ML20151X471
| ML20151X471 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 04/26/1988 |
| From: | COMMONWEALTH EDISON CO. |
| To: | |
| Shared Package | |
| ML20151X467 | List: |
| References | |
| NUDOCS 8805040161 | |
| Download: ML20151X471 (25) | |
Text
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7
.. o INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE
^
3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1
- RECIRCULATION SYSTEM .
Recire';1ation loops.......................................... 3/4 4-1 Jet Pumps.................................................... 3/4 4-2 '
Recirculation Loop Flow.... ................................ 3/4 4-3
- a ===. L 314. ,
m pp ZMW bd',".T.'.9 . .s+a k id. . . . . . . y. ., .. .. ....... .. .. .. .. . . . . . . . q. .
3/4 4.3 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems.................................... 3/4 4-6 Operational Leakage..........................................
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 Dome..........................................., 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 4-22 3/4.4.9 RESIOUAL 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-SHUTDOWN................................................ 3/4 5-6 3/4.5.3 SUPPRESSION CHAPEER......................................... 3/4 5-5 LA SALLE - UNIT 1 VI Amencment No. .
8805040161 DR 880426 ADOCK 0500o373 DCD
INDEX LIST OF FIGURES FIGURE PAGE ;
3.1.5-1 SODIUM PENTABORATE SOLUTION TEMPERATURE / 4 CONCENTRATION REQUIREMENTS ............................. 3/4 1-21 l 3.1.5-2 SODIUM PENTABORATE (Na2B 100 :s 10 H2O)
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
~^
. 2 - r NW .............................. 3/4 2-6 .
3.4.1. 1 C RE THERMAL POWER (% OF RATED) VERSUS TOTAL YC
( 3.4.6.1-1 CORE FLOW (% OF RATED) .................................
MINIMUMREACTORVESSELMETALTESPERATURE
~
3/44f VS. REACTOR VESSEL PRESSURE ............................ 3/4 4-1B 4.7-1 SAMPLE PLAN 2) FOR SNUBBER FUNCTIONAL TEST ............. 3/4 7-32 B 3/4 3-1 REACTOR VESSEL WATER LEVEL ............................. B 3/4 3-7 B 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 1
i l LA SALLE - UNIT 1 XIX Amendment No. '
0 f
. .o i
3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM RECIRCULATION LOOPS LIMITING CONDITION FOR OPERATION i l
3.4.1.1 Two reactor coolant system recirculation loops shall be . operation, j APPLICABILITY: OPERATIONAL CONDITIONS 1* and 2*.
ACTION: j
- a. With one reactor coolant system recirculation loo at in operation:
- 1. Within 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />s:
a) Place the recirculation flow contro stem in the Master Manual mode, and ,
b) Increase the MINIMUM CRITICAL P0 RATIO (MCPR) Safety Limit by 0.01 to 1.08 per Speci ation 2.1.2, and, c) Increase the MINIMUM CRITICA ER RATIO (MCPR) Limiting I Condition for Opergition by per Specification 3.2.3, and, {
' I d) Reduce the MAX ERA LANAR LINEAR HEAT GENERATION RATE (MAPLHGR) mit t value of 0.85 times the two recirculati o op ion limit per Specification 3.2.1, and, e) Reduce t rage er Range Monitor (APRM) Scram and Rod I Block and od B1 Monitor Trip Setpoints and Allowable Valuer those licable to single loop recirculation loop at per cifications 2.2.1, 3.2.2, and 3.3.6.
- 2. When o er ting wi n the surveillance region specified in Figure .1.1-a) With cor ow less than 39% of rated core flow, initiate action hin 15 minutes to either:
- 1) ave the surveillance region within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, or
- 2) Increase core flow to greater than or equal to 39% of rated flow within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, b) th 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 <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, otherwise
- 2) Leave the surveillance region specified in Fig-ure 3.4.1.1-1 within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
/
- See/ ecial Test Exception 3.10.4.
