Text

Proposed Tech Specs Implementing BWROG Enhanced Option I-A Reactor Stability Long Term Solution as Documented in NEDO-32339,Rev 1, Reactor Stability Long-Term Solution, Enhanced Option I-A
	ML20154H369
Person / Time
Site:	River Bend
Issue date:	10/08/1998
From:	; ENTERGY OPERATIONS, INC.
To:	;
Shared Package
ML20154H360	List: ML20154H369; RBG-44593, Application for Amend to License NPF-47,implementing BWROG Enhanced Option I-A Reactor Stability Long Term Solution as Documented in NEDO-32339,Rev 1, Reactor Stability Long-Term Solution,Enhanced Option I-A;
References
	NUDOCS 9810140159
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Fraction of Core Boiling Boundary (FCBB)

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Page 2 of15 9910140159 981006 7' PDR ADOCK 05000458 p PDRa. ,

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, FCBB J'

3.2.4 3.2 POWER DISTRIBUTION LIMITS 3.2.4 Fraction of Core Boiling Boundary (FCBB)

.LC0 3.2.4' The*FCBB shall be s 1.0.

APPLICABILITY: THERMAL POWER and core flow in the Restricted Region as specified in-the COLR.

' MODE 1 when RPS Function 2.b, APRM Flow Biased Simulated Thermal .

Power-High. Allowable Value is " Setup" as specified in the COLR. .6 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. FCBB'not within limit A. Restore FCBB to within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

for reasons other than . limit.

an unexpected' loss-of feedwater heating or unexpected reduction in core flow.

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-1 RIVER BENo 3.2 6 Amendnent No.

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FCBB I

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3.2.5

.3-ACTIONS -(continued)

CGNDITION REQUIRED ACTION COMPLETION TIME B. Required Action and B.1 Initiate action to exit the Immediately associated Completion Restricted Region.

Time of Condition A not met.

AND 1 OR l

.......-NOTE--------- B.2 Initiate action to return i Required Action B.1 and APRM Flow Biased Simulated Immediately !

Required Action B.2 Thermal- Power -High following exit of shall be completed if Allowable Value to "non- Restricted Region this Condition is Setup" value. J entered due to an l unexpected loss of feedwater heating or unexpected reduction in core flow.

FCBB not within limit l' due to an unexpected loss of feedwater heating or unexpected

. reduction in core flow.

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AlVER SEND 3.2 7 Amerxhent No.

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. FCBB 8

3.2.4

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.'.1 4 --------------- NOTE ---------- --- ---

Not required to be performed until 15 minutes after entry into the Restricted Region if entry was the result of an unexpected transient.

Verify FCBB s 1.0. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months AND Once within 15 minutes following-unexpected transient i

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I RIVER BEND 3.2 6 Amendnent No.

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FCBB B 3.2.4 8 3.2 POWER DISTRIBUTION LIMITS B 3.2.4 Fraction of Core Boiling Boundary (FCBB)

BASES BACKGROUND General Design Criterion 12 requires protection of fuel thermal safety limits from conditions caused by neutronic/ thermal hydraulic instability. Neutronic/ thermal hydraulic instabilities result in power oscillations which could result in exceeding the MCPR Safety Limit (SL). The MCPR SL ensures that at least 99.9%

of the fuel rods avoid boiling transition during normal operation and during an anticipated operational occurrence (A00) (refer to the Bases for SL 2.1.1.2).

The FCBB is the ratio of the power generated in the lower 4 feet of the active reactor core to the power required to produce bulk saturated boiling of the coolant entering the fuel channels. The value of 4 feet above the bottom of the active fuel is set as the boiling boundary limit based on analysis describcd in Section 9 of Reference 1. The boiling boundary limit is established to ensure that the core will remain stable during normal reactor operations in the Restricted Region of the poner and flow map defined in the COLR which may otherwise be susceptible to neutronic/ thermal hydraulic instabilities and therefore the MCPR SL remains protected.

Plar.ned operation in the Restricted Regicn is accommodated by manually establishing the " Setup" values for the APRM flow-biased Simulated Thermal Power- High scram and control rod block functions. The " Setup" Allowable Values of tie APRM Flow-Biased Thermal Power-High Function (refer to LC0 3.'s.l.1. Table 3.3.1.1-

'1. Function 2.b.) are consistent with assum(d operation in the Restricted Region with FCBB s1.0. Operatior with the " Setup" values enables entry into the Restricted Region without a control rod block that would otherwise occur Plant operation with the

" Setup" values is limited as much as practical due to the effects on plant operation required to meet the FCBB limit.

UPLICABLE The analytical methods and assumptions used in establishing SAFETY ANALYSES the boiling boundary limit are presented in Section 9 of Reference 1. Operation with the FCBB s1.0 (i.e., a bulk saturated boiling boundary 24 feet) is ex ected to ensure that operation within the Restricted Region wil not result RIVER BEND B 3.2-12 Revision No. O

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i FCBB l

B 3.2.4 ,

BASES L APPLICABLE in reutronic/ thermal hydraulic instability due to either

. SAFETY ANALYSES' steady-state operation or as the result of an A00 which (continued) initiates and terminates entirely within the Restricted Region. !

Analysis also confirms that A00s initiated from outside the !

Restricted Region (i.e., without an initial restriction on FCBB) i which terminate in the Restricted Region are not expected to '

result in instability. The types of transients specifically evaluated are loss of flow and coolant temperature decrease. which are limiting for the onset of instability (Ref.1).

l Although.the onset of instability does not necessarily occur if I ;

the FCBB is greater than 1.0 in the Restricted Region, bulk '

saturated boiling at the 4 foot boiling boundary limit has been

l. adopted to preclude neutronic/ thermal hydraulic instability during operation.in the Restricted Region. The effectiveness of this limit is based on the demonstration (Ref.1) that with the ;

limit met large margin to the onset of neutronic/ thermal !

hydraulic instability exists and all major state parameters that i

, affect stability have relatively small impacts on stability l performance.

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The FCBB satisfies Criterion 2 of the NRC Policy Statement.

LCO Requiring FCBB s 1.0 ensures the bulk coolant boiling boundary is t 2 4 feet from the bottom of the active core. Analysis (Ref. 1) l has shown that for anticipated operating conditions of core ,

l power. core flow, axial and radial power shapes. and inlet t enthalpy, a boiling boundary of 4 feet ensures variations.in these key parameters do not have a significant impact on ,

stability performance. ;

j Neutronic/ thermal hydraulic instabilities result in power l oscillations which could result in exceeding the MCPR Safety Limit (SL). The MCPR SL ensures that at least 99.9% of the fuel L. rods avoid boiling transition during normal operation and during an A00 (refer to the Bases for SL 2.1.1.2).

APPLICABILITY The FCBB limit is used to prevent core conditions necessary for the onset of instability and thereby preclude neutronic/ thermal hydraulic instability while operating in the Restricted Region ;

l defined in the COLR.

, The boundary of the Restricted Region in the Applicability of i this LC0 is analytically established in terms of thermal l

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. RIVER BEND B 3.2-13 Revision No. 0 1 .

FCBB B 3.2.4 r

BASES' APPLICABILITY power and core flow. The Restricted Region is defined by (continued) the APRM Flow Biaseo S4*u'ated 'her :' 0:r "4;" Control Rod l

Block upscale alarm setpoints, which are a function of reactor recirculation drive flow. The Restricted Region Entry Alarm ,

(RREA)signalisgeneratedbytheFlowControlTrigReference (FCTR) card using the APRM Flow Biased S M 'It:d T c-- or:-

W94 Control Rod Block upscale alarm setpoints. As a result. the RREA is coincident with the Restricted Region boundary when the setpoints are not " Setup." and provides indication of entry into the Restricted Region. However. APRM Flow Biased Si u'ated

'h e -": ' " ~ : - '";h Control Rod Block upscale alarm signals

, provided by the FCTR card, that are not coincident with the Restricted Region boundary, do not generate a valid RREA. The Restricted Region boundary for this LCO Applicability is specified in the COLR.

The FCBB limit is also used to ensure that core conditions, while operating with " Setup" values, remain consistent with analyzed transients initiated from inside and outside the Restricted Region.

When the APRM Flow Biased Si uted " c-~2' ": ':- "i;" Control Rod Block upscale alarm setpoints are " Setup" the applicable setpoints used to generate the RREA are moved to the interior boundary of the Restricted Region to allow controlled operation within the Restricted Region. While the setpoints are " Setup" the Restricted Region boundary remains defined by the normal APRM Flow Biased S 4"u' '+ d *:-* ' or~

"d;" Control Rod Block upscale alarm setpoints.

Parameters such as reactor power and core flow available at the reactor controls, may be used to provide immediate confirmation that entry into the Restricted Region could reasonably have :

occurred.

Operation outside the Restricted Region is not susceptible to neutronic/ thermal hydraulic instability when applicable thermal power distribution limits such as MCPR are met.

i RIVER BEND B 3.2-14 Revision No. o l

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i FCBB B 3.2.4 BASES ACTIONS

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AJ If FCBB is not within the required limit. core conditions necessary for the onset of neutronic/ hydraulic thermal instability may result. Therefore, prompt action should be taken to restore the FCBB to within the limit such that the stability of the core can be assured. Following uncontrolled entry into the 'lestricted Region. prompt restoration of FCBB within limit can be expected if FCBB is known to not significantly exceed the limit. Therefore, efforts to restore FCBB within limit following an uncontrolled entry into the Restricted Region are appropriate I if operation prior to entry was consistent with planned entry or the potential for entry was recognized as demonstraced by FCBB being monitored and known to not significantly exceed the limit.

Actions to exit the Restricted Region are appropriate when FCBB !

can not be expected to be restored in a prompt manner. l Actions to restart an idle recirculation loop. withdraw control rods or reduce recirculation flow may result in approaching i unstable reactor conditions and are not allowed to be used to '

comply with this Required Action. The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time is based on engineering judgment as to a reasonable time to restore the FCBB to within limit. The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time is acceptable based on the availability of the PBDS per Specification 3.3.1.3. " Period Based Detection System" and the low probability of a neutronic/ thermal hydraulic instability event.

B.1 and B.2 l

Changes in reactor core state conditions resulting from an l unexpected loss of feedwater heating or reduction in core flow '

(e.g. . any unexpected reduction in feedwater temperature, i recirculation pump trip [r^~ rr'M" p7 rur M '

recirculation pump down shift to slow speed, or flow control valve closure]) require immediate initiation of action to exit the Restricted Region and return the APRM Flow Biased Simulated )

Thermal Power- High Function (refer to LC0 3.3.1.1. Table l 3.3.1.1-1. Function 2.b.) to the "non-Setup" value. Condition B is modified by a Note that specifies that Required Actions B.1 and B.2 must be completed if this Condition is entered due to an unexpected loss of feedwater heating or reduction in core flow.

