Text

Proposed Tech Specs Reflecting Changes Incorporated Into Unit 2 Tech Specs Approved on 831216
	ML20079H418
Person / Time
Site:	LaSalle
Issue date:	01/13/1984
From:	; COMMONWEALTH EDISON CO.
To:	;
Shared Package
ML20079H398	List: ML20079H392; ML20079H418;
References
	NUDOCS 8401230433
	Download: ML20079H418 (212)
	v • d • e

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ATTACHMENT A

SUBJECT:

Update of Unit 1 Technical Specifications to the Issued Unit 2 Technical Specifications for LaSalle County Statien BACKGROUND The Operating License (NPF-18) for LaSalle County Station, Unit 2 was issued on December 16, 1983. The Technical Specifications for LaSalle County Unit 2 differed somewhat from that issued and as amended thruugh 4

Amendment 15 for LaSalle County Unit 1. Commonwealth Edison committed to update the Unit 1 Technical Specifications to that issued for Unit 2 where the changes did not involve specific design differences. This submittal includes those differences as identified by Commonwealth Edison.

DISCUSSION .

1 A. The following is a list of the major changes to the Unit 1 Technical Specification. This list is based on the Unit 2 Technical Specification as issued and on the current Unit 1 Specification:

1. Page XIX through XXIII - added list of Tables and Figures.

2. Page 1-9, *** footnote states moved instead of coupled.

3. Single recirculation loop operation is folded into the body of the Tech Specs, pages 2-1, 2-4, B 2-1, B 2-4, 3/4 2-1, 3/4 2-3,

' 3/4 2-4, 3/4 3-53, 3/4 4-1, la, 2, 3, B 3/4 1-2, B 3/4 2-1, B 3/4 2-3, 8 3/4 4-1.

4. Page 3/4 1 revised 4.1.1.c to allow 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months instead of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

5. Page 3/4 1 added word " withdrawn" to action b.l.a)l).

6. Page 3/4 1 added footnote to allow startup to perform test if necessary.

7. Pages 3/4 1-6, 8, 9, 14 - control rods specifications 3.1.3.2.,

3.1.3.4, 3.1.3.5., and 3.1.3.7 have added "3.0.4 not applicable".

8. Pages 3/4 3-4, 5, 41 - deleted startup test setpoint verification footnote.

9. Pages 3/4 3-11, 14 - added footnote (1) to allow bypass of delta T instruments for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

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PDR ADOCK 05000373 P PDR

10. Page 3/4 3 revised reactor water cleanup ambient and differential temperature setpoints,

11. Pages 3/4 3-15, 16, 17 - deleted startup test setpoint verification footnote.

12. Pages 3/4 3-18, 19 - added _ 5 second time delay reference.

13. Pages 3/4 3-39, 3/4 2-4, 5 - added change to' allow operation if EOC-RPT inoperable.

14. Page 3/4 3 revised APRM calibration frequency to semi-annually.

15. Pages 3/4 3-60, 63 - added footnote for common systems.

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16. Pages 3/4 3-72 and 3/4 9 added footnote to require S/N for source range to be 3 cps.

17. Pages 3/4 3-81, 3/4 11-13, 3/4 11 revised radioactive effluent reporting requirements.

18. Page 3/4 3 specified that isolation of the off gas system was required only during channel calibration.

19. Page 3/4 4 there is no requirement for immediate scram upon loss of both recirc pumps.

for SRV's.

20. Page 3/4 4 revised tolerance from +1% to + or - 1%

21. Page 3/4 4 revised tolerance from + or - 10 to + or - 50 psig.

22. Page 3/4 4 revised withdrawal times for reactor vessel material specimens.

23. Pages 3/4 5-3, 4, 5 - water tight doors specifications for ECCS corner rooms have been added.

24. Page 3/4 5 revised HPCS delta P setpoint.

25. Page 3/4 5 deleted footnote for startup test setpoint verification.

26. Pages 3/4 6-2, 3 - revised calculation method of MSIV leakage rate limit.

27. Pages 3/4 6-8, 9 - revised tendon action statements and surveillance and table clarifications.

28. Pages 3/4 6-15 and 3/4 11 added limitation on using standby gas treatment for purging the primary containment.

29. Pages 3/4 3-70, 3/4 6-16, 17, 18 - deleted SRV test footnote.

30. Table 3.6.3-1. (Pages 3/4 6-24, 26, 27, 28, 32, 34)

a. (PCIS valves) added # (3.0.4 not applicable) to various valves,

b. Revised butterfly valve times af ter first refuel outage.

(Also some VQ valve times have changed immediately.)

c. ICM0238 & 1CM024A have been deleted.

d. lE12-F0998 added.

31. Page 3/4 7 deleted footnote allowing crosstie of 250 volt batteries.

32. Page 3/4 7 revised fire pump parameters and fire suppression water system pressure.

33. Page 3/4 7 (Table 3.7.7-1) revised area temperatures.

34. Pages 3/4 7-27 through 45 - revised total snubber spec. - No Table.

35. Page 3/4 7 deleted calibration requirement for Main Turbine Bypass System and deleted valve position requirement. Also changed applicability to l 25% power. Deleted startup test footnote.

36. Diesels:

a. Page 3/4 8 added explanation to 2A inoperable action f.

to prevent excessive 1A testing when system inoperable.

b. Page 3/4 8 deleted old surveillance item 6.

c. Page 3/4 8 deleted starts on stored air surveillance (old item 13).

37. Page 3/4 8 changed 'and/or' to 'or' for 0 and 1A during shutdown.

38. Page 3/4 8 revised equipment neeoed for Unit 2 Division 1 AC.

39. Page 3/4.8 revisedand/or' to 'or' for AC during shutdown.

40. Pages 3/4 8-14, 15, 17, - deleted Unit 2 Division 1 DC sources and deleted ability to crosstie.

41. Page 3/4 8 revised 'and/or' to 'or' for DC sources during shutdown.

42. Page 3/4 8 added drywell holsts and cranes to-drywell circuits to be deenergized and-deleted them from page 3/4 8-24.

43. Page.3/4 8 added "3.0.4 not applicable" to. thermal overload bypass specification.

44. Page 3/4 8 revised requirements to functionally test the RPS EPA's to only during cold shutdowns l 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

45. Page 3/4'11 deleted P-32 from liquid waste sampling Table 4.11.1-1.

46. Page 3/4 11 revised requirement (sampling).

47. Page 3/4 12 revised number of sample locations (Table.

, 3.12.1-1).

48. Page B 3/4 0 added clarification of time use during required shutdowns.

49. Page B 3/4 5 revised HPCS pump flow basis.

50. Page B 3/4 6 clarified tendon surveillance basis.

51. Pages B 3/4 7-3, 4, 5 - revised snubber basis.

52. Pages B 3/4 11-1, 3 - clarification to radioactive effluent basis.

53. -Page 5 corrected drywell free volume.

L .54. Page 6 revised corporate management figure.

55. Pages 6-13, 14 - new shift manning Table for two units.

'56. Pages 6-28, 29 - added footnote for common PCP and ODCM.

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B. The following.is a list of pages on which the~ change was a minor I

administrative change (eg. typo;. adding comma, parenthesis, etc; wording change for--clarification; etc):

.h

1. Pages II, VIII, XV.

12 . Pages 3/4 1-4, 1-11, 1-19.

3.- Pages 3/4 3-1, 3/4 3-58, 3/4 3-82, 3/4 3-83, 3/4 3-84.

4. Pages 3/A 4-13, 3/4 14, 3/4 17, 3/4 23,'3/4 24;

5. Page 3/4 5-8.

6. Pages 3/4 6-5, 3/4 6-11, 3/4 6-19, 3/4 6-20, 3/4 6-21, 3/4 6-33,.

3/46-35,3/46-36,3/46-37, 3/46-38,3/46-40, 3/4 6-41.

7.. Pages 3/4 7-14, 3/4.7-17, 3/4 7-18, 3/4 7-22, 3/4 7-24.

8 .- Pages 3/4 8-1, 3/4 8-5, 3/4 8-7, 3/4 8-9,=3/4.8-16.

9. Pages 3/4.9-16, 3/4 9-17.

10 . - Page 3/4 11-9.

11. Pages 3/4 12-1, 3/4 12-4.

12. Pages B 3/4 1-1, B 3/4 1-5.

13. Page B 3/4 3-6.

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14. Page B 3/4 12-1.

15. Pages 6-3, 6-20.

CONCLUSION:

Commonwealth Edison Company has reviewed the proposed changes and has concluded-that, because the NRC has previously reviewed and approved all of these items in Unit 2 Technical Specifications, there are no unreviewed 2

safety issues.

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ATTACHMENT B SIGNIFICANT HAZARDS CONSIDERAT_IGN

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Commonwealth Edison has evaluated the proposed T9c'anical Specification Amendment and determined that is does not represcrit a significant hazards [

consideration. Based on the criteria for defining a signficant hazards consideration established in 10 CFR 50.92, opeu pon of LaSalle County Station Unit 1 in accordance with the proposed x..nendment will not:

1) InvolveasignficantincreaseintheprobsoilityorconsequencesrN_

an accident previously evaluated because 'the changes involved in jnis request represent previously reviewed and approved issues which sre already present in the Technical Specification issued for LaSalle County Station Unit 2.

2) Create the possibility of a new or dif ferent kind of accident from, any accident previously r valuated because no new changes not' already e

incorporated in the ism;ed Unit 2 Technical Specifications are included in this change.

3) Involve a significant reduction in the margin of safety because all the changes are as previous issued in the Unit 2 Technical Specifications.

Based on the preceding discussion, it is concluded that the proposed change clearly falls within all acceptable criteria with respect to the system or component, the consequences of previously evaluated accidents aill 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(e), the proposed change does not constitute a significant hazards consideration.

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INDEX

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/ ..

DEFINITIONS SECTION

. OEFINITIONS (Continued) .

PAGE 1.25 OPERABLE - OPERABILITY............................................ 1-4 1.26 OPERATIONAL C0NCIT71H - CON 0ITION.................................

1-4 1.27 PHYSICS TESTS..................................................... 1-4 1.28 PRESSURE BOUNDARY LEAKAGE......................................... 1-5 1.29 PRIMARY CONTAINMENT INTEGRITY..................................... 1-5

1. 30 PROCESS CGNTRO L PR0 GRAM. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-5

~

1.31 PURGE - PURGING................................................... 1-5

1. 3 2 RATED THE RMA L P0WER. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-5

1. 33 REACTOR PROTECTION SYSTEM RESPONSE TIME. . . . . . . . . . . . . . . . . . . . . . . . . . . 1-5 C'

V i 1.34 REPORTABLE OCCURRENCE.............................................

1-6 1.35 R00 0ENSITY....................................................... 1-6 1.36 SECONDARY CONTAINMENT INTEGRITY................................... 1-6 4

1.37 SHUTDOWN MARGIN................................................... ,

1-6

1. 38 SO LI D I F I CAT IO N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-6 1.39 SOURCE CHECK...................................................... 1-7 1.40 STAGGERED TEST BASIS.............................................. 1-7

1.41 THERMAL PCWER..................................................... 1-7 1.42 TURBINE BYPASS RESPONSE TIME.......................: .............. 1-7 s .

I 1 1.43 UNIDENTIFIED LEAKAGE.............................................. 1-7

1.44 VENTILATION EXHAUST TREATMENT SYSTEM.............................. 1-7

. 1.45 VENTING........................................................... 1-7

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INDEX

( ._. LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.7 PLANT SYSTEMS .

3/4.7.1 CORE STANDBY COOLING SYSTEM - EQUIPMENT COOLING WATER SYSTEMS Residual Heat Removal Service Water System................... 3/4 7-1 Diesel Gene ator Cooling Water System........................ 3/4 7-2 Ul timate He at S i n k. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 7-3

- 3/4.7.2 CONTROL ROOM AND AUXILIARY ELECTRIC EQUIPMENT ROOM EMERGENCY FILTRATION SYSTEM.................................. 3/4 7-4 3/4.7.3 REACTOR CORE ISOLATION COOLING SYSTEM. . . . . . . . . . . . I.3/4 . . .7-7

, . 3/4.7.4 SEALED SOURCE CONTAMINATION.................................. 3/4 7-9 .

3/4.7.5 -FIRE SUPPRESSION SYSTEMS O~ Fire Suppression Water System................................ 3/4 7-ii Deluge and/or Sprinkl er Systems. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 7-14 CO Systems.................................................. 3/4 7-17 2

~

Fire Hose Stations........................................... 3/4 7-18 3/4.7.6 FIRE RATED ASSEMBLIES........................................ 3/4 7-22

3/4.7.7 AREA TEMPERATURE MONITORING.................................. 3/4 7-24 3/4.7.8 STRUCTURAL INTEGRITY OF CLASS I STRUCTURES................... 3/4 7-26 3/4.7.9 SNUBBERS...................................................... 3/4 7-27 l

. 3/4.7.10 MAIN TURBINE BYPASS SYSTEM.................................... 3/47-g i j . 33 1

t I i O

LA SALLE - UNIT 1 VIII

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INDEX 8ASE3

,4 4

SECTION PAGE

. 3/4.7 PLANT SYSTEMS ,

I 3/4.7.1 CORE STANOBY CGOLING SYSTEM - EQUIPMENT COOLING

.! WATER SYSTEMS......................................... B 3/4 7-1 3/4.7.2 CONTROL ROOM EMERGENCY FILTRATION SYSTEM. . . . . . . . . . . . . . . . B 3/4 7-1 3/4.7.3 REACTOR CORE ISOLATION COOLING SYSTEM................... B 3/4 7-1

! 3/4.7.4 SEALED SOURCE CONTAMINATION............................. B 3/4 7-2 '

3/4.7.5 FIRE SUPPRESSION SYSTEMS................................ B 3/4 7-2

. _.3/4.1.6 FIRE RATED , A,5,$ EMB LI ES. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 7-3 ,

, 3/4.7.7 AREA TEMPERATURE MONITORING............................. B 3/4 7-3 3/4.7.8 STRUCTURAL INTEGRITY OF CLASS I STRUCTURES.............. 8 3/4 7-3

3/4.7.9 SNU88ERS.....'........................................... B3/47-/Sl 1 .

3/4.7.10 MAIN TUR8INE BYPASS SYSTEM....................-.......... B 3/4 7-5 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 and 3/4.8.2 A.C. SOURCES AND ONSITE POWER DISTRIBUTION SYSTEMS.................................... B 3/4 8-1 .

3/4.8.3 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES. . . . . . . . . . . . . . . . . B 3/4 8-2 -

3/4.9 REFUELING CPERATIONS 3/4.9.1 REACTOR MODE SWITCH..................... ............... B 3/4 9-1

[

i 3/4.9.2 INSTRUMENTATION......................................... 8 3/4 9-1

! 3/4.9.3 CONTROL R00 P0SITION.................................... B 3/4 9-1 l 3/4.9.4 DECAY TIME.....................,......................... B 3/4 9-1 1 3/4.9.5 COMMUNICATIONS.......................................... B 3/4 9-1 f 3/4.9.6 CRANE AND H0!ST......................................... B 3/4 9-1 1 3/4.9.7 CRANE TRAVEL............................................ B 3/4 9-2 i 3/4.9.8 and 3/4.9.9 WATER LEVEL - REACTOR VESSEL l and WATER LEVEL - SPENT FUEL STORAGE P00L............... B 3/4 9-2 i

, 3/4.9.10 CO N T RO L R 0 0 R EM0 VA L . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 9-2 i 3/4.9.11 RESIOUAL HEAT RENVAL AND CCOLANT CIRCULATION. . . . . . . . . . 8 3/4 9-2 s

LA SALLE - UNIT 1 XV

, . * - - - - y-- -,-.w _-r,, 4-+---w-----5y- y-+ vw ' - * ~

LIST OF FIGURES hr- FIGURE

[ PAGE-3.1.5-1 SODIUM PENTABORATE SOLUTION TEMPERATURE /

-_ _ CORCENTRATION REQUIREMENTS ........................ 3/4 1-21

, 3.1.5-2 SODIUM P2NTABORATE (Na2 Bio0 1 s

10 H 2O)

VOLUME / CONCENTRATION REQUIREMENTS ................. 3/4 1-22 3.2.1-3 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLMGR) VERSUS AVERAGE PLANAR EXPOSURE, INITIAL CORE FUEL TYPES 8CR183, 8CR233, AND 8CR711 ............................................ 3/4 2-2 3.2.3-1 MINIMUM CRITICAL POWER RATIO (MCPR) VERSUS T AT RATED FLOW ..........J....................... 3/4 2-5 3.2.3-2 K FACTOR .........................................

f 3/4 2-6

, 3.4.6.1-1 MINIMUM REACTOR VESSEL METAL TEMPERATURE l9 VS. REACTOR VESSEL PRESSURE .'...................... 3/4 4-)F 4.7-1 SAMPLE PLAN 2) FOR SNUBBER FUNCTIONAL TEST . . . . . . . . 3/4 7,)( 32.

B 3/4 3-1 PEACTOR VESSEL WATER LEVEL ........................ S 3/4 3-7 f- B 3/4.4.6-1 CALCULATED FAST NEUTRON FLUENCE (E>1MeV) at 1/4 T (s,g )~ AS A FUNCTION OF SERVICE LIFE ..................... B 3/4 4-7 5.1.1-1 EXCLUSION AREA AND SITE BOUNDARY FOR GASECUS .

