Text

Proposed Tech Specs Pages,Extending Operation of RBS from Current Licensed Power Level of 2894 Mwt to Uprated Power Level 3039 Mwt
	ML20210H290
Person / Time
Site:	River Bend
Issue date:	07/30/1999
From:	; ENTERGY OPERATIONS, INC.
To:	;
Shared Package
ML20137Y972	List: ML20210H290; RBG-45077, Application for Amend to License NPF-47 to Change TS to Extend Operation of RBS from Current Licensed Power Level of 2894 Mwt to Uprated Power Level 3039 Mwt.Encl Proprietary GE Rept NEDC-32778P Withheld;
References
	NUDOCS 9908030324
	Download: ML20210H290 (75)
	v • d • e

ENCLOSURE 3 ENTERGY OPERATIONS, INC.

RIVER BEND STATION (RBS)

PROPOSED TECHNICAL SPECIFICATION (TS) CHANGES LIST OF AFFECTED PAGES INCLUDING NPF-47, APPENDIX A, B, & C AND BASES POWER UPRATE PROJECT I. AFFECTED SPECIFICATIONS Location Description of Change 1.1 Definitions Revise value of rated thermal power definition to uprated power level (3039 MWt) shown in Table 1-2.

2.1.1.1 To maintain the same power value with respect to absolute thermal 1,4-2, 3 power, lower (from 25% to 23.8%) the low flow and pressure Thermal 3.2.1 Power Safety Limit by the ratio of the power increase (1/l.05).

3.2.2 3.2.3 3.3.1.1 3.4.3.1 3.7.5 B2.1.1 ,

Table 3.1.4-1 Because the increase in reactor operating pressure, the higher pressure scram time column pressure basis value should be changed to the value shown in Section 2.5.1 (from 1050 psig to 1059 psig).

3.1.5B & C, As described in Section 2.5.1, increase CRD charging water header SR3.1.5.1 minimum pressure values (from 1520 psig to 1540 psig), to maintain the SR3.9.5.2 same calculated additional (calculated vs. required for scram function) 3.10.8 f scram time design margin.

SR3.10.8.6 B3.1.5 B3.9.5 B3.10.8 3.1.7 Consistent with the power uprate ATWS analysis basis in Section 9.3.1, SR3.1.7.3 increase the (C)(E) product minimum value from 413 to 570.

9908030324 990730 PDR ADOCK 05000458 P PDR

Location Description of Change SR3.1.7.7 As discussed in Section 6.5, to maintain the SLCS test pressure basis, B3.1.7.7 increase the SLCS Surveillance test pressure by the same amount (30 psi) as the SRV setpoint pressure increase shown in Table 5-1 (from 1220 psig to 1250 psig).

Table 3.3.1.1-1 Revise Reactor Vessel Steam Dome Pressure - High scram trip allowable value consistent with the discussion in Section 5.3.1 and analytical limit in Table 5-1 SR3.3.4.2.4 Consistent with the nominal operating reactor dome pressure increase shown in Table 1-2 and the discussion in Section 5.3.2, increase the ATWS recirculation pump trip pressure allowable value (from 1135 psig to 1165 psig), based on the analytical limit in Table 5-1.

SR3.3.6.4.3a, b Consistent with the increased nominal reactor dome pressure shown in SR3.4.4.1 Table 1-2 and the discussion in Section 5.3.3, increase the SRV relief, B3.4.4 LLS function and spring safety lift settings as shown in Table 5-1.

B3.6.1.6 SR3.5.3.3 As discussed in Section 3.8, to be consistent with the nominal reactor dome pressure change, increase the RCIC Surveillance test high pressure value by the same amount as the nominal reactor dome pressure increase shown in Table 1-2 (from 1045 psig to 1075 psig).

Table 3.3.6.1-1 Consistent with the increased rated steam flow shown in Table 1-2 and the discussion in Section 5.3.4, revise the Main Steam Line Flow - High pressure isolation trips, based on the analytical limits shown in Table 5-1.

3.4 As discussed in Section 3.4, based on the ratio of the current rated to 3.4.1 B uprated % power values (100/105), decrease the thermal power value by 100/105 for single loop operation (SLO), to maintain the same SLO absolute thermal power range (83% to 79%).

SR3.4.6.1 For isolation valve leakage, increase the base reactor coolant system pressure by the same amount as the nominal reactor dome pressure increase shown in Table 1-2 (from 1010 psig to 1040 psig, and from 1040 psig to 1070 psig).

Figure 3.4.11-1 Replace the Minimum Reactor Pressure Vessel Metal Temperature vs.

SR3.4.11-1 Reactor Vessel Pressure curves, as discussed in Section 3.3.1.1, with B3.4.11 those shown in Figures 3-2a and 3-2b.

Location Description of Change i

3.4.12 Increase the reactor steam dome operating pressure LCO by the same SR3.4.12.1 amount as the nominal operating dome pressure increase shown in Table B3.4.12 1-2 (from 1045 psig to 1075 psig). This value is the basis for the initial value used in the reactor overpressure protection analysis described in Section 3.2.

I II. AFFECTED PAGE LIST The following pages have been revised.

I Operating License Page 3, Item C.(1)

Technical Specification Page Definitions; Rated Thermal Power 1.0-5 Frequency; Example 1.4-2 & l.4-3 1.0-26,27, & 28 Reactor Core SLs 2.0-1 Control Rod Scram Times 3.1-14 Control Rod Scram Accumulators 3.1-16,17 Standby Liquid Control System 3.1-20,21,22 Average Planar Linear lleat 3.2-1 Generation Rate j Minimum Critical Power Ratio 3.2-2 Linear lleat Generation Rate 3.2-3 I l

Technical Specification Pg

)

RPS Instrumentation 3.3-2, 3, 8 ATWS-RPT Instrumentation 3.3-31 ;

Primary Containment and Drywell 3.3-53 Isolation Instrumentation Relief and LLS Instrumentation 3.3-67 i Recirculation Loops Operating 3.4-1 Jet Pumps 3.4-9 i

1 SRVs 3.4-10 I RCS PIV Leakage 3.4-16

)

RCS P/r Limits 3.4-28,29,32 Reactor Steam Dome Pressure 3.4-33 i

RCIC System 3.5-11 Main Turbine Bypass System 3.7-14 Control Rod Operability-Refueling 3.9-7 SDM Test-Refueling 3.10-19,22 i

I I

l

Bases Pag.e Reactor Core SLs B2.0-3 Control Rod Scram Accumulators B3.1-29, 30, 31 SLC System B3.1-43 Recirculation Loops Operating B3.4-3, 5 SRVs B3.4-19 RCS P/r Limits B3.4-53, 54, 55, 61 Reactor Steam Dome Pressure B3.4-62, 63 Low Low Set Valves B3.6-35 Main Turbine Bypass System B3.7-25 Control Rod Operability B3.9-17 SDM Test-Refueling B3.10-38 III. MARKED PAGES See attached.

(3) E01, pursuant to the Act and 10 CFR Part 70, to receive, l possess and to use at any time special nuclear satorial as j reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the )

Final Safety Analysis Report, as supplemented and amended; (4) E01, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to l receive, possess, and use at any time any byproduct, source and special nuclear material as sealed neutron saurces for reactor startup, sealed sources for reactor instruw .tation and radiation monitoring equipment calibration, W as fission detectors in amounts as required; i

($) E01, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to l I receive, possess, and use in amounts as required any byproduct. I source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (6) E01, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to l I

possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C. This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Maximum 6 weer Level g() 3 @

E01 is authorized to o the facility at reactor core power l 1evels not in excess o megawatts thermal (2005 rated power) in accordance wi e conditions specified herein. The items identified in Attachment I to this license shall be completed as specified. Attachment 1 is hereby incorporated into this license. .

(2) Technical Snecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, at, revised through Amendment No. 70 and the Environmental Protection Plan contained in Appendix 8, are hereby incorporated in the license. E01 shall operate the facility in l accordance with the Technical Specifications and the Environmental Protection Plan.

Amendment No. M,79 I

f Defin nions 1.1 1.1 Definitions (continued)

MAXIMUM FRACTION The MFLPD shall be the largest value of the OF LIMITING fraction of limiting power density in the core.

POWER DENSITY (MFLPD) The fraction of limiting power density shall be the LHGR existing at a given location divided by the specified LHGR limit for that bundle type.

MINIMUM CRITICAL POWER The MCPR shall_be the smallest critical power RATIO (MCPR) ratio (CPR) that exists in the core for each class of fuel. The CPR is that power in the assembly that is calculated by application of the appropriate correlation (s) to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.

MODE A MODE shall correspond to any one inclusive combination of mode switch position, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1 1 with fuel in the reactor vessel.

OPERABLE-OPERABILITY A system, subsystem, division, component, or device shall be OPERA 8LE or have OPERABILITY when it is capable of performing its specified safety function (s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, division, component, or device to perform its specified -

safety function (s) are also capable of performing their related support function (s).

RATED THERMAL POWER RTP shall be a total reactor cor eat transfer (RTP) rate to the reactor coolant of 89 t.

j REACTOR PROTECTION The RPS RESPONSE TIME shall be that time val SYSTEM (RPS) RESPONSE from when the monitored parameter exceeds its RPS TIME trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

(continued)

RIVER BEND 1.0-5 Amendment No. 81

Frequency 1.4 1.4 Frequency "

l EXAMPLES EXAMPLE 1.4 1 (continued)

If the interval as specified by SR 3.0.2 is exceeded while i the unit is not in a MODE or other specified condition in the Applicability of the LCO for which performance of the SR is required, the Surveillance must be performed within the

, Frequency requirements.of SR 3.0.2 prior to entry into the MODE or other specified condition. Failure to do so would result in a violation of SR 3.0.4.

l EXAMPLE 1.4-2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Verify flow is within limits. Once within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after j

= 25% RTP i

)

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter i

Example 1.4-2 has two Frequencies. The first is a one time performance Frequency, and the second is of the type shown in Example 1.4-1. The logical connector ' 3 " indicates that both Frequency requirements must be met. E time h,397 rea, power is increased from a power level < 5% TP to

~

q 12 RTP, the Surveillance must be terformed w urs.

n C.82 The use of "once" indicates a single performance will satisfy the specified Frequency (assuming no other Frequencies are connected by " M "). This type of Frequency does not qualify for the erimion allowed by SR 3.0.2.

(continued) l RIVER BEND 1.0-26 Amendment No. 81 l

Frequency 1.4 1.4 Frequency EXAMPLES EXAMPLE 1.4-2 (continued) 23.8 "Thereafter" indicates future performances must be established per SR 3.0.2, but only after a specified condition is first met (i.e., the "once" performan this example). If reactor power decreases to <Q13)RTP, the measurement of both intervals ops. New intervals start upon reactor power reaching % TP.

EXAMPLE 1.4-3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY l

l

..................N0TE------------------

Not required to be rformed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after a 5% TP.

Perform channel adh stment 7 days l

The interval continues whether or not the unit operation is

< 25% RTP between performances.

As the Note modifies the required perfomance of the Surveillance, it is construed to be part of the "specified Frecuenev." .S the 7 day interval be exceeded while A , this Note allows 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after L #I'Nf operation is <

power reaches TP to perform the Surveillance. The rve111ance m stii considered to be within the "specified C23,82 Frequency." Therefore, if the Surveillan 13,8[j performed within the 7 day interval (plus,ce were not the artensinn allowed by SR 3.0.2), but operation was <Q59RTP, it would not constitute a failure of the SR or failure to meet the LCO. Also, no violation of SR 3.0.4 occurs when changing MODES, even with the 7 day Frequency not met, pro ed operation does not exceed 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months with power a 5 TP. g (continued) 2 RIVER BEND 1.0 27 Amendment No. 81

Frequency 1.4 1.4 Frequency EXAMPLES EXAMPLE 1.4-3 (continued) 23.0 Once the unit reaches 5f,RTP, 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months would be allowed for completing the Surveil nce. If the Surveillance were not ,

performed within this 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months interval, there would then be a failure to perform a Surveillance within the specified Frequency, and the provisions of SR 3.'0.3 would apply.

EXAMPLE 1.4 4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY

..................N0TE------------------

Only required to be met in MODE 1.

Verify leakage rates are within limits. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Example 1.4-4 specifies that the requirements of this Surveillance do not have to be met until the unit is in MODE 1. Th interval measurement for the Frequency of this Surveillance continues at all times, as described in Example 1.4-1. However, the Note constitutes an "otherwise stated" exception to the Applicability of this Surveillance.

