Text

Amend 31 to License NPF-47,revising Tech Specs to Allow Single Recirculation Loop Operation
	ML20206L522
Person / Time
Site:	River Bend
Issue date:	11/22/1988
From:	Calvo J; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20206L528	List: ML20206L522; ML20206L530; ML20206L536;
References
	NUDOCS 8811300043
	Download: ML20206L522 (36)
	v • d • e

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***xq#o, UNITED STATES

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NUCLEAR REGULATORY COMMISSION e

i wAsHmoTow n.c.sosse

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4 GULF STATES UTILITIES COMPANY DOCKET NO. 50-458 RIVER BEND STATION, UNIT 1 i

AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 31 License No. NPF-47 1.

The f;uclear Regulatory Comission (the Commission) h*s found that:

A.

The application for amendment by Gulf States Utilities Company

(+he licensee) dated April 6,1988, as supplerented October 20, 1988, complies with the standards and requirements of the Atemic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Conmission; C.

There is reasonable assurance:

(i) that the activities authorized by this amendment can be conducted without endangering the health j

and safety of the public, and (ii) that such activities will be conducted in compliance with the Corrainsion'I regulations; D.

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

The issuance of this r,mendment is in accordance with 10 CFR part 51 of the Comission's regulations and all applicable requireme,.s have been satisfied.

8811300043 881122 FDR ADOCK 05000458 P

FLC

2-2.

Accordingly, the license is atended by changes to the Technical Specifications as indicated in the attachment to this license arei.v. cont and Paragraph 2.C.(2) of facility Operating License No. NPF-47 is hereby amended to read as follows:

(2) Technical Specifications and Environmental Protection Plan The Technical Sper.ifications contained in Appendix A, as revised through Arendrent No. 31 and the Environmental Protection Plan contained in Appendix B, dre hereby incorporated in the license. GSU shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

The license amendrent is effective as of its date of issuance.

FOR THE hDCLEAR REGULATORY COPr!SSION 7

4 dahu 3%

Iose A. Calvo Director Prcject Directorate - IV Division of Reacter Projects - Ill, IV, V and Special Projects Office of Nuclear Reactcr Regulation Attachrent:

Charges to the Technical Specifications Date of Issuerce:

November 22, 1988 i

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ATTACHMENT TO LICENSE AME!;0 MENT NO. 31 FACILITY OPERATlf;G LICENSE NO. NPF,,4,7 j

1 DOCKET NO. 50d,58 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contains a vertical line indicating the area of change. Overleaf pages are i

provided to raintain document completeness.

Rett 0VE PAGES INSERT PAGES 2-1 2-1 2-4 2-4 2-5 2-5 B 2-1 0 2-1 3/4 2-1 3/4 2-1 i

3/4 2-7 3/4 2-7 3/4 2-7a 3/4 3-62 3/4 3 62 3/4 4-1

' 3/4 4-1 3/4 4-la 1

3/4 *-2 3/4 4 2 3/4 4-ta l

3/4 4-3 3/4 4-3 l

3/4 4-4 3/4 4-4 l

3/4 4-da j

3/4 4-5 3/4 4-5 B 3/4 1-2 B 3/4 1-2 L

B 3/4 2-2 B 3/4 2-2 l

B 3/4 2-3 8 3/4 2-3 B 3/4 2-4 8 3/4 2-4 1

B 3/4 4-1 8 3/4 4-1 l

B 3/4 4-la l

B 3/4 4-2 B 3/4 4-2 l

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2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS THERMAL POWER, low Pressure or low Flow 2.1.1 THERMAL POWER shall not exceed 25% of RATED THERMAL POWER with the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow.

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APPLICABILITY: OPERATIONAL CONDITIONS I and 2.

ACTION:

r With THERMAL POWER exceeding 25% of RATED THERMAL POWER and the reactor vessel i

steam dome pressure less than 785 psig or core flow less than 10% of rated flow, be in at least HOT S'lVTDOWN within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and comply with the requirements of Specification 6.7.1.

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THERMAL POWER, Hiah Pressure and Hiah Flow 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be'less than 1.07 with two recirculation loop operation and shall not be less thar. 1.08 with single recirculation loop operation with the reactor vessel steam dome pressure i

greater than or equal to 785 psig and coro flow greater than or equal to 10% of j

rated flow.

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

ACTION-l With MCPR less than 1.07 with two recirculation loop operation or less than l

1.08 with single recirculation loop operation and the reactor vessel steam dome i

prersure greater than or equal to 785 psig and core flow greater than or equal to 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and comply a

with the requirements of Specification 6.7.1.

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REACTOR COOLANT SYSTEM PRESSURE l

j 2.1.3 Tne reactor coolant system pressure, as measured in the reactor vessel l

steam dome, shall not exceed 1325 psig.

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APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, 3 and 4.

I ACTION:

With the reactor coolant system pressure above 1325 psig, as measured in the j

reactor vessel steam dome, be in at least HOT SHUTDOWN with reactor coolant system pressura less than or equal to 1325 psig within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and comply with i

the reoutrements of Specification 6.7.1.

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RIVER BEND - UNIT 1 2-1 AMENDMENT NO. 22,31 l

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SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SAFETY LIMITS (Continued)

REACTOR VESSEL WATER LEVEL 2.1.4 The reactor vessel water level shall be above the top of the active irradiated fuel.

APPLICABILITY: OPERATIONAL CONDITIONS 3, 4 and 5 ACTION:

i With the reactor vessel water level at or below the top of the active irradiated fuel, manually initiate the ECCS to restore the water level, after depressurizing the reactor vessel, if required.

Comply with the requirements of Specification 6.7.1.

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RIVER BEND - UNIT 1 2-2

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS 2.2.1 The reactor protection system instrumentation setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2.1 1.

APPLICABILITY: As shown in Table 3.3.1-1.

ACTION:

With a reactor protection system instrumentation setpoint less conservative than the value shown in the Allowable Values column of Table 2.2.1-1, declare the channel inoperable and apply the applicable ACTION statement requirement i

of Specification 3.3.1 until the channel is restoied to OPERABLE status with its setpoint adjusted consistent with the Trip Setpoint value.

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RIVER BEND - UNIT 1 2-3 M

3 TA8LE 2.2.1-1 h

REACTOR PROTECTION SYSTEM INSTRUNENTATION SETPOINTS 3,:

FUNCTIONAL UNIT ALLOWA8LE o

TRIP SETPOINT VALUES

[

1.

Intennediate Range Monitor, Neutron Flux-High 5 120/125 divisions 5 122/125 divisions se of full scale of full scale

}

2.

Average Power Range Monitor:

a.

Neutron Flux-High, Setdown

< 15% of RATED

< 20% of RATED

- THERMAL POWER THERMAL POWER b.

Flow Biased Simulated Thermal Power-High

1) Two Recirculation Loop Operation a) Flow Biased

< 0.66 W+48%, with

-< 0.66 W+51%, with a maximum of a maximum of b) High Flow Clamped

-< 111.0% of RATED

< 113.0% of RATED THERMAL POWER

- THERMAL POWER

2) Single Recirculation Loop Operation m

E a) Flow Blased

< 0.66 W+42.7%, with

< 0.66 W+45.7% with a maximum of a maximum of b) High Flow Clamped

< 111.0% of RATED

< 113.0% of RATED THERMAL POWER THERMAL POWER c.