AND LIQUID EFFLUENTS .............................. 5-2 5.1.2-1 LOW P O P U LATI O N ZON 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 O

LASALLE-UNIT [I XIX OCT 4 ::

~' PR00F & BT# COPY (J ! LIST OF TAELES 8a:-: -- PAGE 1.1 SURVEILLANCE FREQUENCY NOTATION ................... 1-8

1. 2 OPERATIONAL CONDITIONS ............................ 19 2.2.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS ......................................... 2-4 E2.1. 2 - UNCERTAINTIES USED IN INE DETERMINATION OF THE FUEL CLA00ING SAFETY LIMIT .................... B 2-4 E2.1.2-2 NOMINAL VALUES OF PARAMETERS USED IN THE STATISTICAL ANALYSIS OF FUEL CLA00ING INTEGRITY SAFE ~Y LIMIT ...................................... B 2-E E2.1.2-3 RELATIVE SUNDLE POWER DISTRIBUTION USED IN THE GEIAB STATISTICAL ANALYSIS , . . . . . . . . . . . . . . . . . . . . . E 2-6 32.1.2-4 R-FACTOR DISTRIBUT ON USED IN GETAS STATISTICAL ANALYSIS ....................... ...... E 2-7 3.3.1-1 REACTOR PROT'ECT!ON SYSTEM INSTRUMENTATION . ....... 3/4 3-2

i p) 3.3.1-2 REACTOR PROTECTION SYS G RESPONSE TIMES .......... 3/4 3-5 t

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SU.'iVEILLANCE REQUIREugg73 ,,,,, ,,, ,,,,,, ,,,,,,, 3/4 3 7 3.3.2-1 ISOLATION ACTUATION lNSTRUMENTATION .. ........... 3/a 3-11 3.2.2-2 ISCLATION AC7dAT!ON NSTRUMENTATICN SETPOINTS . .. 3/4 3-1E 3.3.2-3 ,ISCLATION SYSTEM INSTRUMENTAT:CN RESPCNSE T ME .... 3/a 3-13

- 4.3.2.1-1 ISCLATION SYSTEM INSTRUMENTATION SURVEILLANCE -

l REQUIREMENTS ...................................... 3/4 3-20 3.3.3-1 ESERGENCY CORE COOLING SYSTEM ACTUATION

! INS RUMENTATION ....... .................. ........ 3/4 3-24 l 3.3.3-2 E"ERGENCY CORE COOL NG SYSTEM ACTUATION INSUUMENTATION SETPOINTS ................... ..... 3/4 3-23 l

3.3.3-3 EMERGENCY CORE COOLING SYSTEM RESPONSE TIMES . . . . . 3/4 3-31 1.3.3.1-1 EMERGENCY CORE C0 CLING SYSTEM AC~UATION l INSTRUMENTATION SURVEILLANCE REQUIREMENTS . .... . 3/4 3-32 t

! 3. 3. 4.1- 1 ATd5 RECIRCULATION PUMP TRIP SYSTEM ~

INSTRUMENTATION ......... ....... ... ...... .. . 3/4 3-36 l

i

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("N _

( LASALLE-UNIT /l XX l

l . . . . _ _ . .. . - ..- -

s LIST OF TA?LES (Continued)

G V TABLE PAGE 3.3.4.1-2 ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION S ETPOINTS . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 3-37 4.3.4.1-1 ATWS RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS ......... 3/4 3-38

, 3.3.4.2-1 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION ................................... 3/4 3-41 3.3.4.2-2 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM SETPOINTS ......................................... 3/4 3-42 3.3.4.2-3 END-OF-CYCLE RECIRCULATION ~ PUMP TRIP SYSTEM RESPONSE TIME ..................................... 3/4 3-43 4.3.4.2.1-1 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM SURVEILLANCE REQUIREMENTS .........................

3/4 3-44 3.3.5-3 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION ................................... 3/4 3-46 3.3.5-2 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOINTS ......................... 3/4 3-48

)* 4.3.5.1-1 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS ......... 3/4 3-49 3.3.5-1 CONTROL RCD WITHDRAWAL BLOCK INSTRUMENTATION ...... 3/4 3-51 3.3.5-2 CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION SETPOINTS ................................... ..... . 3/4 3-53 4.3.5-1 CONTROL RCD WITHDRAWAL BLOCK INSTRUMENTATION SURVEI L LANCE REQUIREMENTS " . . . . . . . . . . . . . . . . . . . . . . . . . 3/43,562f4 3.3.7.1-1 RADIATION MONITORING INSTRUMENTATION .............. 3/4 3-5sdI7 4.3.7.1-1 RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS ......................... 3/4 3-59 3.3.7.2-1 SEISMIC MONITORING INSTRUMENTATION ................ 3/4 3-51

~

4.3.7.2-1 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS . . . . . . . . . . . . . . . . . . . . . . . . 3/4 3-62 3.3.7.3-1 METEOROLOGICAL MONITORING INSTRUMENTATION ......... 3/4 3-64 4.3.7.3-1 METEOROLOGICAL MCNITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS . . . . . . . . . . . . . . . . . .. 3/4 3-65 .

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LA SALLE - UNITj Zl XXI [J7 .I ]?:

m Me etm. *M e a , - - --- - --

LIST CF 7;ELES (Continued)

[]

9 TABLE PAGE 3.3.7.4-1 REMOTE SHUTDOWN MONITORING INSTRUMENTATION ........ 3/4 3-57 4.3.7.4-1 REMOTE SHUTDOWN MONITORING INSTRUME'NTATION SURVEILLANCE REQUIREMENTS..................... .... 3/4 3-68 3.3.7.5-1 ACCIDENT MONITORING INSTRUMENTATION ............... 3/4 3-70 4.3.7.5-1 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS ......................... 3/4 3-71 3.3.7.9-1 FIRE DETECTION INSTRUMENTATION .................... 3/4 3-76 3.3.7.10-1 RADICACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION ................................... 3/4 3-82 4.3.7.10-1 RADICACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE, REQUIREMENTS ......... 3/4 3-84 3.3.7.11-1 RADICACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION ................................... 3/4 3-87 4.3.7.11-1 RADI0 ACTIVE GASEQUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS ......... 3/4 3-89

( )' 3.3.E-1 FEEDWATER/ MAIN TUREINE TRIP SYSTEM ACTUATION INSTRUMENTATION ......................... 3/4 3-93 3.3.8-2 FEEDWATER/ MAIN TURSINE TRIP SYSTEM ACTUATION INSTRUMENTATION SETPCINTS ............... 3/4 3-94 -

4.3.3.1-1 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS ........ 3/4 3-95 3.4.3.2-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES .. 3/4 4,26'l 3.4.4-1 REACTOR COO LANT SYSTEM CHEMISTRY LIMITS . . . . . . . . . . . 3/4 4-Jaff 7-4.4.5-1 PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM ....................... .......... 3/4 4-JAIL 6' 4.4.6.1.3-1 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM--

. WITHORAWAL SCHEDULE ............................... 3/4 4-) (19 4.6.1.5-1 TENCON SURVEILLANCE ............................... 3/4 6-11 4.5.1.5-2 TENDON LIFT-OFF FORCE ................. .... ...... 3/4 6-12 3.6.3-1 PRIMARY CONTAINMENT ISOLATION VALVES . . . . . . . . . . . . . . 3/46,2724

(y v

LA SALLE - UNITj fl XXII OCT 4 1953 e- d

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A*TOMATIC d ISOLATION VALVES ........................ 3/4 S-/4 39

, _e. . .

i 3.7.5.2-1 DELUGE AND SPRINKLER SYSTEMS ................. .... 3/4 7-18

. 3.7.5.4-1 FIRE HOSE STATIONS ................................ 3/4 7-19 3.7.7-1 AREA TEMPERATURE MONITORING ....................... 3/4 7 ,2f%4T 4.S.1.1.2-1 DIESEL GENERATOR TEST SCHEDULE .................... 3/4 8-7 4.5.2.3.2-1 BATTERY SURVEILLANCE REQUIREMENTS ................. 2/4 S-IS 3.8.3.2-1 PRIMARY CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES .................... 3/4 S-24 3.S.3.3-1 MOTOR-OPERATED VALVES THERMAL OVERLOAD PROTECTION ................:....................... 3/4 E-27 a . , _. ...,

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TO UNRESTRICTED AREAS IN LIQUID WASTE ............. 3/4 11-2

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?.11.1-1 RADI0 ACTIVE LIOUID WASTE SAMPLING AND ANALYSIS

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Na FnUJ n.e. ................ ......................... 4/? 3..1 4.11.2-1 RADICACT!VE GASEGUS WASTE SAMPLING, AND n:4n,.v.

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4.12.1-1 MINIMUM VALUES FOR THE LOWER LIMITS OF DETECTION .. 3/4 12-7 E3.2.1-1 SIGNIFICANT INPUT PARAMETERS TO THE

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TABLE 1.2 OPERATIONAL CONDITIONS l'~ MODE SWITCH AVERAGE REACTOR CONDITION POSITION COOLANT TEMPERATURE

, 1. POWER OPERATION Run Any temperature

. . 2. STARTUP Startup/ Hot Standby Any t.asperature Shutdown0#E*** > 200*F I

3. HDT SHUTDOWN 4.. COLD SHUTDOWN Shutdown W W "** 1 200*F {

5. REFUELING

ShutdownorRefuel**N $ 140*F l l ['

l 'O -

l l

The reactor mode switch may be placed in the Run or Startup/Het Standby position to test the switch interlock functions provided that the control rods are verified to remain fully inserted by a second licensed operator or other technically qualified member of the unit technical staff.

NThe reactor mode switch may be placed in the Refuel position while a single control rod drive is being removed from the reactor pressure vessel per ,

j Specification 3.9.10.1.

" Fuel in the reactor vessel with the . vessel head closure bolts less than fully tensioned or with the head removed.

- See Special Test Exception 3.10.3

- - The reactor mode switch reay be placed in the Refuel position while a single control rod is being mer;hd provided that the one-rod-out interlock is j g

OPERABLE. N o Ve cl.

LA SALLE - UNIT 1 1-9

,,ee-, n --- -- - , , , - or ,.. --,-

, 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 done 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.

THERMALPOWER.MichkessureandHiahFlow 2.1.2 The MINIMUM CRIT AL POWER RATIO (MCPR) shall not be less than 1.06 -

with the reactor vessel s eam dome pressure greater than 785 psig and core flow greater than 10% of r ed flow.

> APPLICABILITY: OPERATIONAL NOITIONS 1 and 2. im 4-ACTION:

i

\

With MCPR less than 1.06 and the actor vessel steam dome pressure greater f o' l

than 785 psig and core flow greater han 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 w the requirements 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 i

l EM:

l With the reactor coolant system pressurs, 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 coraply with the requiree:ents of Specification 6.4 O'

LA SALLE - UNIT 1 2-1

jggch h~

Oc (Octch <

0 THERMAL POWER. High Pressure and High Flow

' 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall net'be less than 1.06 with two recirculation loop cperation and shall not be less than 1.07 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.

APPLICAEILITY: OPERATIONAL CONDITIONS 1 and 2.

ACTION:

With MCFR less tnan 1.06 with two recirculation loop operation or less than 1.07 witn single recirculation loop operation and the reactor vessel steam come 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 Specificatio'n 6.4.

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4 2 TABLE 2.2.1-1

{ REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS i

1

g FUNCTIONAL UNIT TRIP SETP0lNT*

ALLOWA8LE

, - VALUES i

~*

. 1. Intermediate Range Monitor, Neutron Flux-Higt. $ 120 divisions of

^

full scale i 122 divisions

2. Average Power Range Monitor: of full scale

a. Neutron Flux-High, Setdown 5 15% of RATED THERMAL POWER $ 20% of RATED 9 THERMAL POWER

b. ri si m o si-;1.iw i;m. i T - opx.i.-

, pri; 4) T1.,- T,i-eW . 51% ;t;,

-1 .0. x.

q" _:

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l Fixed Neutron Flux-High 5118% of RATED THERMAL POWER 5120% of RATED

'? 3. THERMAL POWER 4 Reactor Vessel Steam Dome Pressure - High 1 1043 psig i 1063 psig j 4. Reactor Vessel Water Level - Low, Level 3 i 3 >_ 12.5 inches above instrument > 11.0 inches i

zero" , above instrument zero*

5.

I Main Steam Line Isolation Valve - Closure $ 8% closed i 12% closed

6. Main Steam Line Radiation - High i 5 3.0 x full pot.er background $ 3.6 x full power background

7. Prinary Containment Pressure - High 5 1.69 psig ' S 1.89 psig

8. Scram Discharge Volume Water Level - Higli.

1 a 5 767' % " $ 767' %"

i 9. Turbine Stop Valve - Closure 5 $% closed 5 7% closed -

p/ 10. Turbine Control Valve fast Closure, ,

Trip 011 Pressure - Low

> 500 psig > 414 psig 11.

Reactor Mode Switch Shutdown Position NA O ifanual Scriaa NA k .;12.___

NA NA See 13ases Figure 6 3/4 3-1.

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. i 2.1 SAFETY LIMITS-O ..

(

BASES i

l 2.0 The fuel cladding, reactor pressure vessel and primary system oiping '

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 casage 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 1.06. MCPR greater than 1.06 g :: t: : :::::r 1 eMw 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 -

r the cladding, fission product migration from this source is incrementally cumulative and continuously measurable. Fuel cladding perforations, however, can result from thermal stresses which occur from reactor operation signif-icantly 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 cladaing deterioration. Therefore, the fuel cladding Oq 5 r tv '< it i= deriaed wita e raia ta ta caa45tiaa= aica auid Produc-onset of transition boiling, MCPR of 1.0. These conditions represent a signif-j 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 i 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 core THERMAL l

4 PCWER with the following basis. Since the pressure drop in the bypass region l is essentially all elevation head, the core pressure drop at low pcwer and flows will always be greater than 4.5 psi. Analyses show that with a bundle

flow of 28 x 10 lbs/hr, bundle pressure drop'is nearly independent of bundle l power and has a value of 3.5 psi.3 Thus, the bundle flow with a 4.5 psi driving head will be greater than 23 x 10 lbs/hr. Full scale ATLAS test data taken

at pressures frem 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, j this corresponds to a THERMAL P0hER of more than 50% of RATED THERMAL POWER.

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

I

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O LA SALLE - UNIT 1 8 2-1

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. i Bases Table 82.1.2-1 3 ..

UNCERTAINTIES USED IN THE OETERMINATION l Oi' THE FUEL CLADDING SAFETY LIMIT *

~

4

- Standard

. Deviation Quantity (% of Point)

Feedwater Flow . 1.76 Feedwater Temperature 0.76 1

Reactor Pressure 0.5

- Core Inlet Temperature 0.2 ,

o 2.5 CoreTota}s.I% Fig.,gg.s4,,fo[a<4:u wes 6.0 -

ChannelFiowAreawortugu.a s .g 3.0 Friction Factor Multiplier '10.0 -

Channel Friction Factor TI e s ;,y, au 6.

f #

R FactoE Y " * " *" .

. Critical Power 3.6 _

l, e

i

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

l d

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LA SALLE - UNIT 1 3 2-4

y 3/4.1 REACTIVITY CONTROL SYSTEMS h 3/4.1.1 SHUTDOWN MARGIN LIMITING CONDITION FOR OPERATION , .

" ~ ~ ' ' ' " "

?. L 1 The SHUTDOWN MARGIN snall be equal to or greater than:

a. 9.38% delta k/k with the highest worth rod analytically detennined, or ;

b. 0.28% delta :</k with the highest worth rod Jetermined by test.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, 4 and 5.

ACTION:

With the SHUTDOWN MARGIN less than specified:

. a. In OPERATIONAL CONDITION 1 or 2, reestablish the required SHUTDOWN MARGIN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ,

b. In OPERATIONAL CONDITION 3 or 4, immediately verify all insertable control rods to be inserted and suspend all activities that

, could reduce the SHUTDOWN MARGIN. In OPERATIONAL CONDITION 4,

! establish SECONDARY CONTAINMENT INTEGRITY within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

O c.

In OPERATIONAL CONDITION 5, suspend CORE ALTERATIONS

and other activities that could reduce the SHUTDOWN MARGIN, and insert all insertable control rods within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . Establish SECONDAkY CONTAINMENT INTEGRITY within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

SURVEILLANCE REQUIREMENTS 4.1.1 The SHUTDOWN HARGIN shall be determined to be equal to or greater than specified at any time during the fuel cycle:

a. By measurement, prior to or during the first startup after each refueling.

b. By measurement, within 500 MWD /T prior to the core average exposure at which the predicted SHUT 00WN MARGIN, including uncertainties and calculation biases, is equal to the specified limit.

c. Within h hour $after detection of a withdrawn control rod that is l immovable, as a result of excessive friction or mechanical inter-ference, or is untrippable, except that the above required SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for L the withdrawn worth of the immovable or untrippable control rod.

% cept movement of IRMs, SRMs or special movable detectors.

LA SALLE - UNIT 1 3/4 1-1

REACTIVITY CONTROL SYSTEM

) 3/4.1.3 CONTROL R005'

- CONTROL R00 OPERABILITY .

. I.-

LIMITING CONDITION FOR OPERATION '

?

3.1.3.1 All control rods shall be OPERA 8LE.

APPLICABILITY: OPERATIONAL CONDITIONS T and 2.

l ACTION:

l t a. -With one control rod inoperable due to being immovable, as a result j , of excessive friction or mechanical interference, or known to be

untrippable:

> t 1.

i Withinjas hour:

l a) Verify that the inoperable control rod, if withdrawn, is separated from all cther inoperable control rods by at least two control calls in all directions. _ .

l. b) Disarm the associated directional control valves either: 1

, 1) Electrically, or

2) Hydraulically by closing the drive water and exhaust water isolation valves. .

! , c) Comply with Surveillance Requirement 4.1.1.c.

l Othenvise, be in at least H0r SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Q 2.

3. Restare the inoperable control rod to OPERABLE status within

/

t 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT SHUTOOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . ,

j b. With one or more control rods trippable but inoperible for causes i other than addressed in ACTION a, above:

1. If the inoperable control rod (s) is withdrawn:

l~

j i

2) Immediately verify: ' ggg

1) That the inoperrble withdrawn control _ rod (i) i d separate.d froc ali other inoperable 8i:ontrol rod (s) by l at least two control cells in all directions, and

2) The insertion capabi,11ty of the inoperable withdrawn control rod (s) by inserting the control rod (s) at least j one notch by drive water pressure within the normal I

operating range *.* ,

b) l rod (s) and Otherwise, insert the inoperable disarm the associated directionalwithdrawn contro#either:

control valves l

1) Electrically, or

, 2) Hydraulically by closi:ig the drive water and exhaust j water isolation valves I

f *Sne teoceracte :::: col cod :ay tren ::e witharawn to a position no fur:re-

,D i witnorawn than its position onen founo to oe inoperaola.

[*May be rearmed intermittently, under administrative control, to permit testing y ( associated with restoring the control rod to OPERABLE status.

l LA SALLE - UNIT 1 3/4 1-3 1

t

lm. -

" i~

REACTIVITY CONTROL SYSTEM LIMITING CONDITION FOR OPERATION (Continued) _ _ _ _ __

i ACTION (Continued)

2. If the inoperable control rod (s) is inserted

i l a) Within hour disarm the associated directional control valves either: l

1) Electrically, or

2) liydraulically by closing the drive water and exhaust

, ; water isolation valves.

Otherwise, be in at least HOT SHUT 00WN within the next b) 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

B

, c. With more 'than 8 control rods inoperable, be in at leas't HGT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . .

SURVEILLANCE REQUIREMENTS 4.1.3.1.1 The scram discharge volume drain and vent valves shall be demonstrated OPERABLE by: .

a. At leact once per 31 days verifying each valve to be open*? and l ~

b. At least once per 92 days cycling each valve through at least one l complete cycle of full travel.

l 4.1.3.1.2 When above the low power satpoint of the RWM and RSCS, all withdrawn t control rods not required to have their directional control valves disarmed

electrically or hydraulically shall be demonstratad CPERABLE by moving each ,

control rod at least one notch:

a. At least once per 7 days, and

b. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when ar.y control rod is immovable as a l result of excessive friction or mechanical interference.