0 ctor levels A and C of one LPRM string per core octant plus detector levels nd C of one LPRM string in the center region of the core should be monitored.
.A SALLE - UNIT 1 3/4 4-1 Amendment No, 40
- e .,
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 only one (1) reactor coolant system recirculation loop in operation , comply with Specification 3.4.1.5 and:
- 1. Within four (4) hours a) Place the recirculation flow control system in the Master Manual mode or lover, 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 Condition for Operation by 0.01 per Specification 3.2.3, and, d) Reduce the Average Power Range Monitor (APRM)
Scram and Rod Block and Rod Block Monitor Trip Setpoints and Allowable Values to those applicable to single recirculation loop operation per Specificationn 2.2.1, 3. 2. 2, and 3.3.6.
- 2. The provisions of Specification 3.0.4 are not applicable.
- 3. Otherwise, be in at least HOT SHUTDOWN within the next twelve (12) hours,
- b. With no reactor coolant recirculation loops in operation:
- 1. Take the ACTION required by Specification 3.4.1.5, and
- 2. Be in at least HOT SHUTDOWN vithin the next six (6) hours.
LASALLE - UNIT 1 3/4 4-1 Amendment
LIMITING CONDITION FOR OPERATION ,
1:
ACTION: (Continued) l
- 3. The provisions of Specification 3.0.4 are not appl able, j
- 4. Otherwise, be ina at least HOT SHUTDOWN within e next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, I
- b. With no reactor coolant system recirculation loop ln operation, immediately initiate measures to place the unit at least HOT SHUT- -
DOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.4.1.1 Each reactor coolant system irculatio- loop flow control valve shall be demonstrated C?ERABL t le on pe 8 months by:
- a. Verifying that the ce f s "as is" on loss of hydraulic pressure at the hydra u .s. and
- b. Verifying t t the rag at f control valve movement is:
- 1. Less th r to, of stroke per second opening, and -
- 2. Less t 1 1% of stroke per second closing.
4.4.1.2 With re t a ystem recirculation loop not in operation:
- a. Establ e M and LPRM# neutron flux noise level values withi up. ntering the surveillance region of Fig-ure 3 pr ' ded that baseline values have not been established since fu ng.
- b. When operati in the surveillance region of Figure 3.4.1.1-1, verify that the and LPRM# neutron flux noise levels are less than or equal to ee (3) times the baseline values:
l 1. A east once per 12 5ours, and
- 2. thin 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 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.
I c, ..er. operating in the surveillance region of Figure 3.4.1.1-1, verify hat core flow is greater than or equal to 39% of rated core flow at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, t
0 tor levels A and C of one LPRM string per core octant plus detector levels d C of one LPRM string in the center region of the core should be monitored.
SALLE - UNIT 1 3/4 4-la Amendment No. 40
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 control 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
- 12. Less than or equal to 11% of stroke per second closing.
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,. 3/4.4 REACTOR COOLANT SYSTEM
\
3/4.4.1 RECIRCULATION SYSTEM J THERMAL HYDRAULIC STABILITY _ 1 l
LIMITING CONDITION FOR OPERATION 3.4.1.5 Forced core circulation shall be maintained with:
- a. Total core flow greater than or equal to 45% of rated core flow, or
- b. THERMAL POWER within Region III of Figure 3.4.1.5-1, or
- c. THERMAL POWER within Region II of Figure 3.4.1.5-1 AND APRM and noise levels not exceeding the larger of: 1) Three (3)
LPRM times the established baseline noise levels or, 11) 10% peak-to-peak indicated noise level.
APPLICABILITY : OPERATIONAL CONDITION 1 l
i ACTION I
! a. In Region I of Figure 3.4.1.5-1:
l 1. With at least i reactor coolant recirculation loop
) in operation immediately initiato action tot a) Decrease THERMAL POWER by control rod insertion, completing the power decrease within two (2) hours l
' to exit Region I or, 1
b) Increase core flow with the operating Recirculation j
I Loop (s), to exit Region I within two (2) hours.