Neutron Flux-High

< 118% of RATED

< 120% of RATED THERMAL POWER THERMAL POWER d.

Inoperative NA NA 3.

Reactor Vestel Steam Dome Pressure - High 5 1064.7 psig i 1079.7 psig 2-4.

Reactor Vessel Water Level - Low, level 3

> 9.7 inches above

> 8.7 inches above d

instrumer:t zero*

instrument zero 5.

Reactor Vessel Water Level-High, Level 8 1 51.0 inches atmve 1 52.1 inches above g

instrument zero' instrument zero g

6.

Main Stear. Line Isolation Valve - Closure

$ 8% closed 5 12% closed I

"See Bases Figu e 8 3/4 3-1.

P.

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8 TABLE 2.2.1-1 (Continued) 3 l

29 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS i

<=

9 ALLOW # OLE l

1 7

FUNCTIONAL UNIT TRIP SETPOINT VALUE3 E

7.

Main Steam Line Radiation -

High

-< 3.0 x full power

< 3.6 x full power Z

background background I

)

B.

Drywell Pressure - High

< 1.68 psig

< 1.88 psig

~

9.

Scram Discharge Volume Water Level - High i

a.

Level Transmitter - LISN601A and B

< 49"

< 53" i

LISN601C and D 349"

{ 51.7" 3

b.

Float Switches - LSN013A and B

< 48.76"

< 53.50" i

i LSN013C and D 346.88"

{ 49.00"

{

10. Turbine Stop Valve - Closure

< 5% closed

< 7% closed

11. Turbine Control Valve Fast Closure, l

Trip 011 Pressure - Low

> 530 psig

> 465 psig j

12.

Reactor Mode Switch Shutdown Position MA NA

13. Manual Scram NA MA

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F u

See Bases Figure B 3/4 3-1.

.w

-v,ww-

=

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2.1 SAFETY LIMITS BASES 0.0 INTR 00VCTION The fuel cladding, reactor pressure vessel and primary system oiping are the principal barriers to the release of radioactive materials to tie environ 2.

Safety Limits are established to protect the integrity of these barriers during normal 11 ant operations and anticipated transients. The fuel cladding integrity Safety Limit is set t,uch +. hat no fuel damag2 is calculated to ocr.ur if the limit h not violated.

Because fuel damage is not directly observable a step back approachisusedtoestablishaSafetyLimitsuchthattheMCPRIsnotless than 1.07 for two recirculation loop operation and 1.08 for single recirculation loop operation. MCPR greater than 1.07 for two recirculation loop operation and 1.08 for single recirculation loop operation represents a conservative margin relative to the conditions required to maintain fuel a

cladding integrity.

The fuel cladding is one of the separate the radioactive materials from the environs. physical barriers which The integrity of this cladding barrier is related to its relative freedom from perforations or cracking.

Although some corrosion or use related cracking may occur during the life of the cladding, fission product migration from this source is in:rementally cumulative and continuously measurable.

Fuel cladding perfora-tions, however can result from thermal stresses which occur from reactor operationsignIficantlyabovedesignconditionsandtheLimitingSafetySystem Settings. Whilefissionproductmigrationfromcladdingperforationisjustas measurable as that from use related cracking, the thermally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross rather than incremental cladding detsrioration.

Therefore, the fuel cladding Safety Limit is defined with a ma gin to the conditions which would produce onset of transition boiling, MCPR of 1.0.

These conditions represent a significant departure from the condition intended by design for planned operation.

2.1.1 THERMAL POWER, low Pressure or Low Flow The use of the GE Critical Power correlat. ion (Reference 1) is not valid for all critical power calculations at pressures below 785 psig or core flows less than 10% of rated flow.

Therefore, the fuel claddirg integrity Safety Limit is establisned by cther means. This is done by establishing a limiting condition on core THERMAL POWER with the following basis.

Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low power and flows will always be greater than 4.5 psi.

Analyses show that with a bundle flow of 28,000 lbs/hr, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi.

Thus, the bundle flow with a 4.5 psi driving head will be greatet than 28,000 lbs/hr.

Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt. With the design peaking factors, this corresponds to a THERMAL POWER of more than 50% of RATED THERHAL POWER. Thus, a THERHAL POWER limit of 25% of RATED THERMAL POWER for reactor pressure below 785 psig is conservative.

RIVER BEND - UNIT 1 B 2-1 Amendment No. Z2, 31 i

SATETY LIMIT $

BASES 2.1.2 THERMAL POWER. Hich Pressure and Hian Flow The fuel cladding integrity Safety Limit is set such that no fuel damage l

is calculated to occur if the limit is not violated.

Since the parameters l

which result in fuel damage are not directly observable during reactor opera-tion, the thermal and hy/ *aulic conditions resulting its a departure from nucleate boiling have be en used to mark the beginning of the region where fuel 7'

damage could occur.

Although it is recognized that a departure from nucleate belling would not necemrily result in damage to BWR fuel rods, the r.ritical t

pcwer at which boiling transition.ls calculated to occur has been adopted as a convenient. limit.

However, the uncertainties in monitoring the core operating state and in the procedures used to calculate the critical power result in an uncertainty in the value of the critical power.

Therefore, the fuel cladding integrity Safety Limit is defined as the CPR in the limiting fut1 assembly for which aiore than 99.9% of the fuel rods in the core are expected to avoid boiling transition considering the power distribution within the core and all uncertain-l i

ties.

2 The Safety Limit MCPR is determined using a statistical model that combines all of the uncertainties in the operating parameters and in the procedures used to calculate critical power.

The probability of the occurrence of boiling transition is detera:ined using the approved General Electric Critical power correlation.

Details of the fuel cladding inte'rity safety Itait calculation l

are given in Reference 1.

Reference 1 includes a tabulation of the uncertain-l ties ustd in the determination of the Safety Limit MCPR and of the nominal l

values of parameters used in the Safety Limit MCPR statistical analysis, j

j Reference l

L "General Electric Standard Application for Reacter Fuel (GESTAR) "

NEDE-24011-P-A-8.

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RIVER BEND - UNIT 1 8 2-2 Amendment No.12 i

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3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.1 All AVERAGE PLANAR LINEAP. HEAT GCNERATION RATES (APLHGRs) for each type of fuel as a function of AVERAGE PLANAR EXPOSURE shall not exceed the limits shown in Figures 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4, 3.2.1-5, and 3.2.2-6.

The limits of Figures 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4, 3.2.1-5 and 3.2.1-6 shall be reduced to a value of 0.64 times the two recirculation loop operation limit when in single loop operation.

APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER.

i ACTION.

\\

With an APLHGR exceeding the limits of Figure 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4, 3.2.1-5 or 3.2.1-6, initiate corrective action within 15 minutes and l

restore APLHGR to within the required limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

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SURVEILLANCE REQUIREMENTS l

4.2.1 All APLHGRs shall be verified to be equal to er less than the limits determined from Figures 3. 2.1-1, 3.2.1-2, 3. 2.1-3, 3. 2.1-4, 3. 2.1-5 and 3. 2.1-6:

a.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , I

r b.

Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of a THERMAL POWER increase of et i

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' vast 15% of RATED THERMAL POWER, and j

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c.

Initially and at i.4st once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating with a LIMITING CONTROL ROD PATTERN for APLHGR.

1 d.

The provisions of Specification 4.0.4 are not asplicable.

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i RIVER BEND - UNIT 1 3/4 2-1 Amendment No.12, 31 r

[

t l

12 fp$

5

/

M L

3

'l1 g6 1

2

%s E

10 a5 Y$

N

<5

\\g IO Ds 9

L i

3g yz

(

E

\\

5 s\\

J 8

4 7

O 10 20 30 40 50 AVER AGE plt N AR EXPOSURE (GWd/t)

FIGURE 3.2.1 1 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR)

VERSUS AVERAGE Di AMAR EXPOSURE BP85RB094 r

RIVER BEND - UNIT 1 3/42-2

/c.end ent he. 12

POWER DISTRIBUTION _ LIMITS 3/4.2.2 APRM SETPOINTS LIMITING CONDITION FOR OPERATION 3.2.2 The APRM flow biased simulated thermal power-high scram trip setpoint (S) and flow biased neutron flux-upscale control rod block tri shall be established according to the following relationships:p setpoint (SRB) a.

Two Recirculation loop Operation Trip Setpoint Allowable Value 5 1 (0.66W + 48%)T

$ 1 (0.66W + 51%)T Sgg 1 (0.66W + 4 3)T Sgg 1 (0.66W + 45%)T b.

Single Recirculation Loop Operation Tr,ip Setpoint Allowable Value 5 1 (0.66W + 42.7%)T S 1 (0.66W + 45.7%)T S

i (0.66W + 36.7%)T S

1 (0.66W + 39.7%)T RB RB where:

5 and Seg are in percent of RATED THERMAL POWER, W = Loop recirculation f:ow as a percentage of the loop recirculation flow which produces a rated core flow of 84.5 million Ibs/hr.

T = The ratio of FRACTION OF P.ATED THERMAL POWER (FRTP) divided by the CORE MAXIMUM FRACTION OF LIMITIN3 POWER DENSITY (CMFLPD).T is applied only if less than or equal to 1.0.

APPLICABILITY:

OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25T, of RATED THERMAL POWER.

ACTION:

With the APRM flow biased simulated thermal power-high scram trip setpoint end/or the flow biased neutron flux-upscale control rod block trip setpgint less r:onser-vative than the value shown in the Allowable Value column for 5 or Sgg, as above determined, initiate corrective action within 15 minutes and adjust S and/or SRB to be consistent with the Trip Setpoint value

within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

With CMFLPD greater than the FRTP, rather thai adjusting the APPM setpoints, the APRM gain may be adjusted such that the APRM readings are greater than or equal to 100% times CHFLPD, provided that the adjusted APRM read?ng does not exceed 100% of RATED THERMAL POWER, and a notice of the adjustmea.t is posted on the reactor control panel.

RIVER BEND - UNIT 1 3/4 2-7 Amendment No.31

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS I

4.2.2 The FRTP and CMFLPD shall be determined, the value of T calculated, and the most recent actual APRM flow biased simulated thermal power-high scram and flow biased neutron flux-upscale control rod block trip setpoints verified to be within the above limits or adjusted, as required:

a.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b.

Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after comaletion of a THERMAL POWER increase of at least 15% of RATED THERMA. POWER, and Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operet-c.

ing with CMFLPD greater than or equal in FRTP.

d.

The provisions of Specification 4.0.4 are not applicable.

RIVER BEND - UNIT 1 3/4 2-7a A.tendment No.3I

POWER DISTRIBUTION LIMITS 3/4.2.3 MINIMJM CRITICAL POWER RATIO LIMITING CONDITION FOR OPERATION 3.2.3 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be equal to or greater than both MCPR and MCPR limits at indicated core flow and THERMAL POWEA as f

n shown in Figures 3.2.3-l' and 3.2.3-2.

i APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER.

1 ACTION:

With MCPR less than tht applicable MCPR limit shown-in Figures 3.2.3-1 and 3.2.3-2, initiate corrective action within 15 minutes and restore MCPR to within the required limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

_ SURVEILLANCE REQUIREMENTS 4.2.3 MCPR shall be determined to be equal to or greater than the MCPR limit determined from Figures 3.2.3-1 and 3.2.3-2:

a.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , b.

Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months aft 2r completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and I

Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating c.

with a LIMITING CONTROL R00 PATTERN for MCPR.

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d.

The provisions of Specification 4.0.4 are n.ot applicable.

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RIVER BEND - UNIT 1 3/4 2-8

i TABLE 3.3.6-1 (Continued)

)

CONTROL ROD BLOCK INSTRUMENTATION ACTION ACTION 60 Declare the RPCS inoperable and take the ACTION required by Specification 3.1.4.2.

ACTION 61 With the number of OPERABLE Channels:

a.

One less than required by the Minimum OPERABLE Channels per Trip Function requirement, restore the inoperable channel to OPERABLE status within 7 days or place the looperable channel in tue tripped condition within the next hour.#

b.

Two or more less than required by th6 Minimum OPERABLE Channels per Trip function requirement, place at least one inoperable channel in the tripped condition within one hour.#

With the number o'f OPERABLE channels less than required by the ACTION 62 Minimum OPERABLE Channels per Trip Function requireseat, place the inoperabid channel in the tripped condition within one hcur.f NOTES With more than one control rod withdrawn.

Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

OPERABLE channels must be associated with SRM required OPEF.ABLE per Specification 3.9.2.

The provisions of Specification 3.0.4 are not applicable.

(a) This function shall be automatically bypassed it detector count rate is

> 100 cps or the IRM channels are on range 3 or higher.

(b) This function shall be automatically bypasseo when the associated IRM channels are on range 8 or higher.

(c) This function shall be automatically bypassed when the IRM channels are

,n range 3 or higher.

l (d) This function shall be automatically bypassed whM the IRM channels are on range 1.

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RIVER dEND - UNIT 1 3/4 3-61 5

TABLE 3.3.6-2 CONTROL ROD DLOCK~IN5T U U TATION SETPOINTS 5 TRIP FUNCTION TRIP SETPUIKT ALLOWABLF VALUE M

[

1.

R00 PATTERN COMTROL SYSTEM a.

Low PoweVSetpoint 27.513% of RATED THERML POWER 27.5 7.5% of RATED THERML g

80WER o

b.

!'agh Power Setpoint 62.5 i 3% of RATED THERMAL POWER 62.5 i T.ST, of RATD THEREL POWER

[

2.

APRM a.

Flow Blased Neutron Flux Upscale

[

1) Two Recirculation loop Operation 1 0.66W + 42%*

< 0.66W + 45%d

2) Single Recirculation Loop Operation 5 0.66W + 36.7%*

< 0.66W + 39.7%*

b.

Inoperative NA NA c.

Domscale

>5% of RATED THERMAL POWER

> 3% of RATED THERMAL F0WER d.

Neutron F'ex - Upscale

~

Startup

$ 12% of RATED THERMC P(NER

$ 14% of RATED THERMAL P0hER 3.