4.1.3.1.3 All control rods shall be demonstrated GPERABLE by performance of Surveillance Requirements 4.1.3.2, 4.1.3.4, 4.1.3.5, 4.1.3.5 and 4.1.3.7.

i Mhese valves may be clcsed intermittently fcr testirg under administrative f control.

' bs $iMay be rearmed intermittently, under administrative control, to permit testing l associated with restoring the control rod to OPERABLE status.

LA SALLE - UNIT 1 3/4 1-4 t

y -.c _ - . . . - - _ ,-,--m-_, - ,.

1

r % REACTIVITY CONTROL SYSTEM

~ .

SURVERLANCE REQUIREMENTS (Continued)

~

WEr1W The scram discharga volume shall be determined OPERA 8LE by

denom,trating: ,

a. The scram discharge volume driin and vent valves OPERABLE, when i control rods are scram tested from a normal control rod configura-tion of Ue s than or equal to 50% 200 DENSITY at least once per 18 montheby verifying that the drain and vent valves: {

1. Close within 30 seconds after receipt of a signal for control rods to scram, and

2. Open after the scram signal is reset.

b. Proper float response by performance of a CHANNEL FUNCTIONAL TEST of the scras discharge volume scram and control rod bicek level instrumentation after each scram from a pressurize'd condition.

i .

i

' ,m 1

-l

?

I l

i .

'

The pr6 visions of Specification 4.0.4 are not applicable for entry into OPERATIONAL' CONDITION 2 provided the surveillance is performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after achieving less than or equal to 50% ROD DENSITY.

'~

l g O' LA SALLE - UNIT 1 3/4 1-5 ,

~

l .

i . --

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. ,- - --w -- - - - - - - - -, - - - - ---- -

s

~ ~

REACTIVITY CONTROL SYSTEM .

(

.f' - CONTROL R00 MAXIMUM SCRAM IhSERTION TIMES ,

LIMITINGCONDITIONFOROPERAf!ON 3.1.3.2 The maximum scram insertion time of each control rod from the fully

'; withdrawn position to notch position 05, based on de-energization of the scram pilot valve solenoids as time zero, shall not exceed 7.0 seconds.

l APPLICABILITY

OPERATIONAL CONDITIONS 1 and 2.

ACTION:

G. With the maximum scram insertion time of one or more control rods exceeding l

7.0 seconds

~

f.l. Declare the control rod (s) with the slow insertion time inoperable, I and ,

g.Q. Perform the Surveillance Requireneilts of Specification 4.1.3.2.c at I least once per 60 days when operation is continued with three or -

more control rods with maximum scram insertion times in excess of 7.0 seconds.

Otherwise, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

! Q [, b. %e pAlaus of Spe htou 3. o.4 am m+ a g ;u l ble. , )

SURVEIL l.ANCE REQUIREMENTS 4.1.3.2 The maximum scram insertion time of the control rods shall be

- demonstrated through measurement with reactor ccolant pressure greater than or equal to 950 psig and, during single control rod scram time tests, the control rod drive pumps isolated from the acct.milators:

t a. For all control rods prior to THERMAL POWER exceeding 40% of RATED

} THERMAL POWER following CORE ALTERATIONS

or after a reactor 8 shutdown that is greater than 120 days, t

b. For specifically affected individual control rods following maintenanca on or modification to the control rod or control rod

drive system which could affect the scram insertion time of those

' specific control rods, and

! c. For at least 10". of the control rods, on a rotating basis, at least once per 120 days of operation.

'Exca;; c.ecar. of SRM, IRM or specia' :ovaels detect: s ce norma' ::ntrol rod to'.saent.

o O .

LA SALLE - UNIT 1 3/4 1-6 .

, _ _ _ . - . . _ , . , _ _ . _ _ _ . . . _ _ . , _ . . _ . , . . , , . , , , _ _ . _ . . ...-.e, ,-.--_.--_...,_7- . , - - _ - - ,

j REACTIVITY' CONTROL SYSTEM S .

FOUR CONTROL R00 GROUP SCRAM INSERTION TIMES LIMITING CONDITION FOR OPERATION 3.1.3.4 The average scram insertion time, from the fully withdrawn position,

for the three fastest control rods in each group of four control rods arranged

. in a two-by-two array, based on deenergization of the scram pilot valve solenoids i as time zero, shall not exceed any of the following:

Position Inserted From Average Scram Inser-Fully Withdrawn tion Time (Seconds) 45 . 0.45 39 0.92 -

25 2.05 05 3.70 -

APPLICA0!LITY: OPERATIONAL CONDITIONS 1 and 2.

ACTION:

.m i d. With the average scram insertion times of control rods exceeding the above (

limits:

ptl. Declare the control rods with the slower than average scram insertion (

timss inoperable until an analysis is performed to determine that required scram reactivity remains for the slow four control rod group, and b.Q. Perform the Surveillance Requirements of Specification 4.1.3.2.c at

, l l

~

least .once per 60 days when operation is continued with an average

- scram insertion time (s) in excess of the average scram insertion time limit.

Othenvise, be in at least HOT SHUT 00WN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

' b. Th pmas;ous of ' Spe: fica 4w 3.o.4 a re. wt n pglicate. .

}

l '

SURVEILLoNCE REOUIREMENTS l

l li 4.1.3.4 All control rods shall be demonstrated OPERABLE by scram time testing j from the fully withdrawn position as required by Surveillance Raquirement 4.1.3.2.

l t

l O' ; LA SALLE - UNIT 1 3/4 1-8 l

l

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i REACTIVITY t0NTROL SYSTEM 0S CONTROL R00 SCRAM ACCUMULATORS LIMITING CONDITION FOR OPERATION

! 3.1.3.5 All control rod scram accumulators shall be OPERA 8LE. ,,

APPLICABILITY: OPERATIONAL'CONCITIONS 1, 2 and 5*.

, ACTION:

"+ a. In OPERATIONAL CONDITION 1 or 2:

,l 1. M$h one control rod scram accumulator inoperable:

a) Within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , either:

1) Restore the inoperable accumulator to OPERABLE s'tatus, or

2) Declare the control rod associated with the inoperable accumulator inoperable. .

b) Othenvise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . -

2. With more than one control rod scram accumulator inoperable, declare the associated contro'l rod inoperable and: ,

a) If the control rod associated with any inoperable scram accumulator is withdrawn, immediately verify that at least I

m. one CR0 pump is operating by insercing at least one with-crawn control rod at least one notch by drive water j>

h pressure within the normal operating range or place the reactor mode switch in the Shutdown position, b) Insert the inoperable control rods and disarm the associated directional control valves either:

1) Electrically, or ,

2) Hydraulically by closing the drive water and exhaust water isolation valves.

Othendsa, be in at least HOT SH'?T00VN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

! b. In OPERATIONAL CCNDITION 5 with:

r i l ; 1. One withdrawn control rod with its associated scram accumlator I

- fnoperable, insert the affected control rod and disarm the j associated directional control valves wthin pe hour, either: l

~

a) Electrically, or .

i b) Hydraulically by closing the drive water and exnaust water l isolation valves.

2. .Wre than one withdrawn control red with the associated scram accumulator inoperable or with no control rod drive pump crerating, immediately plac the reactor :ce rdtc% S the

- 55;td:ar :osition.

! , C. ~Tl e pregh.ous' of SpecsQca tioa L o.' A re wT Applicable.

' "At least tne abrumulator associ&ted with each withcrawn control fod. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

O' -

LA SALLE - UNIT 1 3/4 1-9 I

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

REACTIVITY CONTROL SYSTEM

, CONTROL RCD ORIVE COUPLING

'O LIMITING CONDITION FOR OPERATION 3.1.3.6 All control rods shall be coupled to their drive mechanisms.

I

. APPLICABILITY: OPERATIONAL CONDITIONS 1, 2,and 5*.

ACTION:

a. In OPERATIONAL CONDITION 1 and 2 with one control rod not coupled to its associated drive mechanism:

Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , either: '

~-

1. .

a) If permitted by the RWM and RSCS, insert the control rod drive mechanism to accomplish recoupling and verify recoupling by withdrawing the control tod, and:

1) Observing any indicated response of the nuclear instrumentation, and

2) Demonstrating that the control rod will not go to the overtravel position.

b) If recoupling is not accomplished on the first attempt or, if not permitted by the RWM or RSCS then until permitted by the RWM and RSCS, declare the control rod inoperable and insert the control rod and disarm the associated directional control valves ** either:

1) Electrically, or

2) Hydraulically by closing the drive water and exhaust water isolation valves.

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

b. In OPERATIONAL CONDITION 5* with a withdrawn control rod not coupled to its associated drive cachani ', within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , either:

1. Insert the control. rod to accomplish recoupling and verify recoupling by withdrawing the control rod and demonstrating thist the control rod will not go to the overtravel position, or

2. If recoupling is not accomplished, insert the control rod and disarm the associated directional control valves ** either:

a) Electrics,lly, or b) Hydraulically by closing the drive water and exhaust water isolation valves.

"At least each withdrawn control rod. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

- May be rearmed intermittently, under administrative control, to permit testing

~' associated with restoring the control rod to OPERABLE status.

LA SALLE - UNIT 1 3/4 1-11

g REACTIVITY ' CONTROL 5YSTEM LIMITING CONDITION FOR OPERATION (Continued)

' ACTION (Continued)

'! 2. With one or more control rod " Full-in" and " Full-out" position i indicators inoperable.

{ a) Either: ,

. 1) When THERMAL POWER is within the low power setpoint i, of the RSCS:

. , (a) Within one hour:

(1) Determine the position of the control rod (s) by:

(a) Moving thd control rod, by single notch movement, to a position with an OPERA 8LE l position indicator,

, ; (b) Returning the control rod, by single notch -

movgeant, to its original position, and

', (c) Verifying no control rod drift alars at least per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , or (2) Move the control red to a position with an .

OPERA 8LE position indicator, or (3) Declare the control rod inoperable.

O (b) Verify the position and bypassing of control roos with inoperable " Full-in" and/or " Full-out" position indica-tors by a second licensed operator or other technically qualified member of the unit technical staff.

2) When THERMAL POWER is greater than the low power setpoint of the RSCS, determine the position of the control rod (s)

per ACTION a.2.a) 1)(a)(1), above.

b) Otherwise, be in at least HOT SHUTOOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

l j b. In OPERATIONAL CONOITION 58 with a withdrawn control rod position indicator inoperabl.e, move the control rod to a position with an

, , OPERABLE position indicator or insert the control rod.

I SURVEILLANCE REQUIRENENTS C

Proi.s:ws .oq a.o.4 a w_ mt /2ppih 4k le. , .

I i

4.1.3.7 The control rod position indication system shall be determined OPERABLE by verifying: -

a. At least once per 24 hou's that the position of each control rod is indicated,

b. That the indicated control rod position changes during the movement of the control rod drive when performing Surveillance Requirement 4.1.3.1.2, and

~

I c. That the c:ntr:1 red positfen indicator correspo,ts to the ccatrol

! rce positi n ' .cicatad by tne " Full out" cositice indicator w s, I

f performing Suneiiiance Requirement A. l.3.g. j "At least eacn withdrawn control rod not applicable to control rods removed per Specifications 3.9.10.1 or 3.9.10.2.

(

LA SALLE - UNIT 1 3/4 1-14 i

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- i O. . REACTIVITY *CONTRO L SYSTEM

. 3/4.1.5 STAN08Y LIQUIO CONTROL SYSTEN LINITING CONDITION FOR OPERATION 3.1.5 The standby liquid control system shall be OPERABLE.

APPLICA81LITY: OPERATIONAL CONDITIONS f, 2, ar.d 5*.

ACTION:

a. In CPERATIONAL CONDITION 1 or 2:

1. With one motor operated suction valve, one pump and/or one explosive valve inoperable, restore the inoperable suction valve, pump and/or explosive valve to OPERA 8LE status within 7 days or be in at least HOT SHUTDQWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

2. With the standby liquid control system inoperable, restore the system to CPERABLE status within a hours or b.e in at least HOT SHUTOOWN within the next 12 bours,

b. In OPERATIONAL CONDITION 5*: -

1. With one r.stor operated suction valve, one pump and/or one

! - explosive valve inoperable, restore the inoperable suction valve, pump and/or explosive valve to OPERA 8LE status within 30 days or insert all insertable control rods within the next hour.

2. With the standby liquid control system inoperable, insert all insertable control rods within jes hour. I i

i l

SURE!LLANCE REQUIREMENTS l

4.1.5 The standby liquid control system shall be demonstrated OPERABLE:

i l a. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying that;

] 1. The available volume and temperature of the sodiu:n pentaborate solution are within the limits of Figures 3.1.5-1 and 3.1.5-2, and

, 2. The heat tracing circuit is CPERABLE-by verifying the indicated temperature to be > 60*/ on the local indicator. ,

i !

j . "With any control red withdrawn. Not apolicable to control rods removed per i Soeci fication 3.9.10.1 or 3.7.10.2.

l s .

1 LA JALLE - UNIT 1 3/4 1-19

- - -- --_ .- _,___m_

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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 I shown in Fi Aall be. ndeced 4ea. valw.

4 0.R5 Fw;gure 3 2.1-1. The (;u +s ( p;3ve L2.1-141s 4tue weirckin+ieu leeg j i APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or

! equal to 2 3 of RATED THERMAL POWER.

ACTION:

) .

With an APLHGR exceeding the I'mits of Figure 3.2.1-1, initiate 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 IS". of RATED THERNAL POWER within the next 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 3. .

n lO" SURVEILLANCE REQUIREMENTS l.

4.2.1 All APLHORs shall be verified to be equal to or less than the limits _.

determined from Figure 3.2.1-1:

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

b. Withi612 hourt after completion of a THERMAL POWER increase of at l

i least 1S*. of RATED THERMAL POWER, and i i i 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 1 1 i O LA SALLE - UNIT 1 3/4 2 I i .

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. POWER DISTRIBUTION LIMITS 3/4.2.2 APRM SEUCINTS

't,IMITING CONDITION FOR OPERtYION 3.2.2 The APRM flow biased simulated thermal power-upscale scram trip setpoint (S) and flow biased simulatad thermal ;iower-upscale control rod block trip

' setpoint (SR8) shall be established according to the following relationships:

r 5 !!N '""" ?" :;2:I t; 'O.!E' ' 5!%)F T C 'e?? *" r er :; r! t: (15" ' "*')T A g, _ na where: 5 and 5 are in percent of RATED THERMAL POWER, W=Lochgrecirculation flow as a percentage of the loep recirculation flow which produces a rated core flow of 108.5 million Ibs/hr, T = Lowest value of the ratio of FRACTION OF RATED THERMAL POWER divided by the MAXIMUM FRACTION OF LIMITING POWER DENSITY. T .

is always less than or equal to 1.0.

APPL!fA8ILITY: OPERATIONAL CONDITION 1, when THERMAL PCWER-is greater than or

. equal to 25". of RATED THERMAL POWER. -

ACTION: .

With the APRM flow biased simulated thermal power-upscale scra.a trip satpoint and/or the flow biased simulated thermal power-upscale control rod block trip setpoint set less conservatively than 5 or S , as above determined, initiate O. correceive actioa withia iS miautes aad restpe 5 and/or S to withia 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 toRYass 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 -

l

' 4.2.2 The FRTP and the MFLPD for each class of fuel shall be determined, the l

value of T calculated, and the most recent actual APRM flow biased simulated thermal power-uoscale scram and control rod block trip setpoint verified to be within the above limits or adjusted, as required:

a. At least once per 2a hours, t 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 I

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 MFLPD greater than or equal to FRTP.

i

! Nith MFLPD greater than the FRTP during power ascension up to 90% of RATED THERMAL POWER, rather than adjusting the APRM setpoints, the APRM gain may be

! adjusted such that APRM readings are great 9r than or equal to 100% times MFLPO,

, provided that the adjusted APRM reading does not exceed 100% of RATED THERMAL i POWER, the required gain adjustrent increment does . mot exceed 10% of RATED ThER."AL PC'..ER, and a nctice of :na ujust a n 's ::Ia :n .ne rasctor : ntrei

, Panal.

' s 1 s

V !

LA SALLE - UNIT 1 3/4 2-3 .

fHSe27 PA6E y/4/ A -3 TE Recirculation Loop Operation Q a.

S less than or equal te (0.66W + 51%)T S g less than or equal to (0.66W + 42%)T

b. Single Recirculation Loop Operation 5 less than or equal to (0.66W + 45.7%)T S g less than or equal to (0.66W + 36.7%)T O

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POWER DISTRIBUTION LIMITS POWER DISTRIBUTION LIMITS t

~/4.2.3 3 NINIMUM CRITICAL POPER RATIO

~

LIMITING CONDITION FOR OPERATION

3.2.3 The MINIMUM ITICAL POWER RATIO (MCPR) shall be equal to or greater than the MCPR limit d armined from Figure 3.2.3-1 times the Kp detarmined from Figure 3.2.3-2.

APPLICABILITY:

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

ACTION With MCPR less than the' applicable MCPR 1 it determined from Figures 3.2.3-1 and 3.2.3-2, initiate corrective action wit. n 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 r uce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 ho .

SURVEILLANCE REQUIREMENTS 4.2.3 MCPR, with:

l a.

t*** = 0.86 prior to performance of the initial scram time measur'ements for the cycle in accordance with Specification 4.1.3.2, or

b. t,y, 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 l

surveillance test required by Specification 4.1.3.2, l

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 10WER increase of I at least 15". of RATED THERMAL POWER, and 1

i 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.

A O- -

LA SALLE - UNIT 1 3/4 2-4 .

i

Inseer easiF 3/y 3-4 3.2.3 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be equal to or greater t than the MCPR limit determined from Figure 3.2.3-1 times the K, determined from Figure 3.2.3-2 for two recirculation loop operation and shall be equal O to or greater than the MCPR limit determined from Figure 3.2.3-1 + 0.01 times the K, determined from "gure 3.2.3-2 for single recirculation loop operation provided that the end ur-cycle recirculation pwp trip (EOC-RPT) system is OPERABLE per Specification 3.3.4.2.

APPLICABI'LITY:

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

, ACTION a.

With the end-of-cycle recirculation pump trip system inoperable per Specification 3.3.4.2, operation may continue and the provisions of Specification 3.0.4 are not applicable provided that, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , MCPR is determined to be' equal to or greater than the MCPR limit shown in Figure 3.2.3-1 EOC-RPT inoperable curve, times the K shown in Figure 3.2.3-2.

y

b. 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 .

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) 3/4.3 INSTRUMENTATI05'

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3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION j LIMITING CONDITION FOR OPERATION 3.3.1 As a minicum, the reactor protection system instrumentation channels shown in Table 3.3.1-1 shall be OPERABLE with the SEACTOR PROTECTION SYSTEM RESPONSE TIME as shown in Table 3.3.1-2.

APPLICABILITY: As shown in Table 3.3.1-1.