Therefore, if the Surveillance were not performed within the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (plus the extension allowed by SR 3.0.2) interval, but the unit was not in MODE 1, there would be no failure of the SR nor failure to meet the LCo. Therefore, no violation of SR 3.0.4 occurs when changing MODES, even with the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency exceeded, provided the M00E change was not made into MODE 1. Prior to entering MODE 1 (assuming again that the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency were not met), SR 3.0.4 would require satisfying the SR.

~

RIVER BEND 1.0-28 Amendment No. 81

SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 With the reactor steam dome pressure < 785 osig or core '

flow < 10% rated core flow:

25.

THERMAL POWER shall be s 5% RTP.

2.1.1.2 With the reactor steam dome pressure a 785 psig and core flow a 10% rated core flow:

MCPR shall be a 1.13 for two recirculation loop operation or a 1.14 for single recirculation loop operation.

2.1.1.3 Reactor vessel water level shall be greater than the top of active irradiated fuel.

2.1.2 Reactor Coolant System Pressure SL Reactor steam dome pressure shall be s 1325 psig.

2.2 SL Violations With any SL violation, the following actions shall be completed:

2.2.1 Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , notify the NRC Operations Center, in accordance with 10 CFR 50.72.

2.2.2 Within 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months s:

2.2.2.1 Restore compliance with all SLs; and 2.2.2.2 Insert all insertable control rods.

2.2.3 Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , notify the plant manager and the corporate executive responsible for overall plant nuclear safety.

(continued) l

Values applicable to Cycle 8 operation only.

RIVER BEND 2.0-1 Amendment No. 84 M 99

Control Red Scram Times 3.1.4 Table 3.1.4-1 Control Rod Scram Times

.....................................N0TES....................................

1. OPERABLE control rods with scram times not within the limits of this Table are considered " slow." ,

2. Enter applicable Conditions and Required Actions of LCO 3.1.3, " Control Rod OPERABILITY," for control rods with scram times. > 7 seconds to notch position 13. These control rods are inoperable, in accordance with SR 3.1.3.4, and are not considered " slow."

SCRAM TIMES (a)(b)

(seconds)

REACTOR REACTOR STEAM DOME PRESSURE (C) STEAM PRESSURE (C)

NOTCH POSITION 950 psig sig 43 0.30 0.31

/059 29 0.78 0.84 13 1.40 1.53 S

(a) Maximum scram time from fully withdrawn position, based on de.energization of scram pilot valve solenoids as time zero.

(b) Scram times as a function of reactor steam done pressure when < 950 psig are within established limits.

(c) For intermediate reactor steam dome pressures, the scram time criteria are determined by linear interpolation.

3.1 14 Amenoment No. 81 RIVER BEND

Congrei Rod Scram Accumulators 3.1.5 ACTICNS (continued)

REQUIRED ACTION COMPLETION TIME

, CONDITION B. Two or more control B.1 Restore charging 20 minutes from rod scram accumulators discovery of water 4^ der pressure Condition B inoperable with to a 152U "sig.

reactor steam dome concurrent with pressure z 600 psig. charging water 1540 he pressure

< pstg 8.!E B.2.1 ----.-.-NOTE---------

Only applicable if the associated l control rod scram l

time was within the li its of Table 3.1.4-1 during ,

the last scram time Surveill ance.

Declare the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months associated control rod scram time

" slow."

9.8 B.2.2 Declare the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months associated control rod inoperable.

Verify all control Ismediately upon C. One or more control C.1 discovery of rod scram accumulators rods associated with inoperable with inoperable charging water accumulators are hea pressure reactor steam dome psig pressure < 600 psig. fully inserted. <

gg IFYO (continued) 3.1-16 Amendment No. 81 RIVER BEND

l Centrei Roc

cram Accumulators 3.1.5 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME

c. (continued) C.2 Declare the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months associated control rod inoperable.

D. Required Action and 0.1 - - - - - - - . NO T E - - - - - -- - -

associated Completion Not applicable if all Time of Required inoperable control Action 8.1 or C.1 not rod scram met. accumulators are associated with fully inserted control rods.

Place the reactor lamediately mode switch in the shutdown position.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.5.1 Verify each con 1 rod scram accumulator 7 days pressure is a psi .

iS 4 0 3.1-17 Amendment No. 81 RIVER BEND

\

St.C System 3.1.7 3.1 REACTIVITY" CONTROL SYSTEMS 3.1.7 Standby Liquid Control (SLC) System LCO 3.1.7 Two SLC subsystems shall be OPERABLE.

AFFt.!C.n.!!LITY: MODES 1 and 2.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (C)(E) < 4 A.1 Restore (C)(E) 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 7 73 F70 E 10 days from discovery of failure to meet the LC0 B. One SLC subsystem B.1 Restore SLC subsystem 7 days inoperable for reasons '

to OPERA 8LE status.

other than M -

Condition A.

10 days from discovery of failure to meet the LCO C. Two SLC subsystems C.1 Restore one SLC 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months inoperable for reasons subsystem to 0PERA8LE other than status.

Condition A.

D. Required Action and 0.1 Se in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time not met.

RIVER BEND 3.1-20 Amendment No. 81

SLC System 3.1.7

/

SURVEILLANCE REOUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.7.1 -------------------NOTE--------------------

The minimum recuired available solution '

volume is determined by the performance of SR 3.1.7.5.

Verify available volume of sodium 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months pentaborate solution is greater than or equal to the minimum required available solution volume.

SR 3.1.7.2 Verify temperature of sodium pentaborate 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months solution is a 45'F.

SR 3.1.7.3 ------------------NOTE--------------------

Sodium Pentaborate Concentration (C), in weight percent, is determined by the performance of SR 3.1.7.5. Boron-10 enrichment (E), in atos percent, is determined by the performance of SR 3.1.7.9.

Verify that the SLC System satisfies the 31 days following equation:

(C)(E) a SR 3.1.7.4 Verify continuity of explosive charge. 31 days (continued)

RIVER BEND 3.1-21 Amendment No. 81 1

SLC System 3.1.7 SURVEILLANCERE0if!REMENTS (continued)

SURVE!LLANCE FREQUENCY SR 3.1.7.5 Verify the available weight of Baron-10 is 31 days a 143 lbs. and the percent weight concentration of sodium pentaborate in A, N,,2 solution is s 9.57. by weight, and determine the minimum required available Once within solution volume. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after water or baron is added to solution M!E Once within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after solution temperature is e,stdred to

15'F SR 3.1.7.6 Verify each SLC subsystem manual, power 31 days operated, and automatic valve in the flow path that is not locked, sealed, or otherwise secured in position, is in the correct position, or can be aligned to the .

correct position.

SR 3.1.7.7 Verify each pump develops a flow rate In accordance a 41. gym at a discharge pressure with the a 4, psig. Inservice Testing Program l

SR 3.1.7.8 Verify flow through one SLC subsystem from 18 months on a pump into reactor pressure vessel. STAGGERED TEST BASIS (continued)

RIVER BEND 3.1-22 Amendment No. 81 i

APLHGR 3.2.1 3.2 POWER DISTRIBUT!0N LIMITS 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)

LCO 3.2.1

~

All APLHGRs shall be less than or equal to the limits spectfied in the COLA.

APPLICABILITY: THERMAL POWER = RTP.

23.5 ACTIONS CON 0! TION REQUIRED ACTION COMPLETION TIME A. Any APLHGR not wtthin A.1 Restore APLHGR(s) to limits. within limits.

2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months B. Required Action and 8.1 Reduc associated Completion ERMAL POWER 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to < RTP.

Time not met.

23.E SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.1.1 Verify all APLHGRs are less than or equal Once within to the limits specified in the COLR. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after a RTP y 23.8' 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter RIVER BEND 3.2-1 Amendment No. 81 ewe e e N

MCPR 3.2.2 3.2 POWER DISTRIOl1710N LIMITS 3.2.2 MINIMM CRITICAL POWER RATIO (MCPR)

LCD 3.2.2 All MCPRs shall be greater than or equal to the MCPR operating limits specified in the CDLA. -

APPLICABILITY: THERMAL POWER = 2 RTP.

23.T ACTIONS CON 0!T!0N REQUIRED ACTION COMPLETION TINE A. Any MCPR not within A.1 Restore MCPR(s) to limits. 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months within limits.

B. Required Action and 8.1 Redu ERMAL POWER 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months associated Completion to < RTP.

Time not met.

g, g SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.2.1 Verify all MCPRs are greater than or equal Once within to the limits specified in the COLR. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after a Q RTP 23 8 E

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter RIVER BEND 3.2-2 Amendment No. 81

-.,q,---

LHGR 3.2.3 3.2 POWER DISTRIBUTION LIMITS 3.2.3 LINEAR HEAT GENERATION RATE (LHGR)

LCO 3.2.3 All LHCRs shall be less than or equal to the limits specified in the COLR.

APPLICABILITY: THERMAL POWER a 2 RTP.

23.B ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Any LHGR not within A.!

limits, Restore LNGR(s) to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months within limits.

B. Required Action and 8.1 Redu associated Completion ERMAL POWER 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to < RTP.

Time not met.

23 3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.3.1 Verify all LNGRs are less than or equal to once within the limits specified in the COLR. 12 rs after a RTP 23.8 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter RIVER BEND 3.2 3 Amendment No. 81

4 RPS Instrumentation 3.3.1.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D.1 Enter the Condition Imediately associated Completion referenced in Time of Condition A, Table 3.3.1.1-1 for B, or C not met, the channel.

E. As required by E.1 Reduce THERMAL POWER 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Required Action D.1 to < 405 RTP.

and referenced in Table 3.3.1.1 1.

F. As required by F.1 Reduc ERMAL POWER 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Required Action D.1 to < RTP.

and referenced in 23 3 Table 3.3.1.1-1.

G. As required by G.1 Se in MODE 2. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Required Action D.1 and referenced in Table 3.3.1.1-1.

H. As required by H.1 Be in MDDE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Required Action 0.1 and referenced in Table 3.3.1.1-1.

1. As required by 1.1 Initiate action to Immediately Required Action 0.1 fully insert all and referenced in insertable control Table 3.3.1.1-1. rods in core cells containing one or more fuel assemblies.

1 RIVER BEND 3.3 2 Amendment No. 81

RPS fnstrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS

...............-:...........--------. NOTES------------ ---------.---------. --

1. Refer to Table 3.3.1.1 1 to determine which SRs apply for each RPS Function.

2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and RMuired Act1ons may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associate Function maintains RPS tr1D capability.

SURVEILLANCE FREQUENCY SR 3.3.1.1.1 Perform CHANNEL CHECK. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.3.1.1.2 -.- ---- -------.

NOTE----------------..-

Not required to be perfo 11.12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after THERMAL POWER a 5 RTP.

verify the absolute difference between 23 6 7 days the average power range monitor (APRM) channels and the calculated power s 25 RTP.

Once within SR 3.3.1.1.3 Adjust the flow control trip reference 7 days after card to conform to reactor flow. reaching equilibrium conditions following refueling outage.

SR 3.3.1.1.4 --- -

NOTE - ---------------

Not required to be performed when entering MODE 2 from MODE 1 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering MODE 2.

Perform CHANNEL FUNCTIONAL TEST. 7 days RIVER BEND 3.3 3 Amendment No. && '. : 1^6 4

RPS Instrumenta8 ion 3.3.1.1 !

febte 3.3.1.1 1 (sepe 2 of 3)

Reacter Protectten Systee instrumentation APPL 1CASLE COWIT10Ns NEDEs Os REGUtaO REFERENCED Ofuta CMAmeELS Fe(pl SPECIFIED Pla falP attuleED suaVEILLANCE ALLOWAsLE .