This action to exit the Restricted Region is required following unplanned events that occur while operating in the region and can result in significant loss of stability margin. During such unplanned events, adherence to the FCBB limit cannot be assured.

Therefore. continued operation in the restricted Region is not appropriate. The completion of Required Actions B.1 and B.2 is required even though FCBB may be calculated and determined to be within l

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RIVER BEND B 3.2-15 Revision No. O t

FCBB s B 3.2.4 BASES ACTIONS B.1 and B.2 (continued) limit. Core conditions continue to change after an unexpected loss of feedwater heating or reduction in core flow due to transient induced changes with the potential that the FCBB may change and the limit not be met. The potential for changing core conditions, with FCBB not met. is not consistent with operation in the Restricted Region or with the APRM Flow Biased Simulated Thermal Power- High Function " Setup" Therefore. actions to exit the Restricted Region and return the APRM Flow Biased Simulated Thermal Power- High Function to the "non-Setup" value are

! required to be completed in the event Condition B is entered due to an unexpected loss of feedwater heating or an unexpected reduction in core flow.

If operator actions to restore the FCBB to within limit are not l successful within the specified Completion Time of Condition A, reactor operating conditions may be changing and may continue to i l change such that core conditions necessary for the onset of I

neutronic/ thermal hydraulic instability may be met. Therefore, l in the event the Required Action and associated Completion Time of Condition A is not met, immediate action to exit the i Restricted Region and return the APRM Flow Biased Simulated l Thermal Power- High Function to the "non-Setup" value is )

required. '

Exit of the Restricted Region can be accomplished by control rod insertion and/or recirculation flow increases. Actions to restart an idle recirculation loop. withdraw control rods or reduce recirculation flow may result in approaching unstable reactor conditions and are not allowed to be used to comply with this Required Action. The time required to exit the Restricted Region will depend on existing plant conditions. Provided l efforts are begun without delay and continued until the

! Restricted Region is exited, operatinn is acceptable.

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FCBB B 3.2.4 BASES (continued)

SURVEILLANCE SR 3.2.4.1 REQUIREMENTS-Verifying FCBB s 1.0 is required to ensure the reactor is operating within the assumptions of the safety analysis. The I boiling boundary limit is established to ensure that the core '

will remain stable during normal reactor operations in the Restricted Region of the power and flow map defined in the COLR which may otherwise be susceptible to neutronic/ thermal hydraulic instabilities.

FCBB is required to be verified every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months while operating in the Restricted Region defined in the COLR. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency is based on both engineering judgment and recognition of the slow rate of change in power distribution during normal operation.

The second Frequency requires FCBB to be within the limit within 15 minutes following an unexpected transient. The verification of the FCBB is required as a result of the possibility that the unexpected transient results in the limit not being met. The 15 minute frequency is based on both engineering judgment and the ;

availability of the PBDS to provide the operator with information l regarding the potential imminent onset of neutronic/ thermal hydraulic instability. The 15 minute Frequency for this SR is i not to be used to delay entry into Condition B following an l unexpected reduction in feedwater heating, recirculation pump l trip, [ :-:4 ra' at" pu7 rrr M :' recirculation pump down shift j to slow speed, or significant flow control valve closure (small changes in flow control valve position are not considered significant)]. The action to exit the Restricted Region in Condition B is required following unplanned events that occur while operating in the region and can result in significant loss of stability margin. During such unplanned events adherence to the FCBB limit cannot be assured. Therefore, continued operation in the restricted Region is not appropriate.

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FCBB' B 3.2.4 l

t This Surveillance is modified by a Note which allows 15 minutes to verify FCBB following entry into the Restricted Region if the entry was the result of an unexpected transient (i.e., an

! unintentional or unplanned change in core thermal power or core

flow). The 15 minute allowance is based on both engineering I

judgment and the availability of the PBDS to provide the operator with information regarding the potential imminent onset of neutronic/ thermal hydraulic instability. The 15 minute allowance of the Note is not to be used to delay entry-into Condition B if the entry into the Restricted Region was the result of an unexpected reduction in feedwater heating, recirculation pump-L trip. [ ::4 : ':tde p' ; :- b::E recirculation pump down shift ot'o slow speed. or significant flow control valve closure (small changes in flow control valve position are not considered significant)].

l l- BASES (continued)

REFERENCE 51.NEDO 32339 A. Revision 1. " Reactor Stability Long Term Solution:

Enhanced Option I-A." - December 1996 l

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B-B 3.2-18 I

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RPS Instrumentation i

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RPS Instrumentation ~ !

3.3.1.1

' SURVEILLANCE REQUIREMENTS i

......-------..----NOTES------.----...------------..----..-.

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

2. When'a channel is placed in an inoperable status solely for performance of required '

' Surve111ances, 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 RPS trip 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 l

SR 3.3.1.1.2 ----.---...--..-.-NOTE-.--..----....-----

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after THERMAL POWER 2 25% RTP.

Verify the absolute difference between the average 7 days power range monitor (APRM) channels and the l_ calculated power s 2% RTP.

SR 3.3.1.1.3 Adje:t th: ch nn:! t: confer- t: : :libr:ted '!: 7-days 649#415-Adjue.t the flow control trip reference card to Once within 7 days conform to reactor flow. after reachina eauilibr1wn conditions followina refuelina outaae SR 3.3.1.1.4 .----.-.... ..----N3TE..-----------..----

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 aftt'r entering MODE 2.

Perform CHANNEL FUNCTI0hAl TEST.

7 days (continued)

SR 3.3.1.1.5 Perform CHANNEL FUNCTIONAL TEST. 7 days RIVER BEND 3.3-3 Amendment No. 84 100 i

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RPS Instrumentation 1 3.3.1.1 l l

SURVEILLANCE REQUIREMENTS (continued) i SURVEILLANCE FREQUENCY l SR 3.3.1.1.6 Verify the source range monitor (SRM) and Prior to i intermediate range monitor (IRM) channels overlap. withdrawing SRMs I from the fully ,

inserted position l i

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SR 3.3.1.1.7 ------------------NOTE--------------...-. i Only required to be met during entry into MODE 2 I from MODE 1. i Verify the IRM and APRM channels overlap. 7 days l

1 SR 3.3.1.1.0 CaliLrate the local power range monitors. 1000 MWD /T average 1 core exposure l l

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SR 3.3.1.1.9 Perform CHANNEL FUNCTIONAL TEST. 92 days I

SR 3.3.1.1.10 Calibrate the trip units. 92 days l

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SR 3.3.1.1.31 ...--------------NOTES------------------- '

1. Neutron detectors and flow reference transmitters are excluded.

2. For Function 2.a. not required t1 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 af ter entering M0'sE 2.

3. For Function 2.b. the dioital components of the flow control t ri o 184 days reference cards are excluded.

Perform CHANNEL CALIBRATION.

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! SR 3.3.1.1.12 Perform CHANNEL FUNCTIONAL TEST. 18 months RIVER BEND 3.3 4 Amendment No. 81 l

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RPS Instrumentation

3.3.1.1 !

SURVEILLANCE REQUIREMENTS (continued) '

. SURVEILLANCE FREQUENCY i

SR -3.3.1.1.13 .-.....-.---.-.--NOTES.--...--.--.---- ..

1. Neutron detectors are excluded. .

2. For IRMs not required to be performed when entering MODE 2 from MODE 1 untti 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months I after entering MODE 2.

Perform CHANNEL CALIBRATION.

'13 months SR 3.3.1.1.14- Verify the APRM Flow Biased Simulated Thermal 18 months Power-High time constant is within the limits specified in the COLR. I SR. 3.3.1.1.15 Perform LOGIC SYSTEM FUNCTIONAL TEST. 18 months i-(continued)

'SR 3.3.1.1.16 Verify Turbine Stop Valve Closure and Turbine . 18 months Control Valve Fast Closure Trip 011 Pressure - Low functions are not bypassed when THERMAL POWER is a 40% RTP.

j. SR 3.3.1.1.17; Calibrate the flow reference transmitters, 18 months

,SR' 3.3.1.1.18 .........-------..N0TF.S..----....-..-----

1. Neutron detectors are excluded.

2. For Functions 3, 4, and 5 in Table 3.3.1.1 1, the channel sensors are excluded.

.3. For Function 6. "n" equals 4 channels for the 1 H purpose of detennining the STAGGERED TEST ]

BASIS Frequency. 18 months on a )

...............................-........- STAGGERED TEST ;

BASIS Verify the RPS RESPONSE TIME is within limits.

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RIVER BEND- 3.3-5 Amendment No. 81

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'RPS Instrumentation- l

'3.3.1.1

. Table 3.3.1.1-1 (page 1 of 3)

Reactor Protection System Instrumentation APPLICABLE' . CONDITIONS MODES OR REQUIRED' REFERENCED-0THER- -CHANNELS FROM SPECIFIED PER TRIP- REQUIRED SURVEILLANCE ALLOWABLE FUNCTION _ CONDITIONS SYSTEM ACTION D.1 REQUIREMENTS VALUE

1. ' Intermediate Range Monitors

a. Neutron Flux - High 2 3 H SR s 122/125-3.3.1.1.1 divisions SR of full 3.3.1.1.4 scale SR 3.3.1.1.6

'SR 3.3.1.1.7 SR 3.3.1.1.13 SR 3.3.1.1.15 5(a). 3- I SR s 122/125 3.3.1.1.1 divisions SR . of full.

3.3.1.1.5 scale SR -3.3.1.1.13 SR 3.3.1.1.15 b.'Inop 2 3 H SR NA 3.3.1.1.4 SR 3.3.1.1.15 5(a) 3 I SR NA 3.3.1.1.5 SR 3.3.1.1.15

2. Average Power Range Monitors

n. Neutron Flux - High, 2 3 H SR s 20% RTP Setdown 3.3.1.1.1 SR 3.3.1.1.4 SR 3.3.1.1.7 SR 3.3.1.1.8 SR 3.3.1.1.11 L SR 3.3.1.1.15 l

f I-l RIVER BEND 3.3-6 Amendment No. -

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RPS Instrumentation 3.3.1.1-APPLICABLE ~ CONDITIONS l-MODES OR REQUIRED REFERENCED l

OTHER CHANNELS FROM-SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREMENTS VALUE

b. - Flow Biased 1 3 G SR : 0.55 "

l Simulated Thermal 3.3.1.1.1 57t n!P Power - High SR and 3.3.1.1.2 4-4444 SR g (b) 3.3.1.1.3 SR ;

3.3.1.1.8 l SR l 3.3.1.1.9- !