ACTION: .

a. With the number of OPERABLE channels less than required by the Minimum OPERA 8LE Channels per Trip System requirement for one trip system, place that trip system in the tripped condition" withinghour, The provisions l .

of Specification 3.0.4 are not applicable.

b. With the the number of CPERABLE channels less than required by the Minimum OPERABLE Channels per Trip fystem requirement for both trip systems,

. place at least one trip system ** in the tripped condition within.pne hour 1

l~

and take the ACTION required by Table 3.3.1-1.

s O .?

SURVEILLANCE REOUIREMENTS l 4.3.1.1 Each reactor protectiot, system instrmentation channel shall be demonstrated GPERABLE by the performance of the CHANNEL CHECX, CHANNEL FUNCTIONAL j '

TEST and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITICNS and at the frequencies shcwn in Table 4.3.1.1-1.

f i 4.3.1.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of j, all channels shall be performed at least once per 18 mo'nths.

4.3.1.3 The REACTOR PROTECTION SYSTEM RESPONSE TIME of eacn reactor trip

} functional unit shcwn in Table 3.3.1-2 shall be demonstrated to be within its

limit at least once per 18 months. Each tes( shall include at least one channel i per trip system such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a s?ecific reactor trip system. .

"With a design providing only one channel per trip system, an inoperable channel need not be picced in the tripped condition where this would cause the Trip Function to occur. In these cases, the inoperable channel shall be restored

{-

to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or the ACTION required by Table 3.3.1-1

. for that Trip Function shall be taken.

ss If : ore cr.3rae's t a ire;:er!!11e in :na tri system than in the otner, select that trip system to place in tne tripped condition, except wnen this would U cause the Trip Function to occur.

LA sat.LE - UNIT 1 3/4 3-1

4 m - .

Q^ .

REACTOR PROTECTION SYSTEM INSTRUMENTATION

! ACTION i?

a.w-

ACIION - Be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4 1 ,

! ACTION 2 -

Verify all insertable control rods to be inserted in the core i and lock the reactor mode switch in the Shutdown position within

._. j_ . . _ . - . _ _ _ one hour.

{ ACTION 3 -

Suspend all operations involving CORE ALTERATIONS

and insert all

insertable control rods within one hour.-

, ACTION 4 -

Be in at least STARTUP within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 5 -

Se in STARTUP with the main steam line isolation valves closed ~

within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ACTION 6 -

Initiate a raduction in THERMAL POWER withi 15 minutes and

! reduce turbine first stage pressure to < 14 psig, equivalent l t to THERMAL PohER less than 30% of RATED THERMAL POWER, within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

l. A

~. ACTION 7 -

Verify all insertable control rods to be insertad within,ons i hour.

ACTION 8 - Lock the reactor mode switch in the Shutdown position within'

, , one' hour. l .

t i i ACTION 9 -

Suspend all operations involving CORE ALTERATIONS,* and insert -

] all insertable control rods and lock the .^eactor mode switch in g the SHUTOOWN position within,gne hour. I i i .

I "Ex ept movement of IRM, SRM"or special movable detectors, or replacement of LPRM strings provided SRM instrumentation is,0PERABLE per Specification 3.9.2.

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i o 55i 90%f-test wJetiom -

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<O O LA SALLE - UNIT 1 3/4 3-4

.l

l . . - _ . . . . . . .

I!

m O, '

. -.) TABLE 3.3.1-1 (Continued) t REACTOR PROTECTION SYSTEM INSTRUMENTATION .

'I i 'l TA8LE NOTATIONS i

2 2 e

l (a) A channel may be placed in an inoperable status for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for

- I -

required surveillance without placing the trip system in the tripped condition provided at least one OPERA 8LE channel in the same trip system is monitoring that parameter.

.f.

I (b) The " shorting ifnks" shall be removed from the RPS circuitry prior to and i during the time any control rod is withdrawn" and during shutdown margin

demonstrations performed per Specification 3.10.3.

(c) An APRM channel is inoperable if there are less than 2 LPRM inputs per .

. ; level or less than 14 LPRM inputs to an APRM channel.

i (d) This function is not required to be OPERA 8LE when the reactor pressure vessel head is ut: bolted or removed per Specification 3.10.1.

.i j (e) This function shall be automatically bypassed when the reactor mode ~

switch is not in tha Run position.

~

l 3 (f) This function is not required to be OPERA 8LE when PRIMARY CONTAINMENT 4

INTEGRITY is not required.-

(g) Also actuates the standby gas treatment system.

! (h) With any control rod withdrawn. Not applicable to control rods removed i per Specification 3.9.10.1 or 3.9.10.2.

I

{ (i) Thisfunctionshagbeautomaticallybypassedwhenturbinefirststage j t

~

pressure is < 140 psig, equivalent to THERMAL POWER less than 30". of I RATED THERMAE PCWER.

J (j) Also actuates the EOC-RPT system.

l l

"Not requirea for control rods removed per Specifications 3.9.10.1 or l 3.9.10.2.

itial-s i l a . ... etpoint.

.a w.... F % ., !--s,etpc

.w ,

at te be dets 4aed

.....<..,u.nu. ...u_4.deria;

. a .. startup

.w. teet ; c; m -

__:,,4.. 4.u,.

'"- - '"' ~ " ' - '

an' i..'2' L.. ;_.'_._3 f'._ C._._ E .- "' ' -~ ~ " " '

,% P

~

LA SALLE - UNIT 1 3/4 3-5 I

i 9

- .~ . . .

. T. . . _ . _ . _ . . . . . . . . . . .. .

. _M TABLE 3.3.2-1

! y '

l- ISOLATION ACTUATION INSTRUMENTATION VALVE C'200PS NINIMUM OPERA 8LE APPLICABLE E OPERATED BY CHANNELS PER OPERATIONAL U TRIP Fl#4CTION SIGNAL (a) TRIP SYSTEM (b) CONDITION ACTION ,

l A. AllI0MATIC INITIATION f ~

! 1. PlflMARY CONTAINMENT ISOLATION .

l .. Reactor Vessel Water Level j (1) Low, Level 3 7 2 1,2,3 20 7-i.

(2) Low Low, Level 2 1, 2, 3 2 1, 2, 3 20 1 b. Drywell Pressure - High 2, 7 2 1,2,3 2G i c. Main Steam Line I 1 1) Radiation - liigh 1 2 1,2,3 21 j 3 2 1, 2, 3 22

. -> ia 2) Pressure - Low 1 2 1 23 I i ~

D 3) Flow - High 1 2/line(d) 1, 2, 3 21 I

Y al. Main Steam Line Tunnel -

! O -

Temperature - High 1 2 1,2,3 21

e. Main Steam Line Tunnel A Temperature - Higli 1 2 If d.k.3d) 21 k.

i (. Condenser.V.icuum - Low 1 2 1, 2", 3* 21 l 2. 51CONDARY CD:ITAINMENT ISOLATION

! a. Reactor Building Vant Exhaust 4 ICII*)

Plenum Radiation - HIGh 2 $,,2,,3,and** 24 i la. Drywell Pressure - High 4(c)(e) 2 1,2,3 24 l c. Reactor Vessel Water

{ Level - tow Low, Level 2 4 ICII*) 2 1, 2, 3, and 24 it. Fuel Pool Vent Exhaust 2 1, 2, 3, and ** 24 Radiation - liigh 4(c)(e) .

I t

G s

g

~

( - TA8tE 3.3.2-1 (Continued)

ISOLATION ACTUATION INSTRUMENTATION ACTION i

l! !

ACTION 20 -

Be in at least HOT SHUTDOWN within 12 ho'urs and in COLD SHUTOOWN with the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

l ACTION 21 -

Be in at least STARTUP with the associated isolation valves' t

closed within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in at least HOT SHUTOOWN withic 3 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLO SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Close the affected system isolation valves within pe' hour and l

.l ACTION 22 -

declare the affected system inoperable. d, ACTION 23 -

Be in at least STARTUP within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 24 - Establish SECONDARY CONTAINHENT INTEGRITY with the standby gas I

treatment system operating withinione hour. g.

ACTION 25 - Lock the affected system isolation valves closed within pat hour [

and declare the affected system inoperable.

ACTION 26 - Provided that the manual initiation function is OPERA 8LE for j each other groun valve, inboard or outboard, as applicable, in each line, restore the manual. initiation function to OPERABLE j status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ; otherwise, restore the manual initiation function t.2 OPERA 8LE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ; otherwise:

a. Be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in 1 COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or

b. Close the affected system isolation valves within the next l j t Q' hour and declare the affected system in operable.

NOTES .

l

Nay be bypassed with reactor steam pressure i 1043 psig and all turbine stop valves closed. ~

l

- When handling irradiated feel in the secondary containment and during CORE -

t ALTERATIONS and operations with a potential for draining the reactor vessel.

Ouring CORE ALTERATIONS anc operations with a potential for draining the 4 reactor vessel.

l_

(a) See Specification 3.6.3, Table 3.6.3-1 for valves in each valve group.

(b) A channel may be placed in an ineperable status for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for

- required surveillance without placing the trip system in the tripped condition provided at least one other OPERABLE channel in the same trip

.]

c j system is monitoring that parameter.

(c) Also actuates the standby gas treatment system.

l j (d) A cnannel is OPERABLE if 2 of 4 instruments in that' channel are OPERA 8LE.

5 (e) Also actuatas secondary containme1t ventilation isolation dampers per Table 3.6.5.2-1.

(f) Closes only RWCU system inlet outbaard valve.

i (g) Requires RCIC steam supply pressurt-low coincic'ent with drywell j pressure-high.

I ; (h) Manual initiation isolates 1(51-F008 only and only with a coincident i reac:cr vessel water level-1cw, level 3. d ;nal. .

l' ([3 h& CLcmeds of coek -hrlp sysles Mgy be. 71 Aced ,*" ON /NogeraWe f p\ / g%s, fee op de 4 ho@s for regelref. 6eackc boildlac Welab f,lfer- cloge i .

W A k: emL %

( Add dAyr Cycbg WNW pAcig &.4r.g sysie +y-w O- u SAttE . UNIT 1 3,4 3 14 r.u u A a.<~

cbmk ja +1s sue ing 4s us m d

' C tr# WG. .

. .~ . .. . .. -.

' ~

i D D .

[.)

Ii

. TA8LE 3.3.2-2 l 1 ISOLATION ACTUATION INSTRUMENTATION SETPOINTS.

l u.

! # ALLOWASLE VALUE i

i !~i TRIT' IllllCTION TRIP SETPOINT A. AUTOHAlIC INITIATION c -

. 1. PHlHARY CONTAINMFNT ISOLATION 5.4 4 p., 4. Reactor Vessel V;.ier level

1) Low, tevei 3 > 12.5 inches * > 11.0 inches *

. 2) Low tow, Level 2 5 -50 inches * . I -57 inches * '

i b. Drywell Pressure - High 51.69psig. 51.89psig .-

c. Main Steam Line l

1) Radiation - H10h 1 3.0 x full power background i 3.6 x full bactground j l 2) Pressure - Low > 854 psig 1 834 psig , ,

i 116 psid

3) Flow - liigh 1 111 psid

d. Main St<*as Line Tunnel Temperature - High i 140*F8 1 146* 8

c. Main Steam Line Tunnel

) a Temperature - Hioh 1 24"F 1 30* .

i., t. Condenser Vacuum - Low > 7 inches Hg vacuum > 5.5 laches Hg vacuum ,

e'.

2. SECONDARY CONTAINHENT ISOLATION .

a. Reactor Building Vent Exhaust Plenum Radiation - liigh 1 10 mr/h 8 1 15 mr/h 8 .

I

h. Drywell Pressure - liigh 1 1.69 psig i 1.89 psig I c. Reactor Vessel Water l

Level - Low Low, Level 2 1 -50 inches

1 -57 inches *

d. Fuel Poos Vent Exhaust Radiation - liigh 1 10 nr/h 8 a 1 15 mr/h

3. I!l' ACTOR WATER CLEANUP SYSTEM 150tATION a.g Aflow - High 1 70 DP8 1 87.5 gym

h. Ricat Exchanger Area Temperature l31 12 1

- liigh 1)Sh*F@ ~

$ )Si'F F

c. lleat Exchanger Area Ventilation qi -

AT - liluh i l& 13 eta 12 2.

d. Pump Area lemperature - Il10h 1 M8" Pump Area Ventilation AT - High ii 36*F J E @F ,

c. i Ja*F

1. SLCS Initiation NA 13 , NA L [g i 9 Reactor Vessel Water Level -

Low Low, level 2 > -50 inches

l 1 -57 inches * '

l sw

0 'D ')

g TABLE 3.3.2-2 (Continued) i 10 ISOLATION ACTUATION INSTRUNENTATION SETPolNTS l .

F

" ALLOWA8tE -

i TRIP FUNCTION TRIP SETPolNT VALUE

! =

j i q 4. HfACIOR CORE ISOLATION COOLING SYSTEM ISOLATION .

d. RCIC Steam Line Flow - High 1 290% of rated flow, 178" H 2O 1 295% of rated flow, 18,5" H 2O ;

b. RCIC Steam Suppiy Pressure - Low 1 57 psig 1 53 psig

c. RCic Turbine Exhaust Olaphragm #

i 20.0 psig i Pressure - !!!gh 1 10.0 psig l d. RCIC Equipment Roon l Temperature - High .

I 200*y M i 206* & .

e. RCIC Steam Line Tunnel ,

j lemperature - High < 200*19S

~

-< 206*F 8 j

f. RCIC Steam Line Tunnel

. A Temperature - High < 117*f N < 123*8 4 -

R g. Drywell Pressure - liigh $1.69psig 31.69psig 4

5. HilR SYSTEM SIEAM CONDENSING M00E ISOLATION -

4 m .

a. RilR Equipment Area. '

j A Temperature - High 1 50*f @ .

1 56*f W

b. RilR Area Cooler Temperature -

liigh 1200*M i 206*f Y

r. . RilR lleat Excl$ anger Steam Supply fluw - High 1 123" H 2O , 1 128" H 2O

. Seeo g a

m . ..-

s) .) J I

TABLE 3.3.2-2 (Continued) ",

C .1

y. ISOLATION ACTUATION INSTRUMENTATION SETPOINTS 2 I i l, ALLOWABLE j TRIP FU_IICTION TRIP SE1PolNT VALUE l i c-

, t 5-i

6. RilR SYSTEM SiluTDOWN COOLING MODE ISOLAT10N

~ a. Reactor Vessel W ter Level - '

Low, Level 3 1 12.5 inches

1 11.0 inches *

b. Reactor Vessel i (RilR Cut-in Permissive) i Pressure - High i135psig**[ i145psig*d

)

c. HilR Ptap Suction flow - liigh 1 180" H 2O i 186" H 2O .

I

d. RilR Area Cooler Temperature -

5:'

liigli 1 200*f @ $ 206*F @

';* e. RilR Equipment Area AT - liigh i 50*F Y i 56*fW l C .

Not Applicable B. 11AllilAL INIIIATION . Not Appilcable .

1. Inboard Valves .

! 2. Outboard Valves

3. Inboard Valves

4. Outboard Valves .

5. Inboard Valves -

6. Outboard Valves . .

! /. Outboard Valve -

l l

Tim 'Itases fTside B 3/4 3-1.

Mn o i.ii seipuli.t. f ta:1 ;;tpc! t te be 6t:=is 4 4. iia; ;tertup 12;i. ris,vii.e.. ".; , i g ' sed -r h W

.tkit -8:eint sh II bc :;;iclited t: 10. E h:;!:n .:! thin "O ty:; c' te ;t : ;-1 th -

/^^Correc.ted for cold water head with reactor vessel flooded.

I

i-TA8LE 3.3.2-3 ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME TRIP FUNCTION RESPONSE TIME (Seconds)# ,,

A AUT0pmTIC INITIATION

1. PRIMARY CONTAINMENT ISCLATION ,

a. Reactor Vessel Water Level

1) Low, Level 3 NA

2) Low Low,' Level 2 < 1. I D. Drywell Pressure - High {13g*f<13,).. L

c. Main Steam Line

1) Radiation - High(b) < 1.0*/< 13(a)**

2) Pressure - Low I 1.0*/7 13(a)..

3) Flow - High I 0.5*/7 13(a),.

~

d. Main Steam Line Tunnel Temperature - High NA

e. Condenser Vacuus - Low NA

f. Main Steam Line Tunnel a Tempera +.ure - High -NA

2. SECONDARY CONTAINMENT ISOLATION

a. ReactorBuilding(g3ntExhaustPlenua, < 13(,)

Radiation - High r k. Orywell Pressure - High 7 13(a)

c. Reactor Vessel Water Level - Lcw, Levelg gI i 13(a)

Q d. Fuel Pool Vent Exhaust Radiation - High 313(a)

3. REACTOR WATER CLEANUP SYSTEM ISOLATION

a. A Flow - High < 13(a)##

b. Heat Exchanger Area Temperature - High NA -

c. Heat Exchanger Area Ventilation AT-High NA

d. Pump Area Temperature - High NA

e. Pump Area Ventilationr AT - High NA

f. SLCS Initiation NA l

g. Reactor Vessel Water Level - Low Low, Level 2 < 13(a)

4. REACTCR CORE ISOLATION COOLING SYSTEM ISOLATION J

a. RCIC Stuas Line Flow - High < 13(&) # #

i ,

b. RCIC Steam Supply Pressure - Low I 13(')

c. RCIC Turbine Exhaust Diaphragm Pressure - High RA

d. RCIC Equipment Room Temperature - High NA

e. RCIC Steam Line Tunnel Temperature - High NA

f. RCIC Steam Line Tunnel A Temperature - High NA

. g. Drywell Pressure - High NA

5. EH9 SYSTE'i STEAM CONDENSING COE ISOLATION

a. Riin Equipment Area a Tppdra ure - High NA

b. RHR Area Cooler Temperature - High NA y

c. RHR Heat Exchanger Steam Supply Flow High NA l

O LA SALLE - UNIT 1 3/4 3-18 ,

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

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

4

~

(

TABLE 3.3.2-3 (Continued)

ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME .

TRIP FUNCTION RESPONSE TIME (Seconds)#

6. RHR SYSTEM SHUT 00WN COOLING MODE I$0LATION

a. Reactor Vessel Water Level - Low, Level 3 < 13(a)

~

b. Reactor Vessel (RHR Cut-In Permissive) Pressure - High .AlA N.A.

c. RHR Pump Suction Flow - High NA4/.A.

d. RHR Area Cooler Temperature High NA A/. A.

e. RHR Equipment Area .iT High NA A/. A.

9

8. MANUAL INITIATION "-t ."ppi kcW A/. A .

1. Inboard Valves

2. Outboard Valves .

3. Inboard Valves

4. Outboard Valves

. 5. Inboard Valves .

