Text

Amend 40 to License NPF-11,revising Tech Specs to Support Operation at full-rated Power During Cycle 2 Operation
	ML20197D832
Person / Time
Site:	LaSalle
Issue date:	05/09/1986
From:	Adensam E; Office of Nuclear Reactor Regulation
To:	;
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ML20197D838	List: ML20197D832; ML20197D855;
References
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	Download: ML20197D832 (22)
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UNITED STATES NUCLEAR REGULATORY COMMISSION o

h WASHINGTON, D. C. 20555

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COMMONWEALTH EDISON COMPANY DOCKET NO. 50-373 LA SALLE COUNTY STATION, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 40 License No. NPF-11 1.

The Nuclear Regulatory Comission (the Comission or the NRC) having found that:

A.

The application for amendment filed by the Commonwealth Edison Company (the licensee), dated October 22, 1985, as supplemented on March 21, 1986, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the regulations of the Commission; C.

There is reascnable assurance: (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted g#

in comoliance with the Comission's regulations set forth in 10 CFR Chapter I; D.

The issuance of this amendment will not be inimical'to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all apolicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifica-tions as indicated in the enclosure to this license amendment and paragraoh 2.C.(2) of the. Facility Operating License No NPF-11 is hereby amended to read as follows:

(2) Technical Specifications and Envirnnmental Protection Plan The Technical Specifications contained in Appendix A, as revised throuah Amendment No. 40, and the Environmental Protection Plan contained in Appendix R, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Techni' cal Specifications and the Environmental Protection Plan.

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This amendment is effective upon startuo following the first refuelina.

FOR THE NUCLEAR REGULATORY COMMISSION A = ^~

Elinor G. Adensam, Director BWR Pro.iect Directorate No. 3 Division of RWR Licensino

Enclosure:

Changes to the Technical Specifications Date of Issuance:

May 9,1986 9

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ENCLOSURE TO LICENSE AMENDMENT NO. 40 FACILITY OPERATING LICENSE NO. NPF-11 DOCKET NO. 50-373 Replace the following pages of the A cendix "A" Technical Specifications with p

the enclosed paces.

The revised rages are identified by Amendment number and contain a vertical line indicating the area of change.

REMOVE INSERT VI VI XIX.

XIX 2-1 2-1 B2-1 B2-1 B2-4 82-4 B2-5 B2-5 R2-6 B2-6 R2-7 82-7 3/4 2-1 3/4 2-1 3/4 2-2 3/4 2-2 3/4 2-2(a) 3/4 2-4 3/4 2-4 3/4 2-5 3/4 2-5 i

3/4 2-6 3/4 2-6 3/4 3-39 3/4 3-39 I.

3/4 4-1 3/4 4-1 3/4 4-la 3/4 4-la 3/4 4-4b 3/4 4 Ab B3/4 4-1 B3/4 4-1 4

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i INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM Recirculation loops..........................................

3/4 4-1 Jet Pumps....................................................

3/4 4-2 Recirculation Loop Flow......................................

3/4 4-3 Idle Recirculation loop Startup..............................

3/4 4-4 Thermal Hydraulic Stability..................................

3/4 4-4a l 3/4.4.2 SAFETY / RELIEF VALVES.........................................

3/4 4-5 3/4 4.3 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems....................................

3/4 4-6 O pe ra ti o nal Le a kage..........................................

3/4 4-7 3/4.4.4 CHEMISTRY....................................................

3/4 4-10 3/4.4.5 SPECIFIC ACTIVITY............................................

3/4 4-13 3/4.4.6 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System.......................................

3/4 4-16 Reactor Steam 0 cme...........................................

3/4 4-20 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES.............................

3/4 4-21 3/4.4.8 STRUCTURAL INTEGRITY.........................................

3/4 422 3/4.4.9 RESIDUAL HEAT REMOVAL Hot Shutdown.................................................

3/4 4-23 Cold Shutdown................................................

3/4/4-24 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ECCS-0PERATING...............................................

3/4 5-1 3 /4. 5. 2 ECCS-SHUT 00WN................................................

3/4 5-6 3/4.5.3 SUPPRESSION CHAM 8ER.........................................