SR 3.3.1.1.11 l SR j 3.3.1.1.14 l SR l 3.3.1.1.15 SR' 3.3.1.1.17 SR 3.3.1.1.18 (continued) 3 I

(a) With any control rod withdrawn from a core cell containing one or more fuel assemblies.

l (b) ^ " rr:51 : V:!ue 1: : 0.55 " Sit R" - h:n rc::t fer cin;1 10 p Operetten per LCO 3.' !. l 7

"Retirce?:tten L p: Oper: ting." Allowable Value$ SDecified in COLR. Allowable value j modification reautred by the COLR due to reouction in feedwater temperature may be l delayed for up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

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RPS Instrumentation B 3.3.1.1 l

BASES APPLICABLE 2.b. Average Power Range Monitor Flow Biased Simulated SAFETY ANALYSES, Thermal Power-High LCO and APPLICABILITY The Average Power Range Monitor Flow Biased Simulated t

(continued) Thermal Power-High Function monitors neutron flux to l

approximate the THERMAL POWER being transferred to the reactor coolant. The APRM neutron flux is electronically l filtered with a time constant representative of the fuel l heat transfer dynamics to generate a signal proportional to the THERMAL POWER in the reactor. The trip level is varied as a function of recirculation drive flow (4.c. , at 'e':-

c c c er:: the :ctpcirt 4: reduced prcpertic" ' te the reductic" da pe":c evperie ced 2: care f'ev' ir reduced "4th 3 #4ved contre' red "2tter") but is clamped at an upper limit that is always lower than the Average Power Range l Monitor Fixed Neutron Flux-High Function Allowable Value.

The Average Power Range Monitor Flow Biased Simulated Thermal Power-High Function provides a general definition of the licensed core power / core tlow operating domain.

prei det protectier againct tranciente '+ere mg- ogy[p snernscnc - 1 mm11, <c,,rh sc +hn ince n r enn4.., + n r he,+4rn i c' 5-f5 Td 7 e Ie d the l ub' l ii3d4;d 5 5 ? If [ 5 d -4 g

! +hs+ +hn MF3D Cf 4r nn+ nven.And ni t ri nn +hnen nirnn+c thn y[bu{opoiL-333 3793qqfir4fr{~, fig {hf~' ~

cut ea uv e:perte and, bec:ure of a 'c':cr trip retpe4rt us11 4r4+4s+n s crrsm hnfern +hn hinh n n, i+ rnr fli e v crrsm

" etyn tc +

( ,c" Tif f bb ^{ bni 'CD smc"3 2

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I NA b + bl e iv (J4 Ce nr+4nN 0i1 tri dN erNsm

( e4mns1 hnfnrn +hn As,nrsnm o ~ .1n r omnnn Mnn 4 + nr C 1 nu Disend f4his+gA Th .h. rm 1 ok N 11 nh rejh knr cg 4n+

crecded.

l During continued operation with only one recirculation loop in service, the APRM flow biased setpoint is required to be l conservatively set (refer to the Bases for LC0 3.4.1, l " Recirculation Loops Operating" for more detailed discussion). The setpoint modification may be delayed .fw up to '.? hour" in accordance with the allowances of LC0 l 3.4.1. After this time, the LC0 3.3.1.1 requirement for l APRM OPERABILITY will enforce the more conservative i setpoint.

(continued) i RIVER BEND B 3.3-8 Revision No. O i

RPS Instrumentation B 3.3.1.1 ,

BASES.

BASES- ,

APPLICABLE 2.b. Average Power Range Monitor Flow Biased Simulated :

SAFETY ANALYSES. Thermal Power-High (continued) l

-LCO. and APPLICABILITY The Average Power Range Monitor Flow Biased Simulated Thermal Power - High Function is not associated with a limiting safety system setting. Operating limits established for the licensed operating domain are used to develop the Average Power Range Monitor Flow Biased Simulated Thermal Power - High Function Allowable Values to provide pre-em]tive reactor scram and prevent gross l

violation of t1e licensed operating domain. Operation outside the licensed operating domain may result in anticipated operational occurrences and Jostulated accidents being initiated from conditions beyond t10se assumed in the safety analysis. Operation within the licensed operating l domain also ensures compliance with General Design Criterion i 12.

General Design Criterion 12 recuires protection of fuel l thermal safety limits from concitions caused by neutronic/

l thermal hydraulic instability. Neutronic/ thermal hydraulic l instabilities result in )ower oscillations which could result in exceeding the iCPR SL.

The area of the core power and flow operating domain L susceptible to neutronic/ thermal hydraulic instability can L

be affected by reactor parameters such as reactor inlet feedwater temperature. lwo complete and independent sets of Average Power Range Monitor Flow Biased Simulated Thermal Power-High Function Allowable Values are specified in 1]Ie COLR. Set 1 (Normal Trip Reference Set) 3rovides profEcfion against neutronic/ thermal hydraulic insta)ility during expected reactor operations. Set 2 (Alternate Trip Reference Set) 3rovides protection against neutronic/ thermal hydraulic insta]ility during reactor operating conditions

requiring added stability protection and is conservative with respect to Set 1. Feedwater temperature values requiring transition between flow control trip reference card sets are specified in the COLR.

(continued) i RIVER BEND B 3.3-9 Revision No. 0

-- . . _ _ _ . _ . . _ .._ ____ __ _ _ _ ._ _ ..__.~ __

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RPS Instrumentation B 3.3.1.1 BASES BASES APPLICABLE 2.b. Average Power Range Monitor Flow Biased Simulated SAFETY ANALYSES, Thermal Power-High (continued) t LCO, and APPLICABILITY In the event of a feedwater temperature reduction. Allowable Value modification (from the Normal Trip Reference Set to the Alternate Trip Reference Set) is required to preserve i the margin associated with the potential for the onset of ,

neutronic/ thermal hydraulic instability which existed prior ?

to the feedwater temperature reduction. The Allowable Value I modification required by the COLR may be delayed up to 12

hours to allow time to adjist and check the adjustment of !

each flow control trip ref.:rence card. At the end of the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months period, the Allowable Value modifications must be complete for all of the required channels or the applicable l Condition (s) must be entered and the Required Actions taken.

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months time period is acceptable based on the low probability of a neutronic/ hydraulic instability event and ,

the continued protection provided by the flow control trip l

reference card. In addition, when the feedwater temperature i reduction results in operation in either the Restricted 1 Region or Monitored Region, the recuirements for the Period.

Based Detection System (LCO 3.3.1.2. Period Based Detection System (PBDS)) provide added protection against neutronic/

thermal hydraulic instability during the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months time 3 period.

The area of the core power and flow operating domain 1

susceptible to neutronic/ thermal hydraulic instability is affected by the value of Fraction of Core Bolling Boundary (LCO 3.2.5. FCBB). "Setu )" and normal ("non-Setup") Average

Eer Range Monitor Flow 31ased Simulated Thermal Power-High Function Allowable Values are specified in the COLR. The normal ("non-Setup") value provides protection against neutronic/ thermal hydraulic instability by preventing operation in the susceptible area of the i operating domain when operating outside the Restricted

Region specified in the COLR with the FCBB limit not required to be met. When the " Setup" value is selected, meeting the FCBB limit provides protection against ;

instability.

(continued)

I RIVER BEND B 3.3-10 Revision No. 0

RPS Instrumentation a

B 3.3.1.1 .

BASES l

BASES- !

~ '

APPLICABLE 2.b. Average Power Range Monitor Flow Blased Simulated SAFETY ANALYSES. Thermal Power-High (continued) !

LCO, and '

APPLICABILITY !

" Setup" and "non-Setup" values are selected by operator manipulation of a Setup button on each flow control trip i reference card. Selection of the " Setup" value is intended only for planned operation in the Restricted Region as ,

specified in the COLR. Operation in the Restricted Region !

with the Average Power Range Monitor Flow Biased Simulated Thermal Power-High Function " Setup" requires the FCBB limit i to be met and is not generally consistent with normal power operation.

The Average Power Range Monitor Flow Biased Simulated {

Thermal Power-High Function uses a trip level generated by the flow control trip reference card based on recirculation I loop drive flow. Proper trip level generation as a function -

of drive flow requires drive flow alignment. This is accomplished by selection of appropriate dip switch i positions on the tlow control trip reference cards (Refer to SR 3.3.1.1.3). Changes in the core flow to drive flow '

l functional relationship may vary over the core flow o]erating range. These changes can result from both gradual l clanges in recirculation system and core components over the l reactor life time as well as specific maintenance performed l

.on these components (e.g.. jet pump cleaning). ]

The APRM System is divided into two grouas of channels with four APRM inputs to each trip system. T1e system is designed to allow one channel in each trip system to be bypassed. Any one Average Power Ra,ge Monitor channel in a trip system can cause the associate' trip system to trip.

Six channels of Average Power Range Monitor Flow Biased Simulated Thermal Power-High, with three channels in each tri) system arranged in one-out-of-three logic, are required to 3e OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal.

In addition, to provide adequate coverage of the entire core, at least 11 LPRM inputs are required for each APRM channel, with at least two LPRM inputs from each of the four axial levels at which the LPRMs are located.

(continued) l l

l RIVER BEND B 3.3-11 Revision No. 0 l

RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 2.b. Average Power Range Monitor Flow Biased Simulated SAFETY ANALYSES. Thermal Power-High (continued)

LCO, and APPLICABILITY Each APRM channel receives one total drive flow signal reprer^"tative c' tote' cerc c'!. The recirculation loop drive flow signals are generated by eight flow units. One flow unit from each recirculation loop is provided to each APRM channel. Total drive flow is determined by each APRM by summing up the flow signals provided to the APRM from the two recirculation loops.

The THERMAL POWER time constant of 6.6 seconds is based on the fuel heat transfer dynamics and provides a signal proportional to the THERMAL POWER.

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pYb0iS6: aEIh.3f+"3tirp"Op3Eti5"5tbtMursusf '6 ggD 9 The Average Power Range Monitor Flow Biased Simulated Thermal Power-High Function is required to be OPERABLE in c'mccrive MODE 1whenthereisthe)ossibilityofgere"3ti""fity.

7"EoM*t 00MED neutronic/tlermal hydraulic instabi The potential to exceed 3".d pct ^"ti 3 ' !y c'rcedi":g the SL a)plicable to high pressure and core flow conditions (MCPR S_). which provides fuel cladding integrity protection.

exists if neutronic/ thermal hydraulic instability can occur.