FusCflou Couplf!Dus SYSTEM ACTION 0.1 afoulaE4Nis VALUE l l

2. Average Power aanse mentters (continued)

c. Fised moutron 1 3 sa 3.3.1.1.1 G s 12C1 atP F t um - nish sa 3.3.1.1.2 Sa 3.3.1.1.8 st 3.3.1.1.9 sa 3.3.1.1.11 sa 3.3.1.1.15 ,

sa 3.3.1.1.18 !

d. Inno 1,2 3 m sa 3.3.1.1.8 NA flO3,7 )

f* 3.3.1.1.9 !

sa 3.3.1.1.15

3. aeoctor vessel Stees Dame 1,2 2 sa 3.3.1.1.1 s'?'d rig N

I pressure - m ish sa 3.3.1.1.9 se 3.3.1.1.10 at 3.3.1.1.13 sa 3.3.1.1.15 se 3.3.1.1.18

6 neoctor vessel water 1,2 2 sa 3.3.1.1.1 N a 8.7 inches j Leven - Low, Level 3 sa 3.3.1.1.9 sa 3.3.1.1.10 M 3.3.1.1.13 2 }"h sa 3.3.1.1.15 sa 3.3.1.1.18

5. aeoctor vesset water a t hfp 2 sa 3.3.1.1.1 F s 52.1 inches i,evel - mism, Level 8 sa 3.3.1.1.9 sa 3.3.1.1.10 sa 3.3.1.1.13 sa 3.3.1.1.15 sa 3.3.1.1.18

6. Main steen leetetten 1 8 G Sa 3.3.1.1.9 s 121 closed va t we - Ciseure se 3.3.1.1.13 5 3.3.1.1.15 Ed 3.3.1.1.18 7 erwelt Pressure - Nipi 1,2 2 M sa 3.3.1.1.1 s 1.88 seid M 3.3.1.1.9 se 3.3.1.1.10 sa 3.3.1.1.13 M 3.3.1.1.15 (continued)

RIVER BEND 3.3-8 Amendment No. 81

ATW5-RPT Instrumentation 3.3.4.2

~

SURVEILLANCE FREQUENCY Perform CHANNEL FUNCTIONAL TEST.

92 days SR 3.3.4.2.2 Calibrate the trip units. 92 days SR 3.3.4.2.3 Perform CHANNEL CAL 1BRATION, The 18 months SR 3.3.4.2.4 A11ewable Values shall be:

a. Reactor Vessel Water Level-Low Low, Level 2: 3 -47 inches; and

b. Rea 5 team Does Pressure-High:

8 Psig. ,

77g Perfore LOGIC $YSTEM FUNCTIONAL TEST, 18 months SR 3.3.4.2.5 including breaker actuation.

3.3 31 Amendment No. 81 RIVER SEND'

Primary Containment and Drywell Isolation Instrumentation 3.3.6.1 Table 3.3.6.1 1 (page 1 of 5)

Primary Contairmant and Drywell Isolation Instrumentation APPLICABLE CONDITION $

MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVE!LLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION C.1 REQUIREMENTS VALUE i

1. Main Steam Line Isolation l

a. Reactor vessel Water 1,2,3 2 0 SR 3.3.6.1.1 = -147 inches Level - Low Low Lew, SR 3.3.6.1.2 Level 1 SR 3.3.6.1.3 SR 3.3.6.1.5 SR 3.3.6.1.6 SR 3.3.6.1.7

b. Main h?sa Line 1 2 E SR 3.3.6.1.1 = E" reis Pressure - t ou SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.5 SR 3.3.6.1.6 SR 3.3.6.1.7 q

c. Main Steam Line f low - Hi gh 1,2,3 2 per MSL 0 SR 3.3.6.1.1 SR 3.3.6.1.2 skt Line A j gc/

SR 3.3.6.1.3 s id, SR 3.3.6.1.5 Li SR 3.3.6.1.6 s id, N0/

SR 3.3.6.1.7 Li j

Li l

d. Condenser Vacuun - Low 1,2(83, 2 D SR 3.3.6.1.1 2 7.6 inches SR 3.3.6.1.2 Hg vacuun 3g,3 SR 3.3.6.1.3 i SR 3.3.6.1.5 SR 3.3.6.1.6

e. Main Steam Tmnel 1,2,3 2 D SR 3.3.6.1.1 s 148.5'F Tenperature - High SR 3.3.6.1.2 SR 3.3.6.1.5 SR 3.3.6.1.6

f. Main steam Tunnel Area 1,2,3 2 D SR 3.3.6.1.1 s 145.3'F Tenperature - High (EL. SR 3.3.6.1.2 95ft) SR 3.3.6.1.3 SR 3.3.6.1.5 SR 3.3.6.1.6

g. Main steam Te nel Area 1,2,3 2 D SR 3.3.6.1.1 s 145.3'F Temperature - Migh (El. SR 3.3.6.1.2 114ft) SR 3.3.6.1.3 SR 3.3.6.1.5 SR 3.3.6.1.6

h. Main Steam Line Turbine 1,2,3 2 D SR 3.3.6.1.1 s 111.3*F shield Wall SR 3.3.6.1.2 Tenperature Migh SR 3.3.6.1.3 SR 3.3.6.1.5 SR 3.3.6.1.6 (continued)

(a) With any turbine stop valve not closed.

RIVER BEND 3.3-53 Amendment No. 81

Relief and LLS Instrumentation 3.3.6.4 SURVEILLANCEREhi!REMENTS

.....................................N0TE--.---.------------------------......

When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , provided the associated Function maintains LLS.or relief initiation capability, as applicable. '

SURVEILLANCE FREQUENCY SR 3.3.6.4.1 Perfore CHAfstEL FUNCTIONAL TEST. g2 days SR 3.3.6.4.2 Calibrate the trip unit. 92 days SR 3.3.6.4.3 Perform CHANNEL CALIBRATION. The 18 months Allowable Values shall be:

a. Relief Function ,

i Low: 133 a 15 psig l Medium: 1 1 '+ 3 Htt 6 15 psig l High: g 15 psig

b. LLS Function

Low open: M&& : 15 psig 9 5'(,

close: 94 15 psig gg Medium open: Wpt- a 15 psig close: 9M e 15 psig M High open: M H e 15 psig g143 close: 946 a 15 psig 9g ,

SR 3.3.6.4.4 Perfom LOGIC SYSTEM FUNCTIONAL TEST. 18 months RIVER BEND 3.3-67 Amendment No. 81

r~ .

Recirculation Locos Coerating 3.4 1 3 4 - REACTOR COO.LMT SYSTEM (RCS) l 3 4.1 Recirculation Loops Operating !

LCO 3.4.1 A. Two recirculation loops shall be in operation with matched flows.

QB l B. One recirculation loop shall be in operation with:

1. THERMALPOWERshRTP:

2. Total core flew within limits:

3. LCO 3 2.1. " AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR).* single loop operation limits spec 1 fled in the COLR.

4 LCO 3.2.2. " MIN!M)M CRITICAL POWER RATIO (MCPR)." single loop oueration limits spec 1 fled in the COLR: and i

5. LCO 3.3.1.1 "keactor Protection System (RPS) l Instrumentat1on." Funct1on 2.b (Average Power Range Monitors Flow Blased Simulated Themal Power. High).

A.llowable Value for single loop operation as specified '9 1 the COLR.

AP8L*CABILITY: MODES 1 and 2.

AC'!CNS CCN0! TION REQUIRED ACTION COMPLETION TIME A rec 1rculation loop Jet A.1 Shutcown one recirculation 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months pump flow mismatch not loop.

within limits.

B. THERMAL POWER TP 8.1 R uce THERMAL POWER to s I hour curing single 1 RTP.

eperation. 77 (contin,ec)

.4IVER BEND 3.4 1 Amenament No. M F '.06 4-

Jet Pumps 3.4.3 ,

1 SURVEILLANCE REQU_IREMENTS SURVEILLANCE FREQUENCY 1

l SR 3.4.3.1 ---------

N0TES-------------------

1. Not required to be perforised until .

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after associated recirculation loop is in operation.

2. Not required to performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after > _

RTP. _23,g l Verify at least two of the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months criteria (a, b, and c) are satisfied for each operating recirculation loop:

a. Recirculation loop drive flow versus flow control valve position differs by s I M from established patterns.

b. Recirculation loop drive flow versus total core flow differs by s IM from established patterns.

c. Each jet pump diffuser to lower plenum differential pressure differs by s 2 M from established patterns, or each jet pump flow differs by s .1M from established patterns.

RIVER BEND 3.4-9 Amendment No. 81 t

S/RVs 3.4.4 3.4 REACTOR C000U(T SYSTEM (RCS) 3.4.4 Safety / Relief Valves (S/RVs)

LCO 3.4.4 The safety function of five S/RVs shall be OPERABLE, E

The relief function of four additional S/RVs shall be OPERABLE.

I APPLICA81LITY: MODES 1, 2, and 3.

ACTIONS CON 0! TION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Se in' MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months S/RVs inoperable.

M e

A.2 Se in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months 4

l SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.4.1 Verify the safety function lift setpoints In accordance of the required S/RVs are as follows: with the Inservice Number of Setpoint Testing Program ,

S/RVs insio) 7 !!. 7 r.: e iia d I 36 1 5 a ;;%. n.: : :"O >

12cf t M I

: 155.2 :-f : !! l 4 i, y (continued)

RIVER BEND 3.4-10 Amendment No. 81

RCS PlV Leakage 3.4.6 1

SURVEILLANCE REQUIREMENTS i

SURVEILLANCE FREQUENCY SR 3.4.6.1 ------------------ NOTE--------------------

Only required to be performed in MODES 1 and 2.

Verify equivalent leakage of each RCS PIV In accordance is s 0.5 gpa per nominal inch of valve with Inservice size up to imum of 5 g an RCS Testing Program pressure psig and 1040/psig.

a l

/04() l#W RIVER BEND 3.4 16 Amendment No. 81

L

]

1 RCS P/T Limits 3.4.11 1

ACTIONS (continued) 1 1 ColeITION REQUIRED ACTION COMPLETION TIME

)

C. ........-NOTE--------- C.1 Initiate action to Immediately '

Required Action C.2 restore parameter (s) shall be completed if to within limits.

this Condition is entered. Alf C.2 Determine RCS is Prior to Requirements of the acceptable for entering MODE 2 l LC0 not met in other operation, or 3 than MODES 1, 2, and 3.

l SURVEILLANCE REQUIREMENTS .

SURVEILLANCE FREQUENCY l

SR 3.4.11.1 -------------------NGTE--------------------

Only required to be perfonned during RCS heatup and cooldown operations, and RCS inservice leak and hydrostatic testing. ,

Verify: 30 minutes

a. RCS pressure and RCS temperature are within the limits of Figure 3.4.11-1, and I b. RCS heatup and cooldown rates are l s 100*F in any one hour period 4 i

for core not ceWcw\ atw d ecc e, (continued)

Cn k ed b'mi-hs,.

8C.$ kesivp4HM Coc(d6WM PA4ff 0+

a't S 20*F e n any out. hout P*T'8d

& in sce vic e, \ea.K And y o sMC-4esNng limits, . d RIVER BEND 3.4 28 Amendment No. 81

('

RCS P/T Limits 3.4.11 l SURVEILLANCE REQUIRENENTS (continued)

SURVEILLANCE FREQUENCY SR 3.4.11.2 ----.........-.--- NOTE--------------------

Only required to be met during control rod withdrawal for the purpose of achieving criticality.

Verify RCS pressure and RCS temperature are once within within the E-tt ..:::2) limits specified in 15 minutes Figure 3.4.ml-1. prior to e n rol rod cort cr'k-cctl withdrawal for

=

the purpose of achieving criti:ality SR 3.4.11.3 -------------------NOTE--------------------

only required to be met in MODES 1, 2, 3, and 4 with reactor steam done pressure a 25 psig during recirculation pump start.

Verify the difference between the bottom Once within ;

head coolant temperature and the reactor 15 minutes pressure vessel (RPV) coolant temperature prior to each is as 100* F. sta-tup of a '

recirculation pump i SR 3.4.11.4 -------------. ..--NOTE--------------------

only required to be met in MODES 1, 2, 3, and 4 during recirculation pump start.

Verify the difference between the reactor Once within coolant temperature in the recirculation 15 minutes loop to be started and the RPV coolant prior to each temperature is as 50*F. startup of a recirculation pump (continued) 3.4 29 Amendment No. 81 RIVER BEN 0

NEDC-32778P GE Proprietcry Information Class 111 A B A' C B' C' 1400 ,

I f, INITIAL RTndt VALUES ARE 1300 - - - - - -

-50*F FOR BELTLINE

-- - - - f -- -20*F FOR UPPER VESSEL, AND 1200 -- - - - - --

l ---

10*F FOR BOTTOM HEAD

$ BELTINE CURVES E.1100 -- -- - f- -

I ADJUSTED AS SHOWN:

d EFPY SHIFT (*F) 14 132 b 1000 .