3/4 5-8 LA SALLE - UNIT 1 VI Amendment No. 40

INDEX LIST OF FIGURES FIGURE PAGE

.3.1.5-1 SODIUM PENTABORATE SOLUTION TEMPERATURE /

CONCENTRATION REQUIREMENTS.............................

3/4 1-21 3.1.5-2 SODIUM PENTABORATE (Na2B o0 s

10 H O) i 1

2 VOLUME / CONCENTRATION REQUIREMENTS......................

3/4 1-22 3.2.1-1 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS AVERAGE PLANAR EXPOSURE, INITIAL CORE FUEL TYPES 8CRB176, 8CRB219, AND 8CRB071................................................

3/4 2-2 3.2.1-2 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHER) VERSUS AVERAGE PLANAR EXPOSURE, FUEL TYPE BP8CRB299L...................................

3/4 2-2a 3.2.3-1 MINIMUM CRITICAL POWER RATIO (MCPR) VERSUS t AT RATED FLOW........................................

3/4 2-5 3.2.3-2 K FACTOR..............................................

3/4 2-6 g

3.4.1.1-1 CORE THERMAL POWER (% OF RATED) VERSUS TOTAL CORE FLOW (% OF RATED).................................

3/4 4-lb 3.4.6.1-1 MINIMUM REACTOR VESSEL METAL TEMPERATURE VS. REACTOR VESSEL PRESSURE............................

3/4 4-18

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4.7-1 SAMPLE PLAN 2) FOR SNUBBER FUNCTIONAL TEST............

3/4 7-32 8 3/4 3,1 REACTOR VESSEL WATER' LEVEL.............................

B 3/4 3-7 8 3/4.4.6-1 CALCULATED FAST NEUTRON FLUENCE (EJ1MeV) at 1/4 T AS A FUNCTION OF SERVICE LIFE..........................

B 3/4 4-7 5.1.1-1 EXCLUSION AREA AND SITE BOUNDARY FOR GASEOUS AND LIQUID EFFLUENTS...................................

5-2 5.1.2-1 LOW POPULATION ZONE....................................

5-3 6.1-1 CORPORATE MANAGEMENT...................................

6-11 6.1-2 UNIT ORGANIZATION.......................................

6-12 6.1-3 MINIMUM SHIFT CREW COMPOSITION.........................

6-13 LA SALLE - UNIT 1 XIX Amendment No. 40

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.

APPLICABILITY:

OPERATIONAL CONDITIONS 1 and 2.

ACTION:

With THERMAL POWER exceeding 25% of RATED THERMAL POWER and the reactor vessel steam dome pressure less than 785 psig or core flow less than 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 with the requirements of Specification 6.4.

THERMAL POWER, High Pressure and High 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 than 1.08 with single recirculation loop operation with the reactor vessel steam dome pressure greater than 785 psig and core flow greater than 10% of rated flow.

APPLICABILITY:

OPERATIONAL CONDITIONS 1 and 2.

ACTION:

With MCPR less than 1.07 with two recirculation loop operation or less than 1.08 with single recirculation loop operation and the reactor vessel steam dome pressure greater than 785 psig and core flow greater than 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 with the require-ments of Specification 6.4 REACTOR COOLANT SYSTEM PRESSURE 2.1.3 The reactor coolant system pressure, as measured in the reactor vessel steam dome, shall not exceed 1325 psig.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, 3 and 4.

ACTION:

With the reactor coolant system pressure, as measured in the reactor vessel steam dome, above 1325 psig, be in at least HOT SHUTDOWN with reactor coolant system pressure 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 the requirements of Specification 6.4..

LA SALLE UNIT 1 2-1 Amendment No. 40

2.1 SAFETY LIMITS BASES 2.0 The fuel cladding, reactor pressure vessel and primary system piping, are the principal barriers to the release of radioactive materials to the environs.

Safety Limits are established to protect the integrity of these barriers during normal plant operations and anticipated transients. The fuel cladding integrity Safety Limit is set such that no fuel damage is calculated to occur if the limit is not violated.

Because fuel damage is not directly observable, a step-back approach is used to establish a Safety Limit such that the MCPR is not less than 1.07.

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 cladding integrity.

The fuel cladding is one of the physical barriers which separate the radioactive materials from the environs.

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 incre-mentally cumulative and continuously measurable.

Fuel cladding perforations, however, can result from thermal stresses which occur from reactor operation significantly above design conditions and the Limiting ~ Safety System Settings.