2.
With no reactor coolant recirculation loops in operrtions a) Immediately reduce CORE THERMAL POWER by inserting control rods, observing the indicated APRM and LPRM noise levels, and complete power reduction to below 36% of RATED CORE THERMAL POWER within two (2) hours, and b) If indicated LPRM or APRM noise levels exceed 10%
peak-to-peak, immediately place the reactor mode switch in the SHUTDOWN position, c) Comply with Specification 3.4.1.1 ACTION b.2 I. A s A U . E u m r t 3 /4 il- % Amend men + No.
l l
1 l neutron In Region II of Figure 3.4.1.5-1, with APRM or1)LPRM b.
flux noise levels exceeding the larger of: orThree (3) li) 10 %
l times the established baseline noise levels, peak-to-peak noise indication.
- 1. Immediately initiate corrective action by inserting control rods or increasing core flow to restore the noise levels to within the required limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, otherwise
- 2. Insert control rods to reduce THERMAL POWER and/or increase core flow to enter Region III of Figure 3.4.1.5-1 j
' within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
SURVEILLANCE REQUIREMENTS 1
l l verify 4.4.1.5 When operating within Region II of Figure 3.4.1.5-1,
- 1. That the APRM and LPRM neutron flux times noise levels do not the established exceed the larger of: 1) Three (3) t baseline levels or, 11) 10% peak-to-peak indicated noise
/ level l
At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and a.
- b. Initiate the survie11ance within 15 minutes after enteting the region or completing an increase of at least 5% of RATED THERMAL POWER, completing the surveillance within the next 30 minutes.
- 2. That core flow is greater than or equal to 39% of rated core flow at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
- Detector levels A and C of one LPRM string per core octant plusin the center reg detector levels A and C of one LPRM string the core should be monitored.
b LA SALL6 UMT 1 3/4 4-4b A mera med vo.
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3 I. Restricted Region p 60
!(Restricted Region III II. Stability Monitoring
- 5 (Allowable) Allowable Region 3 3 50 @
III Attowable Region i V i i 5 30 ;
.c III t- ! !
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k 35 4S i
b 20 30 40 50 60 70 80 90 100 105 k Total Core Flow (% of Rated) k y Figure 3.4.1.5-1
i ll 3/4.4- REACTOR COOLANT SYSTEM ,
BASES q
3/4.4.1 RECIRCULATION SYSTEM i
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 mode.
In order to prevent undue stress on the vessel nozzles and b'ottom 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. 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.
The possibility of thermal hydraulic instability in a BWR has been inves-Based on tests and analytical models, tigated since the startup of early BWRs.it has been identified that the high po This region maybe encountered map is the region of least stability margin.
during startups, shutdowns, sequence exchanges, and as a result of a recircula-tion pump (s) trip even,t- <
To ensure stability, single loop operation is limited in a Single designatedloop
)I restricted region (Figure 3.4.1.1-1) of the power-to-flow map.
operation with a designated surveillance region (Figure 3.4.1.1-1) of the ower-to-flow map requires monitoring of APRM and LPRM noise levels.
~Og./
3/4.4.2 SAFETY / RELIEF VALVES The safety valve function.of the safety-relief valves operate to prevent g gg g the reactor coolant system from being pressurized A totalabove the Safety Limit of 18 OPERABLE safetyof/
1325 psig in accordance with the ASME Code.
A 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.
B 3/4 4-1 Amendment No. 40 LA SALLE-UNIT 1
INSERT A To~LaSalle Unit 1 BASES Page B 3/4 4-1 BASES Sect. 3/4.421 Region I of Figure 3.4.1.5-1 represents a region of the power / flow map where instability in neutron flux have been observed. Operation in this region is prohibited to ensure that stable reactor conditions are maintained. Actions to immediately exit Region I are intended to prevent lower priority (i.e., non-emergency) concerns from delaying exit from the region. Observation of neutron flux indications, while not requiring formal surveillance, is needed to avoid reliance on automatic protective systems. A manual reactor scram is required if instabilities are evidenced in Region I with no recirculation pumps operating.