SOURCE RANGE MONITORS a.

Detector not full in MA NA

{

b.

Upscale

< 1 x 10 cps

< 1.6 x 10 cp, 5

5 m

c.

Inoperative NA NA d.

Downscale 1 0.7 cps 3 0.5 cps **

4.

INTERE DIATE RANGE MONITORS a.

Detector not full in MA NA b.

Upscale

-< 108/125 d! vision of full

< 110/125 division of full scale

- scale c.

Inoperative NA NA d.

Downscale

> 5/125 division of full

> 3/125 division of full

- scale

- scale 5.

SCRAM DISCHARGE VOLUNE E

a.

Water Level-High - LISN602A

< 18.00"

< 21.12" i:_ 18.00" 7 21.60" LIS%028 g

6.

REACTOR COOLANT SYSTEM RECIRCULATION FLOW

=3 a.

Upscale 1 IOC or rated ficw

$ 111% of Pat 4d flow o.

S

The Average Power Range Monitor rod block function is varied as a function of recirculation loop flow (W).

The trip setting of '.his function must be maintained in accordance with Specification 3.2.2.

- Provided signal t.. mise ratio is > 2, otherwise setpoint of 3 cps and allowable 1.8 cps.

e e

=

e rw

a 3444 REACTOR COOLANT SYSTEM 3 /4._4.1 RECIRCULATION SYSTEM I

RECIRCULATION LOOPS l

)

LIMITING CONDITION FOR OPERATION 3.4.1.1 The reactor coolant system rectreulation loops shall be in operation and in Region ! 68 specified 'n Figure 3.4.1.1-1 with nither:

[

a.

Two recirculation it, ops operating with limits and setpointr per Specifier. ions 2.1. 2, 2. 2.1, 3. 2.1, J. 2. 2, 3. 3. 6, o r b.

A single loop operating with:

l 1.

Volumetric recirculation loop flow rate less than or equal to I

33,000 gpm, and i

2.

The recirculation loop flow contro1 system in the loop Manual

(

(Position Control) Mode, and

{

3.

THERMAL POWER less than or equal o 70% of RATED THERML POWER, and i

I 4.

Limits and setpoints for single recir%1stien loop operation

[

per Specifications 2.1.2, 2.2.1, 3.2.1, 3.2.2, and 3.3.6.

i APPLICABILITY: OPERA 110NAL CONDITIONS l' and 2*

ACTION a.

N ing ingle loop operation, with volumetric recirculation loop flow rate greater than 33,000 gra, immediately initiate t.orrective

(

iction to reduce flow to lens than or equal to 33,0no gpa within I hour, i

i b.

During sing 1" loop operation, with the recirculation fW control i

system not in'the Loop Manual mode, impediately initiate corrective action to place the recirculation flow control system in the Loop i

Manuel mode within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

c.

During single loop operation, with THERMAL POWER greater than 70% of 5

RATED THERMAL POWER, immediately initiate corrective action to reduce THERMAL POWER to less than or equal to 70% of RATED THERMAL POWER within I hour, d.

Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months upon entry into single loop operation, verify that the operating limits in Specification 3.2.1 have beer appropriately adjusted for single loop operation, j

l

See Special Exception 3.10.4 l

RIVER BEND - UNIT 1 3/4 4-1 Amendment No. 31

REACTOR C0OLANT SYSTEM LIMITING CONDITION FOR OPERATION Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months upon entry into single loop operation, verify that e.

the setpoints in Specifications 2.2.1, 3.2.2 and 3.3.6 are within appropriate Ifmits.

f.

During single loop operation with either THERML POWER 130iof RATED THERMAL POWER or recirculation loop flow in the operating loop is 1 50% of rated recirculation loop flow and temperature differences excoeding the limits in Surveillance Requirement 4.4.1.1.4, suspend THERMAL POWER or recirculation loop flow increase..*

l g.

With one or two reactor ccolcnt system recirculation loops in i

operation and total core fin + greater than 39% and less than 45% of f

rated core flow and THERMAi /0WER greater than the limit specified in Region !! of Ifgure 3.4.1.1 1:

1.

Determine the APRM and LPRM** noise levels (Surveillance 4.4.1.1.2):

a)

At least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , and b)

Within 30 minutes after completion of a THERMAL POWER increase of at least 5% of RATED THERMAL POWER.

2.

With the APRM or LPRM** neutron flux noise levus greater than three tires their established baseline noise levels, immediately initiate corrective action to restore the noise levels within the required limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months by increasing core flow to greater than or equal to 45% of rated core flow or by reducing THERMAL POWER to less than or equal to the limit specified in Region !! of Figure 3.4.1.1-1.

h.

With one or two reactor coolant system recirculation loops in operation and total core flow less than 39% of rated core flow and THERMAL POWER greater than the limit specified in Region !!! of Figure 3.4.1.1-1, irnediately within 15 minutes initiate corrective action to increase core flow to greater than or equal to 39% of rated core flow or reduce THERMAL POWER to less than the limit specified in Region !!! of Figure 3.4.1.1-1 within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

"With one recirculation loop not in operation and isolated, the differential temperature requirements of Surveillance Requirement 4.4.1.1.4b and c are not applicable, and the provisions of Specification 3.0.4 are not spplicable with respect to Surveillance Requirement 4.4.1.1.4b and c.

- Detector levels A and C of one LPRM string in the center of the core should be monitored.

RIVER BEND - UNIT 1 3/4 4-la Amendment No. 31

i i

REACTOR COOLANT SYSTEM SURVElLLANCE REQUIREMENTS f

i 4.4.1.1.1 Each reactor coolant systen recirculation loop flow control valve i

shall be demonstrated OPERABLE at least once per 18 months by:

}

a.

Verifying that the control valve fatis "as is" on loss of hydraulic pressure at the hydraulic control unit, and b.

Verifyingthattheaveragerateofcontrolvalvemovementih:

l 1.

Less than or equal to 11% of stroke per second opening, and i

2.

Less than or equal to 11% of stroke per second closing l

4.4.1.1.2 Establish a beseline APRM and LPRM* neutron flux noise valve within i

the regions for which monitoring is required (Specification 3.4.1.1 ACTION c) within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of entering the region for which monitoring is required unless baselining has previously been performed in the region since the last

{

refueling outage.

i f

4.4.1.1.3 Initially, within I hour upon entry into single loop operation and I

once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereaf ter, verify that:

l a.

THERMAL POWER is less than or equal to 70% of RATED THERMAL POWER, and b.

The recirculation flow control s,fstem is in the Loop Manual (Position Control) mode, and c.

The volumetric recirculation flow rate is less than or equal to 33,000 gpm.

4.4.1.1.4 With one reactor coolant system recirculation loop not in i

operation, and either THERMAL POWER less than or equal to 30% of RATED THERMAL 6

POWER or the recirculation loop f1w in the operating loop is less than or equal to 50% of rated recirculation loop flow, within 15 minutes prior to an

)

increase in THERMAL POWER or recirculation loop flow, verify that the follo. wing differential temperature requirements are met:

j a.

< 100'F between reactor vessel steam space coolant and bottom head

' drain line coolant, and

Detector levels A and C of one LPFJi string per core octant plus detectors A f

and C of one LPRM string in the center of core should be monitored.