During MODES 2 and 5. Other IRM and APRM Functions provide protection for fuel cladding integrity.

2.c. Average Power Range Monitor Fixed Neutron Flux-High l

The APRM channels provide the primary indication of neutron flux within the core and respond almost instantaneously to neutron flux increases. The Average Power Range Monitor Fixed Neutrca Flux-High Function is capable of generating a trip signal to prevent fuel damage or excessive RCS pressure. For the overpressurization protection analysis of Reference 2, the Average Power Range Monitor Fixed Neutron Flux-High Function is assumed to terminate the main steam l isolation valve (MSIV) closure event and. along with the (continued)

[ RIVER BEND B 3.3-12 Revision No. 0 l

RPS Instrumentation B 3.3.1.1 l

BASES SURVEILLANCE SR 3.3.1.1.2 l REQUIREMENTS (continued) To ensure that the APRMs are accurately indicating the true !

core average power, the APRMs are calibrated to the reactor l

l power calculated from a heat balance. The Frequency of once per 7 days is based on minor changes in LPRM sensitivity, which could affect the APRM reading between performances of SR 3.3.1.1.8.

A restriction to satisfying this SR when < 25% RTP is

! arovided that requires the SR to be met only at 2 25% RTP 3ecause it is difficult to accurately maintain APRM l indication of core THERMAL POWER consistent with a heat

! balance when < 25% RTP. At low power levels, a high degree i of accuracy is unnecessary because of the large inherent '

l margin to thermal limits (MCPR and APLHGR). At 2 25% RTP, the Surveillance is required to have been satisfactorily performed within the last 7 days in accordance with SR 3.0.2. A Note is provided which allows an increase in THERMAL POWER above 25% if the 7 day Frequency is not met per SR 3.0.2. In this event, the SR must be performed i within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reaching or exceeding 25% RTP. Twelve l hours is based on operating experience and in consideration

of providing a reasonable time in which to complete the SR.

l l

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+h0 acou rgr-tign g ggo r 3 t g] rgfiggtg thg rgggirgy ggtpg4rt 4,,

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I l

I i RIVER BEND B 3.3-28 Revision No. 3-10 i

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l RPS Instrumentation 1 B 3.3.1.1 BASES 1

l

The Average Power Range Monitor Flow Biased Simulated Thermal Power-High Function uses a trip level generated by the flow control trip reference card based on the recirculation loop drive flow. The drive flow is ad: iusted by a digital algorithm according to selected drive flow alignment dip switch settings. This SR sets the flow control trip reference card to ensure the drive flow l

alignment used results in the appropriate trip level being generated from the digital components of the card.

l The Frequency of once following a refueling outage is based on the expectation that any change in the core flow to drive flow functional relationship during power operation would be {

j gradual and that maintenance on recirculation system and ;

core components which may impact the relationship is l expected to be perfonned during refueling outages. The completion time of 7 days af ter reaching equilibrium conditions is based on plant conditions required to perform the test and engineering judgment of the time required to collect and analyze the necessary flow data and the time required to adjust and check the adjustment of each flow control trip reference card. The completion time of 7 days ;

after reaching equilibrium conditions is acceptable based on the low probability of a neutronic/ hydraulic instability event.

(continued)

RIVER BEND B 3.3-29 Revision No. 3-10

A m - _ - , - - L -,e m - hA.6 -+ei cma1 - Ja 4.M i, .a_

RPS Instrumentation.

B 3.3.1.1

! BASES SURVEILLANCE SR 3.3.1.1.11. SR 3.3.1.1.13 and SR 3.3.1.1.17 '

REQUIREMENTS (continued) A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel l adjusted to account for instrument drifts between successive l

calibrations consistent with the plant specific setpoint methodology.

l ;

! For Functions 9 and 10 the CHANNEL CALIBRATION shall include the turbine first stage pressure instruments.

Note 1 states that neutron detectors and flow reference transmitters are excluded from CHANNEL CALIBRATION because ,

! of the difficulty of simulating a meaningful signal l Changes in neutron detector sensitivity are' compensated for I by performing the 7 day calorimetric calibration 1 i

(SR 3.3.1.1.2) and the 1000 MWD /T LPRM calibration against '

! the TIPS (SR 3.3.1.1.8). Calibration of the flow reference l i transmitters is performed on an 18 month Frequency (SR l

3.3.1.1.17). A second Note is provided that requires the

APRM and IRM SRs to be performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months of entering l

MODE 2 from MODE 1. Testing of the MODE 2 APRM and IRM Functions cannot be performed in MODE 1 without utilizing jumpers, lifted leads or movable links. This Note allows entry into MODE 2 from MODE 1 if the associated Frequency is not met per SR 3.0.2. Twelve hours is based on operating experience and in consideration of providing a reasonable

, time in which to complete the SR. The Frequency of l SR 3.3.1.1.11, SR 3.3.1.1.13. and SR 3.3.1.1.17 is based u)on the assumption of the magnitude of equipment drift in t1e setpoint analysis. Note 3 states that the digital components of the flow control trip reference card are l excluded from CHANNEL CALIBRATION of Function 2.b. Average '

Power Range Monitor Flow Biased Simulated Thermal Power-Hi gh . The analog output potentiometers of the flow

, control trip reference card are not excludeu. The flow l control trip reference card has an automatic self-test feature which periodically tests the hardware which performs the digital algorithm. Exclusion of~ the digital components i of the flow control trip ref erence card f rom CHANNEL i CALIBRATION of Function 2.b is based on the conditions required to perform the test and the likelihood of a change in the status of these components not being detected.

(continued)

RIVER BEND B 3.3-33 Revision No.1

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[ Pen.o d Based Detection System (PBDS) r f

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.y . - . - - - , .

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, 3.3:-INS'TRUMENTATION L

3;3.1.3 Period Based' Detection System (PBDS) l LLCO 3.3.1.3

.One channel of PBDS instrumentation shall be OPERABLE.

-AND Each OPERABLE channel of.PBDS instrumentation shall not indicate Hi-Hi DR Alarm, i

I APPLICABILITY: THERMAL POWER and core flow in.the Restricted Region specified in the COLR, l

THERMAL' POWER'and' core flow in the Monitored Region specified in

-the COLR.

i ACTIONS ~

l CON 0lTION REQUIRED ACTION COMPLETION TIME A. Any OPERABLE PBDS A.1: Place the reactor mode Immediately channel indicating Hi- switch in the shutdown Hi DR-Alarm. position.

l (continued) !

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M i N R BEND 3.3-17 Amenchent No.

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

p& .

c PBDS ;

3.3.1.3 i 1

L ' ACTIONS (continued)

CONDITION- REQUIRED ACTION COMPLETION TIME l

B. Required PBDS channel B.1 --------NOTE;-------- l

! inoperable while in -

1 1.

the Restricted Only applicable if RPS l Region. Function 2.b. APRM Flow l- Biased Simulated Thermal l Power-High. Allowable 1 l Value is " Setup".

L ...................... 1 Immediately l t Initiate action to exit .

the Restricted Region-l OR

~ ~ '

. Immediately B.2 Place the reactor !

mode switch in the shutdown position.

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C. -Required PBDS channel C.1 Initiate action to 15 minutes 1 inoperable while in exit the Monitored the Monitored Region. Region.

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l NRBEND '

3.3 18 Amenchent No.

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.. PBDS 3.3.1.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.1.3.1 Verify each OPERABLE channel of PBDS 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months instrumentation not in Hi-Hi DR Alarm.

SR 3.3.1.3.2 Perform CHANNEL CHECK. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months l

SR- 3.3.1.3.3 Perform CHANNEL FUNCTIONAL TEST. 24 months i

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1 i U N R SEND 3.3-19 Amerwhent No.

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PBDS B 3.3.1.3 B 3.3 INSTRUMENTATION B 3.3.1.3 Period Based Detection System (PBDS)

BASES BACKGROUND General Design Criterlon 12 recuires protection of fuel thermal safety li.its from concitions caused by neutronic/

thermal hydraulic instability. 'Neutronic/ thermal hydraulic instabilities can result in power oscillations which could result in exceeding the MCPR Safety Limit (SL). The MCPR SL ensures that at least 99.9% of the fuel rods avoid boiling transition during normal operation and during an anticipated operational occurrence (A00) (refer to the Bases for SL 2.1.1.2).

The PBDS provides the operator with an indication that conditions consistent with a significant degradation in the stability performance of the reactor core has occurred and the potential for imminent onset of neutronic/ thermal hydraulic instability may exist. Indication of such degradation is cause for the operator to initiate an immediate reactor scram if the reactor is being operated in either the Restricted Region or Monitored Region. The Restricted Region and Monitored Region are defined in the COLR.

The PBDS instrumentation of the Neutron Monitoring System consists of two channels. Each of the PBDS channels includes input from a minimum of.8 local power range monitors (LPRMs) within the reactor core. These inputs are continually monitored by the PBDS for variations in the neutron flux consistent with the onset of neutronic/ thermal hydraulic instability. Each channel includes separate local j indication, but share a common control room Hi-Hi DR Alarm. l While this LC0 specifies OPERABILITY requirements only for one monitoring and indication channel of the PBDS. if both are OPERABLE. a Hi-H' JR Alarm from either channel results i in the need for the operator to take actions. 4 1

RIVER BEND B 3.3-40 Revision No. 0 1 1

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I PBDS B 3.3.1.3 BASES' BACKGROUND The primary PBDS component is a card in the Neutron (continued) Monitoring System with analog inputs and digital-processing.

l d' The PBDS card has an automatic self-test feature to periodically test the hardware circuit. The self-test ,

i functions are executed during their allocated portion of the executive loop sequence. Any self-test failure indicating loss of critical function results in a control room alarm.

The inoperable condition is also displayed by an indicating light on the card front panel. A manually initiated internal test sequence can be actuated via a recessed push button. This internal test consists of simulating alarm and inoperable concitions to verify card OPERABILITY, Descriptions of the PBDS are provided in References 1 and 2.

Actuation of the PBDS Hi-H1 DR Alarm is not postulated to occur due to neutronic/ thermal hydraulic instability outside the Restricted Region and the Monitored Region. Periodic 3erturbations can be introduced into the thermal hydraulic 3ehavior of the reactor core from external sources such as recirculation system components.and the 3ressure and feedwater control systems. These perturaations can potentially drive the neutron flux to oscillate within a

~

frequency range expected for neutronic/ thermal hydraulic instability. The presence of such oscillations would be recognized by the period based algorithm of the PBDS and j potentially result in a Hi-Hi DR Alarm. Actuation of the PBDS Hi-Hi DR Alarm outside the Restricted Region and the i Monitored Region would indicate the presence of a source ;

external to the reactor core and are not indications of neutronic/ thermal hydraulic instability.