, , 6. Outboard Valves

! 7. Outboard Valve l

O <> ra 4 it4#=>=t<tatr# atati#r recorded as a part of the ISOLATION SYSTEM RESPONSE TIME.

si nit 6 Isolation

rd #a l

system instrumentation response time specified includes the delay for diesel generator starting assumed in the accident analysis.

(b) Radiation detactors are exempt from response time testing. Response time i shall be measured from detector output or the input of the first electronic component in the channel.

t

Isolation s'ystem instrumentation response time for MSIVs only. No diesel l

generaccr delays assumed.

4 l

- Isolation system instrumentation response time for associated valves

except MSIVs.

i

Isolacion systec instrument.ation response time specified for the Trip 4

i Function actuating each valve group shall be added to isolation time i snown in Table 3.6.3-1 and 3.5.S.2-1 for valves in each valve group

. to obtain ISOLATION SYSTEM RESPCNSE TIME for each valve.

I # 'Jithout 45+1 second time delay.

~ *

. :r#.g WsthouT 4 S Ssund hme dely.

0 N. A . US A llcOlb l O '

LA SALLE - UNIT 1 3/4 3-19

, _ . . . _ - . , . . , . _ _ - . . - , . _ _ _ . _ . . . - _ - . . , . - - - . ,.__-_._..-m.- . - - - - - - - - , . . . , --

^ ^ ~

e

. . i INSTRUMENTATICN .

O .~ ~ END-OF-CYCLE RECIRCULdTION PUMP TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERA ION 3.3.4.2 The end-of-cycle recirculation, pump trip (ECC-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 shown in3.3.4.2-2andwiththe(RECIRCULATIONPUMPTRIPSYSTEMRESPON Table 3.3.4.2-3. En D - o r- c'r c.LE

, APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to ac:5 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 OPERA 8LE status with the channel setpoint adjusted consistent with the Trip Setpoint value.

b. With the number of OPERA 8LE channels one less than required by the Minimum GPERABLE Channels per Trip System requirement for one or r both trip systems, place the inoperable channel (s) in the tripped condition within y hour. l

.. c. With the numoer of OPERABLE channels two or more less than cequired 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 -

t 1

channel and one turbine step valve channel, place both inoperable channels in the tripped condition within j ae hour. l l .$.

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

d. With one trip system inoperable, restore the inoperable trip system '

to wOPERASLE me ,r oar:n mestatus yu q ao within athin 72tha hours am 5orheu-e.

rette M"'%

ehe. 7CaER Se R :.a. A

rugviird % Spac;b_hoa 3.2.3 I i e. With both trip systems ingperable, restore at least one trip system i t to OPERASLE status within2pae hour or d: :: T1'.'1. CaER w i===

i

! -tt.= 27. J "ATC 'ER.".' L "0SE. 4 thi a the ccxt 5 h: r . h ita. 42 [4(IIOM 9 $i ch 3.2.3.

regvired i

1

~

l OL LA SALLE - UNIT 1 3/4 3-39 .

. _ , _ . - . , - . . _ - , - - , , 7 ..y, __._ _ ,

,m,,p.m,,

, ,y.,, -f, , ,_ , _ _ _ .

~~

, c .

.i t

I TABLE 3.3.4.2-1 - 4 C

m END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION ;

N -

$ 081NIMUM

, OPERABLECilANNEgy ,

c- TRIP fuuCTION PER TRIP SYSTEM '

e

1. furbine Stop Valve - Closure 2(b) 2 IDI

2. Tushine Control valve - Fast Closure

? .

R w (a)A L:ip system may be placed in an inoperable status for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for required survet11ance ,

a girovided that the other trip system is OPERABLE.

(b)Ihis function s. hall be automatically bypassed when turbine.first stage pressure is less than or equal to 14bpsig, equivalent to TilERHAL POWER less than 30% of RATED THERMAL POWER. .

I

Sinal-setpoint t 5: dete rfaed d--85 -* r* ;

st ;-- -

?.; :: ! sed-eksage-4e- I i tato M <-ymint

-ti115 Tatp~oTnt1tratl-be-submitted to tha Commission withier ^a, -,. .e t..i. sm ht'-- {

i l

O e

e W'Be e 4.

O , D .

, 9 TABLE 3.3.6-2

~.

, g

! CONTROL ROD WITHDRAWAL BLOCK INSTRMENTATION SETPOINTS k TRIP FUNCTION .

TRIP SETPOINT ALLOWABLE VALUE

[ 1. R00 SLOCK MONITOR 5 . Upscale e j e. 55 Y ^ '"

. Inoperative x 0.;; ;; ; G j 13,.te c. Downscale

.HA. Al.A. .NA.#, A.

2. 1 5% of RATED THEIDEL POWER ]

APRM 135 of RATED THEM4L POWER a.

! Flow Blased Simulated .

Thermal Power-Upscale I

1 g ,37 Inoperative < 5. 55 M ^ '?*

-3Ar A7. A. > 9 M Y ^ di t.*.L

c. Downscale .N4 As.A .

d. Neutron Flux-High > 5% of RATED THERMAL POWER l

> 35 of RATED THf MAL POWER

3. 3125ofRATEDTHERMALPOWER l SOURCE RANGE MONITORS 314% of RATED TMEM4L POWER R

a. Detector not full in-

b. Upscale N ArxN10

.Acps.

.5 Y c. Inoperative <2 -N4 M.A. 5 j 1", -HAr 12. A .

< 5 x 10 cp, ,

d. Downscale E4-p.4 . I 1 0.7 cps

4. 1 0.5 cps INTERMEDIATE RANGE MONITORS

a. Detector not full in

{ b. Upscale NA M.A.

-NArAl 4.

j

c. Inoperat.ive 5 108/125 of full scale 5 110/125 of full scale l

d. Downscale M d.A.

. MAN.A.

i 5.

1 5/125of full scale 1 3/125 of full scale g i, SCRAM DISCHARGE VOLUME

a. Water Level-High

b. Scram Discharge Volume 1 765' % "

j

$ 765' %"

Switch in Bypass JiA d.A.

l g 6. -N4NA REACTOR COOLANT SYSTEM RECIRCULATION FLOW l

~

,$ a. Upscale g b. Inoperative $ 108/125 of full scale .

x c. Comparator

-NAN,4. 5 111/125 of full scale O .M4-A7 4.

1 10% flow deviation -

! 5 11% flow deviation '

h

The (W). Average Power Range Monitor rod block function is varied as a function of recircu The trip setting of this function mur.t be maintained in accordance with Specification ...

322

. l

TM5e2TS FOR PA6E 3/g 3-53 0 .

TUSetT A

1) Two Recirculation < 0.66 W + 43%

Loop Oparation < 0.66 W + 40%

2) Single Recirculation < 0.66W + 37.7%

Loop Operation 1 0.66W + 34.7%

JNSedT G

. 1) Two Recirculation < 0.66 W + 45%*

Loop Operation < 0.66 W + 42%*

Single Recirculation Q 2).

Loop Operation 5 0.66W + 36.7%*

< 0.66W + 39.7%*

O

,.------,,..m- ,,-ne, -

l . .. . . . . . _- -- - - - . . . - . .. .- .

I I k! -

l t

i* '

TABLE 4.3.6-1 i t g *

,,, CONTROL R0D WITilDRAWAL BLOCK INSTRilNENTATION SURVEILLANCE REQUIREMENTS

' # CilANNEL OPERATIONAL j

i G CHANNEL CONDITIONS FOR 1AllCH L CHANNEL FUNCTIONAL TRIP IllllC110N CilECK lEST CALIBRATION I ,I SultVEILI ANCE ItEQUIRED i 4 ! c.

h I. IJ)l,_BtOCK Jl NDN110R I

" IJpscale b)(c) c) q ja !

! a. NA S/U((b)(c), c) .yg,y,g, ja l i

i i h. Inoperat.ive NA S/U Nd, d j c. Downscale NA S/U , Q 1* ,

I 2. APill! _

d. EloW Olased Simulated '

l lhermal Power-Upscale -

NA S/U((b) y y3 y .

l h. Inoperative NA S/U b),M , -NA-N 1, 2, 5

! s. . Ilownscale NA S/U ,M -4~ SA 1 1 it. lieutron flux-liigh NA S/U ,M g SA 2, 5 ;

,Util:CE ltANGL HUNilDRS I

$ 3.

a. NA -NA.N .A . 2 '5 I- .

4 *;>

h.

lletector not full in Upscale NA S/Ufb)W S/U

,W 2, 5

f g , Q I

c. Inoperative NA S/UIb),W -NA-al . A . 2, 5 .

I d. Downtcale NA S/U ,W Q 2, 5 )

! 4. Illillmtill ATE RANGE NONITORS .

I

.. Detector not full in NA S/U ,W 414-N. A . 2, 5 j h. Upscale NA W Q 2, 5

t. . Inoperative NA S/U(b),W -41A-it).A. 2, 5 I

S/U(b),W

d. Downscale NA S/U , , Q 2, 5

5. SCIUJ4 DISCilARGE VUluME

a. Water Level-High NA Q R' 1, 2, 5**

b. Scram Discharge Volume Switch in Hypass NA M .JiA-N. A . 5** j 6.

i El'AC10R COOLANT SYSILM RECIRCULATION FLUW

a. Upscale NA S/U(b) y , q y

h. Inoperative NA S/U .M -- NAr M A . 1

c. ComparaLor NA S/U ,M Q 1 ;

l '

O. ,.- ..

., TABLE 3.3.7.1-1 (Continued)

RADIATIdMMONITORINGINSTRUMENTATION .

2- -- - - - - - -

ACTION

' ACTION 70

a. With one of the required monitors inoperable, place the inoperable channel in the downscale tripped condttion i~

withinmihour; restore the inoperable channel to I OPERABLE status within 7 days, or, within the next 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months , initiate and e.aintain operation of the controi Som emergency filtration system in the pressurization mode of operation.

b. With both of the required monitors inoperable, initiate .

and maintain operation of the control rcos emergency filtration system in the pressurization mode _of operation

. within ,osw hour. g

.1 t

t

b i

l .

4

0. '

. LA SALLE - UNIT 1 3/4 3-58 t

+ .

- _ .- - -e --

m.e---p.M -ee6*N- m

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

. i INSTRUMENTATION .

i #

SEISMICMONITORINGINSYRUMENTATION

! LIMITING CONDITION FOR OPERATION 3.3.7.2 The seismic monitoring instrume'ntation shown in Table 3.3.7.2-1 shall be OPERA 8LE.*# l APPt.ICA8ILITY: At all times.

ACTION:

a. With one or more seismic monitoring instruments inoperable for more than 30 days, in lieu of any other report required by Specification 6.6.8, prepare and submit a Special Report to the Commission pursuant to Specification 6.6.C within the next 10 days outlining the cause -

4 of the malfunction and the plans for restoring the instrument (s) to OPERA 8LE status. -

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.7.2.1 Each of the above required seismic monitoring instruments shall be .

l demonstrated CPERA8LE by the performance of the CHANNEL CHECK, CHANNEL j . FUNCTIONAL TEST and CHANNEL CALIBRATION operations at the frequencies shown in {

Table 4.3.7.2-1).

l l

. 4.3.7.2.2 Each of the above required seismic monitoring instruments actuated

] during a seismic event greatar than or equ?'s to 0.02g shall be restored to. (

)

OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and a CHANNEL CALIBRATION performed within '

5 days following the seismic. event. Data shall be retrieved from actuated instruments and analyzed to determine the magnitude of the vibratory ground motion. In lieu of any other report required.by 5pecification 6.6.8, a Special

! Report shall be prepared and submitted to the Commission pursuant to Specifica-tion 6.6.C within 10 days describing the magnitude, frecuency spectrum and resultant effect upon u6ft features important to safety.,

t l

! * ^ine nor :al or emergency power source may be inoperable in OPERATIONAL j CONDITION 4 or 5 or when defueled.

I

%M s (, M od.To d E s h M a d tuli N $7 s fM 'E . SkMF1 beltne2~ LASalle. Ch1. cad LaSalte. Un.+ 9, (v 1 i LA SALLE - UNIT 1 3/4 3-60

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1 ._. ..

. _ - -- __ ______ ..__.._.-- _. _ . ._ ..____- _.s

_- s

~

i.

h- .

INSTRUNENTATION e

w/ METEOROLOGICAL MONITORING INSTRUMENTATION s .

i.INITING CONDITION FOR OPERATION 3.3c7.3 The meteorological monitoring instrumentation channels shown in Table l i 3.3.7.3-1 shall be OPERASLE.*

I APPLICABILITY: At all times.

ACTION:

i i a. With one or more meteorological monitoring instrumentation channels

' inoperable for more than 7 days, in lieu of any other report required

'by Specification 6.6.8, prepare and submit a Special Report to the

Commission pursuant to Specification 6.6.C within the next 10 days outlining the cause of the malfunction and the plans for restoring the instrumentation to OPERABLE status.

i b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

O ' SURVEILLANCE REQUIREMENTS l

I 4.3.7.3 Each of the above required meteorological monitoring instrumentation -

! , channels shall be demonstrated OPERABLE by the performance of the CHANNEL i CHECK and CHANNEL CALIBRATICN operations at the frequencies shown in Table 4.3.7.3-1. .

I 4"The normal or emergency power' source may be inoperable in OPERATICNAL

! CONDITION 4 or 5 or when defueled.

1 I

! 8 MeWorol0*Al Med.4er:m .Tvs4ra.w4 max Sys0m is Sko ud.

' baks,La%$Un;&Ao h % lb. U a.'+ 9 .

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LA SALLE - UNIl 1 3/4 3-63

. . . - . . . - - . - -. . ~ . - - - - - . ~ . . - - . .-.

. - - . ...u.. -.. .. - . .

b j .

i . .

i TABLE 3.3.7.5-1

% ACCIDENT MDNITORING INSTRUNENTATION

, m i 2-F

REQUIRED MINIIR41 2 i NulBER OF CilANNELS CilANNELS OPERA 8LE .

l ' E e-o

. -1 1. Iteactor Vessel Pressure 2 1 i H .

2. Reactor Vessel Water Level 2 1 l.

3. Suppression Chamber Water Level 2 1 l

4. Suppression Chamber Water Temperature 7,1/wel 7,1/wel I 1

i 5. Supression Chanbar Air Temperature 2 1 l

\ '

6. 1)rywell Pressure 2 1 s" 2 1

! = 7. Drywell Air Temperature ,

, u y 8. niidell Oxygen Concentration" 2 1

~

. 9. Drjwell Hydrogen Concer.tration Analyzer

and Monitor 2 1 ll I a

l i 10. Primary Containment Gre:s Gamma Radiation 2 1 1 .

i 11. Safety / Relief Valve Position Indicators 1/ valve 1/ valve '

'l t

12. Huble Gas Honitor, Main Stack 1 1 l ,

i i.

t!.

13. Ilohle Gas Monitor, Standby Gas Treatment System Stack 1 1 f I

4 g

'I Actuated after LOCA.

g. .

.i Ir.it!;' requirement. -Final requirement-to be-determined-aftee-itemanstratten-of-4:errelatten-of-peel-bulk s L :. , .:ter: : :n.r:d by 02:5 d!"! !: t: ; :1 bulb

eratur :: !:rred by-be*h divistensc-Results I,

.of_ h.munstration and-necessa y che~;es te this specif!satien rbr!! bc :"-ittr ' t t'r f -!s 8 -- - '- -

90-daywf toonshath ,

h.

. s INSTRUNENTATION

'. 500PCE RANGE NONITORS 1

LINITING C0 5! TION FOR OPERATION 4

3.3.7.6 At least three source range monitor channels shall be OPERA 8LE.

I 3 APPLICABILITY: OPERATIONAL CONDITIONS 28, 3)and4.

ACTION:

t

a. In OPERATIONAL CONDITION 2* with one of the above required source .,4 bee, g range monitor channels inoperable, restcre at least f source range acnitor channels to OPERAdLE status within 4 nours or be in at least H0T SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ,

b. In OPERATIONAL CONDITION 3 or 4 with two or more of the above required source range monitor channels inoperable, verify all insertable control rods to be inserted in the core and lock the reactor mode switch in the Shutdown position within,gne hour. , l 1

SURVEILLANCE REQUIRE *1ENTS G

4.3.7.6 Each of the above required source range monitor channels shall be demonstrated CPERABLE by:

a. Performance of a:

. 1. CHANNEL CHECK at least once per: -

a) 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months in CONDITION 2*, and b) 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in CONDITION 3 or 4.

2. CHANNEL CALIBRATION ** at least once per 18 months.

I

b. Perfortsance of a CHANNEL FUNCTIONAL TEST:

1. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to moving the reactor mode switch from

( the Shutdown position, if not performed within the previous l 7 days, and

2. At least once per 31 days.

c. Verifying, prior to withdrawal of control rods, that the SRM count l

rate is at least 0.7 cps 1fwith the detector fully inserted. l l

l "With IRM's on range 2 or below.

! ** Neutron detectors may be excluded from CHANNEL CALIBRATION.

ItVMedl. S WI -40 40isc rcc4i0- i s 2 2., 04herw.sa. , 3 e s. (

LA SALLE - UNIT 1 3/4 3-72 Amendment No. 2 i

.e . i_

INSTRUMENTATION

. RADI0ACTIVELIQUIDEFFi.UENTMONITORINGINSTRUMENTATION LIMITING CONDITION FOR OPERATION

/

3.3.7.10 The radioactive liquid effluent monitoring instrumentation channels shown in Table 3.3.7.10-1 shall be OPERA 8LE with their alare/ trip setpoints set to ensure that the limits of Specification 3.11.1.1 are not exceedec. The alars trip setpoints of these channels shall be determined in accordance with the Offsite Dose Calculation Manual (00CM).

APPLICA8ILITY: At all times.

ACTION:

a. With a radioactive liquid affluent monitoring instrumentation channel -

alars/ trip setpoint less conservative than required, immediately suspend the release of radioactive liquid effluents monitored by the affected channel or declare the channel inoperable. .

b. With less than the minimum number of radioactive ifquid effluent .

monitoring instrumentation channels OPERA 8LE, take the ACTION shown i in Table 3.3.7.10-1.

}; c. The provisions of Specifications 3.0.3 and 3.0.4-are.not appitcable.

SURVEILLAhCE REQUIREMENTS 4.3.7.10 Each radioactive ifquid afflutat monitoring instrumentation channel shall be demonstrated OPERA 8LE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations at the i frequencies shown in Table 4.3.7.10-1.

l l '

t . . Restore the inoperatle tastrumentation to I

OPERABLE status within the time specified in the ACTION or, in lieu of a Licensee Event Report, explain in the next Semiannual Radioactive Effluent Release Report why this inoperability was not corrected within the time specified.

i i

D i

, LA SALLE - UNIT 1 3/4 3-81 e

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g TABLE 3.3.7.10-1 h RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION l;;

$ MINIMUM .