[

i RIVER BEND - UNIT 1 3/4 4-2 Amendment No.31

(

i l

4 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) i b.

< 50'F between the reactor coolant within the loo Ind the coolant in the reactor pressure vessel **,p not in operation and

< 50*F between the reactor coolant within the loop not in op5 ration c.

i and the operating loop.**

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- With ont recirculation loop not in operation and isolated, the differential l

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temperature require: crit of Surveillance Requirement 4.4.1.1.4b and c are I

not applicab!a and the provision of Surveillance Requirement 4.0.4 are not l

applicable with respect to Surveillance Requirement 4.4.1.1.4b and e l

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l RIVER BEND - UNIT 1 3/4 4 2a Amendment No.31 k

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FIGURE 3.4.1.1-1 THERMAL POWER VERSUS CORE FIDW l,

i 1

i i

RIVER BEND - UNIT 1 3/4 4 3 A.endrent No. 31

REACTORCOOLANTS3T3 JET PUMPS LIMITING CONDITION FOR OPERATION 3.4.1.2 All jet pueos shall be OPERA 8LE.

APPLICABILITY: OPERATIONAL CONDITI0E 1 and 2.

,NTION:

Withoneormorejetpumpsinoperable,beinatleastHOTSHUTDOWNwithin 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

[

SURVEILLANrr REQUIREMENTS 4.4.1.2.1 During two recirculation loop operation each of the above required

(

jet pumps shall be demonstrated OPERABLE prior to THERMAL POWER exceeding 25%

of RATED THERMAL POWER, and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months while greater than 25%

of RATED THERMAL POWER, by determining recirculation loop flow, total core flow and diffuser-to-lower plent.m differential pressure for each jet pump and verifying that no two of the following conditiens occur when both recirculation I

loop indicated flows are in complianco with Specification 3.4.1.3.

The indicated reeirculation loop flow differs by more than 10% from a.

the established flow control valve position-loop flow characteristics.

(

[

b.

The indicated total core flow differs by more than 10% from the established total core flow value derived from recirculation loop

(

flow reasurements, The indicated diffuser-to-lower plenum differential pressure of any c.

individual jet pump differs from established patterns by more than l

10%.

i d.

The provisions of Specification 4.0.4 are not Opplicable provided that this surveillance is performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after ext:eeding l

25% of RATED THERML POWER.

4.4.1.2.2 During single recirculation loop operation, each of the required I

jet pumps in the operating recirculation loop shall be demonstrated OPERABLE at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months while greater than 25% of RATED THERMAL POWER, by determining recirculation loop flow in the operating loop, total core flow and diffuser-to-lower plenum differential pressure for each jet pump in the operating loop and verifying thst no two of the following conditions occur:

a.

The indicated recirculation loop flow in the operating loop differs by more than 10% from the established

single recirculation flow control valve positions - loop flow characteristics.

To be determined dun ng initial use of single loop operation.

Survaillance Requirements of 4.4.1.2 are not required to allow determination of characteristic curves.

e RIVER BEND - UNIT 1 3/4 4 4 A.mendment No. 33 i

].

REACTOR COOLANT SYSTEM I

j SURVEILLANCE REQUIREMENTS (Continued) b.

The indicated total core flow differs by more than 10% from the j

{

established *j6tpumpf1w/recirculationpumpflowcharacteristic l

for the operating loop, i

1 1

The individual diffuser to-lower plenum differential pressure of any c.

individual jet pump di*fers from established

single recirculation loop patterns by more whan 10%.

i b

d.

The provisions of speciff,'ation 4.0.4 are not applicable provided that this surveillance is performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding i

l 25% of RATED THERMAL POWER.

l 4

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1 i

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I j

1 7

d i

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l I

f i

1 I

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l

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i f

I

'To be dettrained during initial use of single loop operation.

Surveillance Requirements of 4.4.1.2 are not required to allow determination of characteristic curves, j

i RIVER BEND - UNIT 1 3/4 4-4a A eneent No. 31 1

i I

L i

t I

l I

l

i REACTOR COOLANT SYSTEM RECIRCULATION LOOP FLOW LIMITING CONDITICN FOR OPERATION

)

j t

3.4.1.3 Recirculation loop flow mismatch shall be maintained within:

4

)

a.

5% of rated recirculation flow with core flow greater than of equal to 70% ot' rated core flow.

4 I

b.

10% of rated recirculation flow with core flow less than 70% of rated core flow, t

t I

APPLICA8!LITY: OPERATIONAL CONDITIONS 18 and 2* during two recirculation loop I

i operation.

l 4

l ACTION:

i i

With recirculation loop flows different by more than the specified limits, i

}

either*

L l

a.

Restore the recirculation loop flows to within the sp2cified limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , or i

J l

)

b.

Shatdown one of the recirculation loops and take the ACTION required by Specification 3.4.1.1.**

i i

l l

SURVEILLANCE REQUIREMENTS k

(

i 4.4.1.3 Recirculation icop flow mismatch shall be verified to be dthin the l

)

limits at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

J l

I J

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i

See Special Test Exception 3.10.4.

i

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i

- The provisions of Specification 3.0.4 are not Applicable.

l l

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RIVER BEND - UNIT 1 3/4 4,,

Amendment No. 31 l

l

[

REACTOR COOLANT SYSTEM 10LE RECIRCULATION LOOP STARTUP LIMITING CONDITION FOR OPFRATION r

3.4.1.4 An idle recirculation loop shall not be started unless the temperature differential between the reactor pressure vessel steam space coolant and the bottom head drain line coolant is less than or equal td 100*F," and:

\\

a.

When both loops have been idle, unless tha temperature differential between the reactor coolant within the id'.e loop to be started up and the coolant in th) reactor pressure vessel is less than or equal to 50*F, or b.

When only one loop has been idle, unless the temperature differential between the reactor coolant within the 'J1e and operating recirculation loops is less than or equal to 50'F and the operating loop flow rate is less than or equal to 50% of rated loop flow.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3 and 4.

l ACTION:

With temperature differences and/or flow rates exceeding the above limits, suspend startup of any idle recirculation loop.

l

}URVE!LLANCE WEQUIREMENTS I

4.4.1.4 The temperature dif ferentials and flow rate shall be determined to be within the limits within 15 minutes prior to startup of an idle recirculatico loop.

I l

1 1

I 1

"8elow 25 psig, this temperature differential is not applicable.

I i

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4 1

RIVER REND - UNIT 1 3/4 4 6

)

. - - - _ - ~.

f I.

f j

_ REACTOR COOLANT SYSTEMS i

BASES j

1/4.1.3 CONTROL R005

[

ihe specifications of this section (1) ensure that,the minious 5HUTDOWN MAGIN is asintained and the control rod insertion times are consistent with i

l uose used in the safety analyses, and (2) limit the potential effects'of l

the rod drcp accident.

The ACTION statements permit variations from the basic

{

requirements but impose more restrictive criteria for continued operation. A liettation on inoperable rods is set soch that the reswitant effer.t on total Pod worth and scram shape will be kept to a minimum. The requirements for the i

j various scram time measurements ensure that any indication of systematic pro-blems with rod drives will be investigated on a timely basis.