While fission product migration from cladding perforation is just as 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 deterioration.

Therefore, the fuel cladding Safety Li nit is defined with a margin to the conditions which would produce onset of transition boiling, MCPR of 1.0.

These conditions represent a signif-icant departure from the condition intended by design for planned operation.

2.1.1 THERMAL POWER, Low Pressure or Low Flow The use of the GEXL correlation 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 cladding integrity Safety Limit is established by other means.

This is done by establishing a limiting condition on ccre 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 x 103 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 greater than 28 x 103 lbs/hr.

Full scale ATLAS test data taken at pres-sures 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 THERMAL POWER.

Thus, a THERMAL POWER limit'of 25% of RATED THERMAL POWER for reactor pressure below 785 psig is conservative.

l l

l LA SALLE - UNIT 1 B 2-1 Amendment No. 40 l

Bases Table 82.1.2-1 UNCERTAINTIES USED IN THE DETERMINATION OF THE FUEL CLADDING SAFETY LIMIT

Standard Deviation Quantity

(% of Point)

Feedwater Flow 1.76 Feedwater Temperatura 0.76 Reactor Pressure O.5 Core Inlet Temperature 0.2 Core Total Flow 2.5 Two recirculation Loop Operation Single recirculation Loop Operation 6.0 Channel Flow Area 3.0 Friction Factor Multiplier 10.0 Channel Friction Factor Multiplier 5.0 TIP Readings 8.7 l

Two Recircula. tion Loop Operation Single Recirculation Loop Operation 6.8 R Factor 1.6 l

Critical Power 3.6

The uncertainty analysis used to establish the core wide Safety Limit MCPR is based on the assumption of quadrant power symmetry for the reactor core.

The values herein appply to both two recirculation loop operation and single recirculation loop operation, except as noted.

LA SALLE - UNIT 'l B 2-4 Amendment No. 40

Bases Table B2.1.2-2 NOMINAL VALUES OF PARAMETERS USED IN THE STATISTICAL ANALYSIS OF FUEL CLADDING INTEGRITY SAFETY LIMIT THERMAL POWER 3293 MW l

Core Flow 102.5 M1b/hr l

Dome Pressure 1010.4 psig R-Factor 1.038 - 0 GWD/t 1.031 - 7 GWD/t 1.030 - 15 GWD/t 1.033 - 20 GWD/t LA SALLE - UNIT 1 B 2-5 Amendment No. 40

Bases Table B2.1.2-3 RELATIVE BUNOLE POWER DISTRIBUTION USED IN THE GETA8 STATISTICAL ANALYSIS Percent of Fuel Bundles Within Range of Relative Bundle Power Power Interval 1.375 to 1.425 5.1 1.325 to 1.375 7.3 1.275 to 1.325 7.8 1.225 to 1.275 9.8 1.175 to 1.225 7.3 1.125 to 1.175-11.8 1.075 to 1.125 4.7 1.025 to 1.075 4.7

<1.025 41.5

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150 U LA SALLE - UNIT 1 B 2-6 Amendment No. 40

Bases Table B2.1.2-4 R-FACTOR DISTRIBUTION USED IN GETAB STATISTICAL ANALYSIS 8x8 Rod Array R-Factor Rod Sequence No.

1.038 1

1.038 2

1.037 3

1.037 4

1.035 5

1.035 6

1.030 7

11.030 8 through 64 4

LA SALLE - UNIT 1 B 2-7 Amendment No. 40

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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 LINEAR HEAT GENERATION RATES (APLHGRs) for each type of fuel as a function of AVERAGE PLANAR EXPOSURE shall not exceed the limits 1shown in Figures 3.2.1-1 and 3.2.1-2.

The limits of F.igures 3.2.1-1 and l

3.7.1-2 shall be reduced to a value of 0.85 times the two recirculation loop fc,seration limit when.in single r,ecirculation loop operation.

APPLICABILITY:

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

ACTION:

i j

With an APLHGR exceeding the limits of Figures 3.2.1-1 and 3.2.1-2, initiate l

corrective action within 15 minutes and 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 .

i I

3 1

SURVEILLANCE REQUIREMENTS l

4.2.1 All APLHGRs shall be verified to be equal to or less than the limits determined from Figures 3.2.1-1 and 3.2.1-2:

a.