HEATUP/COOLDOWN f

y 900 -

-t- - l-j

--f /

20*F/HR F R CURVE A, 100*F/HR FOR CURVES B&C 800 --- -- --

- /- --- ----

A', B', C'- CORE BELTLINE g / '

/ A, B, C - NON-BELTLINE o 700 - - - -

-/- -

-/-

$ / / A - PRESSURE TEST WITH b

E 600 -

./ - - - -f ---

FUEL IN THE VESSEL i

l E j [

B - NON-NUCLEAR t;: 500 d HEATUP/COOLDOWN f~

7 --

E CORE NOT CRITICAL a

g 400 --

y y 7,u.

-k ,

--~

C - HEATUP/COOLDOWN b h ' ^

l312 PSIG ~ ~ ~

NON-BELTLINE b

200 --- - -_ . . . . . . BELTLINE AT 14 EFPY BOLTUP CURVES A', B',C' 100 . _ , _ . _ _

70*F ARE VALID UP TO 14 EFPY OF OPERATION CURVES A, B, C 0 50 100 150 200 250 300 350 400 ARE VALID UP TO EOL FOR NON-BELTUNE MINIMUM REACTOR VESSEL METAL TEMPERATURE ('F)

Figure 3-2a. Minimum RPV Metal Temperatura vs. Reactor Vessel Pressure For 14 EFPY 3-29

NEDC-32778P GE Proprietery Information Class III AB CA' B' C' 00 INITIAL RTndt VALUES ARE t j j -50*F FOR BELTLINE

' -20*F FOR UPPER VESSEL, j f ,

1300 -

q--- i -4!- - / --_L__ AND 10*F FOR BOTTOM HEAD 1200 -- -- -- -

-t'- ---- BELTINE CURVES 8'

_ ADJUSTED AS SHOWN:

.5P / / EFPY SHIFT (*F) 8.1100 -- -- - -' l- - - - +~ 32 159 !

E / ,

HEATUP/COOLDOWN j

g 1000 -- - - -- - -

i J~

RATE l

a. # 20*F/HR FOR CURVE A, O 900 '

I- 100*F/HR FOR CURVES B&C a / / -I /

g j j j A', B', C' CORE BELTLINE v) 800 -- -- -- L 1 A, B, C - NON-BELTLINE

g ./

I

/ i

/1-7 i A - PRESSURE TEST WITH g 700 --

/- -/ [ FUEL IN THE VESSEL U !

I b

E 600 --- - -

- /--/ B - NON-NUCLEAR t ,1 HEATUP/COOLDOWN E

/ / CORE NOT CRITICAL b 500 - -

E / C - HEATUP/COOLDOWN CORE CRIMAL h400 y- y l{- u.

S 312 PSIGl [ NON-BELTLINE

$g 300 ,

u-

. BELTLINE AT 32 A- EFPY 200 - - -

CURVES A', B',C' ARE VALID UP TO 32 EFPY BOLTUP OF OPERATION 100 _ __

70*F CURVES A, B, C l ARE VALID UP TO 0

EOL FOR NON-BELTLINE O 50 100 150 200 250 300 350 400 MINIMUM REACTOR V5SSEL METAL TEMPERATURE (*F)

Figure 3-2b. Minimum RPV Metal Temperature vs. Reactor Vessel Pressure For 32 EFPY l

3-30

ace perumas aA11 l

/ r A* C-E 1,400

!8 , ! Ra. p l a c.<- mis 4 ig u re. v., ;n

'l e .',!. !!

e e Fr'lu,e s. 3 a s </ 3-LA erf Nii b e - 3 2- 857 P.

.ll. ?* !

: : :: N

!!!'Ni / \

Y I ** '

1.000) -

i: . :l

. le ll l : .: ll l 5 i !! ..!! !! i \ / .

.. i.

l r. r. e . cons 1,,, TLast

/ ll ,

e, *./ .. . .

e #

ann gg,y 1

agg

[e* # e W E8 ETer M.

ano 7 t, [;

. e

a. ,,svun um

= ,- , ,.e.

wrrw l

f / (Colla CHmCMJ .

I 400 I /

sA

.s.

taats e e e ConsMTLSELSETS .

3g3g wrrntarv mar?

A/ aumes s.r.

aspreor ranswue ron tum.

19 ,

200 k

eunwes A,s,ase c musrom I esrpror osanation.

~ ' '

O 100 ED 300 400 000 000 S

GEnimum Reestor Vessef Ifetal Tenipereews (T)

Mgure 3A111(page1 et1) tilnisnuun Teniporature Required vs. RCS Pressure SJ 32 Amendment No. ST. 9-Rrygn agND

RCS Pfr Limits A B A' C B' C' 1500 INITIAL RTndtVALUES ARE 1400 --

i- -

-50*F FOR BELTLINE

' 7

-20*F FOR UPPER VESSEL, f

j 1300

_. f.'-_lf - il -

AND 10*F FOR BOTTOM HE AD g,

- 1200 __ '_ _ _j_ .J RFI TINE CURVES O i ADJUSTED AS SHOWN:

f 1100- ___

. [_. j__fi EFPY SHIFT (*F)

n. [ l j 14 132

.l__ ncm vemvvuuvvvr.

$1000 _

l 1(_ _[I RATE

.J

/_ f__/ 20*F/HR FOR CURVE A 900 _

/ / / 100*F/HR FOR CURVES B&C 800 _ . _ [___ __/ ^ E' C'- CGE EELTUNE

/ / A B, C - NON-BELTLINE h '

E 700 _/ __f _.[ A- PRESSURE TEST WITH g / / j

/

/ FUEL IN THE VESSEL a: 600 / _T E . /

t- B - NON-NUCLEAR 500 __

..[_ _ ./

E / HEATUPEOOLDOWN CORE NOT CRITICAL 400 $ _S _

S _ y O " LOO" 300 .

w 'N9tEERITlat4E a:

200 -

. . . . . . . BELTLINE AT 14 EFPY 100 BOLTUF _

NES A', @

70*F AREVALD UPTO14 EFPY 0 '

CURv'E5 A,5. G 0 50 100 150 200 250 300 350 400 AREVALD UPTO EOL FOR NON-BELTLNE MINIMUM REACTOR VESSEL METAL TEMPERATURE (*Fl Fignre 3.4.11-1 Mininium RPV Metal Temperature vs. Reactor Vessel Pressure For 14 EFPY River Bend 3.4- Amendment

RCS Pfr Limits A B CA' B' C' 1500 INITIAL RTndt VALUES ARE

-50*F FOR BELTLINE 1400 -20*F FOR UPPER VESSEL,

__y

10*F FOR BOTTOM HEAD 1300 ~ --

CELT;NC OUrX0

.. t-I I

f. ADJUSTED AS SHOWN:

$1200 0

/-] ( [

f EFPY SHIFT (*F) 32 159 1100 ._ _, _ ) ,

/ Nt;AIUPLUULUUVVN

( t RATE g

1000 _

_ ,f- --j '

20*F/HR FOR CURVE A a / [ 100*F/HR FOR CURVES B&C

$ 900 _ _j _l--/ A' B'. C'- GORE BELTLINE h 800 / l / A B, C - NON-BELTLINE

_/_ __[ / ]

5 / / / A- PRESSURE TEST WITH o 700 /

_i

/ /

/ FUELIN THE VESSEL g /

E soo .!__/ B - NON-NUCLEAR E *

.. f / / HEATUP/COOLDOWN

/ / CORE NOT CRITICAL h 500 a -

tu a u. u., C HEATUP/COOLDOWN g 400 m ..m

f. f'._

h l312 PSG h h h g MNb

& 300 - -

E . . . . . . . BELTLINE AT 32 EFPY a.

200

~

CURVES A', B',C AREVALD UPTO32 EFP(

100 BOLTUP _ _

OF OPERATON 70*F CURVES A, B, C 0 AREVALD UPTO O 50 100 150 200 250 300 350 400 EOL FOR MBELM MINIMUM REACTOR VESSEL METAL TEMPERATURE (*F) 1 J

Figure 3.4.11-2 Minimum RPV Metal Temperature vs. Reactor Vessel Pressure For 32 EFPY River Bend 3.4- Amendment

Reactor Steam Deme Pressure 3.4.12 3.4 REACTORC0b"LANTSYSTEM(RCS) l 3.4.12 Reactor Steam Dome Pressure 147ir5 LCO 3.4.12 The reactor steam done pressure shall be 04 psig.

APPLICABILITY: MODES 1 and 2. --

ACTIONS REQUIRED ACTION COMPLETION TIME CONDITION A. Reactor steam done A.1 Restore reactor steam 15 minutes pressure not within done pressure to limit, within limit.

Required Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months B.

associated Completion Time not met.

SURVEILLANCE REQUIREMENTS i i

SURVEILLANCE FREQUENCY i

1 Verif reactor steam done pressure is 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.4.12.1 s psig.

i

/

( -

3.4 33 Amendment No. 81 RIVER BEND

RC8C Systeo 3.5.3 SURVEILLANCE R'EOUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.3.1 Verify the RCIC System piping is filled 31 days -

with water from the pump discharge valve to the injection valve.

SR 3.5.3.2 Verify each RCIC System manual, power 31 days operated, and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the ,

correct position.

SR 3.5.3.3 --

.---------.-NOTE----.---.-----------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months -

after reactor steam pressure and flow are adequate to perform the test.

ion Verifv. with RCIC steam supply pressure 92 days sMpsig and a 920 psig, the RCIC pump canTevelop a flow rate a 600 gpa against a system head corresponding to reactor pressure.

SR 3.5.3.4 ----------.----..-.N0TE---.-..-----.-------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure and flow are adequate to perform the test.

Verify, with RCIC steam supply pressure 18 months s 165 psig and a 150 psig, the RCIC pump

t. e develop a flow rate a 600 gpa against a >; sten head corresponding to reactor pressure.

(continued)

RIVER BEND 3.5-11 Amancaent No. 81

Main Turbine Sypass System I 3.7.5 3.7 PLANT SYSTDtS l

3.7.5 Main Turbine 8ypass System LCO 3.7.5 The Main Turbine Bypass System shall be OPERA 8LE.

i i

APPLICA8ILITY: THERMAL POWER a 2 RTP.

l 2.3.3 ACTIONS CONDITIGN REQUIRED ACTION COMPLETION TIME A. Main Turbine Bypass A.1 Restore Main Turbine System inoperable. 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Bypass System to OPERA 8LE status.

8. Required Action and 8.1 Reduc associated Complacion ERMAL POWER 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to < RTP.

Time not met.

23.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.5.1 Verify one complete cycle of each main 31 days turbine bypass valve.

SR 3.7.5.2 Perform a system functional test. 18 months SR 3.7.5.3 Verify the TUR81NE BYPASS SYSTEM RESPONSE 18 months TIME is within limits.

RIVER SEND 3.7 14 Amendment No. 81 g -

Contrei Ace GPERABI'.; Y-Refusiina 3.9.5 3.9 REFUELINGdPERATIONS 3.9.5 Control Rod OPERABILITY-Refueling LCO 3.9.5 Each withdrawn control red shall be OPERABLE.

APPLICABILITY: M00E 5.

ACTIONS t REQUIRED ACTION COMPLETION TIME CONDITION A.1 Initiate action to immeciately A. One or more withdrawn control roos fully insert inoperable. inoperable withdrawn control rods.

l l

l t

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY

. i l ,

SR 3.9.5.1 ------------------NOTE ------------------- !

Not required to uc performed until 7 days after the control rod is withdrawn. 3 i

7 days l Insert each withdrawn control rod at least '

l one notch.

l l

Verify each withdrawn contr rod scram 7 days SR 3.9.5.2 ;

accumulator pressure is a psig.