Operation within a designated surveillance region (Region II of Figure 3.4.1.5-1) requires monitoring of APRM and LPRM noise levels. Observed instabilities require immediate corrective action due to the potential for increasing oscillations.
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INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.4 REACTOR COOLANT SYSTEM 1/4.4.1 RECIRCULATION SYSTEM Recirculation loops.......................................... 3/4 4-1 Jet Pumps.................................................... 3/4 4-3 Recirculation Loop F1ow...................................... 3/4 4-4
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3/4 4.3 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems.................................... 3/4 4-7 Operational Leakage.......................................... 3/4 4-8
(
- 3/4.4.4 CHEMISTRY.................................................... 3/4 4-11 3/4.4.5 SPECIFIC ACTIVITY............................................ 3/4 4-14 3/4.4.6 PRESSURE / TEMPERATURE LIMITS R e a cto r C oo l ant 5y s tem. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-17 Reactor Steam Dome........................................... 3/4 4-21 3/4.4.7 MAIN STE AM LINE I SO LATION VA LVES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-22 3/4.4.8 STRUCTURAL INTEGRITY......................................... 3/4 4-23 3/4.4.9 RESIDUAL HEAT REMOVAL Hot Shutdown................................................. 3/4 4-24 Cold Shutdown................................................ 3/4/4-25 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ECCS-0PERATING............................................... 3/4 5-1 3/4.5.2 ECCS-SHUTDOWN................................................ 3/4 5-6 3/4.5.3 S U P P R E S S IO N CHAMB E R . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 5- 8 LA SALLE - UNIT 2 VI
. LIST OF FIGURES-FIGURE PAGE 3.1.5-1 S0DIUM PENTABORATE SOLUTION TEMPERATURE /
CONCENTRATION REQUIREMENTS ........................ 3/4 1-21 3.1.5-2 S0DIUM PENTABORATE (Na 2 10 B 016 " 10 H20)
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 3.2.3-1 MINIMUM CRITICAL POWER RATIO (MCPR) VERSUS I AT RATED FLOW .................................. 3/4 2-5 3.2.1-2 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS AVERAGE PLANAR EXPOSURE, FUEL TYPE BP8CRB299L. ....................................... 3/4 2-2(a)
, N TOR ........................ ........ 3/4 2-6 3.4.1. -1 CORE THERMAL POWER (% OF RATED) VERSUS TOTAL CORE SC FLOW (% OF RATED) .................................. 3/44-h 3.4.6.1-1 MINIMUM REACTOR VESSEL METAL TEMPERATURE VS. REACTOR VESSEL PRESSURE ....................... 3/4 4-19 4.7-1 SAMPLE PLAN 2) FOR SNUBBER FUNCTIONAL TEST ........ 3/4 7-33 8 3/4 3 1 REACTOR VESSEL WATER LEVEL ........................ B 3/4 3-7 0 3/4.4.6-1 CALCULATED FAST NEUTRON FLUENCE (E>1MeV) at 1/4 T AS A FUNCTION OF SERVICE LIFE ..................... B 3/4 4-7 5.1.1-1 EXCLUSION AREA AND SITE B0UNDARY FOR GASE0US AND LIQUID EFFLUENTS .............................. 5-2 5.1.2-1 LOW PO P U LATION ZO N E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 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 2 XIX Amendment No. 32
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 sh e in operation.