1 5

H CHANNELS INSTRUMENT OPERABLE ACTICN

1. GAMA SCINTILLATION MONITOR PROVIDING ALARM AND AUTOMATIC

] TERMINATION OF RELEASE I

a. Liquid Radwaste Effluent Line (0010-0002) 1 100 l

2. GAMA SCINTILLATION MONITORS PROVIDING ALARM BUT NOT PROVIDING AUTOMATIC TERMINATION OF RELEASE a m a. Service Water System Effluent Line -(1010-X000)- (Unit O 1 101 1

b. Effluent Line (1010-X50')- 101

) RHR Service Water (Line A) 1 m c. RHR Service Water (Line B) Effluent Line-(1010 K665) 1 101 {

a d. Service Water System Effluent' Line ffHH6-X000)- (Ug+ g) 1 101 l l N

3. FLOW RATE MEASUREMENT DEVICES -

a. Liquid Radwaste Effluent Lir.e -(0 FIT 'T 017 . .J 010)- 1 102 l

b. River Discharge - Blowdown Pipe {0F'i '4001)- 1 102 i

I N

3 3 -

r Ch

o

-g INSTR W ENTATION ,

TA8 G 3.3.7.10-1 (Continued) p TAs u NOTATION ACTION 100 - With the number of QPERABE channels less than required by the Minimum Channels OPERA 83 requirement, effluent releases may continue for up to 14 days provided that prior to initiating a release

4

a. At least two independent samples are analyzed in accordance with Specification 4.11.1.1.3, and l

b. At least two technically qualified members of "the f

Facility Staff independently verify the release rate calculations and discharge line valving; 1 .

Othentise, suspend release of radioactive affluents via

' this pathway. ,_

I l

ACTICM 101 - With the number of channels OPERA 8 G 1ess than required

. by the Minimum Channels OPERAB M requirement, effluent -

releases via this pathway any continue for up to 30 days

\_' provided that, at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , grab samples are collected and analyzed at a limit of detection of at least i ( ,'1

(

10-7aicrocurisj'/miorgammaspectrometricanalysis.

, ACTIGN 102 - With the number of channels OPERABLE less than esquired

~! by the Minimus Channels OPERA 8LE requireetat, effluent releases via this pathway may continue for up to 30 days i provided the flow rate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during actual releases. Pump curves for Instru-

] g sent 3a, or for kn1wn valve positions for Instrument 3b, i may be used to estimate flow.

.i l s s

vv .

e LA SA RE - UNIT 1 3/4 3-83 .

-- .-m-w,,, ,,w- -

A f

j g TABLE 4.3.7.10-1

% RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS E

y, . ..

c.

CHANNEL 5 CHANNEL SOURCE FUNCTIONAL CHANNEL H INSTRUMENT CHECK CHECK TEST CALIBRATION s

1. GAMA SCINTILLATION MONITOR PROVIDING ALARM i AND AUTOMATIC TERMINATION OF RELEASE i

a. Liquid Radwaste Effluents Line C P Q(1) R(3)

2. GAMA SCINTILLATION MDM TORS PROVIDING ALARM i BUT NOT PROVIDING AUTOMATIC TERMINATION OF RELEASE l

' $ a. Service Water System Effluent Line (Uv;4 f) .

, -{ !918-K599)- D M Q(2) R(3) l.

b. RHR Service Water (Line A) Effluent Line D M Q(2) R(3)

I c. RHR Service Water (Line B) Effluent Line D M Q(2) R(3)

d. Service Water System Effluent Line (()p;f 2) -

l 1 -(2010 K666T D M Q(2) R(3) 1 l 3. FLOW RATE MEASUREMENT DEVICES

a. Liquid Radwaste Effluent Line 0(4) N.A. Q R j b. River Olscharge - Blowdown Pipe D(4) N.A. Q R i

1

, a i

y .

a m

) -

4

i' Oc - -

1N5 = ,.NTAT . .

TABLE 4.3.7.11-1 (Continued)

. TABLE NOTATION

,.t .

At all times. ,

i ** During mein condenser offgas treatment system operation.

During operation of the main condenser air ejector. -

N Ouring operation of the 58GTS.

4kt.

(1) The CHANNEL FUNCTIONAL TEST shall also demonstrate.thet automatic isoTation cyle;ldg of this pathway,and%6ntrol room alarm annunciation occurs if any of the fk"*" Yk";hf , AN.

1. Instrument indicates measured fevels above the alarz/ trip setpoint. -

2. Loss of power. , _

3. Instrument alarms on downscale failure.

- *"2a**"L?BL"?LS*!

(2) The CHANNEL FUNCTIONAL TEST for N1og scale monitor shall also

'" *L*@,3."c'22:WT.* W*'- t .

1 demonstrate that control room alarm annunciation occurs if any of the following conditions exists: -

Q. 1. Instrument indicates measured levels above the alarm setpoint.

2. Loss of power.

3. Instrument alarms on downscale failure.

4. Instrument controls not set in Operate or High Voltage mode.

i (3) The initial CHANNEL CALIBRATION shall be performed using one or more of

the reference radioactive standa'rds certified by the National Bureau of g3 standard 4 cr using standards that have been obtained from suppliars that f

participate in measurement assurance activities with N85. These standards

shall permit calibrating the system over its intended range of energy and

measurement range. For subsequent CHANNEL CALIBRATION, the initial i reference radioactive standards or radioactive sources that have been i . related to the initial calibration shall be used.

l .

(4) The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal:

! 1. One volume percent hydrogen, oslance nitrogen, and

2. Four volume percent hydrogen, balance nitrogen.

I (5) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room i alarm annunciation occurs if any of the following conditions exists:

j 1. Inf.trument indicates measured levels abcve the alara set;:oint.

. 2. Circuit failure.

3. Instrument controls not set in the Operate mode.

i LA SALLE - UNIT 1 3/4 3-90 ,

_ _ . , , , , - .,y-_--_.. ,,,..,w,m._._, ._,_-._,__..,y_m, yy-, ,,,-m-,,,_,,--- -,_,,--,,.,n,_-

3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RE RCULATION SYSTEM "JW58# d 4 RECIRCULATION PS LIMITING CONDITION OR OPERATION

\

3.4.1.1 Two reactor co lant system recirculation loops shall be in operation.

APPLICABILITY: OPERATION CONDITIONS 1* and 2*.

ACTION:

I

a. With one reactor coolant stem recirculation loop not in operation, be in at least HOT SHUT ithin the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. With no reactor coolant system ecirculation loops in operation, place the reactor mode switch i the Shutdown position.

~

l4

,./ SURVEItLANCE REQUIREMENTS 4.4.1.1 Each reactor coolant system recirculation loop flow control valve shall be demonstrated 00ERABLE at least once per 10 nths by:

'a. Verifying that the control valve fails "as i " on loss of hydraulic pressure at the hydraulic power unit, and

b. Verifying that the average rate of control valve ovement is:

1. Less than or equal to 11% of st:oke per secon opening, and

2. Less than or equal to 11% of stroke per second c sing.

l "See Special Test Exception 3.10.4.

l l

(O LA SALLE - UNIT 1 3/4 4-1

~

- ~- .

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

l OPERATIONAL CONDITIONS 1" and 2".

APPLICABILITY:

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) Reduce THERMAL POWER to 5 50% of RATED THERMAL POWER, and, r- c) Increase the MINIMUM CRITICAL POWER RATIO (MCPR) Safety Limit by 0.01 to 1.07 per Specification 2.1.2, and,

- Ci d) Increase the MINIMUM CRITICAL POWER RATIO (MCPR) Limiting

- Condition for Operation by 0.01 per Specification 3.2.3, 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, and, l

f) Reduce the Average Power Range Monitor (APRM) Scram and Rod Block and Rod Block Monitor Trip Setpoints and Allowable Values to those applicable for single loop recirculation loop operation per Specifications 2.2.1, 3.2.2, and 3.3.6.

At least once per 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months s:

f' 2.

a) Verify that the APRM flux noise averaged over 30 minutes does not exceed 5% peak to peak; otherwise, reduce the recirculation loop flow until the APRM flux noise is less than the 5% peak to peak limit, and, b) Verify that the core plate .iP noise does not exceed 1 psi peak to peak; otherwise, reduce the recirculation loop flow until the t.P noise is less than the 1 psi limit.

f' Os "See Special Test Exception 3.10.4.

LASALLE-UNIT /1 3/4 a-1

-_ -.,--.___,..7-m--oy--,,-, - , , g g. - -

I i - REACTOR COOLANT SYSTEN LINITING CONDITION FOR OPERATION (Continued) ll ACTION: (Continued)

,. 3. The provisions of Specification 3.0.4 are not applicable.

4. Otherwise, be in at least HOT 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 locps in operation, immediately initiate measures to place the unit in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

l 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 unit, and

.O*

Verifying that the averag'd rite of control valve movement is:

b.

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.

e I

LA SALLE - UNITg3 3/44-fla.

i REACTOR COOLANT SYSTEM JET PUMPS LIMITING CONDITION FOR OPERATION

3.4.1.2 All jet pumps shall be OPERABLE.

n _ -APPLICA8ILITY: OPERATIONAL CONDITIONS 1 and 2.

ACTION:

With one or more jet pumps inoperable, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

l l

SURV51LLANCE REQUIREMENTS 4.4.1.2 Each of the above requ ed jet pumps shall be demonstrated OPERABLE prior to THERMAL POWER exceeding $5% of RATED THERMAL POWER and at least once e

per 24 nours by measurin, and rec 6rdin, each of the below specified parameters v

.(O and verifying that no two of the fbilowing conditions occur when the recircula-tion loops are operating at the s flow control valve position.

a. The indicated recirculation loop flow differs by more than 10% from the established flow control valve position-loop flow characterisi.ics.

l ~

l j b. The indicated total core flow differs by more than 10% from the established total core flow vat e derived from either the:

jpserD goJ9 1. Established THERMAL POWER-c ,re flow relationship, cr N 2. Established core plate differ ntial pressure-core flow relationship.

! c. The indicated diffuser-to-lower plent- differential pressure of any individual jet pump differs from estab ished patterns by more than 10%.

LA SALLE - UNIT 1 3/4 4-2

r 4.4.1.2.1 Each of th3 above requircd jet pumps shall be demonstrated OPERABLE prior to THERMAL POWER exceeding 25% of RATED THERMAL POWER and at least ence per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by measuring and recording each of the below specified parameters and verifying that no two of the following conditions occur when both recircula-tion loops are operating at the same flow control valve position.

'I a. The indicated recirculation loop flow differs by more than 10% from the established flow control valve position-loop flow characteristics j for two recirculation loop operation.

fI b. The indicated total core flow differs by more than 10%.from the established total core ficIw value derived from either the:

1. Established THERMAL POWER-core flow relationship, or

2. Established core plate differential pressure-core flow relationship for two recirculation loop operation.

c. The indicated diffuser-to-lower plenum differential pressure of any individual jet pump differs from established two recirculation loop operation patterns by more than 10%.

, 4.4.1.2.2 During single recirculation loop operation, each of the above required jet pumps shall be demonstrated OPERABLE at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying that no two of the following conditions occur:

a. The indicated recirculation loop flow in the operating loop differs by more that 10% from the established single recirculation 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 from single recirculation loop O flow measurements.

c. The indicated to-lower plenum differential pressure of any individual jet pump differs-from established single recirculation loop y more than 10%. -

optabak f b5 2 A;5er.T Foe PA6a 3/44-9.

l t

0

=-- - . - - - - - , _ x. .

[. , . . . . . .. .

\

9 I . .

REACTOR COOLANT SYSTEM T

1,

. vp . -

I RECIRCULATION LOOP F1.0W .

.s s

LIMITING CONDITION FOR OPERATION

! 3.4.1.3 Recirculation loop flow mismatdh shall be asintained within:

I' a. 55 of rated recirculation flow with core flow greater than or equal to 705 of rated core flow.

! b. 105 of rated recirculation flow with core flow less than 70% of I

rated core flow.

, APPLICA8ILITY: OPERATIONAL CONDITIONS 1 and 2. ~Jurean he r ceirculahott loop eye rden. . \

e ACTION: . . .

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

a. Restore the recirculation loop flows to within the specified limit ~

, within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , or

b. Declare the recirculation loop with the lower flow not in operation and take the ACTION require by Specification 3.4.1.1.

, SURVEILLANCE REQUIREMENTS i

i 1

4.4.1.3 Recirculation loop flows shall be verified to be within the limits at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

l 1

l b .

I

' A SALLE - UNIT 1

. 3/4 4-3 ,

Oh

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.,S REACTOR COOLANT SYSTEM

}

3/4.4.2 sac' TTY / RELIEF VALVES .

4 l Tiuunu CONDITION FOR OPERATION

. 3.4.2 The safety valve function of eigtjteen reactor coolant system safety /

l I relief valves s l

lift settings.pil be OPERA 8LE with the specified code safety valve function

a. 4 safe lief valves 9 1205 psig i N

b. 4 safe lief valves 91195 psig

3

c. 4 safe lief valves 91185 psig $ N

d. 4 safetyGjeliefvalves91175psigtM

. a. 2 safetyWrelief valves 91146 psig 3 3 APPLICA8ILITY: 0 TIONAL CONDITIONS 1, 2 and 3.

- ~

ACTION:

a. With the safety valve fun:: tion of one or more of the above muired l safety / relief valves inoperable, be in at least HOT SHUTDOWN within i 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

I

b. With one or more safety / relief valves stuck open, provided that suppression pool average water teeperature is less than 110*F, close the stuck open relief valve (s); if unable to close the open valve (s)

~

I ., j within 2 minutes or if suppression pool average water temperature is 110*F or greater, place the reactor mode switch in the Shutdown i

position.

, c. With one or more safety / relief valve stem position indicatorp i inoperable, restore the inoperabic stes position indicatioW to l OPERA 8LE status within 7 days or be in at least HOT SHUTDOWN within

[ the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

i

! SURVEILLANCE REQUIREMENTS l

! 4.4.2.1 The safety / relief valve stem position indicators of each safety / relief j valve shall be demonstrated OPERABLE by performance of a:

I j 4 a.

b.

CHANNEL CHECK at least once per 31 days, and a CHANNEL CALIBRATION at least once per 18 months.** ,

4.4.2.2 The low low set function shall be demonstrated not to interfere with the OPERA 8ILITY of the safety relief valves or the A05 by performance of a CHANNEL CALIBRATION at ,least once per 18 months I

"The lift setting pressure shall correspond to ambient conditions of the valves at nominal operating temoeratures ano-pressures.

Ol /', ; ,,

<ua == t*e < c= r>aie veives =>v ee =eoiecee ita aere aea^a's veives ita lower setpoints until the next' refueling outage.

l ~ **The provisions of Specification 4.0.4 are not applicable provided the surveil-

! lance is performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure is adequate to perform the test.

LA SALLE - UNIT 1 3/4 4-5 .

.- , - - - - - - . . , - - - - - , - - , , - , . - - - - - , ,,-,,wmc-.-. -...-c., , , , , -,.- _.

? REACTOR COOLANT SYSTEM

.7,

(- .. OPERATIONAL LEAKAGE

~

I

' ;.q - 'p LIMITING CONDITION FOR OPERATION f '![- 3.4.3.2 Reactor coolant system leakage shall be limited to:

g.

9~' - -

a. No PRESSURE BOUNDARY LEAKAGE.

b. 5 gpa UNIDENTIFIED LEAKAGE.

c. 25 gpa total leakage averaged over any 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period. gg

- d. 1 gpa leakage at a reactor coolant system pressure at 1000 1 ,16 psig \

from any reactor coolant system pressure isolation valve specified j in Table 3.4.3.2-1.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3.

ACTION:

a. With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

(

b. With any reactor coolant system leakage greater than the limits in b and/or c. above, reduce the leakage rate to within the limits within [

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and

- in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . ,

c. With any reactor coolant sys' tem pressure isolation valve leakage greater than the above limit, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by s,_

use of at least two closed valves, or be in at least HOT SHUTDOWN 1

I within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . -

d. With one or more high/ low pressure interface valve leakage pressure' monitors inoperable, restore the inoperable monitor (s) to OPERABLE status within 7 days or verify the pressure to be less than the alarm setpoint at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by local indication; restore the inoperable monitor (s) to OPERABLE status within 30 days l or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 12 hcurs.

l

. SURVEILLANCE REQUIREMENTS 4.4.3.2.1 Tne reactor coolant system leakage shall be demonstrated to be within each of the above limits by:

l a. Monitoring the primary containment atmospheric particulate and

. gaseous radioactivity at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ,

b. Monitoring the primary containment sump flow rate at least once per I 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , and

c. Monitoring the primary containraent air coolers condensate flow rate at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

LA SALLE - UNIT 1 3/4 4-7 i

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= : x ._:--

t i [ REACTOR COOLANT SYSTEM

. N .

._/ 3/4.4.5 SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION

~

~ ', . ! 1.4.5 The specific activity of the primary coolant shall be limited to:

a. Less than or equal to 0.2 microcurip per gram DOSE EQUIVALENT I

- I-131,and l

l b. Less than or equal to 1004 sierocuries per gram.

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

ACTION:

a. In OPERATIONAL CONDITIONS 1, 2 or 3 with the specific activity of the primary coolant;

1. Greater than 0.2 sierocuries per gram 1)CSE EQtlIVALENT I-131 but less than or equal to 4.0 microcuries per gram, operation say continue for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months provided that the cumulative opsrat-ing time under these circumstances does not exceed 800 Murs in any consecutive 12 month period. With the total cumulative operating time at a primary coolant specific activity greater than or equal to 0.2 aicrocuripf per gram DOSE EQUIVALENT I-131 I-O~ .~ , exceeding 500 hours0.00579 days 0.139 hours 8.267196e-4 weeks 1.9025e-4 months in any consecutive six month period, prepare and submit a special report to the Commission pursuant to l

Specification 6.6.C within 30 days indicating the number of i hours of operation above this limit. The provisions of Specification 3.0.4 are not applicable.

2. Greater than 0.2 microcuries per gras DOSE EQUIVALENT I-131 for g

more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months during one continuous time interval or for more than 800 hours0.00926 days 0.222 hours 0.00132 weeks 3.044e-4 months cumulative c>erating time in a consecutive

! 12-month period, or greater than or equal to 4A microcuries per I i gram, be in at least HOT SHUTDOWN with the main steam ifne i isolation valves closed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

I 3. Greater than 100d microcuries per ! ras, be in at least HOT SHUTDOWN with the sain steaaline is 1ation valves closed within j

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

l j b. In OPERATIONAL CONDITIONS 1, 2, 3 or 4, with the specific activity of the primary coolant greater than 0.2 microcurief per gram COSE l EQUIVALENT I-131 or greater than 100 I! microcuries per gram, perform the sampling and nalysis requirements of Item 4a of Table 4.4.5-1 until the specific activity of the primary coolant is restored to within the Ifmit. A REPORTABLE OCCURRENCE shall be prepared and l

) i submitted to the Commission pursuant to Specification 6.6.8. This I

i report shall contain the results of the specific activity analyses

' and the time duration when tha scecific activity cf the cectant exceeded 0.2 aicrocurief ::er gram 00SE EQUIVALslT I-131 tege .'.ar )

with the foll: wing adoitional infor=atica.