Damage within the control rod drive mechanism could be a generic problem.

l Therefore, with a cor. trol rod immovable because of excessive friction or i

.l mechanical interference, operation of the reactor is limited to a time peried that is long enough to permit determining the cause of the inoperability yet j

prevent operation with a large number of inoperable control ads.

i Control rods that are inoperable for other reasons are pa.eltted to be taken out of service provided that those not fully inserted are consistent with the SHUTDOWN MRGIN zquirements.

The number of control rods permitted to be inoperable could be more'than l

the eight allowed by the specification, but the occurrence of eight inoperable rods could be indicative of a generic problem and the reactor must be st. 4 down for investigation and resolution of the problem.

s The centrol rod system is designed to bring the reactor subcritical at a

{

rate fast enough to prevent the MCPR from becoming less than the fuel cladding, safety ifmit during the limiting power transient analyzed in Section 15.0 of f

the FSAR.

This analysis shows that the negative reactivity rates, resulting i

froa the scram with the average response of all the drives as given in the specifications, provide the equired protection and MCPR remains greater than the fuel cladding safety limit. The occurrence of scram times longer then I

those specified should be viewed as an indication of a systeetc problem with l

the red drives and, therefore, the surveillance interval (A reduced in order to prevent operation of the reactor for long periods of time with a potentially serious probles.

I The scram discharge volume is required to be OPERA 8LE so that it will be l

available when needed to accept discharge water from the control rods during a l

reactor scram and will isolate the reactor coolant system fro ine :ontainment t

when required, i

Control rods with inoperable accumulators are declares te.

le and I

s Specification 3.1.3.1 then applies.

This prevents a pattern ei i sperable f

l accumulators that would result in less reactivity insertion on a scram than has been analyzed even though control rods with inoperable accumulators say i

l stiil be inserted with normal drive water pressure. Operability of the accumulaar ensures that there is a means available to insert the control rods i

even under the most unfavorable depressurization of the reactor.

{

RIVER BEND UNIT 1 0 3/4 1-2 Amendment No. 22,31

(

f i

f

3/4.1 REACTIVITY CONTROL SYSTEMS BASES s

3/4.1.1 SHUTOOWN MARGIN A sufficient SHUTDOW Koti!N ensures that 1) the reactor can be made sub-critical from all operating cunditions, 2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and 3) the reactor will be maintained sufficiently suberitical to preclude inadvertent criticality in the shutdown condition.

Sincecorereactivityvalueswillvarythroughcorelifeasafdctionof fuel depletion and poison burnup, the demonstration of SHUTOOWN MARGIN will be performed in the cold, xenon-free condition and shall show the core to be suberitical by at least R + 0.38% delta k/k or R + 0.28% delta k/k, as appro-priate. The value of R in units of % delta k/k is the dif.ference between the cniculated value of maximum core reactivity during the operating cycle and the calculated beginning-of-life core reactivity.

The value of R must be positive or zero and must be determined for each fuel loading cycle.

Two different values are supplied in the Limiting Con,ition for Operation to provide for the different methods of demonstration of the SHUTDOWN MARGIN.

The highest worth rod may be de.termined analytically or by test.

The SHUTDOWN MARGIN is demonstrated by an in-sequence control rod withdrawal at the beginning-of-life fuel cycle conditions and, if necessary, at any future time in the cycle if the first demonstration ir,dicates that the required sargin could be reduced as a function of exposure.

Observntion of suberiticality in this condition assures suberiticality with the most raketive control rod fully withdrawn.

This reactivity characteristic has been a basic assumption in the analysis of plant perfornance and can be best demonstrated at the time of fuel loading, cut the eargin must also be determined any time a t.ontrol rod is incapable of.

insertfon.

3/4.1.2 REACTIVITY ANOMALIES Since the SHUTDOWN MARGIN requirement for the reactor is small, a careful comparison of actual conditions ;o the predicted conditions is necessary, and the changes in reactivity can be inferred from these comparisons of rod patterns.

Sin:;e the comparisons are easily done, frequent checks are not an imposition on normal operations. A 1% change is larger than is expected for normal operation so a change of this magnitude should be thoroughly evaluated.

A change as large as 1% would not exceed the design conditions of the reactor and is on the safe side of the postulated transients.

RIVER BEND - UNIT 1 8 3/4 1-1 A,-

3/4.2 POWER DISTRIBUT!ON LIMITS BASES 1

The specifications of thiJ section assure that the peak cladding temper-ature following the postulated design basis loss-of-coolant accident will not l

exceed the 2200*F limit specified in 10 CFR 50.46.

3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE The peak cladding temperature (fCT) following a postulated loss-of-coolant accident is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any exial location and is dependent only second-arily on the rod to rod power distribution within an assembly. The peak clad l

temperature is calculated assuming a LHGR for the' highest powered rod which is equal to or less than the design LHGR corrected for densification.

This LHGR l

times 1.02 is used in the heatup code along with the exposure-dependent steady state gap conductance and rod-to-rod local peaking factor.

The Technical Specification AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) is this LHGR of the highest powered rod divided by its local peaking factor. The limiting value for APLHGR is shown in Figures 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4, 3.2.1-5 and 3.2.1-6.

The daily requirement for calculating APLHGR when THERMAL POWER is greater than or equal to 25% of RATED THERMAt POWER is sufficient since power distribu-tion shifts are very slow when there have not been significant power or control red changes.

The requirement to calculate API.HGR within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after the com -

pietion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER ensures thermal limits are met after power distribution shifts while still allotting time for the power distribution to stabilize. The requirement for calculating APLHGR after initially c'etermining a LIMITING CONTROL ROD PATTERN ex!sts ensures that APLHGR will be known following a change in THERMAL POWER or power shape that could place operation into a condition exceeding a thermal limit.

The calculational procedure used to establish the APLHGR shown on Figures 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4, 3.2.1-5 and 3.2.1-6 it based on a loss-of-coolant accident analysis.

The analysis was performed using General Electric (GE) calculational roodels which are consistent with the requirem?nts of Appendix K to 10 CFR 50.

A complete discussion of each code employed in the analysis is presented in NEDE-20566(1)

Differences in this analysis compared to previous analyses can be broken down as follows.

a.

Input Chances 1.

Corrected Vaporization Calculation - Coefficients in the vaporization correlation used in the REFLOOD code were corrected.

2.

Incorporated more accurate bypass areas - The bypass areas in the top guide were recalculated using a more accurate technique.

RIVER BEND - UNIT 1 B 3/4 2-1 Amendment No. 12

POWER DISTRIBUTION LIMITS BASES AVERAGE PLANAR LINEAR HEAT GENERATION RATE (Continued) 3.

Corrected guide tube thermal resistance.

4.

Correct heat capacity of reactor internals heat nodes.

b.

Model Change 1.

Core CCFL pressure differential - 1 psi - Incorpora'.e the assumption that flow from the bypass to lower plenum must overcome a 1 psi pressure drop in core.

2.

Incoporate NRC pressure transfer assumption - The assumption used in the SAFE-REFLOOD pressure transfer when the pressure is increasing was changed.

A few of the changes affect the accident :alculation irrespective of CCFL.