APPLICABILITY: OPERATIONAL CONDITIONS 18 and 2*. ,
ACTION: /
l a. With one reactor coolant system recircula loop not in operation:
- 1. Within 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />s:
a) Place tne recir o f l 'i ontrol system in the Master Manual mode, an b) Increa AL POWER RATIO (MCPR) Safety
- l Limit by Specification 2.1.2,.and, c) Increase INIMU ITICAL POWER RATIO (MCPR) Limiting i Conditio r n by 0.01 per Specification 3.2.3, and, d) Red e M XI ' VERAGE PLANAR LINEAR MEAT GENERATION RATE I l (M )I i t' a value of 0.85 times the two recirculation lo r o mit per Specification 3.2.1, and, e) *e age Power Range Monitor (APRM) Scram and i d k. Rod Block Monitor Trip Setpoints and AllowsMe
\ ues t ose applicable for single loop recirculation lo op ion per Specifications 2.2.1, 3.2.2, and 3.3.6.
- 2. cera within the surveillance region specified in
- 3. .1-1:
Wi . ore flow less than 39% of rated core flew, l 1 iate action within 15 minutes to either:
Leave the surveillance region within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, or
) Increase core flow to greater than or equal to 39% of rated flow within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
With the APRM and LPRM neutron flux noise level greater than three (3) times their established baseline noise levels:
^$ee $p ial Test Exception 3.10.4.
- 0et .or levels A anc C of one LPRM string per core octant plus detector leveis A d C of one LPRM string in the center region of the core should be conitored.
LA SALLE - UNIT 2 3/4 4-1 Amendment No. 32
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.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 only one (1) reactor coolant system recirculation loop in operation , comply with Specification 3.4.1.5 and:
- 1. Within four (4) hours a) Place ~the recirculation flow control system in the Master Manual mode or lower, 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 Condition for Operation by 0.01 per
[
Specification 3.2.3, and, d) Reduce the Average Power Range Monitor (APRM)
Scram and Rod Block and Rod Block Monitor Trip Setpoints and Allowable Values to those applicable to single recirculation loop operation per Specifications 2.2.1, 3.2.2, and 3.3.6, and e) Reduce the MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) limit to a value of 0.85 times the two recirculation loop operation limit per Specification 3.2.1.
- 2. The provisions of Specification 3.0.4 are not applicable.
- 3. Otherwise, be in at least HOT SHUTDOWN within the next twelve (12) hours.
b.. With no reactor coolant recirculation loops in operation
- 1. Take the ACTION required by Specification 3.4.1.5, and
- 2. Be in at least HOT SHUTDOWN within the next six (6) hours.
LASALLE - UNIT 2 3/4 4-1 Amendment
LIMITING CONDITION FOR OPERATION (Continued) 4 ACTION: (Continued)
- 1) Initiate corrective action within 15 es to restcre the noise levels to within the requ imit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, otherwise
- 2) Leave the surveillance region s fied in Figure 3.4.1.1-1 within the n hours.
- 3. The provisions of Specification 3.0.4 not applicable.
- 4. Otherwise, be in at st HOT SHUTC within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
- b. With no reactor coolan s tem recir tion loops in operation, immediately in iate as es t p the unit in at least HOT SHUTDOWN withi e t ho 2.
SURVEILLANCE REOUIREMENTS 4'.4.1.1 Each reactor nt stem ' rculation loop flow control valve shall be demonstrated ABLE 1 ence per 18 months by:
- a. Ver ving th cent -valve fails "as is" on loss of hydraulic pres. ea e dra power unit, and i b. Verify he age rate of control valve movement.is:
- 1. $ than or al to 11% of stroke per second opening, and
- 2. ss's,an qual to 11% of stroke per second closing.
l 4,4.1. Wi ne ea coolant system recirculation loop not in operation:
- a. f5 b iine APRM and LPRMW neutron flux noise level values t 4" s upon entering the surveillance region of Figure 3.4.1.1-1 p
e t the baseline values have not been established since st r i r,g .
. hen rating in the surveillance region of Figure 3.4.1.1-1, verify tha ,e APRM and LPRM# neutron flux noise levels are less than or e to three (3) times the baseline values:
V At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 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.