QO ,

! LA SALLE - UNIT I. 3/4 4-13

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5

, ;},

f _

REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION (Continued)

ACTION (Continued) .

c. In 0FERATIONAL CONDITION 1 or ,2, with:

1. THERMAL POWER changed by more than 15% of RATED THERMAL POWER in oji hour *, or l l!

2. The off gas level, prior to the holdg line, increased by more than 25,000 microcuries per second in one hour during steady state operation at relears rates less than 100,000 microcuries per second, or

3. The off gas level, prior to the holdup line, increased by more than 15% ini p w hour during steady state operation at relaise I ,

rates greater than 100,000 microcuries per second, I perform the sampling and analysis requirements of~Itas 4h of Table 4.4.5-1 until the specific activity of the primary coolant is

' restored to within its limit. Prepare and submit to the Commission a Special Report pursuant to Soecification 6.6.C at least once per i . 92 days containing the results of the specific activity analysis together with the below additional information for each occurrence.

] ~ '

Additional Information

1. Reactor power history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to:

The first sample in which the limit was exceeded, and/or I a) b) The THERMAL POWER or off gas level change.

! 2. Fuel burnup by core region. .

l ! 3. Clean-up flow history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to:

l a) The first sample in which the Itait was exceeded, and/or b) The THERMAL POWER or off gas le, vel change.

l i 4. Off gas level starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to:

I c) The first sample in which tne limit was exceeded, and/or b) The THERMAL POWER or off gas level change.

i,,

', SURVEILLANCE REQUIREMENTS

l 4.4.5 The specific activity of the reactor coolant shal: be demonstrated to l be within the limits ey cerformanca of the sarpling and a.talysis progrr: of Table 4.4.5-1.

,wl -

t "Not applicable during the Startup Test Program.

l Li SALLE - UNIT 1 3/4 4-14 y, -.__.,_ , _. _ _ _ . . . _ . . - . _ . _ - _ , _ _ _ . _ . _ _ _ , _ . , , . , , _ _ _ . . _ _ . . . - - , _ - , , _ . . . _ _ _ _ _ , , , _ _ _ _ , _ _ , _ , . _ , . _ , . , , , , , _

. .-e~

s I

REACTOR COOLANT SYSTDI ,

. SURWILLMICE REQUIREMENTS (Continued) l 4.4.6.1.2 The reactor coolant system temperature and pressure shall be determined to be to the right of the criticality limit line of Figure 3.4.6.1-1

, curves C and C' within 15 minutes orier to the withdrawal of control rods to kring the reactor ta critica11ty.

j

'i 4.4.6.1.3 The reactor vessel material specimens shall be removed and examined

to determine reactor pressure vessel fluence as a function of time and THERMAL POWER as reauind tw 10 CFR,50, Appendix H in act;prdance with the schedule in l dable 4.4.6.1.3-L The results of these fluence determinations shall be used

., Part to update the curves of Figuref 3.4.6.1-1r,0.4.0.1-2 -4 0.0.0.1 3. l 4.4.6.1.4 The reactor vessel flange and head flange temperature shall be ~

verified to be greater ttan or equal to 80*F:

a. In OPERATIONAL CONDITION 4 when the reactor coolant temperature is:

1. i 100*F at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . .

2. S 85*F at least once per 30 minutes.

O b. Within 30 minutes prior to ar.d at least once per 30 minutes during h/ tensioning of the reactor vessel head bolting s*.uds.

(

i 1 .

I 1

l i

i

! 9

)

Q. l

! LA SALLE - UNIT 1 3/4 4-17

.. l

(- '..

9 Table 4.4.6.1.3-1 k Reactor Vessel Material Surveillance Program Withdrawal Schedule e

E y Specimen holder Vessel location Lead factor Withdrawal time (ScT"*ce Y1:=) (EFr'TCTWE PULL-FbwnR H'ARS.)

117C4936G010 300* 0.6 ,.le- b 117C4936G011 120* 0.6 34V/S i 117C4936G012 30* 0.6

Spare

] Neutron Dosimeter 30* 1st Refuel Outage I

R

=

G k

d N

i a

n LD r

g eF%e e O m

.N___ -

, *l

~

i REACTOR COOLANT SYSTEM -

'l x ,/ ., 3/4.4.9 RESIDUAL HEAT REMOVAL -

i

, IqiSHUTDOWN t .

LIMITING CONDITION FOR OPERA." ION ih 3.4.9.1 Two# shutdown cooling mode loops of the residual heat removal (RHR) systas shall be OPERA 8LE and at least one shutdown cooling mode loop shall be inoperation*N with each loop consisting of at least: I

' i i

a. One OPERA 8LE RHR pump, and

b. One OPERA 8LE RHR heat exchanger.

1 APplICA8ILITY: OPERATIONAL CONDITION 3, with reactor vessel pressure less than the RHR cut-in permissive setpoint. -

I ACTICN: ,,

a. With less than the above required RHR shutdown cooling mode loops OPERABLE, immediately OPERA 8LE status initiateascorrective soon as possible. action toWithin,pa retur@nhour the required loops and at least onceto l per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter, demonstrate the operability of at least one alternate method capable of decay heat removal for each inoperable RNR s shutdown cooling mode loop. Se in at least COLD SHUTDOWN within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .**

l

' b. With no RHR shutdown cooling mode loop in operation, immediately initiate corrective action to return at least one loop to operation as soon as r possible. Within hour establish reactor coolant circulation by an l alternate methoa and monitor reactor coolant temperature and pressure at least once per br,ur. ~

! SURVEILLANCE REQUIREMENTS -

i a

4.4.9.1 At least one shutdown cooling mode loop of the residual heat removal i system or alternata method shall be determined to be in operation and circulating

+

reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

l i #0ne RHR shutdown cooling mode loop may be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing provided the other loop is OPERABLE and in operation.

The shutdown cooling pump may be removed from operation for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months

per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period provided the other loop is OPERABLE.

l NThe RHR shutdown cooling mode loop may be removed from operation during

} hydrostatic testing.

< 't

} **Whenever tw or more RHR subsystems are inocerable, if unable to attain COLD 1 SMUTCC'nN as required by this AC'"I':i, t.aintain react::r ceciant tamperaturs as

! ,. , low as practical by use of alternata heat removal methocs.

.' ) e d

LA SALLE - UNIT 1 3/4 4-23 I

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_. , . . , . ,, . ,,,g.c.,7,.,..uu,,,

REACTOR COOLANT SYSTEM

.i , .

COLD SHUTDOWN . .

o ? -

l! LIMITING CONDITION FOR OPERATION

'i ,

-, -3.4.9.2 Two 0shutdown cooling mode loops of the resiaual heat removal (RHR) system shs11 be OPERA 8LE* and at least one shutdown cooling mode loop shall be I in operation

with each loop consisting of at least: I

a. One OPERAELE RHR pump, and

b. One OPERA 8LE RHR heat exchanger.

APPLICABILITY: OPERATIONAL CONDITION 4.

i- .

ACTION:

a. With less than the above required RHR shutdown cooling mode loops OPERABLE, within hour and at least once per*24 hours thereafter, demonstrate l the operability of at least one alternate method capable of decay heat removal for each inoperable RHR shutdown cooling mode loop. -

.1.

, b. With no RHR shutdown cooling mode loop in operation. within and hour i 1 - estabitsh reactor coolant circulation by an alternate method and monitor

' reactor coolant temperature and pressure at least once per hour.

s SURVEILLANCE REQUIREMENTS t

4.4.9.2 Atleastoneshutdowncoolingmodeloopoftheresidualheatremoval

! system or alternate method shall be determined to be in operation and i circulating reactor coolant at least orce per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

I i

~

! y

One RHR shutdown cooling mode loop may be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for

! surveillance testing provided the other loop is OPERABLE and in operation.

The normal or emergency power source may be inoperable.

l l

**The shutdown cooling pump may be removed from operation for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period provided the other loop is OPERABLE.

l

MThe shutdcun eco11:1g mode loop :nay be removed from operation during hydrostatic tasti.g.

/

}

O. 4 l

l .

LA SALLE - UNIT 1 3/4 4-24

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~

1,-

EMRGENCY CORE COOLING SYSTEMS

~

/- .

LDf! TING CONDITION FOR OpfRATION (Continued)

, ACTION: (Continued)

. , 4 d. For ECC5 divisions 1 and 2, provided that ECCS division 3 is OPER48G:

l*l l ;

1. With LPCI subsystem "A" and either LPCI subsystem "B" or "C" inoperable, restore at least the inoperable LPCI subsystem "A" l, e or inoperable LPCI subsystem "8" or "C" to OPERA 83 status within j 72 nours.

! 2. With the LPCS system inoperable and either LPCI subsystems "B" or i "C" inoperable, restore at least the inoperable LPCS system or inoperable LPCI subsystsa "B" or "C" to OPERA 8M status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

i 3. .. Ottigmise. be in at lesst HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months l1 and in COLIT 5HUTDOWN within thurfollowing 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months *. .

I I

, ; e. For ECC5 divisions 1 and 2, provided that ECCS division 3 is

OPERA 8E and divisions 1 and 2 are othemise OPERA 8M

l ,

L With one of the above required A05 valves inoperable, restore t ; the inoperable ADS valve to OPERA 8M status within 14 days or -

l i be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and reduce reactor steam done pressure to 1122 psig within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

2. With two or more of the above required A05 valves inoperable, l,

be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and reduce reactor steam done pressure to 3122 psig within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

I f. With an ECCS discharge line " keep filled" pressure alarm instrumenta-tion channel inoperable, perform Surveillance Requirement 4.5.1.a.1 at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

l g. With an ECCS header delta P instrumentation channel inoperable, restore the inoperable channel to OPERA 8LE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months

.I or determine ECCS header delta P locally at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ;

_j otherwise, declare the associated ECCS inoperable.

h. With Surveillance Requirement.4.5.1.d.2 not performed at the required l interval due to low reactor steam pressure, the provisions of Specifi-cation 4.0.4 are not appifcable provided the surveiliance is performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure is adequate to perform the test. .

! f. In the event an ECCS system is actuated and injects water into the Reactor Coolant System, a Special Peport shall be prepared and

{ submitted to the Commission pursuant to Specification 6.6.C within 90 days describing the ci cumstances of the actuation and the total

accumulated actuation cycles to date. The current value of the 1, .

,, usage factor for each affected safety infection nozzle shall be

provided in tais Special Re;or: mever its value excee s 0.70. g,,.c T %nanever :wo or core TsHR subsystems are inoperable, if ur.aoie to attain COLD N

>J SHUTDOWN as required by this ACTION, maintain re ctor coolant temperature as

. low as practical by use of alternate heat removal methods. ]6 I LA SAL M - UNIT 1 3/4 5-3

?

. . . - . _ , , - . . , . , , , . - - - , .m .--,..m.__ . . , _ . . _ _ . , . . _ , _ _ , ~ ~ . , , _

IAJSetJ Fo P PA6-E 3/4 g.3 O

, j W ith one or more ECCS corner room watartight doors inoperable,'resto h all the inoperable ECCS corner room watartight doors to OPERABLE '"* '!

-- status within 14 days, otherwise, be in at least HOT SHUTDOWN'within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the followihg 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .'

O I

O

4 .

... , : n. 62 e. w:.p.: . . . .e. : .. :. . ... .

. . . .~ . . % . . - ..

a

- I

b. , DERGENCY CORE COOLING SYSTEMS

'i -

SURVEILLANCE REQUIREMENTS t ..

! . 4.5.1 ECCS divisions 1, 2 and 3 shall be demonstrated OPERA 8LE by:

i

- P a. At least once per 31 days for the LPC5, LPCI and HPCS systems: -

I Verifying by venting at the high poirt vents that the system j 1.

piping from the pump discharge valve to the system isolation i; valve is filled with water.

2. Performance of a CHANNEL FUNCTIONAL. TEST of the:

a) Discharge line " keep filled" pressure alars instrumentation, i -

and . .

j b) Header delta P instrumentation.

3. Verifying that each valve, manual, power operated or automatic, in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position. -

b. Verifying that, when tested pursuant to Specification 4.0.5, each:

i- 1. '.PCS pump develops a flow of at least 6350 gpm against a test line pressure greater than or equal to 290 psig. .

l ,

2. LPCI oump develops a flow of at least 7200 gpa against a tast line pressure greater than or equal to 130 psig.

3. HPCS pump develops a flow of at least 6250 gpa against a tact line pressure greater than or equal to 370 psig.

c. For the LPC5, LPCI and HPCS systems, at itast onca per 18 months:

1. Performing a system functional test wnich includes simulated eutomatic actuation of the system throughout its emergency operating sequence and verifying that each automatic valve in l the flow path actuates to its correct position, Actual injection of coolant into the reactor vessel may ba excluded from this test. #

l

4. Verifying that each ECCS corner room watertight door is closed, except during entry to and exit from tne room.

t L J

i LA SALLE - UNIT 1 3/4 5-4

+ __ _.

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umr

- - . ___;2 - - - -

4

, i.

Os i==E*Cv CORE C=u= svs-s - -

3.. -

SURVEILLANCE REQUIREMENT 5 (Continued)

! 4

' ~

2. Performing a CHANNEL CALIBRATION of the:

a) Discharge line " keep filled" pressure alars instrinentation and verifying the:

1) High pressure setpoint and the low pressure setpoint of the:

(a) LPCS system to be 1 500 psig and 1 55 psig, respectively.

(b) LPCI subsystems to be 1 400 psig and 1 55 psig, respectively. -

2) Low pressure setpoint of the HPCS system to be 1 63 l

psig.

b) Header detta P instrumentation and verifying the setpoint .

of the: ,

1) LPCS system and LPCI subsystems to be's 1 psid.

' ~

! 2) HPCS system to be 5 s.f,4 psid greater than the l normal indicated AP.

3. Verifying that the suction for the HPCS system is automatically transferred from the condensate storage tank to the suppression chamber on a condensata storage tank low water level s,ignal and i en a suppression chamber high water level signal.

d. For the A05 by: #

At least once per 31 cays, performing a CHANNEL FUNCTIONAL TEST

. 1.

l ,

of the accumulator backup compressed gas systes low pressure i alarm system.

I

2. At least once per 18 months: ,

i

a) Performing a system functional test which includes simulated l i automatic actuation of the system thn.ughout its emergency i operating sequence, but excluding actual valve actuation.

i j b) Manually opening each ADS valve and observing the expected 1

change in the indicated valve position.

i ~ ' c) Perforaing a 0!AN!!EL CALIBRATICS of the ac umulator ::ackuo

.- compressea gas system low pressure alarm system and verifying an alarm setpoint of dOO + 40, - O psig on decreasing pressure.

t

LA SALLE - UNIT 1 3/4 5-5 ,

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TNSeCT Fo2 FAs.s 3A 5-g

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4. Visually inspecting the ECCS corner room watertight door seals

j and room penetration seals and verifying no abnormal degradation, i4 damage, or obstructions. ,

1 9

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anr.m em 1 CORE COOLING SYSTEMS ,

i'N -

a

,' 3/4.5.3 SUPPRESSION CHAMBER" Lin m nG CONDITION FOR OPERATION

. s 3.5.3 The suppression chamber shall be 4PERA8LE:

a. In OPERATIONAL CONDITION 1, 2 or 3 with a contained w at least 128,800 ft*,equivalenttoalevelof2F21:$atervolumeofi9. l m

! + 4+

b. In OPERATIONAL CONDITION 4 or 5* with a contained water volume of at

. least 70,000 fta, equivalent to a leval of 14foip except that the j suppression chamber level may be less than the limit or may be i

drained in OPERATIONAL CONDITION 4 or 5* provided that:

L No operations are perfonned that have a potential for draining -

the reactor vessel,

. 2. The reactor mode switch is locked in the Shutdown or Refuel l

position,

3. The condensate storage tank contains at least 135,000 available .

I i gallons of water, equivalent to a level of 14.5 feet, and O -' 4. The a,CS system is 0, ERA 8tE per Specification 3.5.2 with an OPERABLE flow path capable of taking suction from the condensate storage tank and transferring the water through the spray sparger to the reactor vessel.

APPLICA8ItITY: OPERATIONAL CONDITIONS 1, 2, 3, 4, and 58 f f '

I ACTION:

a. In OPERATIONAL CONDITION 1, 2, or 3 with the suppression chamber water level less than the above limit, restore the water level to within the Ifmit within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLO SHUT 00WN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

j b. In OPERATIONAL CONDITION 4 or 58 with the suppression chamber water i level less than the above limit or drtined and the above required I conditions not satisfied, suspend CORE ALTERATIONS and all operations t

- that have a potential for draining the reactor vessel and lock the I reactor mode switch in the Shutdown position. Establish SECONDARY CONTAINMENT INTEGRITY within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .,

l t

i #5ee Specification 3.6.2.1 for pressure suppression requirements.

! *The suceression chamber is not recuired to be OPERABLE orovided that the

reactor vessel head is removed, na cavity is floed a er :eing f'occed t

fr:7 the sucaression pool, the s:ent fuel pool gatas ar? removed wnen the cavity 11 flooded, and the water level is maintained witnin the limits of O.' 1

' . Specifications 3.9.8 and 3.9.9.

I LA SALLE - UNIT 1 3/4 5-8 .

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O, BERGENCY CDRE COOLING SYSTEMS

~T LIMITING COMITING FOR OPERATION (Continued)

).

i, ACT15N (Continued)

I

c. With' one suppression chamber water level instrumentation channel inoperable, restore the inoperable channel to OPERA 8LE status within i ; 7 days or verify the suppression chamber wster level to be gr4ater j then or equal to r 14"0L as applicable, at least once per i

+ 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by local i ication

! . d. With bef.h suppression chamber water level instrumentation channels

' inoperable, restore at least one inoperable channel to OPERA 8LE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or be in at least HOT SHUTDOWN within the'next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and verffy the suppression chamber water level to be greater than or equal to or14j0*w as applicable, at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months j i

by Tocal i cation I

8 . -

l SURVEILLANCE REQUIREMENTS

' 4.5.3.1 The suppression chamber shall be determined OPERA 8LE by verifying: -

.., s.- The water level to be greater than or equal to, as applicable:

~'

1. 28,'2(,at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . l

, )

~' j 2. 14fo{at 1, east once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . l

- ~

t

b. Two suppression chamber water level instrumentation channels OPERABLE l

by performance of a:

i i

! 1. CHANNEL CHECK at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , e

i ' 2. CHANNEL FUNCTIONAL TEST at least once per 31 days, and i CHANNEL CALIBRATION at least once per 18 months, j 3.