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

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 at least 15% of RATED THERMAL POWER, and c.

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 with a LIMITING CONTROL R00 PATTERN for APLHGR.

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i LA SALLE UNIT 1 3/4 2-1 Amendment No. 40 l

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POWER DISTRIBUTION LIMITS 3/4.2.3 MINIMUM CRITICAL POWER RATIO LIMITING CONDITION FOR OPERATION 3.2.3 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be equal to or greater than the MCPR limit determined from Figure 3.2.3-1 times the K determined f

from Figure 3.2.3-2 for two recirculation loop operation and shall be equal to or greater than the MCPR limit determined from Figure 3.2.3-1 + 0.01 times the K determined from Figure 3.2.3-2 for single recirculation loop operation.

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APPLICABILITY:

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

ACTION With MCPR less than the applicable MCPR limit determined from 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 i

4.2.3 MCPR, with:

a.

t

= 0.86 prior to performance of the initial scram time measurements ave for the cycle in accordance with Specification 4.1.3.2, or b.

t determined within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of the conclusion of each scram time ave surveillance test required by Specification 4.1.3.2, shall be determined to be equal to or greater than the applicable 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 after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and c.

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 with a LIMITING CONTROL R00 PATTERN for MCPR.

LA SALLE UNIT 1 3/4 2-4 Amendment No. 40

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INSTRUMENTATION END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.4.2 The end-of-cycle recirculation pump trip (EOC-RPT) system instrumenta-tion channels shown in Table 3.3.4.2-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.4.2-2 and with the END-0F-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME as shown in Table 3.3.4.2-3.

APPLICABILITY:

OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 30% of RATED THERMAL POWER.

ACTION:

a.

With an end-of-cycle recirculation pump trip system instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.4.2-2, declare the channel inoperable until the channel is restored to OPERABLE status with the channel setpoint adjusted consistent with the Trip Setpoint value.

b.

With the number of OPERABLE channels one less than required by the Minimum OPERABLE Channels per Trip System requirement for one or both trip systems, place the inoperable channel (s) in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

c.

With the number of OPERABLE channels two or more less than required by the Minimum OPERABLE Channels per Trip System requirement (s) for one trip system and:

1.

If the inoperable channels consist of one turbine control valve channel and one turbine stop valve channel, place both inoperable channels in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

2.

If the inoperable channels include two turbine control valve channels or two turbine stop valve channels, declare the trip system inoperable.

d.

With one trip system inoperable, restore the inoperable trip system to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or reduce THERMAL POWER to less than 30% of RATED THERMAL POWER within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

e.

With both trip systems inoperable, restore at least one trip system to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or reduce THERMAL POWER to less than 30% of RATED THERMAL POWER within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

1 l

l LA SALLE - UNIT 1 3/4 3-39 Amendment No. 40 l

3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM RECIRCULATION LOOPS LIMITING CONDITION FOR OPERATION 3.4.1.1 Two reactor coolant system recirculation loops shall be in operation.

APPLICABILITY:

OPERATIONAL CONDITIONS 1* and 2*.

ACTION:

a.

With one reactor coolant system recirculation loop not in operation:

1.

Within 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months s:

a)

Place the recirculation flow control system in the Master Manual mode, and b)

Increase the MINIMUM CRITICAL POWER RATIO (MCPR) Safety Limit by 0.01 to 1.08 per Specification 2.1.2, and, c)

Increase the MINIMUM CRITICAL POWER RATIO (MCPR) Limiting i

Condition for Operation by 0.01 per Specification 3.2.3,

and, d)

Reduce the MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION I

RATE (MAPLHGR) limit to a value of 0.85 times the two recirculation loop operation limit per Specification 3.2.1,

and, e)

Reduce the Average Power Range Monitor (APRM) Scram and Rod i

Block and Rod Block Monitor Trip Setpoints and Allowable Values to those applicable to single loop recirculation loop operation per Specifications 2.2.1, 3.2.2, and 3.3.6.

2.