16 40 l l

l 3.9 7 Amendment No. 81 RIVER BEND

50M Test- Aefueline 3.10.8 3.10 SPECIA1. OPDATIONS 3.10.B SHUTOOWN MARGIN (SDM) Test-Refueling LCO 3.10.8 The reactor mode switch position specified in Table 1.1 1 for MODE 5 may be changed to include the startue/ hot standby position, and operation considered not to be in MODE 2. to allow SDM testing, provided the following requirements are 1 met:

a. LC0 3.3.1.1, " Reactor Protection System (RPS)

Instrumentation," MODE 2 requirements for Function 2.a and 2.d of Table 3.3.1.1-1:

b. 1. LCO 3.3.2.1, " Control Rod Block Instrumentation,'

MODE 2 requirements for Function 1.b of Table 3.3.2.1-1, 91

2. Conformance to the approved control rod sequence for the SDM test is. verified by a second licensed operator or other qualified member of the technical staff;

c. Each withdrawn control rod shall be coupled to the associatad CRD;

d. All control rod withdrawals during out of sequence l control rod moves shall be made in single notch ,

withdrawal mode;

e. No other CORE ALTERATIONS are in progress; and

f. CRD charging water header pressure ahpsig. ,

% 29 ,

M 5 with the reactor mode switch in startup/ hot standby (

APPLICABILITY: l position.

f l1 1

1 Amendment No. 81 l RIVER BEND 3.10-19 j

SOM Test- Refueline 3.10.8 SURVEILLANCE REQUIREMENTS (continued)

FREQUENCY SURVE!LLANCE Verify each withdrawn control roo oces not Each L.se the SR 3.10.8.5 control red is go to the withdrawn overtravel position. withdrawn to

" full out" position ME Prior to satisfying LCD 3.10.8.c requirement after work on control rod or CRD System that could affect coupling i

i 7 days SR 3.10.8.6 Verif CRD charging water header pressure a psig i

Amendment No. 81 RIVER SEND 3.10-22

k M , 2.

Reactor Core SLs B 2.1.1 BASES APPLICABLE 2.1.1.1 Fuel Claddino Inteority (continued) '

SAFETY ANALYSES indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt. With the design peaking factors, this corresponds to a THERMAL POWER > 50% TP. Thus, a THERMAL POWER limit of RTP for eactor pressure < 785 psig is conservat v .

l 0

2.1.1.2 MCPR The fuel cladding integrity SL is set such that no significant fuel damage is calculated to occur if the limit is not violated. Since the parameters that result in fuel damage are not directly observable during reactor operation, the thermal and hydraulic conditions that result in the onset of transition boiling have been used to mark the beginning of the region in which fuel damage could occur.

Although it is recognized that the onset of transition boiling would not result in damage to BWR fuel rods, the critical power at which boiling transition is calculated to occur has been adopted as a convenient limit. However, the uncertainties in monitoring the core operating state and in the procedures used to calculate the critical power result in an uncertainty in the value of the critical power.

Therefore, the fuel cladding integrity SL is defined as the critical power ratio in the limiting fuel assembly for which more than 99.9% of the fuel rods in the core are expected to avoid boiling transition, considering the power distribution l

within the core and all uncertainties.

The MCPR SL is determined using a statistical model that combines all the uncertainties in operating parameters and the procedures used to calculate critical power. The probability of the occurrence of boiline transition is determined using the approved General Electric critical power correlations. Details of the fuel cladding integrity SL calculation are given in Reference 2. Reference 2 also includes a tabulation of the uncertainties used 1n the determination of the MCPR SL and of the nominal values of the parameters used in the MCPR SL statistical analysis.

(continued) g ,0 4

"'"'*"'a o9 m.1 Davition Nn. 0

o -

I

-- Control Red Scram Accumulators l

83. 5 i

BASES ACTIONS A.1 and A.2 (continued) would already be considered " slow" and the further degradation of scram performance with an inoperable accumulator could result in excessive scram times. In this event, the associated control red is declared inoperable (Required Action A.2) and LCO 3.1.3 entered. This would result in requiring the affected control rod to be fully inserted and disarmed, thereby satisfying its intended function in accordance with ACTIONS of LC0 3.1.3.

The allowed Completion Time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is considered reasonable, based on the large number of control rods 4 available to provide the scram function and the ability of l the affected control rod to scram only with reactor pressure at high reactor pressures.

8.1. B.2.1. and 8.2.2 With two or more control rod scram accumulators inoperable and reactor steam done pressure a 600 psig, adequate i pressure must be supplied to the charging water header.

With inadequate charging water pressure, all of the accumulators could become inoperable, resulting in a potentially severe degradation of the scram performance.

Therefore, within 20 sing a from discovery of charging water header pressure < & 4Bpsig concurrent with O, condition B, acequate charging water header pressure must be l restored. The allowed Completion Time of 20 minutes is considered a reasonable time to place a CRD pump into service to restore the charfling header pressure, if required. This Completion 'ine also recognizes the ability of the reactor pressure alone to fully insert all control rods.

The control rod may be declared " slow." since the control rod will still scram using only reactor pressure, but may not satisfy the times in Table 3.1.4 1. Required Action B.2.1 is modified by a Note indicating that declaring the control rod " slow" is only applicable if the associated control scram time was within the Otherwise, limits of Table 3.1.4-1rod the control during the last scram time test.

(continued)

Eb\

- -- -- m -

o i i.to Revision No. O

.l i Jot 1troi *bd kram A sumul C a: . r 63 i i

BASES ACTIONS B.1, 3.2.! and 8.2.2 (continued) would already be considered " slow" and the further degradation of scram performance with an inoperaDie {

accumulator could result in excessive scram times. In this event, the associated control rod is declared inoperable (Required Action 8.2.2) and LCO 3.1.3 entered. This wnuld result in requiring the affected control rod to be fully inserted and disarmed, thereby satisfying its intended function in accordance with ACTIONS of LCO 3.1.3.

The allowed Completion Time of I hour is considered reasonable, based on the ability of only the reactor pressure to scram the control rods and the low probability of a DBA or transient occurring while the affected accumulators are inoperable.

C.1 and C.Z With one or more control rod scram accumulators inoperable and the reactor steam done pressure < 600 psig the pressure supplied to the charging water header must be adequate to ensure that accumulators remain charged. With the reactor steam dome pressure < 600 psig, the function of the accumulators in providing the scram force becomes much more important since the scram function could become severely degraded during a depressurization event or at low reactor pressures. Therefore, inmediatel u on discovery of ,

charging water header oressure < psig, concurrent with

'I 40 Condition C. all control rods associated with inoperable accumulators must be verified to be fully inserted.

Withdrawn control rods with inoperable scram accumulators may fail to scram under these icw pressure conditions. The associated centrol rods must also be declared inoperable within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . The allowed Completion Time of I hour is reasonable for Required Action C.2, considering the low ,

probability of a DBA or transient occurring during the time the accumulator is inoperable.

fcontinued) .

8 3.1-30 Revision No. 0 l RIVER BENO

I 1

.. Control Rod Scram Accumulators B 3.1.5 SASES ACTIONS Q.d ,

(continued)

The reactor mode switch must be immediately placed in the shutdown position if either Required Action and associated Completion Time associated with the loss of the CRD pump (Required Actions B.1 and C.1) cannot be met. This ensures l that all insertable control rods are inserted and that the reactor is in a condition that does not require the active function (i.e., scram) of the control. rods. This ReQutred Action is modified by a Note stating that the Required Action is not applicable if all control rods associated with the inoperable scram accumulators are fully inserted, since the function of the control rods has been performed.

SURVEILLANCE SR 3.1.5.1 REQUIREMENTS SR 3.1.5.1 requires that the accumulator pressure be checked every 7 days to ensure adequate accumulator pressure exists to provide sufficient scram force. The primary indicator of accumulator OPERA 81LITY is the accumulator pressure. A minimum accumulator pressure is specified, below which the capability of the accumulator to perform its intended function becomes degraded and the accumulator is g dered inoperable. The minimum accumulator pressure of umv stg is well below the expected pressure of 1750 psig (rte . 2).

Declaring the accumulator inoperable when the minimum pressure is not maintained ensures that significant ,

degradation in scram times does not occur. The 7 day Frequency has been shown to be acceptable through operating experience and takes into account indications available in the control room.

REFERENCES 1. USAR, Section 4.3.2.5.5.

USAR, Section 4.6.1.1.2.5.3. 1590 2.

3. USAR, Section 5.2.2.2.3.

4. USAR, Section 15.4.1.

otvro aryn 3 3,1 31 Revision No. O

l l

SLC System B 3.1.7 BASES SURVEILLANCE SR 3.1.7.7 REQUIREMENTS .

(continued) Demonstrating each SLC System pump dev a flow rate !

= 41.2 gpm at a discharge pressure a psig ensures that pump performance has not degraded ing the fuel cycle. This minimum pump flow rate requirement ensures that, when combined with the sodium pentaborate solution concentration requirements, the rate of negative reactivity l insertion from the SLC System will adequately compensate for the positive reactivity effects encountered during power 3 j

reduction, cooldown of the moderator, and xenon decay. This i test confirms one point on the pump design curve, and is indicative of overall perforinance. Such inservice inspections confirm component OPERA 8ILITY, trend performance, and detect incipient failures by indicating i abnormal performance. The Frequency of this Surveillance is !

in accordance with the Inservice Testing Progrge.

l SR 3.1.7.8 This Surveillance ensures that there is a functioning flow path from the boren solution storage tank to the RPV, including the firing of an explosive valve. The replacement charge for the explosive valve shall be from the same manufactured batch as the one fired or from another batch i that has been certified by having one of that batch .

I successfully fired. Other administrative controls, such as l those that limit the shelf life of the explosive charges, must be followed. The pump and explosive valve tested should be alternated such that both complete flow paths are tested every 36 months, at alternating 18 month intervals.

The Surveillance may be perfor1med in separate steps to prevent injecting baron into the RPV. An acceptable method for verifying flow from the pump to the RPV is to pump domineralized water from a test tank through one SLC subsystem and into the RPV. In order to pump this water, the test valve 1C41*F031 is open. A system initiation signal (which normally signals the 1C41*F001 storage tank suction valve) is generated with the test valve open and verification is made that the storage' tank suction valve I remains closed. The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the Operating experience has shown these reactor at power.

g (continued) ovveo seyn , , , . , e..4-4-- o- '

Recirculation Loops Op2 rating B 3.4.1 i

BASES APPLICA8LE The recirculation system is also assumed to have sufficient SAFETY ANALYSES flow coastdown characteristics to maintain fuel thermal (continued) margins during abnormal operational transients (Ref. 2). ,

which are analyzed in Chapter 1S of the USAR. I A plant specific LOCA analysis has been performed assuming -

only one operating recirculation loop. This analysts has demonstrated that, in the event of a LOCA caused by a pipe break in the operating rec 1rculation loop. the Emergency Core Cooling System response will provide adequate core cooling, provided the APLHGR requirements are modified accordingly (Ref. 3).

The transient analyses of Chapter 15 of the USAR have also been performed for single rectreulation loop operation l

(Ref. 3) and demonstrate sufficient flow coastdown characteristics to maintain fuel thermal margins during the abnormal operational transients analyzed provided the MCPR recuirements are modified. During single recirculation loop operation. modif1 cation to the Reactor Protection System .

average power range monitor (APRM) instrument setpoints is also required to account for the different relationships between recirculation drive flow and reactor core flow. The APLHGR and MCPR limits for single loop operation are specified in the COLR. The APRM flow biased simulated thermal power setpoint is in LCO 3.3.1.1. " Reactor Protect 1on System (RPS) Instrumentation."

Recirculation loops operating satisfies Criterion 2 of the NRC Policy Statement.

LCO Two recirculation loops are normally required to be in operation with their flows matched within the limits specified in SR 3.4.1.1 to ensure that during a LOCA caused by a break of the piping of one recirculation loop the assumptions of the LOCA analysis are satisfied.

l . Alternatively. with only one recircu' on 1 - in W operation. THERMAL POWER must be TP. th M core flow limitations identified above t be met. modifications to the required APLHGR limits (LCO 3.2.1. " AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)"). MCPR limits (LCO 3.2.2. " MINIMUM l

(continued) l RIVER BEND B 3.4 3 Revson No. 4-8 l .

1 l

Recircolatson Loops Operating 8 3.4.1 BASES ACTIONS B.1

(continued) l Should RTP.duringsingleloopoperationthecoreresponsemaynot! THE a LOCA or trans1ent occur with t

be bounded by the safety analyses. Therefore. only a ted time is allowed to reduce THERMAL POWER to s The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time is based on the low probability of an accident occuring dur1ng this t1me period, on a reasonable time to complete the Required Action, and on frecuent core monitoring by operators allowing changes In THERMAL POWER to be quickly detected.

(continued) l

( RDER BEND B 3.4 5 Revoon No. 4 8

~

l il

- S/RVs B 3.4.4 BASES 23% l

/ \

LC0 .

of the nominal setpoint to account for potential '

(continued) setpoint drift to provide an added degree of conservatism.

Operation with fewer valves OPERABLE than specified, or with setpoints outside the ASME limits, could result in a more severe reactor response to a transient than predicted, possibly resulting in the ASME Code limit on reactor pressure being exceeded.