APPLICABLE Analysis, as described in Section 4 of Reference 1. ,

SAFETY ANALYSES confirms that A00s initiated from outside the Restricted !

Region without stability control and from within the

. Restricted Region with stability control are not expected to result in neutronic/ thermal hydraulic instability, The stability control applied in the Restricted Region (refer to LC0 3.2.5, " Fraction of Core Boiling Boundary (FCBB)") is established to prevent neutronic/ thermal hydraulic (continued) i RIVER BEND B 3.3-41 Revision No. O !

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. _ . _ _ _ _ _ . . _ . _ . _ . _ _ _ _ _ _ _ _ i L PBDS l B 3.3.1.3 BASES

APPLICABLE

l SAFETY ANALYSES L (continued) instability during-operation .in the Restricted Region.

Operation in the Monitored Region is only susceptible to instability under hypothetical operating conditions beyond those analyzed in Reference 1. The types of transients specifically evaluated are loss of flow and coolant temperature decrease which are limiting for the onset of

-instability.

The initial conditions assumed in the analysis are L reasonability conservitive and the immediate post-event .

reactor conditions are sifgnificantly stable. Howerever, these assumed initial conditions do not bound each indiviual parameter which impacts stability performance (Ref.1). The l PBDS instrumentation provides the operator with an f

indication that conditions consistent with a significant i degradation in the stability performance of hte reactor core has occured and the potential for imminent onset of neutronic/ thermal hydraulic instability may exist. Such conditions are only postulated to result from events initiated from initial conditions beyond the conditions I assumed in the safety analysis (refer to Section 4. Ref. 1).

L The PBDS has no safety function and is not assumed to function during any FSAR-design basis accident or transient analysis. However, the PBDS provides the only indication of the imminent onset of neutronic/ thermal hydraulic l instability during operation in regions of the operating

! domain potentially susceptible to instability. Therefore.

L the PBDS is included in the Technical Specifications.

LC0 One PBDS channel is required to be OPERABLE to monitor reactor neutron flux for indications of imminent onset of

^ neutronic/ thermal hydraulic instability. OPERABILITY requires the ability for the operator to be immediately alerted to a Hi-Hi DR Alarm. This is accomplished by the instrument channel control room alarm. The LCO also requires reactor operation be such that the Hi-Hi DR Alarm is not actuated by any OPERABLE PBDS instrumentation channel.

j RIVER BEND B 3.3-42 Revision No. O o

PBDS B 3.3.1.3 APPLICABILITY At least one of two PBDS instrumentation channels is required to be OPERABLE during operation in either the Restricted Region or the Monitored Region specified in the COLR. Similarly, operation with the PBDS Hi-Hi DR Alarm of any OPERABLE PBDS instrumentation channel is not allowed in the Restricted Region or the Monitored Region. Operation in these regions is susceptible to instability (refer to the Bases for LCO 3.2.5 and Section 4 of Ref. 1). OPERABILITY of at least one PBDS instrumentation channel and operation with no indication of a PBDS Hi-Hi DR Alarm from any 0PERABLE PBDS instrumentation channel is therefore required during operation in these regions.

The boundary of the Restricted Region in the Applicability of this LC0 is analytically established in terms of thermal power and core flow. The Restricted Region is defined by the APRM Flow Biased Sim"'ated 7"c-- c':c- "ig" Control Rod Block upscale setpoints, which are a function of reactor recirculation drive flow. The Restricted Region Entry Alarm (RREA) signal is generated by the Flow Control Trip Reference (FCTR) cam sing the APRM Flow Biased Si u'ated 7"er 2' or:c- %b r.rol Rod Block upscale alarm setpoints. As c . u uh , the RREA is coincident with the Restricted Regic c .Jary when the setpoints are not

" Setup." and provides indication of entry into the Restricted Region. However. APRM Flow Biued S"u'ated The-- or!c- "ig" Control Rod Block upscale alarm signals provided by the FCTR card, that are not coincident with the Restricted Region boundary, do not generate a valid RREA.

The Restricted Region boundary for this LCO Applicability is specified in the COLR.

When the APRM Flow Biased Simu'ated T"c-~2' o cc c- "ig" Control Rod Block upscale alarm setpoints are ~ Setup" the a]plicable setpoints used to generate the RREA are moved to t1e interior boundary of the Pestricted Region to allow controlled operation within the Restricted Region. While the setpoints are " Setup" the Restricted Region boundary remains defined by the normal APRM Flow Biased S"u'ated 7"c-~2' or :c- "4;" Control Rod Block upscale alarm setpoints.

n uma u; B-B 3.3-M

1 l PBDS l B 3.3.1.3 BASES (Continued)

Parameters such as reactor power and core flow available at the reactor controls, may be used to provide immediate confirmation that entry into the Restricted Region could reasonably have occurred. ;

The Monitored Region in the Applicability of this LCO is !

analytically established in terms of thermal aower and core :

flow. However, unlike the Restricted Region 30undary the 1 l Monitored Region boundary is not specifically monitored by i plant instrumentation to provide automatic indication of j region entry. Therefore. the. Monitored Region boundary is ;

defined in terms of thermal power and core flow. The )

Monitored Region boundary for this LCO Applicability is i specified in the COLR. l

)

Operation outside the Restricted Region and the Monitored Region is not susceptible to neutronic/ thermal hydraulic instability even under extreme postulated conditions.

ACTIONS A.1 If at any time while in the Restricted Region or Monitored Region, an OPERABLE PBDS instrumentation channel indicates a valid Hi-Hi DR Alarm, the operator is required to initiate an immediate reactor scram. Verification that the Hi-Hi DR Alarm is valid may be performed without delay against 4 another output from a PBDS card observable from tha reactor I controls in the control room prior to the manual reactor i scram. This provides assurance that core conditions leading to neutronic/ thermal hydraulic instability will be _

mitigated. This Required Action and associated Completion i Time does not allow for evaluation of circumstances leading to the Hi-Hi DR Alarm prior to manual initiation of reactor ;

scram. :

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(ColllifiUPd}

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FBDS B 3.3.1.3 B.1 and 8.2 Operation with the APRM Flow Biased Simulated Thermal i Power-High Function (refer to LCO 3.3.1.1. Table 3.3.1.1-1.

Function 2.b.) " Setup" requires the stability control applied in the Restricted Region (refer to LCO 3.2.5) to be l met. Requirements for operation with the stability control l met are established to prevent reactor thermal hydraulic l instability during operation in the Restricted Region. With !

the required PBDS channel inoperable, the ability to monitor i conditions indica +ing the potential for imminent onset of neutronic/ thermal hydraulic instability as a result of unexpected transients is lost. Therefore, action must be immediately initiated to exit the Restricted Region. While i the APRM Flow Biased S "u'ated W 3' o' r "i? Control !

Rod Block upscale alarm setpoints are " Setup." operation in the Restricted Region may be confirmed by use of plant parameters such as reactor power and core flow available at the reactor controls.

Exit of the Restricted Region can be accomplished by control rod insertion and/or recirculation flow increases. Actions to restart an idle recirculation loop, withdraw control rods or reduce recirculation flow may result in unstable reactor conditions and are not allowed to be used to comply with this Required Action. '

The time required to exit the Restricted Region will depend on existing plant conditions. Provided efforts are begun without delay and continued until the Restricted Region is ,

exited. operation is acceptable based on the low probability i of a transient which degrades stability performance 1

(contmued)

U- ti M~ la

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. PBDS B 3.3.1.3 i I

BASES l

1 ACTIONS B.1 and B.2 (continued) i occurring simultaneously with the required PBDS channel l

~ inoperable. 1 Required Action B.1 is modified by a Note that specifies-

.. ~ that initiation of action to exit the Restricted Region only L applies if the APRM Flow Biased Simulated Thermal Power-High L

Function is "Setu)" Operation in the Restricted Region E without the APRM : low Biased Simulated Thermal Power-High l Function " Setup" indicates uncontrolled entry into the Restricted Region. Uncontrolled entry is consistent with the occurrence of unex)ected transients, which, in combination with the a)sence of stability controls being met may result in significant' degradation of stability performance.

WMn the APRM Flow Biased Stu'ated Ther 2' oc':c- "h

! ControlRodBlockupscalealarmsetpointsarenot"SetEJp" I

uncontrolled entry into the Restricted Region is identified by receipt of a valid RREA. Immediate confirmation that the RREA is valid and indicates an actual entry into the Restricted. Region may be performed without delay. Immediate confirmation constitutes observation that plant parameters immediately available at the reactor controls (e.g., reactor power and core flow) are reasonably consistent with entry into the Restricted Region. This immediate confirmation may also constitute recognition that plant parameters are rapidly changing during a transient (e.g.. a recirculation Jump trip) which could reasonably result in entry into the Restricted Region.

For uncontrolled entry into the Restricted Region with the I required PBDS instrumentation channel inoperable, the ability to monitor conditions indicating the potential for imminent onset of neutronic/ thermal hydraulic instability is lost and continued operation is not justified. Therefore.

Required Action B.2 requires immediate reactor scram.

PBDS B 3.3.1.3 C.1 (antinued) b-b d j'Ib f

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m y 99y- 9 .r.ny neq- p e- +w w y vp A4ee "" 'e'

l

, In the Monitored Region the PBDS Hi-Hi DR Alarm provides l -

indication of degraded stability performance. Operation in the Monitored Region is susceptible to neutronic/ thermal l hydraulic instability under postulated conditions exceeding I

those previously assumed in the safety analysis. With the required PBDS channel inoperable, the ability to monitor conditions indicating the potential for imminent onset of neutronic/ thermal hydraulic instability is lost. Therefore, l ac': ion must be initiated to exit the Monitored Region.

Actions to restart an idle recirculation loop, withdraw ccntrol rods or reduce recirculation flow may result in approaching unstable reactor conditions and are not allowed to be used to comply with this Required Action. Exit of the Monitored Region is accomplished by control rod insertion and/or recirculation flow increases. However, actions M reduce recircu bti - cw are allowed provided the Fraction of Core Boiling Boundary (FCBB) is recently (within 15 minutes) verified to be 5: 1.0. Recent verification of '

FCBB being met provides assurance that with the PBDS inoperable, planned decreases in recirculation drive f1nt should not result in significant degradation of core stability performance.