When operating in the surveillance region of Figure 3.4.1.1-1, verify that core flow is greater than or equal to 3% of rated core flow at ,
least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
l l
[DetectorlevelsAandCofoneLPRMstringpercoreoctantplusdetector levels A and C of one LPRM string in the center region of the core should be monitored.
i LA SALLE - UNIT 2 3/4 4-2 Amendment No.32 l
a SURVEILLANCC REQUIREMENTS 4.4.1.1 Each reactor coolant system recirculation loop flow control val *re shall be demonstrated OPERABLE at least once per 18 months by:
- a. Verifying that the control 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 etroke per second opening, and
- 2. Less than or equal to 11% of stroke per second closing.
I l
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. 3/4.4 REACTOR' COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM THERMAL HYDRAULIC STABILITY LIMITING CONDITION FOR OPERATION 3.4.1.5 Forced core circulation shall be maintained withs
- s. Total core flow greater than or equal to 45% of rated core flow, or THERMAL POWER within Region III of Figure 3.4.1.5-1, or b.
- c. THERMAL POWER within Region II of Figure 3.4.1.5-1 AND APRM and noise levels not exceeding the larger of: 1) Three (3)
LPRM times the established baseline noise icvels or, 11) 10% peak-to-peak indicated noise level. APPLICABILITY : OPERATIONAL CONDITION 1 ACTION
- a. In Region I of Figure 3.4.1.5-1:
- 1. With at least i reactor coolant recirculation loop in operation immediately initiate action to l n) Decrease THERMAL POWER by control rod insertion, completing the power decrease within two (2) hours to exit Region I or, l
b) Increase core flow with the operating Recirculation Loop (s), to exit Region I within two (2) hours. 2. With no reactor coolant recirculation loops in operation l a) Immediately reduce CORE THERMAL POWER by inserting control rods, observing the indicated l l APRM and LPRM noise levels, and complete power reduction to below 36% of RATED CORE THERMAL POWER within two (2) hours, and b) If indicated LPRM or APRM noise levels exceed 10% peak-to-peak, immediately place the reactor mode switch in the SHUTDOWN position. c) Comply with Specification 3.4.1.1 ACTION b.2 LA S A LLE u ^>lT '2 3/4459 MW w,
,"e neutron In Region II of Figure 3.4.1.5-1, with APRM or1)LPRM b.
flux noise levels exceeding the larger of: orThree (3)
- 11) 10 %
times the established baseline noise levels, peak-to-peak noise indication.
- 1. Immediately initiate corrective action by inserting control rods or increasing core flow to restore the noise levels to within the required limit within 2 hours, otherwise
- 2. Insert control rods to reduce THERMAL POWER and/or increase core flow to enter Region III of Figure 3.4.1.5-1 within the next 2 hours.
SURVEILLANCE REQUIREMENTS verify: 4.4.1.5 When operating within Region II of Figure 3.4.1.5-1, That the APRM and LPRM 1) neutron flux noise levels do not 1. exceed the larger of: Three (3) times the established baseline levels or, 11) 10% peak-to-peak indicated noise levels At least once per 12 hours, and a.
- b. Initiate the survie11ance within 15 minutes after entering the region or completing an increase of at least 5% of RATED THERMAL POWER, completing the surveillance within the next 30 mir.utes.