These changes are listed below, a.

Input Change 1.

Break Areas - The DBA break area was calculated more accurately, b.

Model Change 1.

Improved Radiation and Conduction Calculation - Incorporation of CHASTE-05 for heatup calculation.

A list of the significant plant input parameters to the loss-of-coolant accident enalysis is presented in Bases Table B 3.2.1-1.

For plant operation with a single recirculation loop, the MAPLHGR limits of figures 3.2.1-1 through 3.2.1-6 are multiplied by 0.84.

The constant factor 0.84 is derived from LOCA analyses initiated from single recirculatic.n loop operation to account for earlier boiling transition at the limiting fuel mode compared to the standard LOCA evaluations.

3/4.2.2 APRM SETPOINTS The fuel cladding integrity Safety Limits of Specification 2.1 were based on a power distribution which would yield the design LHGR at RATED THERMAL POWER.

The flow biased simulated thermal power-high scram trip setpoint and the flow biased neutron flux-upscale control rod block trip setpoints of the APRM instru-ments must be adjusted for both two recirculation loop operation and single recirculation loop operation to ensure that MCPR does not become less than the fuel cladding safety limit or that > 1% plastic strain does not occur in the degraded situation.

The scram settTngs and rod block sottings are adjusted in accordance with the formula in this specification, when the combination of THERMAL POWER and CHFLPD indicates a peak power distribution, to ensure that an LHGR transient would not be increased in degraded conditions.

RIVER BEND - UNIT 1 B 3/4 2-2 Amendment No. 31

POWER DISTRIBUTION LIMITS Bases Table B 3.2.1-1 SIGNIFICANT INPUT PARAMETERS TO THE LOSS-OF-COOLANT ACCIDENT ANALYSIS Plant Parameters; Core THERMAL POWER....................

3015 Mwt* which corresponds to 105% of rated steam flow 6

Ves sel Steam Output...................

13. 08 x 10. lbm/hr which corresponds to 105% of rated steam flow Vessel Steam Dome Pressure.............

1060 psia Design Basis Recirculation Line Break Area for:

2 a.

Large Breaks 2.2 ft,

2 b.

Small Breaks 0 09 ft,

Fuel Parameters:

PEAK TECHNICAL INITIAL SPECIFICATION DESIGN HINIMUM LINEAR HEAT AXIAL CRITICAL FUEL ASSEMBLY GENERATION RATE PEAXING POWER FUEL TYPE GEOMETRY (kw/ft)

FACTOR RATIO l

1.4 1.17**

l Initial Core 8x8 13.4 A more detailed listing of input of each model and its source is presented in Section II of NEDE 20566(1) and subsection 6.3.3 of the FSAR.

l "This power level meets the Appendix X requirement of 102%. The core heatup calculation assumes an assembly power consistent with operation of l

the highest powered rod at 102% of its Technical Specification LINEAR HEAT GENERATION RATE limit.

- For single recirculation loop operation, loss of nucleate boiling is assumed I

at 0.01 after LOCA regardless of initial MCPR.

RIVER BEND - UNIT 1 B 3/4 2-3 Amendment No. 31

POWER DISTRIBUTION LIMITS BASES 3/4.2.3 MINIMUM CRITICAL POWER RATIO The required operating limit MCFRs at steady state operating conditions as specified in Specification 3.2.3 are derived from the established fuel cladding integrity Safety Limit MCPR of 1.07 and an analysis of abnorgal operational transients.

For any abnormal operating transient analysis, with the initial condition of the reactor being at the steady state operating limit, it is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any time during the transient assuming instrument trip settings given in Specification 2.2.

To assure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnormal operational transient, the most limiting tran-sients have been analyzed to determine which result in the largest reduction in CRITICAL POWER RATIO (CPR).

The type of transients evaluated were loss of -

flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease.

The limiting transient yields the largest delta MCPR.

When added to the Safety Limit MCPR of 1.07, the required minimum operating limit MCPR of Specification 3.2.3 is obtained and is presented in Figure 3.2.3-1.

Analysis of transients occurring during single recirculation loop operation indicates that the maximum operating limit MCPR will be bounded by the limits in Specification 3.2.3.

The power-flow map of Figure B 3/4 2.3-1 shows typical regions of plant operation.

The evaluation of a given transient begins with the system initial param-eters identified in Reference 2 that are input to a GE core dynamic behavior transient computer program.

The codes used to evaluate transients are described in Reference 2.

The principal result of this evaluation is the reduction in MCPR. caused by transient.

The purpose of the MCPR and MCPR of Figures 3.2.3-1 and 3.2.3-2 is to f

p define operating limits at other than rated core flow and power conditions.

At less than 100% of rated flow and power the required MCPR is the larger value of the MCPR and MCPR at the existing core flow and power state.

The MCPR s f

p f

are established to protect the core from inadvertent core flow increases such that the 99.9% MCPR limit requirement can be assured.

The MCPR s were calculated such that, for the maximum core flow rate and f

the corresponding THERMAL POWER along the 105%-of-rated steam flow control line, the limiting bundle's relative power was adjusted until the MCPR was slightly above the Safety Limit.

Using this relative bu'1dle power, the MCPRs were calcu-lated at different points along the 105% of-rated steam flow control line corresponding to different cora flows.

The calculatad MCPR at a given point of core flow is defined as MCPR.

f RIVER BEND - UNIT 1 B 3/4 2-4 Amendment No. 22, 31

3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 RECIRCULATION SYSTEM The impact of single recirculation loop operation upon plant safety h&s been assessed and single recirculation loop operation is permitted if the MCPR fuel cladding safety limit is increased as noted by Specification 2.12; APRM scram and control rod block setpoints are adjusted as noted in Tables 2.2.1-1 and 3.3.6-2, respectively; MPLHGR limits are decreased by the factor given in Specification 3.2.1 (Reference 3). MCPR operating limits are adjusted per specification 3/4.2.3, for both single and two recirculation loop operation.

Additionally, surveillances on the volumetric flow ra'te of the operating recirculation loop is imposed to exclude the possibility of excessive core internal vibration.

The surveillance on differ.ential temperatures below 30%

THERMAL POWER or 50% rated recirculation loop flow is to mitigate the undue thermal stress on the vessel nozzles, recirculation pump and vessel bottom head during extended operation in the single recirculation loop mode.

An inoperable jet pump is not, in itself, a sufficient reason to declare a recirculation loop inoperable, but it does, in case of a design basis accident, increase the blowdown area and reduce the capability of reflooding the core; thus, the requirement for shutdown of the facility with a jet pump inoperable.

Jet pump failure can be detected by monitoring jet pump performance on a prescribed schedule for significant degradation.

During single loop operation the jet pump operability surveillances are only performed for the jet pumps in the operating recirculation loop, as the loads on the inactive jet pumps are expected to be very low due to the low flow in the reverse direction through the jet pumps.

Recirculation loop flow mismatch limits are in compli '

ance with ECCS LOCA analysis design criteria for two recirculation loop operation.

The limits will ensure an adequate core flow coastdown from either recirculation loop following a LOCA.

In the case where the mismatch limits cannot be maintained during two recirculation loop operation, continued operation is permitted in a single recirculation loop operation mode.