When operating within the surveillance region specified in Figure 3.4.1.1-1:

a)

With core flow less than 39% of rated core flow, initiate action within 15 minutes to either:

1)

Leave the surveillance region within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , or 2)

Increase core flow to greater than or equal to 39% of rated flow within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

b)

With the APRM and LPRM# neutron flux noise level greater than three (3) times their established baseline noise levels:

1)

Initiate corrective action within 15 minutes to re-store the noise levels to within the required limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , otherwise 2)

Leave the surveillance region specified in Fig-ure 3.4.1.1-1 within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

l

See Special Test Exception 3.10.4. -

[

etector levels A and C of one LPRM string per core octant plus detector levels D

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

l LA SALLE - UNIT 1 3/4 4-1 Amendment No. 40

a.

REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION ACTION:

(Continued) 3.

The provisions of Specification 3.0.4 are not applicable.

4.

Otherwise, be ina at least H0T SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b.

With no reactor coolant system recirculation loops in operation, immediately initiate measures to place the unit in at least HOT SHUT-00WN within the next 6 h6urs.

SURVEILLANCE REQUIREMENTS 4.4.1.1 Each reactor coolant system recirculation loop flow control valve shall be demonstrated OPERABLE at least once per 18 months by:

a.

Verifying that the c.ontrol valve fails "as is" on loss of hydraulic pressure at the hydraulic power units, and b.

Verifying that the average rate of control valve movement is:

1.

Less than or equal to 11% of stroke per second opening, and 2.

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

4.4.1.2 With one reactor coolant system recirculation loop not in operation:

a.

Establish baseline APRM and LPRM# neutron flux noise level values within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months upon entering the surveillance region of Fig-ure 3.4.1.1-1 provided that baseline values have not been established since last refueling.

i b.

When operating in the surveillance region of Figure 3.4.1.1-1, verify that the APRM and LPRM neutron flux noise levels are less than or equal to three (3) times the baseline values:

1.

At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , and 2.

Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after completion of a THERMAL POWER increase of at least 5% of RATED THERMAL POWER, initiating the surveillance within 15 minutes of completion of the increase.

l c.

When operating in the surveillance region of Figure 3.4.1.1-1, verify that core flow is greater than or equal to 39% of rated core flow at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

etector levels A and C of one LPRM string per core octart plus detector levels l

0 l

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

t LA SALLE - UNIT 1 3/4 4-la Amendment No. 40 l

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3/4.4 REACTOR COOLANT SYSTEM BASES 4

3/4.4.1 RECIRCULATION SYSTEM Operation with one reactor core coolant recirculation loop inoperable has been evaluated and been found to be acceptable provided the unit is operated in accordance with the single recirculation loop operation Technical Specifi-cations herein.

An inoperable jet pump is not, in itself, a sufficient reason to declare a recirculation loop inoperable, but it does present a hazard in case of a design-basis-accident by increasing the blowdown area and reducing 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 scheduled for significant degradation.

Recirculation loop flow mismatch limits are in compliance with the ECCS LOCA analysis design criterion.

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

Where the recircu-lation loop flow mismatch limits can not be maintained during the recirculation loop operation, continued operation is permitted in the single recirculation loop operation made.

j In order to prevent undue stress on the vessel nozzles and bottom head j

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.

Since the coolant in the bottom of the vessel is at a lower temperature than the water in the upper regions of the core, undue stress on the vessel would result if the temperature difference was greater than 145*F.

i The possibility of thermal hydraulic instability in a BWR has been inves-i tigated since the startup of early BWRs.

Based on tests and analytical models, it has been identified that the high power-low flow corner of the power-to-flow map is the region of least stability margin.

This region maybe encountered during startups, shutdowns, sequence exchanges, and as a result of a recircula-tion pump (s) trip event.

To ensure stability, single loop operation is limited in a designated restricted region (Figure 3.4.1.1-1) of the power-to-flow map.

Single loop operation with a designated surve.illance region (Figure 3.4.1.1-1) of the power-to-flow map requires monitoring of APRM and LPRM noise levels.

3/4.4.2 SAFETY / RELIEF VALVES The safety valve function of the safety-relief valves-operate to prevent the reactor coolant system from being pressurized above the. Safety Limit of 1325 psig in accordance with the ASME Code.

A total of 18 OPERABLE safety /

relief valves is required to limit reactor pressure to within ASME III allowable values for the worst case upset transient.

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.

LA SALLE-UNIT 1 B 3/4 4-1 Amendment No. 40

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ML20197D832

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