The specified Completion Time of 15 minutes ensures timely operator action to exit the region consistent with the low probability that reactor conditions exceed the initial l conditions assumed in the safety analysis. The time required to exit the Monitored Region will depend on i existing plant conditions. Provided efforts are begun I within 15 minutes and continued until tie Monitored Region is exited, operation is acce) table based on the low probability of a transient w1ich degrades stability performance occurricq simultaneously with the required PBDS channel inoperable.

(conhnued)

H-n u-h

PBDS B 3.3.1.3 BASES SURVEILLANCE SR 3.3.1.3.1 l REQUIREMENTS During o]eration in the Restricted Region or the Monitored i Region t1e PBDS Hi-Hi DR Alarm is relied upon to indicate '

conditions consistent with the imminent onset of neutronic/

thermal hydraulic instability. Verification every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months provides assurance of the proper indication of the alarm 1 during operation in the Restricted Region or the Monitored Region. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency supplements less formal. but i more frequent. checks of alarm status during operation. 1 SR 3.3.1.3.2 Performance of the CHANNEL CHECK every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ensures that a gross failure of instrumentation has not occurred. This CHANNEL CHECK is normally a comparison of the PBDS indication to the state of the annunciator, as well as comparison to the same parameter on the other channel if it is available. It is based on the assumption that the ;

. instrument channel indication agrees with the immediate !

indication available to the operator, and that instrument '

channels monitoring the same parameter should read similarly. Deviations between the instrument channels could be an indication of instrument component failure. A CHANNEL CHECK will detect gross channel failure: thus, it is key to j verifying the instrumentation continues to operate properly between each CHNiNEL FUNCTIONAL TEST. Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability.

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is based on operating experience that demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of I

I aonhnued) 11- 11 1 'r ld l

l l

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

1

--o.- PBDS :

B 3.3.1.3 i

i BASES I SURVEILLANCE SR 3.3.1.3.2 (continued) i REQUIREMENTS channels during normal operational use of 1he displays i associated with the channels required by the LCO.

i SR 3.3.1.3.3 ;

I A CHANNEL FUNCTIONAL TEST is performed for the PBDS to ;

ensure that the entire system will perform the intended function. The CHANNEL FUNCTIONAL TEST for the PBDS includes i' manual initiation of an internal test sequence and ;

verification of appropriate alarm and inop conditions being i reported. ;

Performance of a CHANNEL FUNCTIONAL TEST at a Frecuency of i 24 months verifies the performance of the PBDS anc i associated circuitry. The Frequency considers the plant ,

l conditions required to perform the test.'the ease of !

( performing.the test, and the likelihood of a change in the ,

l system or. component status. The. alarm circuit is designed i to operate for over 24 months with sufficient accuracy on j signal amplitude and signal. timing considering environment, !

initial calibration and accuracy drift (Ref. 2). j i

L REFERENCES _1. NEDO 32339. Revision 1. " Reactor Stability Long ]

i Term Solution: Enhanced Option I-A," December 1996.

2. NED0-32339P-A.~ Supplement 2, " Reactor Stability I Long Term Solution: Enhanced Option I-A Solution {

Design." December 1996.

l i

1 1

f (contmued) l' if-tj .id -PJ k - _ _ _

7.

Recirculation Loops Operating i

l l

I i

i i

l i

i j

l l

Page 5 of 15

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

i I i'

- i L EIA TS inserts ;

i

!. Technical Specification 3.4.1 l

. I LCO l

A. Two recircluation loops shall be in operation with matched flows. l t

OR_

B. One recirculation loop may be in operation provided the following limits are applied when the associated LC0 is applicable:

1. THERMAL POWER < 83% RTP- !

f

2. Total core flow is within limits I

3. LC0 3.2.1. " AVERAGE PLANAR LINEAR HEAT GENERATION RATE l (APLHGR)." single loop operation limits [specified in the COLR]; ;

4. LCO-3.2.2, " MINIMUM CRITICAL POWER RATIO (MCPR)." single !

loop operation limits [specified in the COLR]: and i

5. LC0 3.3.1.1 " Reactor Protection System (RPS)

L Instrumentation." Function 2.b (Average Power Range Monitors !

l Flow Biased Simulated Thermal Power-High). Allowable Value !

c' 78'c 2.2.1.1 1 1: rc:ct for single loop operation as_ l specified in the COLR.

ACTIONS' t l

l i L -

.C. Requirements B.3. B.4 or C.1 Staisfy the requirements 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months B.5 of the LC0 not met. of the LC0 ' j L.

D.

required actions and-

. associated completion times D.1 Be in Mode 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months of conditions A. B or C not met. 1 i

OR-4 No recirculation loops in '

operation.

L I

l'

- _l

l

l 1

Recirculation Loops Operating 3.4.1

3.4 REACTORCOOLANTSYSTEM(RCS) l 3.4.1 Recirculation Loops Operating i

LC0 3.4.1 A. Two recirculation loops shall be in operation with+

4r Matched flowst-and )

i

o. T..a e. .s 1 a.n. .

a

. #.1. a.w .s . A. T..u.rDM

. . A.f D. A.u r.D. ..w. 4. +. L 4. m.

1 4. m.4 4.

. . e. .

OR t

B. One recirculation loop shall be in operation with:

I l l 1. THERMAL POWER s 83% RTP; i

2. Total core flow and THER"^.L P0"ER within limits; t

l 3. Required 14-it: modified for : ingle recirculation l

100p Operation :: Specified in the COLR; and l

4. LC0 3.3.1.1, "R cter Protection Sy:ter (PPS)

Dawa=

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Required !!rit and setpoint =0di'icati n: for single i recirculation 100; Operatien m. y be delayed for up t0 12 hour: :#ter transitter, from t c recirculation 100p !

cperst10n t cingle recirculatten 100p operation.

I APPLICABILITY: MODES 1 and 2.

1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Recirculation loop jet A.1 Shutdown one 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months l pump flow mismatch not recirculation loop.

within limits, i

(continued) 3 f RIVER BEND 3.4 1 Amendment No. 81 97

(

l

, Recirculation Loops Operating 3.4.1 l ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. THERMAL POWER > 83% B.1 Reduce THERMAL POWER 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months RTP during single loop to s 83% RTP.

operation.

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(continued)

RIVER BEND 3.4-2 Amendment No. 81 97

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

? . ,

e ?- ,

L Recirculation Loops Operating H

i 3.4.1 ;

ACTIONS (continued) !

, CONDITION REQUIRED ACTION COMPLETION TIME 1, i

h t

F. '

L " ree'r:21: tic: 1:0p: F.1 Initi t: ::tice t h r: i

4. ---= 44mm . J a.n TurDMal DALa rn l "r"'-"'""' .

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! net perf =^d. ::tpei-t(:) net ::t. !

i !

L SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY L

SR 3.4.1.1: -------------------NOTE--------------------

! Not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months l- after both recirculation loops are 'in i operation.

L _______________________ ..____ .___________ ,

t 1 l

L Verify recirculation loop jet pump flow 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months l mismatch with both recirculation loops in l l operation is:- !

L j l1 a. s 10% of rated core flow when operating o

at < 70% of rated core flow; and ,

1

b. s 5% of rated core flow when operating !

at a 70% of rated core flow.

i, (continued) i r.

RIVER BEND 3.4-3 Amendment No. 81 l-i

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

Recirculation Loops Operating 3.4.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SP. 3.'.'.? Yer'f9: 34-hours

. Tet:' c0re ex : 'St r:ted core f'0 ;

er t

.k . T. u. r.D.M.A.I.

. . - D A.u. r D. .s n. A. +. a. +. .s 1 ..

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D.am.

. . ._ . l. a n a. f. E4,m,.nam. . a O. . A. . 1. .

l t

l l

p.

l l

t I

s 1

RIVER BEND 3.4-4 Amendment No. 81

. _. m _ _ _ _ . . _ _ _ _ _ _ _ . _ . . . _ _ _ . _ _ _ . _ _ _ _ _ . _ . _ _ -

~

Recirculation Loops Operating a

B 3.4.1 l

l B 3.4 ' REACTOR COOLANT SYSTEM (RCS)

L l

l- ~ B 3.4.1 Recirculation Loops Operating' i

BASES I BACKGROUND The Reactor Coelant' Recirculation System is designed to 3rovide a forced coolant flow through the core to remove leat'from the fuel. The forced coolant flow removes more heat from the fuel than would be possible with just natural circulation. The forced flow, therefore, allows operation -

-at significantly higher power than would otherwise be possible. The recirculation system also controls reactivity over a wide span of reactor power by varying the

-recirculation flow rate to control the void content of the moderator. The Reactor Coolant Recirculation System l consists of two recirculation pump loo)s external to the l reactor vessel. These loo)s provide t1e piping path for the L ' driving flow of water to t1e reactor vessel jet pumps. Each external laop contains a two speed motor driven !

recirculation pump, a flow control valve, and associated piping. jet pumps, valves, and instrumentation. The recirculation loops are part of the reactor coolant pressure i L boundary and are located inside the drywell structure. The

l. jet pumps are reactor vessel internals.

L l The recirculated coolant consists of saturated water from the steam separators and dryers that has been subcooled by i incoming feedwater. This water passes down the annulus J between the reactor vessel wall and the core shroud. A i portion of. the coolant flows from the vessel, through the l'

two external recirculation loops, and becomes the driving flow for the jet pumps. Each of the two external l recirculation loops discharges high pressure flow into an-I external manifold, from which individual recirculation inlet lines are routed to the jet pump risers within the reactor vessel. The remaining portion of the coolant mixture in the annulus becomes the suction flow for the jet pumps. This flow enters the jet pump at suction inlets and is accelerated by the driving flow. The drive flow and suction flow are mixed in the jet Jump throat section. The total flow then passes through t1e jet pump diffuser section into l the area- below the core (lower plenum), gaining sufficient L head in the process to drive the required flow upward

through the~ core.

(continued)

I RIVER BEND B 3.4-1 Revision No. 0 l

1

, , , . , . , -- .l

Recirculation loops Operating B 3.4.1 BASES BACKGROUND The subcooled water enters the bottom of the fuel channels (continued) and contacts the fuel cladding, where heat is transferred to the coolant. As it rises, the coolant begins to boil.

creating steam voids within the fuel channel that continue until the coolant exits the core Because of reduced moderation, the steam voiding introduces negative reactivity that must be compensated for to maintain or to increase reactor power. The recirculation flow control allows o)erators to increase recirculation flow and sweep some of t1e voids from the fuel channel, overcoming the necative reactivity void effect. Thus.'the reason for having variable recirculation flow is to compensate for reactivity effects of boiling over a wide range of power generation (i.e. 55 to 100% RTP) without having to move control rods and disturb desirable flux patterns.