That core fiev is greater than or equal to 39% of rated core flow at least once per 12 hours, 2. l t Detector levels A and C of one LPRM string per core the core should be monitored. 3M Ll-fb QcajA NO, LA sal LG MMT 2 1 l
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4 3/4.4 REACTOR COOLANT SYSTEM BASES . 3/4.4.1 RECIRCULATION SYSTEM Operation with one reactor recirculation loop inoperable has been evaluated and been found to be acceptable provided the unit is operated in accordance with i the single recirculation loop operation Technical Specifications 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 retir-culation loop flow mismatch limits can not be maintained during the recir-culation loop operation, continued operation is permitted in the 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. 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. The possibility of thermal hydraulic instability in a BWR has been investi-gated 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 may be encountered l during startups, shutdowns, sequence exchanges, and as a result of a retircula-tion pump (s) trip event. _ l [ To ensure stability, single loop operation is limited in a designated f restricted region (Figure 3.4.1.1-1) of the power-to-flow map. Single loop I operat b wer ionto-flow with amap designated surveillance requires monitoring region of APRM and(Figure LPRM noise3.4.1.1-1) levels, of the m 3/4.4.2 SA[ETY/RELIEFVALVES dG.k g7 The safety valve function of the safety / relief valves operate to prevent the reactor coolant system from being pressurized above the Safety Limit of A total of 18 OPERABLE safety / g l ! 1325 psig in accordance with the ASME Code. l relief valves is required to limit reactor pressure to within ASME III l 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. l LA SALLE - UNIT 2 B 3/4 4-1 Amend ent No. 32 l k
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. To LaSalle Unit 2-BASES Page B-3/4 4-1 4
6 4 BASES Sect. 3/4.4.1 Region I of Figure 3.4.1.5-1 represents a region of the power / flow map where instability in neutron flux have been observed. Operation in this region is prohibited to ensure that stable reactor conditions are maintained. Actions,to.immediately exit Region I are intended to Prevent lower priority (i.e., non-emergency) concerns from delaying exit from the region. Observation of neutron flux indications, while not requiring formal surveillance, is needed to avoid reliance on automatic protective systems. A manual reactor scram is required if instabilities are evidenced in Region I with no recirculation pumps
, operating.
operation within a designated surveillance region-(Region II of Figure 1 3.4.1.5-1) requires monitoring of APRM and LPRM noise levels. Observed instabilities require immediate corrective action due to the potential for increasing oscillations. 1 4 4 !i. J 4511K d '
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c v .o . ATTACHMENT C TECHNICAL SPECIPICATION CHANGE REOUBST LASALLS COUNTY STATION UNITS 1 AND 2 SIGNIFICANT HAZARDS CONSIDERATION Commonwealth Edison has evaluated the proposed Technical Specification Amendment and determined that it does not represent a significant hazards consideration. Based on the critieria for defining a significant hazards consideration established in 10 CFR 50.92, operation of LaSalle County Station Units l'and 2 in accordance with the proposed amendment will not:
- 1. Involve a significant increase in the probability or consequences of an accident previously evaluated because:
The stability monitoring provisions contained herein are more restrictive (conservative) than the presently approved specifications, and as such, increase the margin of safety during operation of the plant. The proposed revisions assure increased operator awareness of the core, neutron flux and thermal hydraulic status. Significantly more conservative actions are dictated than previous spectheations, including a reactor scram under certain specified conditions. These actions are evaluated to bound all existing safety requirements and therefore will not increase the probability or consequence of an accident previously evaluated.
- 2. Create the possibility of a new or different kind of accident from any accident previously evaluated because:
The proposed revisions do not authorize operation in any new regions or ' configurations, but only provide for increased stability monitoring and reduced time in configurations of possibile instability. No changes to the operational modes of platit systems are involved in these Technical Specification changes.
- 3. Involve a significant reduction in the margin of safety because:
The specification revisions increase the margin of safety by reducing the allowable time inside regions of possible reactor instabilities. No significant changes to acceptance criteria for operation are involved. The provisions of these changes are consistent or more conservative than previously approved Technical Specifications for thermal hydraulic stability monitoring. Based on the preceding discussion, it is concluded that the proposed system change clearly falls within all acceptable criteria with respect to the system or component, the consequences of previously evaluated accidents will not be increased and the margin of safety will not be decreased. Therefore, based on the guidance provided in the Federal Register and the criteria established in 10 CFR 50.92, the proposed change does not constitute a significant hazards consideration. I I 4511K i I}}