APPLICA81LITY In MODES 1, 2, and 3, the specified number of S/RVs must be OPERA 8LE since there may be considerable energy in the reactor core and the limiting design basis transients are assumed to occur. The S/RVs may be required to provide pressure relief to discharge energy from the core until such time that the Residual Heat Removal (RHR) System is capable of dissipating the heat.

..... ..s....

. . . . ,.,.,. m~.t y T T o r 1serstus '~.

In MODE 4, decay heat is low enough for the RHR System to ,

I provide adequate cooling, and reactor pressure is low enough !

that the overpressure limit is unlikely to be approached by !

assumed operational transients or accidents. In MODE 5, the reactor vessel head is unbolted or removed and the reactor is at atmospheric pressure. The S/RV function is not needed during these conditions.

1 ACTIONS A.1 and A.2 With less than the minimum number of required S/RVs '

OPERA 8LE, a transient may result in the violation of the ASME Code limit on reactor pressure. If one or more required S/RVs are inoperable, the plant must be brought to a MODE in which the LCD does not apply. To achieve this

, status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and to MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed e' ~

Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.4.4.1 REQUIREMENTS This Surveillance demonstrates that the required S/RVs will open at the pressures assumed in the safety analysis of Reference 2. The demonstration of the S/RV safety function fcentinued) od B

c RCS P/T Limits

~~ B 3J4.11 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.11. RCS Pressure and Temperature (P/T) Limits 4

BASES BACKGROUND All components of the RCS are designed to withstand effects of cyclic loads due to system pressure and temperature changes. These loads are introduced by startup (hentup) and shutdown (cooldown) operations, power transients, and reactor trips. This LCD limits the pressure and temperature changes during RCS heatup and cooldown, within the design assumptions and the stress limits for cyclic operation.

e a

on/ coola h , @ . j{ g e 3.q-} contains { l y - _ t

                                                                    '                     'd
 'd     M lM         e a nd hy dro slin S t.

55MS t P/T limit curve defines an acceptable region for nonnal

  +e3y core not-                   E
                                     I" C rt'.h C A. I 4 "d Col'f-guidance during heatup or cooldown maneuvering, when c r # +1 c Al op"N'               pressure and temperature indications are monitored and compared to the applicable curve to determine that operation is within the allowable region.

The LCO establishes operating limits that provide a margin to brittle failure of the reactor vessel and piping of the reactor coolant pressure boundary (RCPS). The vessel is the component most subject to brittle failure. Therefore, the LC0 liatts apply mainly to the vessel. 10 CFR 50, Appendix G (Ref. 1), requires the establishment of P/T limits for material fracture toughness requirements of the RCP8 materials. Reference 1 requires an adequate margin to brittle failure during nonnal operation, anticipated operational occurrences, and system hydrostatic tests. It mandates the use of the American Society of Mechanical Engineers (ASME) Code, Section III, Appendix G (Ref. 2). The actual shift in the RT,,, of the vessel material will be established periodically by removing and evaluating the irradiated reactor vessel material specimens, in accordance with ASTM E 185 (Ref. 3) and 10 CFR 50, Appendix H (Ref. 4). (continued) 8 3.4-53 Revision No. O RIVER BEND

I RCS P/T Limits B 3.4.11 BASES BACKGROUND The operating P/T limit curves will be adjusted, * (continued) as necessary, based on the evaluation findings and the recommendations of Reference 5. The P/T limit curves are composite curves established by superimposing limits derived from stress analyses of those portions of the reactor vessel and head that are the most restrictive. At any specific pressure, temperature, and temperature rate of change, one location within the reactor vessel will dictate the most restrictive limit. Across the span of the P/T limit curves, different locations are more restrictive, and, thus, the curves are composites of the most restrictive regions. 9e4 y ca_. _ ______,,__,___m,,_ __, ,,, ,, ,,,,,,,,,, h d hhh;h-h ,Eh'[ da]$. git [h = nbE ,

                          'Mx! ; "'- '- *"::;;h th: = ::1 91' : : = : m d.                               The Y     t'--    21   ;-idi--t -e e-5:1 :!te s 'k-                 m tia= e' t'e t. .; M e
                          .....u..--... __. ...-... ..~.. .... ........ ..,a....
                                                                             . _.'m..

t cr$Yi'CAhTheb-'tdet!'t311mits include the Reference 1 requirement that they be at least 40*F above the heatup curve or the cooldown curve and not lower than the minimum pomissible temperature for the inservice leak and hydrostatic testing. The consequence of violating the LCO limits is that the RCS has been operated under conditions that can result in brittle failure of the RCP8, possibly leading to a , nonisolable leak or loss of coolant accident. In the event these limits are exceeded, an evaluation must be performed to determine the effect on the structural integrity of the RCP8 components. The ASME Code, Section XI, Appendix E (Ref. 6), provides a recommended methodology for evaluating an operating event that causes an excursion outside the limits. APPLICA8LE The P/T limits are not derived from Design Basis Accident SAFETY ANALYSES (08A) analyses. They are prescribed during nomal operation to avoid encountering pressure, temperature, and temperature rate of change conditions that might cause undetected flaws to propagate and cause nonductile failure of the RCPB, a condition that is unanalyzed. Reference 7 establishes the methodology for determining the P/T limits. Since the P/T limits are not derived from any DBA, there are no acceptance

                                                                                       -         (continued) p       RIVER BEND                                    B 3.4-54                                Revision No. O

ATTACHMENT A (Bases 83.4.11, page B3.4-54) Figure 3.4.11-1 shows a set of F-T curves for the best-up and cool-down operating conditions at a given EFFY. These curves apply for both the %T and %T locations. When combining pressure and thermal stresses, it is usually necessary to evaluate stresses at the %T location (inside surface Haw) and the %T location (outside surface Haw). This is because the thensal gradient tensile stress oflaterest is in the inner wall during cool-down and is la the outer wau during best-up. However, as a conservative simpiincation, the thermal gradient stress at the %T is assumed to be tensile for both heat-up and cool-down. This resalts la the approach of applying the maximum tensile stress at the %T location. This approach is consenstive for two reasons: 1) the maximum stres is used regardless of Gaw location, and 2) the irradiation efects cause the nuoweble toughness, Kw at %T to be less than that at

 %T for a gives metal temperature. This approach causes no operational dimculties, since the BWR is at steam saturation conditions during normal operation, satisfying the best-up/ cool-dows curve limits.

e e

RCS P/T Limits B 3.4.11 BASES APPLICABLE limits related to the P/T limits. Rather, the P/T limits , SAFETY ANALYSES are acceptance limits themselves since they preclude (continued) operation in an unanalyzed condition. RCS P/T limits satisfy Criterion 2 of the NRC Policy Statement. LCO The elements of this LC0 are:

a. RCS pressure, temperature, and heatup or cooldown rate are within the limits, during RCS heatup, cooldown, and inservice leak and hydrostatic testing.

b. The temperature difference b 4 een the reactor vessel bottom head coolant and the r . ctor pressure vessel (RPV) coolant is within the limit during recirculation pump startup, and during increases in THERMAL POWER or loop flow while operating at low THERMAL POWER or loop flow.

c. The temperature difference between the reactor coolant in the respective recirculation loop and in the reactor vessel meets the limit during recirculation pump startup, and during increases in THERMAL POWER or l loop flow while operating at low THERMAL POWER or loop flow.

d. RCS nra<<ues and temperature are within the l hereecQ ptinHtllimitspriortoachievingcriticality.

e. The reactor vessel flange and the head flange temperatures are within limits when tensioning the reactor vessel head bolting studs. ,

These limits define allowable operating regions and persit a large number of operating cycles while also providing a wide margin to nonductile failure. The rate of change of temperature limits control the thermal gradient through the vessel wall and are used as inputs for calculating the heatup, cooldown, and inservice leak and (continued) 8 3.4 55 Revision No. O k RIVER BENO ,

r RCS P/T Limits B 3.4.11 BASES SURVEILLANCE SR 3.4.11.8 and SR 3.4.11.9 (continued) REQUIREMENTS ' Plant specific test data has determined that the bottom head is not subject to temperature stratification with natural circulation at power levels as low as 30% of RTP and with any single loop flow rate greater than 50% of rated loop flow. Therefore, SR 3.4.11.8 and SR 3.4.11.9 have been modified by a Note that requires the Surveillance to be met only when THERMAL POWER or loop flow is being increased when the above conditions are not met. The Note for SR 3.4.11.9 further limits the requirement for this Surveillance to exclude comparison of the idle loop temperature if the idle loop is isolated from the RPV since the water in the loop can not be introduced into the remainder of the reactor coolant system. j REFERENCES 1. 10 CFR 50, Appendix G.

2. ASME, Boiler and Pressure Vessel Code, Section III, Appendix G.

3. ASTM E 185-82, " Standard Practice for Conducting Surveillance Tests For Light-Water Cooled Nuclear Power Reactor Vessels," July 1982.

4. 10 CFR 50, Appendix H.

5. Regulatory Guide 1.99, Revision 2, May 1988. .

6. ASME, Boiler and Pressure Vessel Code, Section XI, Appendix E.

7. M"^ 21775 ^, "Tr;;;i;r.t 77;;;;r: "'::: '"::ti ng

                                                       ..;; Ter :"/";,'

fr::=r; T;.p.n;;; 4... 3:: r ;r 1^7".

8. USAR, Section 15.4.4.

I R.G Can %d B. T. Smbd " /o s r. P.u er L(pv.sk valu AW 'Repov T- d., E.d ev g 3 o n mWs L c..h er % A Sh-W. n G E . T4.s.,K N o , # 2., o Reachey V esse.\ Fva c.h re. 7~. 4% e.s , " G G. - bl C, Sn n 3~* s e, C A pf%,.y h q g (G E - N.E - M 7 - 00 b 81 s 3.4-61 Revision No. 3-4 RIVER 8EMO k

Reactor Steam Dome Pressure 8 3.4.12 8 3.4~ REACTOR COOLANT SYSTEM (RCS) 8 3.4.12 Reactor Steam Dome Pressure 8ASES 8ACKGROUND The reactor steam done pressure is an assumed value in the detentination of compliance with reactor pressure vessel overpressure protection criteria and is also en assumed initial condition of Design Basis Accidents (DBAs) and transients. ,g APPLICA8LE The reactor steam dome pressure of s hesig is an SAFETY ANALYSES initial condition of the vessel overpressure protection analysis of Reference 1. This analysis assumes an initial maximum reactor steam dome pressure and evaluates the l response of the pressure relief system, primarily the safety / relief valves, during the limiting pressurization ; transient. The determination of compliance with the overpressure criteria is dependent on the initial reactor steam dome pressure; therefore, the limit on this pressure ensures that the assumptions of the overpressure protectfos analysis are conserved. Reference 2 also assumes an initial reactor steam done pressure for the analysis of 08As and transients used to determine the i;mits for fuel cladding integrity MCPR (see Bases for LC0 3.2.2, " MINIMUM CRITICAL POWER RATIO (MCPR)") and 1% cladding plastic strain (see 8ases for LC0 3.2.1, " AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)"). , Reactor steam dome pressure satisfies the requirements of Criterion 2 of the NRC Policy Statement. LCO The ecified reactor steam done pressure limit of 10 W psig ensures the plant is operated within the assumptions of the vessel overpressure protection analysis. Operation above the limit may result in a transient response more severe than analyzed. APPLICA81LITY In MODES 1 and 2, the reactor steam done pressure is required to be less than or equal to the limit. In these MODES, the reactor may be generating significant st m . and events which may challenge the overpressure limits are possible. (continued) 8 3.4-62 Revision No. 0 RIVER BEMO

c _ Reactor Steam Dome Pressure

~~

B 3.4.12 1 BASES APPLICA61LITY In MODES 3, 4, and 5, the limit is not applicable because (continued) the reactor is shut down. In these MODES, the reactor - pressure is well below the required limit, and no anticipated events will challenge the overpressure limits. ACTIONS M With the reactor steam done pressure greater than the limit, prompt action should be taken to reduce pressure to below the limit and return the reactor to operation within the l bounds of the analyses. The 15 minute Completion Time is i reasonable considering the importance of maintaining the l pressure within limits. This Completion Time also ensurus that the probability of an accident while pressure is greater than the limit is minimal. I M , If the reactor steam done pressure cannot be restored to within the limit within the associated Completion Time, the ; plant must be brought to a MODE in which the LC0 does not apply. To achieve this status, the plant must be brought to . at least MODE 3 within 12 hours. The allowed Completion Time of 12 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems. . SURVEILLANCE SR 3.4.12.1 REQUIREMENTS

           /075       Verifi)ationthatreactorsteamdomepressureispsig
                       %04                                                     ensures tha vessel overpressure protection analysis are met. Operating experience has shown the 12 hour Frequency to be sufficient for identifying trends and verifying operation within safety analyses assumptions.