Each recirculation loop is manually started from the control room. The recirculation flow control valves provide regulation of individual recirculation loop drive flows.

The flow in each loop can be manually or automatically controlled.

APPLICABLE The operation of the Reactor Coolant Recirculation System is SAFETY ANALYSES an initial condition assumed in the design basis loss of coolant accident (LOCA) (Ref. 1). During a LOCA caused by a recirculation loop pipe break, the intact loop is assumed to provide coolant flow during the first few seconds of the accident. The initial core flow decrease is rapid because the recirculation pump in the broken loop ceases to aump reactor coolant to the vessel almost immediately. T1e pump in the intact loop coasts down relatively slowly. This pump coastdown governs the core flow response for the next several seconds until the jet pump suction is uncovered (Ref. 1). The analyses assume that both loops are operating at the same flow prior to the accident. However, the LOCA analysis was reviewed for the case with a flow mismatch between the two loops, with the aipe break assumed to be in the loop with the higher flow. While the flow coastdown and core response are potentially more severe in this assumed case. (since the intact loop starts at a lower flow rate and the core response is the same as if both loops were operating at. the lower flow rate) a small mismatch has been determined to be acceptable based on engineering judgement.

, (continued) l l

i RIVER BEND B 3.4-2 Revision No. O

Recirculation Loops Operating B 3.4.1 BASES APPLICABLE 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 15 of the USAR.

A plant specific LOCA analysis has been performed assuming only one operating recirculation loop. This analysis has demonstrated that, in the event of a LOCA caused by a pipe break in the operating recirculation 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 recirculation loop operation (Ref. 3) and demonstrate sufficient flow coastdown characteristics to maintain fuel thermal margins during the abnormal operational transients analyzed provided the MCPR i requirements are modified. During single recirculation loop operation, modification 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 saecified in the COLR. The APRM flow biased simulated tiermal power setpoint is in LC0 3.3.1.1, " Reactor Protection 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. .W ad,di+icr.

r ~, mr +-+, the, em-mtotr c,c

~s c mrmerma m ~" m rt be :r.,mr+4mm

,e ,

'E! ^*ersted Tureum core Dnbt*D e NN Sn N bnmhNn k I r- k ANN + fiNA in hk N,irN 9 N1 1 fuFou f 6nU[oe m r;"ri W jy iltio d em ." N e~nativeiy. r with only one recirculation loop in operation, THERMAL POWER l must be s 83% RTP. the total core flow limitations identified above must be met, modifications to the required APLHGR limits (LCO 3.2.1. " AVERAGE PLANAR LINEAR HEAT l GENERATION RATE (APLHGR)") MCPR limits (LCO 3.2.2. " MINIMUM (continued)

RIVER BEND B 3.4-3 Revision No. 3-8 l

O Recirculation Loops Operating B 3.4.1 BASES LCO CRITICAL POWER RATIO (MCPR)"), and APRM Flow Biased (continued) Simulated Thermal Power-High setpoint (LCO 3.3.1.1) must be applied to allow continued operation consistent with the assumptions of Reference 3.

The LCO 4: ~edi'4ed by a "^te "hich ' ' ^':: up to lo hour before he'/ ng to put 4" c"ect the required medi'icati^": te 4

required '4mit: and setpc4"t 3+te a change 4" the reacte- I operati"; ce"ditie": # e- t'!c recircu'3 tic" ' cept operati";

to ri";'e recircu'atic" 'ecp ope atic" I' the required

'4-4t and retpcirt: 2re "et 4" cc~p'iance "ith the app'iceb'e require-^^t: et the e"d ^' this pc-ded, the ,

at Ociated equipmert ~utt be dec's ed 4"epe able c- the j

'4-it: ""et retic'4ed." and the ACTIO 5 equired by l re"ce"'c- ?"ce 'ith the app'icable specification: l 4mp'e c"ted. This time 1: p c'/ided due t^ the " cod to Et:bi'ize operatier 4t" one ecircu'at4c" leep, 4"cludi";

the precedu 21 Step: "ece:5 3 y to ' 4 *4 t e" 3"d c'c' cent cl cdc 1" +he operati n ; 'ecp. merite- #cr ^vce: i'/c

^oou and 'ecel pe':e- an;c ~^"4tcr (Loou). "cutron 'luv "ci te ' ~/el t : 2"d the complex 4ty and detai' required to

'ully 4mp'ement and co #4-= the requi red '4-4 + and retpc4rt modifications.

APPLICABILITY In MODES 1 and 2. requirements for operation of the Reactor Coolant Recirculation System are necessary since there is considerable energy in the reactor core and the limiting design basis transients and accidents are assumed to occur.

In MODES 3, 4, and 5, the consequences of an accident are reduced and the coastdown characteristics of the recirculation loops are not important.

ACTIONS A.1 With both recirculation loo s operating but the flows not matched, the recirculation . oops must be restored to operation with matched flows within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . If the flow mismatch cannot be restored to within limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , one recirculation loop must be shutdown.

1 (continued)

! RIVER BEND B 3.4-4 Revision No. 0 l

- - . _ _ - . . - .- - - . . . . -- - =- -- . --- -.

9 Recir Stion Loops Operating B 3.4.1 BASES i ACTIONS' A.1 (continued)

Alternatively, if the single loop requirements of the LCO are applied to operating limits and RPS setpoints, operation with only one recirculation loop would satisfy the requirements of the LCO and the initial conditions of the accident sequence.

The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time is based on the low probability of an accident occurring during this time period, on a reasonable time to complete the Required Action, and on frequent core monitoring by operators allowing abrupt changes in core flow conditions to be quickly detected.

B.1 l Should a LOCA or transient occur with THERMAL POWER > 83%

RTP, during single loop operation the core response may not be bounded by the safety analyses. Therefore, only a limited time is allowed to reduce THERMAL POWER to s 83%

RTP.

The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time is based en the los, probability of an accident occuring during this time period, on a reasonable time to complete the Required Action, and on frequent core monitoring by operator: allowing changes in THERMAL POWER to be quickly detected.

C.1. " 1 and E.1 n ue te ther 3' hydrau'ic stab 4'ity concer": apcoti^^ ef .

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i b" i Ibquiredte hbh-4ttbd . Sb 4 U E~5d N [ 4-~5di5tb~htt nv4+ +hn rnninn i.r4 + b 4 n A hn, er e hs, rnArininn TurDM A t Dn!XD

+h;N, inh n N+rN1 rnN inenr+ kn nN h1, NN;N$c4nN r i rniikk+ 4 nn iLn;73 c ~, ' h i, Nann!N, 4E; iii,;in;;n, T,Lii,;' 'nnN;s+i;,Ji;^

i D;','i nn ' i'T it ;5;L mLN ';,,c;Nn+is,N si i;c4;h4,iC, sEsn n;L;, N;n;;+ L ~, ~ m s ; s ~,m; N;c ' ' ui,1,N ; J;N;;+4n;'4; E c

-bg4bbIEhiibdd"itF5bb~bxNhEIbbthiE4F5idendb?

(continued)

RIVER BEND B 3.4-5 Revision No. 3-8

l Recirculation Loops Operating B 3.4.1 BASES

- -c c, m-o_ . ,. snA n 1 A I NJ -.. _ ._ _.. Continu 4

inc+sh414+u nem ire /4 n ADDM nr ! DDM n m ,+ r nn fliiv

_.__. . ., s . _ _ _ . , s .. . _- .. - _ , _. 1 m.ini 4e

+hrnn +4mne +hn nc+shlichnA hsenlinni thnn nr&4nn 4e ram vi rnA &n nv4& thic rnn4nn .A+hin~,9

- h m, _i r c. Thn. _ s l l nun ,4 7 .. __ .. .. .. .y.,. .. . . .

l'nmn1 nt 4 nn T4mne srn

...( . . . . ,

. . rnsenn. shin hsenA rnc+nrn nisnt. ~ . . , . , - _ , , , .

n_ n .

nnnrs+4nn

,y. v . .g nvnnrinne-n +n nsr,mn+nre +n n n rm. .s l in sn n rAn rl s tmsnnnr .. . , _

7. ..

v i + hm ,,&. ,_ . . .rhs11nnninn nisn& cuc&nme sn

_ . _ . ., . . . .. _..A_ . _ . . . _ .

.a' a r-- s-----

A An4 nrm 4nst4nn nf ADDM snA IDDM n m ,+ r n n fl i e v rnien l m inl c m in r s r O hm ere n rm ,4 An c f ram inn & nnr4nA4r infnrms+4nn

(. _. ._._

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

rni s 6 i sin

+n ne+sh14chnA hsenlinn nnien l m eni c /cnn s.

/'. n n A 4 + 4 an. ni +hst i n A i r s + n.

c+,hin e+nsAu c+s+n nnnestinn A

7. _ . . _ .

An&ncmins&4nn nf &bnen nn4cn l mini c ui+hin On mine,+ne s f+ n r sn 4rernsen nf . . . . .. . , , .

s C& DTD n-nuiAne s mnrn frnm mnt inA4rs+4nn r--

--- - ~~' -

4 v'-- ' ' ' ' ' - ' ' - ' '

o f +hn c & s hi 14 + 1r nf nnnestinn fnl l nt.r4 nn snst c4/mif4rsn+

_. . .. .._ . .g ..

,r- - ' '- --

'a - - - --

nn&nn&4s1 fnr chsnnn nf +hn + hn rm s l ' h u A r s ,' sil 4 r nrnnnr+4ne nf a-

--a- -

r'h' n-- c i r c + nm t

,.,s,.. .

Thnen Ernm innr4 7.

- - nc

' nrmi4

' ' ' ' s An r' ' -- - ' - -

" ' -nsr '" l'ss i'- An&nc+4nnr ' -' r - nf n m i+ r n, n .-

fl e iv ner411s+4nne

.._. .. . ., _- .. enrn. <4rin +n ..

+hnrmsl

. .. m hs .

rAr

.s_.-_...,

siil 4 r inctsh414+4nc Ar& 4 nn mi i c + hn initi,+nA +n r n c + n,., r. n +hn

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

nisnt +n s mnrn c+shln nnun r / fi no rs+4n if c,ich inAirn+inne r---- ' - - - '~ - - ' '-- - - ' ' '

nf limit curln

_s..

nm e+ rnn fl i e v ne r 4 31. s t i nn srn e _. An+nr+nA ._..,.

I 1 I 9 snd C 9 ._ ._

(f 4 + hnn._ r nr 4 r m ,1. s & 4 nn

. .. ~_. ' ~

' - r -'

Innne in nnnrs+4nn r- - ' ' - - - -

thm ,,n4+

ic -

hn h r m inh' t &n s Mnnr in t.:h i ch + h' n Icn Anne nnt g mi4rnA ..+n ~.