In order to prevent undue stress on the vessel nozzles and bottom head region, the recirculation loop temperatures shall be within 50'F of each other prior to startup of an idle loop.

The loop temperature must also be within 50'F of the reactor pressure vessel coolant temperature to prevent thermal shock to the recirculation pump and recirculation nozzles.

Sudden equalization of a temperature difference >100'F between the reactor vessel bottom head coolant and the coolant in the upper region of the reactor vessel by increasing core flow rate would cause undue stress in the reacter vessel bottom head.

The objective of GE BWR plant and fuel design is to provide stable opera-tion witti margin over the normal operating domain.

However, at the high power /

low flow corner of the operating domain, a small probability of limit cycle neutron flux oscillations exists depending on combinations of operating condi-tions (e.g., rod pattern, power shape). To provide assurance that neutron flux limit cycle oscillations are detected and suppressed, APRM and LPRM neutron flux noise levels should be monitored while operating in this region.

RIVER BEND - UNIT 1 B 3/4 4-1 Amendment No. 31

REACTOR COOLANT SYSTEM 3/4.4 REACTOR COOLANT SYSTEM BASES I

RECIRCULATION SYSTEM (Continued)

Stability tests at operating BWRs were reviewed to determine a geheric region of the power / flow map in which surveillance of neutron flux noise levels should be performed.

A conservative decay ratio of 0.6 was chosen, as the basis for determining the generic region for surveillance, to account for the plant-to plant variability of decay ratio with core and fuel designs.

This generic region has been determined to correspond to a core flow of less than or equal to 45% of rated core flow and a thermal power greater than that specified in Figure 3.4.1.1-1 (Reference 1).

Plant-specifi calculations can be performed to determine an applicable region for monitoring neutron flux noise levels.

In this case the degree of conservatism can be reduced since plant-to plant variability would be eliminated.

In this case, adequate margin will be assured by monitoring the region which has a decay ratio greater than or equal to 0.8.

RIVER BEND - UNIT 1 B 3/4 4-la Amendment No. 31

REACTOR COOLANT SYSTEM 3/4.4 REACTOR COOLANT SYSTEM

, BARS dECIRCULATION SYSTEM (Continued)

Neutron flux noise limits are also established to ensure early detection of limit cycle neutron flux oscillations. BWR cores typically operate.'with neutron flux noise caused by random boiling and flow noise. Typical neutron flux noise levels of 1 to 12% of rated power (peak-to peak) have been Yeported for the range of low to high recirculation loop flow during both sing 1'e and dual recirculation loop operation.

Neutron flux noise levels which signifi-cantly bound these values are considered in the thermal / mechanical design of GE BWR fuel and are found to be of neglig'ble consequence (Reference 2).

In addition, stability tests at operating BWRs have demonstrated that when stabil-ity related neutron flux limit cycle oscillations occur they result in peak-to-I peak neutron flux limit cycles of 5 to 10 times the typical values.

Therefore, I

actions taken to reduce neutron flux noise levels exceeding three times the typical value are sufficient to ensure early detection of limit cycle neutron flux oscillations.

l Typically, neutron flux noise levels show a gradual increase in absolute magnitude as core flow is increased (constant control rod pattern) with two reactor recirculation loops in operation. Therefore, the baseline neutron flux noise level obtained at a specific core flow can be applied over a range of core flows.

To maintain a reasonable variation between the low flow and

. high flow ends of the flow range, the.ange over which a specific baseline is applied should not exceed 20% of rated core flow with two recirculation loops in operation. Data from tests and operating plants indicate that a range of 20% of rated core flow will result in approximately a 50% increase in neutron flux noise level during operation with two recirculation loops.

Baseline data should be taken near the maximum rod line at which the majority of operation will occur.

However, baseline data taken at lower rod lines (i.e., lower power) will result in a conservative value since the neutron flux noise level is proportional to the power level at a given core flow.

References (1) "BWR Core Thermal-Hydraulic Stability," Service Information Letter 380, Revision 1, Februar.y 1984.

(2)

G. A. Watford, "Compliance of the General Electric Boiling Water Reactor Fuel Designs to Stability Licensing Criteria," December 1982 (NEDE 22277-P).

(3) "Single-Loop Operation Analysis for River Bend Station, Unit 1,"

NEDO-31441, May 1987.

3/4.4.2 SAFETY / RELIEF VALVES The safety valve function of the safety / relief valves (SRV) is to prevent the reactor coolant system from being pressurized above the Safety Limit of 1375 psig, in accordance with the ASME Code.

A total of 9 OPERABLE safety-relief valves is required to lirit reactor pressure to within ASME III allowable values for the worst case upset transient.

Any combination of 4 SRVs operating in the relief mode end 5 SRVs r,perating in the safety mode is accepi.able.

RIVER BEND - UNIT 1 B 3/4 4-2 Amendment No. 31

REACTOR COOLANT SYSTEM 3/4.4 REACTOR COOLANT SYSTEM BASES SAFETY / RELIEF VALVES (Continued)

Demonstration of the safety-relief valve lift settings will occur.'only during shutdown and will be performed in accordance with the provisions of Specification 4.0.5.

The low-low set system ensures that safety /relicf valve discharges are minimized for a second opening of these valves, following any overpressure transient. This is achieved by automatically lowering the closing setpoint of 5 valves and lowering the opening setpoint of 2 valves following the initial opening.

In this way, the frequency and magnitude of the containment blowdown duty cycle is substantially reduced.

Sufficient redundancy is provided for the low-low set system such that failure of an one valve to open or close at its reduced setpoint does not violate the desi n basis.

~

3/4.4.3 REACTOR COOLANT SYSTEM LEAXAGE 3/4.4.3.1 LEAKAGE DETECTION SYSTEMS The RCS leakage detection systems required by this specification are provided to monitor and detect leaksge from the reactor coolant pressure boundary.

These detection systems are consistent with the recommendations of Regulatory Guide 1.45, "Reactor Coolant Pressure Boundary Leakage Detection Systems", May 1973.

In conformance with Regulatory Guide 1.45, the atmospheric gaseous radioactivity system will have a sensitivity of 10 8 pCi/cc.

3/4.4.3.2 OPERATIONAL LEAKAGE The allowable leakage rates from the reactor coolant system have been based on the predicted and estperimentally observed behavior of cracks in pipes. The normally expected background leakage, due to equipment design and the detection capability of the instrumentation for determining system leakage, was also considered.

The evidence obtained from experfir.ents suggests that, for leakage somewhat greater than that specified for UNIDENTIFIED LEAKAGE, the probability is small that the imperfection or crack associated with such leakage would grow rapidly. However, in all cases, if the leakage rates exceed the values specified or the leakage is located and known to be PRESSURE BOUNDARY LEAKAGE, the reactor will be shut down to allow further investigation and corrective action.

The Surveillance Rcquirements for RCS pressure isolation valves provide added 6ssurance of valve integrity, thereby reducing the probability of gross valve failure and consequent intersystem LOCA.

Leakage from the RCS pressure isolation valves is IDENTIFIED LEAKAGE and will be considered as a portion of the allowed limit.

RIVER BEND - UNIT 1 B 3/4 4-3

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ML20206L522

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