REFERENCES 1. USAR, Section 5.2.2.1.

2. USAR, Section 15.

8 3.4-63 Revision No. O RIVER BEND

LLS Valves B 3.6.1.6 8 3.6 CONTAllmihTSYSTEMS B 3.6.1.6 Low-Low Set (LLS) Valves 1 BASES i BACKGROUND The safety / relief valves (S/RVs) can actuate either in the relief mode, System mode, orthe safety the LLS mode, the Automatic Depressurizati mode. t In the LLS mode (one of the Q-A power ano stemactuated assemblytmodes of operation), a pneumatic diaphragm

  *M %                  valve allows a differential pressure to develop acro main valve piston and thus opens the main valve.

ab4 The main as low as 0 psig. valve can be maintained open with valv p aM The pneumatic operator is arranged so

                  . that its malfunction will not prevent the valve disk from lifting if steam pressure           inlet pressure exceeds the safety mode setpoints.

Five of the S/RVs are equipped to provide the LLS function. The LLS logic causes two LLS valves to be opened at a lower pressure than the relief or safety mode pressure setpoints and causes all the LLS valves to stay open longer, such that actuations. reopening of more than one S/RV is prevented on subseq shortsetpoint. relief duration S/RV cycles with valve actuation at Each S/RV discharges steam through a discharge line and quencher to a location near the bottom of the suppression pool, which causes a load on the suppression pool wall. Actuation at lower reactor pressure results in a lower load. APPLICABLE SAFETY ANALYSES The LLS relief mode functions to ensure that the containment design basis of one S/RV operating on

                      ' subsequent actuations" is met Ref. 1). In other words, sultiple simultaneous openings o(f S/RVs (following the                ;

initial opening) and the corresponding higher loads, are avoided. The safety analysis demonstrates that the LLS ' functions to avoid the induced thrust loads on the S/RV discharge line resulting free " subsequent actuations' of the S/RV during Design Basis Accidents Further:nore, the LLS function justifies the primary c(DRAs). ontainment analysis 1 only on the initial actuation for DRAs. assumption that m Even though five (continued) RIVER BENO 8 3.8 35 Revision No. O p '

Main Turbine Bypass System B 3.7.5 B 3.7 PLANT SYSTEMS B 3.7.5 Main Turbine Bypass System BASES BACKGROUND The Main Turbine Bypass System is designed to control steam pressure when reactor steam generation exceeds turbine requirements during unit startup, sudden load reduction, and cooldown. It allows excess steam flow from the reactor to the condenser without goin hrough the turbine. The bypass capacity of the sys of the Nuclear Steam Supply Aem rated steam flow. udden load reductions within the 9' ga capacity of the steam bypass can be accommodated without reactor scram. The Main Turbine Bypass System consists of a two valve chest connected to the main steam lines between the main steam isolation valves and the turbine stop valves. Each of these valves is sequentially operated by hydraulic cylinders. The bypass valves are controlled by the pressure regulation function of the Turbine Pressure Regulator and Control System, as discussed in the USAR, Section 7.7.1.4 (Ref. 1). The bypass valves are normally closed, and the pressure regulator controls the turbine control valves, directing all steam flow to the turbine. If the speed governor or the load limiter restricts steam flow to the turbine, the pressure regulator controls the system pressure by opening the bypass valves. When the bypass valves open, the steam flows from the bypass chest, through connecting piping, to the pressure breakdown assemblies, where a series of orifices are used to further reduce the steam pressure before the steam enters the condenser. APPLICABLE The Main Turbine Bypass System is assumed to function during SAFETY ANALYSES the design basis feedwater controller failure, maximum demand event, described in the USAR, Section 15.1.2 (Ref. 2). Opening the bypass valves during the pressurization event mitigates the increase in reactor vessel pressure, wnich affects the MCPR during the event. An inoperable Main Turbine Bypass System may result in an MCPR penalty. The Main Turbine Bypass System satisfies Criterion 3 of the NRC Policy Statement. (continued) l RIVER BEND B 3.7-25 Revision No. 0 l

Control Rod OPERABIL;1. efusiing

                    ~~                                                              B 3.9.5 BASES    (continur1)

The withdrawn LCO Each withdrawn control rod must be OPERA 8LE. contr-' rod is considered OPERA 8LE if the scram accumulator ps1g and the control roc is capable of I f VO pressure is a@Hy being automatica inserted upon receipt of a scram s1gnal. ; Inserted control rods have already completed their reactivity control function, and therefore are not required

           -           to be OPERA 8LE.

APPLICA8ILITY During MODE 5, withdrawn control rods must be OPERABLE to ensure that in a scram the control rods will insert and . provide the required negative reactivity to maintain the reactor subcritical. For MODES 1 and 2. control red requirements are found in LCO 3.1.2. " Reactivity Anomalies." LCO 3.1.3. " Control Rod ! OPERA 8ILITY " LCO 3.1.4. " Control Rod Scras Times.* and 3 During MODES LC0 3.1.5. " Control Rod Scram Accumulators.' and 4. control rods are not able to be withdrawn since the reactor mode switch is in shutdown and a control rod block is applied. This provides adequate requirements for control rod OPERASILITY during these conditions: ACTIONS M With one or more withdrawn control rods inoperable, action must be immediately initiated to fully insert the inoperable control rod (s). Inserting the control rod (s) ensures that

l the shutdown and scram capabilities are not adversely affected. Actions must continue until the inoperable control rod (s) is fully inserted.

SURVEILLANCE st 3.s.l.1 and sR 3,s.5.2 REQUIREMENTS Ouring MODE 5. the OPERA 8ILITY of control rods is primarily required to ensure that a withdrawn control rod will automatically insert if a signal requiring a reactor shutdown occurs. Because no explicit analysis exists for the shutdown function automatic shutdown during refueling,l rod is capable of is satisfied automatic inse if the withdrawn controand the associated pressure is a psig. (continued 1

                                                    -- E Revision No. O

, RIVER BEND 8 3.9-17

SOM Test--Refusiing B 3.10.8 BASES SURVEILLANCE SR 3.10.8.6 REQUIREMENTS CRD charging water neader pressure verification is performed (continued)- to ensure the motive fores is availableA to scramaccumulator minimum the control rods in the event of a scraa signal. pressure is specified, below which the capability of the accumulator to perform its intended function becomes 4 The degraded and the accumu11 tor is considered inoperable. '

                                                        """stg 1s well below the minimusaccumulatorpressureof'*'"$e#7   day Frequency has        540 expected pressure of 1750 psig. i                                     i been shown to be acceptable through operating experience and l

takes into account indications available in the control room. 1 l

1. NEDE-24011-P-A, ' General Electrie Standard Application REFERENCES l for Reactor Fuel, GESTAR II' (latest approved l

revision). - l

2. Letter. T.A. Pickens (BWROG) to G.C. Lainas (NRC),

                             ' Amendment 17 to General Electric Licensing Topical Report NEDE-24011-P-A,' August 15, 1988.

Revision No. O RIVER BEMD 8 3.10-33 m

[ l ENCLOSURE 4 ENTERGY OPERATIONS,INC. RIVER BEND STATION (RBS) ENVIRONMENTAL ASSESSMENT POWER UPRATE PROJECT Identification ofProposedAction River Bend Station (RBS) Unit 1 is currently licensed to operate at a core thermal power level of 2894 MWt. This proposed license amendment will increase the licensed core thermal power to 3039 MWt, or 105 percent of the current maximum steady state power level. Need for Proposed Action The proposed action permits an increase in the licensed core thermal power from 2894 MWt to 3039 MWt and provides the flexibility to increase the potential electrical output of RBS. Environmental Assessment of the Proposed Action The proposed amendment allowing power up-rate operation will not have a significant impact on the environment and does not constitute an un-reviewed environmental question. The radiological assessment of power up-rate operation is addressed in Enclosure and is summarized below. Details of the non-radiological assessment of the impact of power up-rate are provided. Radiological Environmental Assessment Adequate margin exists for the proposed power up-rate without exceeding regulatory limits for radiological effects. Power up-rate is not expected to have an adverse impact on the previous radiological environmental analyses; thus, revision of these analyses is not required. Existing Technical Specifications limits on radiological effluents will be maintained. Enclosure provides the power up-rate safety analyses report for RBS, as well as an l assessment of the radiological effects of power up-rate operation during both normal and i postulated accident conditions. Sections 8.1 and 8.2 discuss the potential effect of power l up-rate on the liquid and gaseous radwaste systems. Sections 8.3,8.4 and 8.5 discuss the 4 potential effect of power up-rate on radiation sources within the plant and radiation levels during normal and post-accident conditions. Section 9.2 presents the results of the ! l calculated whole body and thyroid doses at the exclusion area boundary and the low population zone that might result from the postulated design basis radiological accidents. All offsite doses remain below established regulatory limits for power up-rate operation. l L

Non-Radiological Environment Assessment Since power up-rate will not significantly change the methods of generating electricity or of handling any influents from the environment or effluents to the environment, no new or different environmental impacts are expected. The detailed evaluation presented below conci des that non-radiological parameters affected by power up-rate will remain within the boundice conditions cited in the Final Environmental Statement (FES), which states that no significant environmental impact will result from operation of RBS. This conclusion remains valid for power up-rate. Note; The Environmental Report - Operating License Stage (ER-OLS), as part of the original application for the operating license for the operation of RBS, was for Units 1 and 2. The environmental impacts presented in this report were conservatively based on two-unit operation. The environmental operating permit applications and subsequent approved permits are also { based on two-unit operation. The cancellation of Unit 2 was not announced until January 5,1984. i The FES evaluated the non-radiological impact at a maximum design reactor power level of 2894 MWt per unit. The parameters evaluated in the Environmental Report and the subsequent FES (References 1 and 2) were re-evaluated at 3039 MWt to determine whether the proposed change is significant relative to adverse environmental impact (Env). The table provides a comparison of environmental-related operating parameters at I rated and up-rated power. Parameter Permit Level Current Uprated

Flow mgd 8.1 5.9 5.9 Discharge Temp. OF 97/99 91/93 98.5 Chlorine mg/L 0.2 0 0 ave / day Note: Reference Enclosure 7 Table 6.2 The Circulating Water System design flow rate is the primary basis for determining makeup water for the cooling towers. Other factors affecting tower makeup are tower performance and meteorological conditions. Based on the review of cooling tower performance parameters associated with power up-rate, the design flow rate of the cooling towers will not change. Makeup requirements may increase slightly due to increased heat load on the towers and the associated increase in evaporation. The increase in makeup due to consumptive water use (evaporation) will not be significant and is enveloped by the river water withdrawal rates discussed in the FES and the rates approved in the current National Pollutant Discharge Elimination System (NPDES) and Louisiana Department of Enviromnental Quality (LADEQ) permits.

r: Changes in cooling tower blowdown rate and cooling tower chemistry as a result of the uprate are not significant. Any changes in blowdown rate and cooling tower cycles of concentration resulting from uprated power operation are enveloped by the existing

l. design criteria discussed in the FES.

l The losses or discharges to the atmosphere are n_ot o expected to be altered as a result of up- ! rate, except for evaporation from the cooling towers. As a result ofincreased heat rejection from the condensers, there will be an additional heat load on the cooling towers. For the same atmospheric conditions, there will be an increase in tower evaporation and drift. Due to this, there may be slight increases in drift deposition rates. Slight increases in evaporation and drift will not have any adverse environmental impacts nor will they afTect any environmental permits.