7 ~_ .

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

A r-& 4 nn mi i c + in4+4stnA &n m_nlg ju . . . . , . - - . k_ n_ . . . . . .

r n A, , r n. TtJEDM A ? DnU. . r D + n.

hn

_ .A+hin +hn 14 mite An fi nn A.. in +hn /'n. l. D. +n s e r v irn, + hn rm s 1 b u A r m , , , 4 ,- c+sh m +m

..s,,__.

enmenrnc ,rm sAArner,A Thm n sn+ 4r

+hmn enmi4rnA

+n hn nlsemd 4n unne 9 4n c h m,., _, r.-c ' ,r - -na unne 9 . _ _ . -

-7 7._,. . . .. . , . ,.,

in 19 hm irc -- . , . , _ . ..

In thie enrM4+4nn +hn enr 4 em tl 3 + 4 r-n Innne sen r- -

a nn& enmeirnA &n kn nnnrstinn kne .y ss te n n f. + h. n rnAr tenA ' e_nt , _ . .rnrftsu

~ 7 ,.nd m 4 n 4 m , , ,. . .AnnmnAnnem nf nDAc nn +bm r , r- 4 r m . ,. ,._ -s u nn ,nnn r' - ' - - - - --- -- --

~r e n s e + A nt . .n Thn A /'emnin&4nn T4mne srn

- _ _ _ _ _ . . c h s .r s c + n r i c + 4 r c. . - .

..._ _ _ . . . . . ~ s 11 .,_n.. - -- ' r - -- ' - -

rnsconshlm hsend An n n n e s &. 4. n, n Avnneinnen &n ensch MnnI 9

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

g ..r-- - ' - ' -- --- --- -

femm fe t l. l g, nnt.nr ennM4+4nne in me An el i r msnnne ,n A

.s. n. ,. ,.. .j , . . . , . 1.e4 + h.n.ert.

v..

n,.,n+ cuc+ mmc

. , . . ,. ..v. .. _. _ -

c h s , , mn n 4 m,,

. . , , . y.... , ._.

(continued)

RIVER BEND B 3.4-6 Revision No. 1

Recirculation Loops Operating o B 3.4.1 l

BASES l ACTIONS G-1 l (continued) '

i Tf +sn rnemirnA 14m4+ nr en+nm4n+ mmAi#4m,+4nne ,rm nm+

mrhmrmna Srh&b n 19 kmbre h+Nr ir s NN U + h NN fhAN + ,N L;mi;;mi;+ i;L im;; m;;;;+iim~+m'ciL;i;'iimm ;;m;;T4mm +sm i

"6k"4bbb~i'mit:

mmc i

5b9EbtbI"thUE4bhSi{'e'bbt

+ sm 4mmma4,+m,m ammV,rma m,+

553 ~5564h4bd~

mm+ Tsm ommm4-ma am+4mme E

'bb~tb3bbbi5tbbI~Ib5'5bd Ib brb-^-+ 'bh5bbbic'Y" t ~tEAE be teF^"

C.1 With the requirements of the LCO not met. the recirculation loo 3s must be restored to operation with matched flows wit'11n 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . A recirculation loop is considered not in operation when the pump in that loop is idle or when the mismatch between total jet pump flows of the two loops is greater than required limits. The loop with the lower flow must be considered not in operation. Should a LOCA occur with one recirculation loop not in operation, the core flow coastdown and resultant core response may not be bounded by the LOCA analyses. Therefore. only a limited time is allowed to restore the inoperable loop to operating status.

Alternatively, if the single loop requirements of the LC0 are applied to operating limits and RP5 setpoints. operation with only one recirculation loop would satisfy the requirements of the LCO and the initial conditions of the accident sequance.

The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Completion Time is based on the -low probability ;

of an accident occurring during this time period, on a reasonable time to complete the Required Action, and on frequent core monitoring by operators allowing abrupt changes in core flow conditions to be quickly detected.

This Required Action does not require tripping the recirculation pump in the lowest flow loop when the mismatch between total jet pump flows of the two loops is greater than the required limits. However in cases where large flow mismatches occur. low flow or reverse flow can occur in the low flow loop jet pumps, causing vibration of the jet 1 pumps. If zero or reverse flow is detected, the condition l should be alleviated by changing flow control valve position I to re-establish forward flow or by tripping the pump.

(continued)

RIVER BEND B 3.4-7 Revision No. O

Recirculation Loops Operating B 3.4.1 ,

BASES

i BASES D.1

' With no recirculation loops in operation, or the Required Action and associated Completion Time of Condition A not ;

met, the unit is required to be brought to a MODE in which ,

'the LCO does not ap)ly. To achieve this status the plant :

must be brought to iODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . In this !

condition, the recirculation loops are not required to be operating because of the reduced severity of DBAs and j minimal dependence on the recirculation loop coastdown l characteristics. The allowed Completion Time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is reasonable. based on operating experience, to reach MODE 3 l trom fulI power conditions.in an orderly manner and without challenging plant systems.

l SURVEILLANCE SR 3.4.1.1 !

REQUIREMENTS This SR ensures the recirculation loop flows are within the j allowable limits for mismatch. At low core flow (i.e., i

< 70% of rated core flow), the MCPR requirements provide '

larger margins to the fuel cladding integrity Safety Limit such that the potential adverse effect of.early boiling transition during a LOCA is reduced. A larger flow mismatch can therefore be allowed when core flow is < 70% of rated l core flow. The recirculation loop jet pump flow, as used in '

this Surveillance. is the summation of the flows from all of j the jet pumps associated with a single recirculation loop.

The mismatch is measured in terms of percent of rated core i flow. This SR is not required when both loops are not in i operation since the mismatch limits are meaningless during i single loop or natural circulation operation. The Surveillance must be performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after both loo]s are in operation. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency is consistent wit 1 the Frequency for jet pump OPERABILITY verification and has been shown by operating experience to be adequate to detect off normal jet pump loop flows in a timely manner.

(continued) i

! (continued)-

h RIVER BEND B 3.4-8 -Revision No. O f.

_ _. _ _ _ , r _ , .

. . . - .= . - . . -_ _ _ - . .

Recirculation Loops Operating B 3.4.1 BASES BASES SURVEILLANCE REQUIREMENTS So ?A . ' .2 Th4.e CD nn e i e rn e +hn rnsetnr TtJrDM AI DnUCD snA enrn fi ni.,c srn

..,4 + h 4 n ,nnrmnr4s+n ner,mn+mr , 4 mu .- c --+m n r 7~., n +.. ,,n-nn+.. - . . ,n, . , m a

,77 7 ._(. - - -. . --... . .-.

i r AceilInt4Anc

,a.. . _ . , , . . .

ht I nt . f eneiertflSt4nn fl rt!c . . . -s _n. -r4h 4egh.,

. ,, ,. -- -v- ., - --

. p ensetAr thn ensetAr Avh4h4tc ineen. scad estr e n n t i h 4 l i f ts

. .. - . n Ai r. -# -- Ar- -----' -

- CD' - Ann +---r-~'

+n +hnrms1 h u A r , , ,14. c inc&sh414+u Th4c 4 4 fi ne i. ,h nn s

. ..._...y,_ ._ . _ . . . _ .

..g .- -. ._. . ._ ..

thn P AM M 4. f 4 An e ro Thn

.7 mtti ednN i nf Ar4 m _S et. i Ari c . . e n e e c e 3 r. gti . Sen

.y .._ - . - --_ .-

SR 3.4.1.2 (continued) crm,,,mn-u 4e heema mn e n m r , + 4 ., , n n n r 4 m n ,. A mna + hm

_,,..g .- - . . - , -,

r' --

a -'r-- -

operaterc' drherert tme'c:'ed;c'of rea:ter'- ctat=

. -- , irr' M4r; e4 nn4 f 4 csnt chsnanc in TllrDM A f DnUCD nnA enrn fi n. . ,

REFERENCES 1. USAR Section 6.3.3.

2. USAR Section 5.4.1.4.

3. USAR, Section 15.0.6.

(continued)

RIVER BEND B 3.4-9 Revision No. O

O l

Administrative Controls - COLR i

p i

l i

l l

I l

1 1

l i

i

\

i l

1 Page 6 of 15 4

Reporting Requirements

., 5.6 5.6 Reporting Requirement:

5.6.2 Annual Radiological Environmental Operating Report (contiriued) results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing.results. The. missing. data shall be submitted in a supplementary report as soon as possible.

5.6.3 Radioactive Effluent Release Report The Radioactive' Effluent Release Report covering the operation of the unit during the previous calendar year shall be submitted by May 1 of each year. The' report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material 3rovided shall.be consistent with.the objectives outlined in t1e ODCM and process control program and in conformance with 10 CFR 50.36a and 10 CFR 50. Appendix I.Section IV.B.1.

5.6.4 Monthly Operating Reports Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the main steam safety / relief valves, shall be submitted on a monthly basis no later than the 15th of each month following the calendar month covered by the report.

5.6.5- CORE OPERATING LIMITS REPORT (COLR) i

a. Core operating limits shall be established prior to each reload cycle, or 3rior to any remaining )ortion of a reload cycle, and shall 3e documented in the C0_R for the ;

following:

1) LCO 3.2.1. Average Planar Linear Heat Generation Rate ]

(APLHGR). ;

l 2) LC0 3.2.2. Minimum Critical Power Ratio (MCPR) ;

I 3) LC0 3.2.3. Linear Heat Generation Rate (LHGR)

b. The analytical methods used to determine the core operating ,

. limits shall be those previously reviewed and approved by l the NRC, specifically those described in the following documents.

(continued) l i

l RIVER BEND 5.0-18 Amendment No. M 100 l

I

t Reporting Requirements o 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

1) NEDE-24011-P-A. " General Electric Standard Application for Reactor Fuel" (latest approved version);*

2) NEDC-32489P (April 1996). "T-Factor Setdown Elimination Analysis for River Bend Station" (for power and flow dependent limits methodology only as evaluated and approved by Safety Evaluation and License Amendment 100 1 dated October 10. 1997):

3) NEDO-32339P-A. " Reactor Stability Long-Term Solution:

Enhanced Option I-A" including Supplements 1 through 4 ,

(latest approved version).

c. The core operating limits shall be determined such that all applicable limits (e.g.. fuel thermal mechanical limits, core thermal hydraulic limits. Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM. transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR. including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

i RIVER BEND 5.0-19 Amendment No. 84 96 94 100

ML20154H369

Text

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