 250 - 10 EFFD

<O g OCOSEE 2 b@uttnat9 Figure OCONEE 3.5.2-231 NUCLEAR STA 3.5-19

i l

110 OPERATION IN THl$ (133.102) (120.102) (235.102)

REGION iS NOT (300 102) l 100

• ALLO 4E0 #1TH 2 OR 3 PUMPS (168.92) 90 30

- S g%*

y

." SHUT 00NN MARGIN  %

4R

~

limit  ; O S

e o (158.53) 5 50 ~

q.N h 50 (52.50) 4g g@

40 - 4 4g 5  % PERMISSISLE c.

GPERATING REGION 30 - (0.21) 20l-10 -

0 ' ' ' ' ' ' '  ! '  ?  ! I f ,

0 20 40 60 80 100 120 140 150 ISO 200 220 240 250 280 300 Roa incex. ", #0 i f f f 1 f f I f f 0 25 50 75 100 0 25 50 75 100 Graug 5 Graco 7 i i f I f 0 25 50 75 103 Group S RCD POSITICN LIMITS FOR I"40 .C;D T'4RIE PUMP c?ERATION AFTER 250 t 10 EFFD

<- OCONIE 2

~

.'o n0hm, OCONEE NUCLEAR STATION

'@' Figure 3.5.2-232 3.f-19a

Figure 3.5.2-233 Deleted during Oconee Unit 2, Cycle 4 Operation 3.5-19b

0 PowGr % of 2568 MWt RESTRICTED REGION - 110 *

(-13.2,102) ,,

(8.102)

( -l '4. 2,92 ) (8.6.92)

.- - 90

(-202.80) - - 80

- 70 PERMISSIBLE . 60 OPERATING REGION -- 50 40

- 30

- 20

- 10

, , 0 , , ,

-30 -20 -10 0 10 20 30 Core I:ncal ance. *.

t i

l l

[

OPERATIONAL. PC'a~rR IM3ALANCE ENVELOPE FOR OPERATION FRCM 0 TO 250 t 10 E7?D OCONEE 2

'o  :

'M') OCONEE Figure 3.5.2-331N CLEAR STA i 3.5-22 i

Power 5 of 2568 MWt

- 110 RESTRICTED REGION .

(-22.3.102) (9.7.102)

- 100

(-20.7.92) (17.4.92) 90

~ ~

PERMISSIBLE OPERATING REGION --

70 50

- - 50

- 40

- 30

-- 20

- 10

, , i 0 , .

_ f

-30 -20 10 0 10 20 30 Imcatance. ",

OPERATIONAI. PC'w'ER IM3ALANCE I:iVELOPE FOR OPERATION AFTER 250 - 10 EFPD OCONIE 2

, \

aat M*E t OCONEE NUCLEAR STATION DT Figure 3.5.2-332 3.5-22a

Figure 3.5.2-3B3 Deleted During Oconee Unit 2, Cycle 4 Operation 1

i

.s l

I l 3.5-22b t

I

RESTFICTED REGION ,

(6.1.102) (25.5,102) 100 90 - (3.2,92) (27.5.92) 30 (0.80) (29.3.80)

~

x 70 - RESTRICTED

= REGION 2

~ 60 -

~

_0 a - (64.4,50) 5 (100.40) f 40 -

PERMISSIBLE OPERATING 30 - REGION f

20 -

5 10 -

0 I i ' I I I 4 I t 0 10 s 20 30 40 50 50 7C 80 90 100

\

APSR, 5 #0 APSR POSITION LD!ITS FOR OPERATION FRCM 0 TO 250 - 10 EF?D OCONEE 2

. OCONEE NUCLEAR STATION b @aENw Figure '

3.5.2-431 3.5-23f t

RESTRICTED REGION 110- -

( 2.102) 100 -

' )

90 (2.3.92) '

(0,80)

• 80 (33.7.80)

- 70 -

=

RESTRICTED REGION E

.n 60 -

PERMISSISLE OPERATING

- REGION (64.4.50) 50

$ 40 -

(100.40) s-30 -

20 -

IG _

i r - i , , e i i 0

30 40 50 60 70 80 90 100 0 10 20 APSR $. 40 ,

l i

l APSR POSITION 1.!MITS FOR l

OPEFATION AFTER 250 - 10 EF?D 0CONEE 2 lontMwa ' OCONEE NUCLEAR STATION j

3.5-23; h] Figure 3.5.2-a32

Figure 3.5.2-4B3 Deleted during Oconee Unit 2, Cycle 4 Operation I

l e * ,/

3.5-23h

Attachment 2 Oconee Unit 2, Cycle 4 Reload Repor:

SNJ-1491

. .s