' l Cooling tower blowdown temperature associated with power uprate operation will increase slightly, thereby producing a slight increase in average and maximum river discharge temperature. A review of the increase in the river discharge temperature relative to the conclusions of the FES and thermal studies required to support licensing of the plant, indicates the slight temperature increase is not significant and is within the present permit limits. The slight temperature increase does not significantly impact the size of the thermal i mixing zone for the thermal effluent and does not alter the conclusions of the FES ' relative to thermal impacts. An application for a new Louisiana Pollutant Discharge , Elimination System (LPDES) has been submitted to LADEQ requesting an increase in average and maximum water discharge temperature due to present and future power up- ) rates. It is anticipated that this will be approved. No significant change in discharge flow rate, velocity, or chemical composition will occur due to the proposed power up-rate. Power rp-rate does not impact the discharge characteristics upon which the NPDES Permit is based. No notification, changes, or other action relative to the NPDES or LADEQ permits are required. No change in the groundwater withdrawal required to supply the RBS service water or fire protection system will result from the proposed up-rate. The evaluation also considered the flow rate required by the liquid radwaste system due to the proposed up-rate. No significant change in liquid radwaste quantities or activity le /els which would increase the required radwaste dilution flow are expected. 1 , l l l l l

Conclusions l EOI concludes that the proposed up-rate will not result in a significant adverse environmental J impact and is not an unreviewed environmental question. Based on the above evaluation, the plant operating parameters impacted by the proposed power up-rate remain within the bounding conditions on which the FES was based. The FES concluded that no significant environmental impact would result from the operation of RBS. This conclusion remains valid for power up-rate. Referencesi

1. River Bend Station Environmental Report - Operating License Stage

2. Final Environmental Statement - Docket No. 50-458

I N, '~

r 1 l lL l ENCLOSURE 5 i

   'ENTERGY OPERATIONS,INC.
  ' RIVER BEND STATION (RBS)                                                                              )

LIST OF COMMITMENTS , EOI has initiated an engineering requests (ER 97-577 & 97-548) to track and control all changes required to implement power uprate. This request has been initiated in accordance within current design and modification processes. This ER will include the implementation of those changes , included in this amendment request, and changes identified in the associated Safety Analysis Report contained in Enclosure 7 of this request. Significant nuclear safety related actions, included in this ER, which must be completed by EOI prior to implementation of the power uprate include:

     . Completion of the evaluations to insure that the safety related electrical equipment required per 10CFR50.49 is properly qualified for the normal, abncrmal and accident conditions postulated at uprated conditions.

Incorporation of the recommendations in NEDC-31753P SER, to reset the safety function lift setpoints for all tested valves to within 1% of the design lift setpoint and increasing the test J

sample size by two valves for each valve found outside of the i3% safety function lift l setpoint. I e Training of the operators will be conducted to insure that the changes to plant operation resulting from the uprate are identified. Training for the remaining plant staff will be included in the ongoing support staff training as required.

  -e     Procedures will be revised to incorporate changes to the plant operation and design resulting from the uprate.

e - EOP's will be revised to incorporate changes to plant and design resulting from the uprate is incorporated.

     . Startup/ Surveillance testing will be conducted as described in Enclosure 7 section 10.4 to demonstrate the ability of plant systems to perform their designed functions under uprated conditions as defined by the start-up test program.

As identified above this ER has been initiated using current plant processes used to initiate and l document engineering analysis, design changes and modifications at RBS. l l

i ENCLOSURE 6 l ENTERGY OPERATIONS,INC. RIVER BEND STATION (RBS) 1 GE AFFIDAVIT FOR NEDC-32778P (See attached.) J l l 1 l l l I

4 GE Nuclear Energy l Donald A. l*lbbert Power Uprate Project ofanager Phone: 408-925-2718 Fax:408-925-1674 GeneralElectric Conpany A1allCode 772 175 Curtner Avenue San Jose, CA 95125 July 29,1999 RBS-99-088 Mr. Amir Shahkarami Entergy Operations Inc. River Bend Station P.O. Box 220 St. Francisville, LA 70775.

Subject:

River Bend Power Uprate Project SAR Affidavit

Reference:

NEDC-32778P; Safety Analysis Reportfor River Bend 5% Power Uprate; July 1999 Amir: This letter is to document that the proprietary information contained in the River Bend SAR (Reference) requires protection under NRC regulations. The River Bend SAR transmittal contains GE-NE proprietary information which is provided under Entergy i Operations Incorporated /GE-NE proprietary information agreement. GE-NE customarily maintains this information ! in confidence and withholds it from public disclosure. { 1 The attached affidavit identifies that the designated information has been handled and classified as proprietary to GE-NE. Along with the affidavit this information is suitable for review by the NRC, GE-NE hereby requests that the designated information be withheld from public disclosure in accordance with the provisions of 10 CFR 2.790. If you have any questions or would like to discuss this topic further, please contact me. l l Sincerely,

                  /
                     /  .

Donald A. Vibbert I

Attachment:

GE Affidavit l L !

l l l General Electric Company l i AFFIDAVIT l l I, George B. Stramback, being duly sworn, depose and state as follows: (1) I am Project Manager, Regulatory Services, General Electric Company ("GE") and have been delegated the function of reviewing the information described in ! paragraph (?) which is sought to be withheld, and have been authorized to apply for its withhold.ng. l (2) The inforr ation sought to be withheld is contained in the GE proprietary report i NEDC-32',78P, Safety Analysis Reportfor River Bend 5% Power Uprate, Class III (GE Proprietary Information), dated July 1999. This document, taken as a whole, constitutes a proprietary compilation of information, some of it also independently proprietary, prepared by the General Electric Company. The independently j proprietary elements are identified by light gray shading of the text and tables of the specific material. (3) In making this application for withholding of proprietary information of which it is the owner, GE relies upon the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"), 5 USC Sec. 552(b)(4), and the Trade Secrets Act,18 USC Sec.1905, and NRC regulations 10 CFR 9.17(a)(4), 2.790(a)(4), and l 2.790(d)(1) for " trade secrets and commercial or financial information obtained from l a person and privileged or confidential" (Exemption 4). The material for which l exemption from disclosure is here sought is all " confidential commercial l information", and some portions also qualify under the narrower definition of " trade secret", within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Energy Project v. Nuclear Regulatory Commission. 975F2d871 (DC Cir.1992), and Public Citizen Health Research Group

v. FDA,704F2d1280 (DC Cir.1983).

(4) Some examples of categories of information which fit into the definition of j propnetary information are:

a. Information that discloses a process, method, or apparatus, including supporting l data and analyses, where prevention of its use by General Electric's competitors without license from General Electric constitutes a competitive economic advantage over other companies; GBS-99-7-Af RiverBend PUP SAR. doc Affidavit Page 1

I

I

b. Information which, if used by a competitor, would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product;

c. Information which reveals cost or price information, production capacities, budget levels, or commercial strategies of General Electric, its customers, or its suppliers; . l

! d. .Information which reveals aspects of past, present, or future General Electric customer-funded development plans and programs, of potential commercial value to General Electric; i e. Information which discloses patentable subject matter for which it may be i desirable to obtain patent protection. l Both the compilation as a whole and the marked independently proprietary elements incorporated in that compilation are considered proprietary for the reason described l in items (4)a. and (4)b., above. l (5) The information sought to be withheld is being submitted to NRC in confidence. That information (both the entire body of information in the form compiled in this j document, and the marked individual proprietary elements) is of a sort customarily ! held in confidence by GE, and has, to the best of my knowledge, consistently been I held in confidence by GE, has not been publicly disclosed, and is not available in public sources. All disclosures to third parties including any required transmittals to l NRC, have been made, or must be_ made, pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in i confidence. Its initial designation as proprietary information, and the subsequent j steps taken to prevent its unauthorized disclosure, are as set forth in paragraphs (6) and (7) following. (6) Initial approval of proprietary treatment of a document is made by the manager of j the originating component, the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge. Access to such i documents within GE is limited on a "need to know" basis. (7) The procedure for approval of external release of such a document typically requires ; review by the staff manager, project manager, principal scientist or other equivalent authority, by the manager of the cognizant marketing function (or his delegate), and by the 12 gal Operation, for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside GE are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary agreements. GBS-99-7 Af RiverBend PUP SAR. doc Affidavit Page 2

(8) The information identified by bars in the margin is classified as proprietary because it contains detailed results and conclusions from these evaluations, utilizing l analytical models and methods, including computer codes, which GE has developed, ) obtained NRC approval of, and applied to perform evaluations of transient and l accident events in the GE Boiling Water Reactor ("BWR"). The development and ) ! approval of these system, component, and thermal hydraulic models and computer { codes was achieved at a significant cost to GE, on the order of several million j dollars. l The remainder of the information identified in paragraph (2), above, is classified as proprietary because it constitutes a confidential compilation of information, including detailed results of analytical models, methods, and processes, including computer codes, and conclusions from these applications, which represent, as a whole, an integrated process or approach which GE has developed, obtained NRC approval of, and applied to perform evaluations of the safety-significant changes necessary to demonstrate the regulatory acceptability of a given increase in licensed power output for a GE BWR. The development and approval of this overall j approach was achieved at a significant additional cost to GE, in excess of a million l dollars, over and above the very large cost of developing the underlying individual proprietary analyses. l To effect a change to the licensing basis of a plant requires a thorough evaluation of the impact of the change on all postulated accident and transient events, and all other regulatory requirements and commitments included in the plant's FSAR. The I analytical process to perform and document these evaluations for a proposed power uprate was developed at a substantial investment in GE resources and expertise. The results from these evaluations identify those BWR systems and components, and those postulated events, which are impacted by the changes required to accommodate operation at increased power levels, and, just as importantly, those which are not so impacted, and the technical justification for not considering the latter in changing the licensing basis. The scope thus determined forms the basis for GE's offerings to support utilities in both performing analyses and providing licensing consulting services. Clearly, the scope and magnitude of effort of any attempt by a competitor to effect a similar licensing change can be narrowed considerably based upon these results. Having invested in the initial evaluations and developed the solution strategy and process described in the subject document GE derives an important competitive advantage in selling and performing these services. However, the mere knowledge of the impact on each system and component reveals the process, and provides a guide to the solution strategy. (9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to GE's competitive position and foreclose or reduce the availGility of profit-making opportunities. The information is part of GE's comprehensive BWR technology base, and its commercial value extends beyond the original i development cost. The value of the technology base goes beyond the extensive GBS-99-7-Af RiverBend PUP SAR. doc Affidavit Page 3

a physical database and analytical methodology and includes development of the expertise to determine and apply the appropriate evaluation process. In addition, the technology base includes the value derived from providing analyses done with NRC-approved methods, including justifications for not including certain analyses in applications to change the licensing basis. GE's competitive advantage will be lost if its competitors are able to use the results of the GE experience to avoid fruitless avenues, or to normalize or verify their own process, or to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions. In particular, the specific areas addressed by any document and submittal to support a change in the safety or licensing bases of the plant will clearly reveal those areas where detailed evaluations must be performed and specific analyses revised, and also, by omission, reveal those areas not so affected. While some of the underlying analyses, and some of the gross structure of the process, may at various times have been publicly revealed, enough of both the analyses and the detailed structural framework of the process have been held in confidence that this information, in this compiled form, continues to have great competitive value to GE. This value would be lost if the information as a whole, in the context and level of detail provided in the subject GE document, were to be disclosed to the public. Making such information available to competitors without their having been required to undertake a similar expenditure of resources, including that required to determine the areas that are not affected by a power uprate and are therefore blind alleys, would unfairly provide competitors with a windfall, and deprive GE of the opportunity to exercise its competitive advantagc to seek an adequate return on its large investment in developing its analytical process. GBS-99-7-Af RiverBend PUP SAR. doc Affidavit Page 4

   .                                                                                                               )

STATE OF CALIFORNIA )

                                                    )     ss:

COUNTY OF SANTA CLARA ) George B. Stramback, being duly sworn, deposes and says: !

That he has read the foregoing affidavit and the matters stated therein are true and correct to the best of his knowledge, information, and belief. l Executed at San Jose, California, thisg day of b 1999.

                                                                     /c/                                           ;

1

'~

b Gdrge 5. Sframback General Electric Company i l Subscribed and sworn before me this A day of ddV' 1999.

j vlCKY D.SCHROER > O p q ceT h # 122425: Nokry Putsc-Coffomic H VM b. e l santo curo county i NotaryPliblic, State of California

              -     --    WComm.EWuJun122Xnf GBS-99-7-Af RiverBend PUP SAR. doc                                                           Affidavit Page 5

ENCLOSURE 7 ENTERGY OPERATIONS, INC. RIVER BEND STATION (RBS) GE REPORT NEDC-32778P," POWER UPRATE SAFETY ANALYSIS FOR THE River Bend Station NUCLEAR PLANT," MAY/ JUNE 1999 (See attached.) i i l l}}

ML20210H290

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