I with the low water level alars setpoint at greater than or equal to 26#4 j" "* l 4.5.3.2 With the suppression chamber level less than the r_bove limit or drained in OPERATIONAL CONDITION 4 or 58, at least once per 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months s:

l *

a. Verify the required conditions of Specification 3.5.3.b to be satisfied, or l !

i : ,

b. Verify footnete conditions

to be satisffed. -

c' :' : t::f-: :: :: :- ' ;d t " ; :t--~r :::: : 7 :-

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et c71 tich a HkkeA t

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zaiseef Fo

PAGE J/p 5- 9 0 .

l

'! ,ThesuppressionchamberisnotrequiredtobeOPERA8LEpYoiidedEattheraa'ctor

, vessel head is removed, the cavity is flooded or being f1ooded from the suppres- ' '

sion pool,' the spent fuel pool gates are removed when the cavity is floo'ded, and

' the water level is maintained within the limits of Speciff' cations 3'.9'.8 and 3.9.9. -

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. CONTApmENTSYSTD45 G - - - - -

./ PRIMARY CONTAINMENT LEAKAGE - . - - - - - - ---

I LIMITING CON 0!TTON FOR OPERATION

3. 6.1. 2 Primary containment leaks.ge rates shall be Ifrited to:

, a. An overall integrated leakage rate of less,than or equal to L,,

0.635 percent by weight of the containment air per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at P,,

39.6 psig.

l

b. A combined leakage rate of less than or equal to 0.60 L, for all I penetratisns and all valves listed in Table 3.6.3-1, except for main steam isolation valves and valves which are hydrostatically leak .

tested per Table 3.6.3-1, subject.to Type B and C tests inhen pressurized to P,, 39.6 psig. _

! c. *Less than or 1toNscfperhourfor E main steam lives 4 keg k'Y

, isolation valve whest tested at 25.0 psig. I

^

d. A combined leakage rate of less than or equal to 1 gpa times the

, total number of ECCS and RCIC containment isolation valves in hydro-statically tested lines which penetrate the primary containment, when tested at 1.10 P,, 43.6 psig.

APPLICABILITY: When PRIMARY CONTAIPMENT INTEGRITY is required per -

i Specification 3.6.1.1.

ACTION:

.. With: .

t

, a. The measured overall integrated primary containment leakage rate j exceeding 0.75 L,, or i b. The seasured combined leakage rate for all penetrations and all

valves listed in Table 3.6.3-1, except for main steam isolation l valves and valves which are hydrostatically leak tested per Table 3.6.3-1, subject to Type 8 and C tests exceeding 0.60 L , or i The seasured leakage rate exceeding /4d15 scf per hour .ffor -=j r: eoc p(.9 sfe all

. c. ,

j pues isolation valva) or l d. The measured combined leakage rate for all ECCS and RCIC containment isolation valves in hydrostatically tested lines which penetrate the

+

primary containment exceeding i gpm times the total number of such valves, ~

O i

" Exemption to Appendix "J' of 10 CFR 50.

l LA SALLE - UNIT 1 3/4 6-2

! i O- CONTAI194ENT SYSTEMS -

Leti mu CONDITION FOR' OPERATION (ContinuedF ACTION (Continued)

'{ ,

l restore: r

s. The overall integrated leakage rate (s) to less than or equal to 0.75 l , and ,

t a ,

'l b. The combined leakage rate for all penetrations and all valves listed in Table 3.6.3-1, except for main steam isolation valves and valves l which are hydrostatically leak tested per Table 3.6.3-1, subject to i

? Type B and C tests to less than or equal to 0.60 L4, and

! /oe

c. leakage rate to less than or equal to-25-sef per hour forany. all feoc ga.Wthem L ;

< lives b _x : .1 --isolationvalve?and '

i

! ; d. The combined leakage rate for all ECCS and RCIC containment isolation

! - valves in hydrostatically tested lines which penetrate the primary

! containment to less than or equal to 1 gpa times the total number of -

such valves, ,

prior to increasing reactor coolant system temperature above 200*F.

SURVEILLANCE REQUIREMENTS __

I 4.6.1.2 The primary containment leakage rates shall be demonstrated at the following test schedule and shall be determined in confomance with the criteria s

f specified in Appendix J of 10 CFg,"50 using the methods and provisions of ANSI

y. N45.4[-[1972: ,

j a. Three Type A Overall Integrated Containment Leakage Rate tests shall j be conducted at 40110 month intervals during shutdown at P,, _

! 39.6 psig, during each 10 year service period. The third test of each set shall be conducted during the shutdown for the 10 year plant inservice inspection.

i i b. If any periodic Type A test fails to meeto.75 L,, the test schedule I for subsequent Type A tests shall be reviewed and approved by the

! Commission. If two consecutive Type A tests fail to.seeto.75 L3 , a l Type A test shall be performed at least every 18 sonths until two l' consecutive Type A tests mesto.75 L,, at which time the above test I i schedule may be resumed.

I c. The accuracy of each Type A test shall be veaified by a supplemental '

test which:

ll 1. Confirms the accuracy of the test by verifying that the difference between the supplemental data and the Type A test data is within 0.25 L,.

2. Has duration sufficient to establish accurately the change in 1eakage rate between the Type A test and the supplemental test.

l

} 3. Requires the cuantity of gas injected :., : -he contain :ent or bled frc::: the contain int during the s;.:;;iecental tast to ce

.s equivalent to at feast 25.gsesent of the total measured leakage l at P,, 39.6 psig. %

l LA SALLE - UNIT 1 3/4 6-3 I *

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.[ PRIMARY CONTAINMENT AIR LOCKS ll '

,- LINITIN@ CONDITION FOR OPERATION

) .

I l 3.6.1.3 Each primary containment air lock shall be OPERA 8LE with:

l !

I a. Both doors closed except when the air lock is being used for normal transit entry and exit through the containment, then at least one air lock door shall be closed, and l

+ b. An overall air lock leakage rata of less than or equal to 0.05 L, at P,, 39.6 psig.

i APPLICA8ILITY: OPERATIONAL CONDITIONS 1, 2*,and 3. l i

l ACTION:

~

a. With one primary containment air lock door inoperable:

1. Maintain at least the OPERABLE air lock door closed and either .

- [] . .

restore the inoperable air lock door to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or lock the OPERA 8LE air lock door closed.

l S i V 2. Operation may then continue until performance of the next required l

overall air lock leakage test provided that the OPERA 8LE air lock j door is verified to be locked closed at least once per 31 days. ,-

. 3. Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

3

4. The provisions of Specification 3.0.4 are not appitcable.

1

'; b. With the primary containment air lock inoperable, except as a result :2f I i an inoperable air lock door, maintain at least one air lock door closed; l 1 restore the inoperable air lock to 0PERA8LE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT SHUTDOWN within the next .12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

"See Special Test Exception 3.10.1.

l 1 p.

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LA SALLE - UNIT 1 3/4 6-5

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CONTAINMENT 5hsem

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PRIMARY CONTAINMENT STRUCTURAL INTEGRITY e .

~

LIMITING CONDITION POR OPERATION i

t 3.6.1.5 The structure 1 integrity of the primary containment shall be maintained

at a level consistent with the acceptance criteria in Specification 4.6.1.5.

APPLICA8ILITY: OPERATIONAL CONDITIONS 1, 2;and 3.

t ACTION:

. With the structural tagrity of the primary containment not conform-

ing to the above requ nts in that tested tendon lift-off force

! of individual tendons es below the predicted lower limit but is greater than the design tait, perfom an engineering eguation of -:

the primary containment demonstrate its structural integrity within

.: 15 days; if the measured ft-off force of a tendon (s) is less than the design limit, perform engineering evaluation of the primary containment to demonstrate s structural integrity within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; otherwise, be in at least HQ 5HUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and .

I in COLD SHUTDOWN within the f lowing 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

k.\ notWith the structural integrity o the primary containment othe mise h,.s \

conforming to the above requ evaluation of the primary contai nt to demonstrate its structural nts, perform an engineering i -

integrity within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; othemi , be ir. at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in CD SHUTDOWN within the following

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

a. With more than one tendon with an observed lift-off force between the predicted lower limit and 90% of the predicted lower limit or with 2 one tendon below 90% of the predicted lower limit, restore the tendon (s) to the required level of integrity within 15 days and j perfors an engineering evaluation of the containment and provide a

! Special Report to the Commission within 30 days in accordance with 2

Specificatfon 6.6C. or be in at itast HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

I b. With any other abnormal degradation of the structural integrity at j a level below the acceptance criteria of Specification 4.6.1.5, restore the containment vessel to the required hvel of integrity

! t within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and perform an engineering evaluation of the contain-ment and provide a Special. Report to the Commission within 15 days e in accordance with Specification 6.6C. or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following l ,

30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

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t CONTAINMEfff SYSTEMS

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g SURVEILLANCE REQUIREMENTS 3 .

4.6.1.5 Primary Containment Tendons. The primary c neent st uctural integ-rity shall shall be demonstrated at the end of M .onta; theeda Mr years after

the initial structural integrity test (ISIT) and every years thereafter l in accordance with Table 4.6.1.5-1. The structural integrity shall be demon .

strated by: '

l

a. Determining that a representative s' ample of at least 13 tendons, 8 hori-zontal and 5 vertical, selected in accordance with Table 4.6.1.5-1 have a lift-off force 20* r - * ' r i-" :-d ri '-

t' egal de -

values listed in Table 4.6.1.5-2 l

' ,, ,ga. at the first year inspection. For subsequent inspections, for tendons g and periodicities per Table 4.6.1.5-1, th: r ' - - " -" y-- " ' t e "

f: ;;; :h-P bc M-- ::d by th: n t ':1 ';; t Q; '- Y t: f::: m

. Muum x2 ?:;; t g! <a- '::; t;;;; .; ;;d the minia' u n lift-off forces shall be decreased by the amount X2 log t for V tendons and Y2 log t for hoop tendons where t is the time interval in { years from initial tensioni{ng of the tendon to the current testino date and the values X1, X2, Y1 and Y2 are in accord- -

ance with the values listed in Table 4.6.1.5-2 for the surveillance tendon.

4 aaJ to is us. This test shall include essentially a coeplete detensioning of tendons i

I" l selected in accordance with Table 4.6.1.5-1 in which the tendon is deten-sioned to determine if any wires or strands are broken or damaged. Tendons in yeaes M found acceptable during this test shall be retensioned to their observed I

..idl 4cus;.etq lift-off force, 1 3%. During retensioning of these tendons, the change gpyc in load and elongation shall be measured simultaneously at a minimum of d

ggs i three, approximately equally spaced, levels of force between the seating If elongation corresponding to a specific load differs hs+ insf" , orcemore and than zero.5% from that recorded during installation of tendons, an gisspl4 to investigation should be made to ensure that such difference is not related wire failures or slip of wires in anchorages. If the lift-off force Qar* of any one tendon in the total sample population lies between the predicted lower limit and 90% of the predicted lower limit, two tendons, one on each i side of this tendon, shall be checked for their lift-off force. If both l 8 these adjacent tendons are found acceptable, the surveillance program may i proceed considering the.-single deficiency as unique and acceptable. The tendon (s) shall be restored to the required level of integrity. More than one tendon ;t;;; ;- below the predicted bounds out of the original sample }

population or the lift-off force of a selected tendon lying below 90% of l

I the prescribed lower limit is evidence of abnormal degradation of the con-tainment structure.

l I

b. Performing tendon detensioning and material tests and inspections of a previously stressed tendon wire or strand from one tendon of each group, hoop and V, and determining that over the entire length of the removed wire or strand that:

1. The tendon wires or strands are free of corrosion, cracks and damage.

t

2. t. minimum tensile strength value of 240 ksi, the guaranteed ultimate l, strengtn of the tendon materi.il, for at least tnree wire or strand

.. samoles. one t-om each end and one at mid-length, cut from mh I removed wire or strand. Failure of any one of the wire or strand samples to meet the minimum tensile strength test is evidence of O, abnormal degradation of the primary containment structure.

! LA SALLE - UNIT 1 3/4 6-9

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TABLE 4.6.1.5-1

. 4 m TENDON SUEVEILLANCE TEMON NUMERS Years After Ig g aj,.,5,tr g al 1 3 ,

5 10 15 Ig.)[g3.,, H V H V H V H V H V

. _ Visual Inspection 48AC 15C 48AC 15C 4aAC 15C 48AC 15C 48AC 15C of End Anchorages 56C8 15A 2C8 6C 38A 28A 48A 308 50C8 ISA Adjacent Concrete 12C8 204 14AC 17A 128A 23A 41C8 22A 538A 138

. Surface and Pre- 708 47C 248A 32C 21C8 58 50AC 57AC stress Monitor- 20C8 29A 37C5 42C 238A 31C ing Tests IC8 47C8 38C8 12AC

57C8 49AC 568A 608 688 21AC Detensioning and 20C8 47'C 2C8 42C 238A 31C 4BA 22A 50C8 19A Material Tests l

1.

O'.

r TENDON NUMERS Years After Initial Structural 20 25 30 35 40 Integrity Test Type of Inspection H V H V H V H V H V T[$*E' Inspection 48AC 15C 48AC 15C 48AC 15C 48AC 15C 48AC 15C I

, of End Anchorages 39C8 258 18A 38 48C8 78 49CB 25A 36CB 13A l AdjacentConcrete 498A 11A 47AC 12A 51AC 18A 518A 188 488A 278 Surface and Pre- 690 71D 578A 58BA 590 stress Monitor-ing Tests Detensioning and 48A 11A 47AC 38 48C8 18A 518A , 188 3608 13A l Material Tests 1 l

P I

LA SALLE - UNIT 1 3/4 6-11 Amendment 1 I 1

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. !l O ~3 m!cuaanafsvsrias ORYWELL AND SUPPRESSION CHAN8ER PUHGE SYSTEM LDuunu CONDITION FOR OPERATION O

3.8.1.8 The drywell and suppression chamber purge system may be in operation

! with the drywell and/or suppression chamber purge supply and exhaust butterfly isolation valves open for inerting, de-inerting,and pressure control, provided l that each

%rsoeb %s.butterfly Moudk G*s valve Tamis4 blocked mart sys4eu so as notbsteresMel,Q sleMt open more than 4e less #w50*.

or- "Pu si'wg+o 90 hours0.00104 days 0.025 hours 1.488095e-4 weeks 3.4245e-5 months APPLICABILITY: OPERATIONAL ColElITIONS 1, 2 and 3. p4c '36r days.-

M: ,

With a drywell and/or suppression chamber purge supply and/or exhaust butterfly .

isolation valve open for other than inerting, de-inerting,or pressure control, I i

or not blocked to less than or equal to 50* open, close the butterfly valve (s) withink hour or be in at least HOT SHUTDOWN within the nixt 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in l COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE REQUIREMENTS 4.6.1.8.1 When being opened, the drywell and suppression chamber purge supply

' and exhaust butterfly isolation valves shall be verified to be blocked so as i to open to less than or equal to 50' open, unless so verified within the previous 31 days. ,

4.6.1.8.2 Each drywell and suppression chamber purge supply and exhaust

' butterfly isolation valve shall be demonstrated OPERA 8LE at least once per

92 days by verifying that the measured leakage rate is less than or equal to O.05 L,.

4.6.1.8.3 The cumulative time that the drywell and suppression chamber purge sy:; tem has been in operation purging through the Standby Gas Treatment System j

shall be verified te be less than or equal to 90 hours0.00104 days 0.025 hours 1.488095e-4 weeks 3.4245e-5 months per 365 days prior to use in this mode of operation.

q i

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CONTA!!#9ENT SYSTEMS

~s \

3/4.6.2 OEPRESSURIZATION SYSTEMS i SUPPRESSION CHAMBER

; LIMITING CONDITION FOR OPERATION i i'

-~

_3.6.21-~ The suppression chamb'er shall be OPERA 8LE with:

a--- The pool water: <

3 3 j 1. Volume between 131,900 ft and 128,800 ft , equivalent to a

- level between 285 107.,and 26p W , and a l.

y . 2. Maximum average temperature of 100*F" during OPERATIONAL CONDITION 1 or 2, except that the maximum average temperature

~

(. ,

may be pemjged to increase to:

a) 105'F8, during testing whien adds heat to the suppression }

chamber.

. b) 110*FFwith THERMAL POWER less than or equal to I% of i 1 RATEDEERMAL POWER.

j . c) 120*FFwith the main steam line isolation valves closed l -

j following a scian.

! b. Drywell-to-suppression chamber bypass leakage less than or equal to 105 of the acceptable A//E design value of 0.03 fit2 ,

APPLICA8ILITY: OPERATIONAL CONDITIONS 1, 2, and 3.

~

ACTION:

a. With the suppression chamber water level outside the above Ifmits,

. restore the water level to within the limits within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in

h. .-. '

at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ,

b. In OPERATIONAL CONDITION 1 or 2 with the suppression chceber average j water temperature greater than or equal to 100*F, restors the average

temperature to less than or equal to 100*F within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in

' at least NOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , except, as pomitted above:

I 1. With the suppression chamber average water temperature greater

. than 105'F during testing which adds heai; to the suppression chamber, stop all testing which adds heat to the suppression

- chamber and restore the average temperature to less than or equal to 100*F within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following i 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

I 2. With the suppression chamber average water temperature greater j than 110*F, place the reactor mode switch. in the Shutdown i

position and operate at least one residual heat removal loop in the suppression pool cooling mode.

3. With the suppression chamber average water temperature greater than 120*F, depressurize the reactor pressure vessel to less I than 200 psig within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

I j : #See Specification 3.5.3 for ECCS requirements.

! "': 5: - :M'!:d wi:h'- 50 dcy; ;f ;. ;;;tica ;f tt; T"! c.at;ir;;nt 1;a; t;;t "fch ch2!' 5: :::: Mt:d 5:f:r: ME"^L PC 5" ;;;;;d; EC% c' rT : O .2 '. . . ,

^ DNE2 U *eq W a. e cNe~:e te 21: :p i'ic:ti :h:!! ::: pr;;;;;; wittia O, t5: .;.,0 00 de,;.

1. Soe Special Test Exception 3.10.8.

l I

i LA SALLE - UNIT 1 3/4 6-16

. . .-w.-

yew--w g--s- w ww t -v -

e-e eww-'---'e

.ww-w-wewer w - se'v-e w -eevw-e- w vww-v-v--+--wwmww"'-w -- -m' ww-v- -

v

l -.

~:v-.a--

- . . a.. - -

I D-C-TA - S - .

l. .

LDTTING CONDITION FOR OPERATION (Continued)

)l 5CTION: (Continued) [

c With one suppression chamber water lev".1 instrumentation channel '

l inoperable and/or with one suppression pool water temperature instrumentation division @inop'erable, restore the inoperable 1

instrumentation to OPERA 8LE status within 7 days or verify suppres-3 sion chamber water level and/or temperature to be within the limits

{j at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by local indication.

! d. With both suppression chamber water level instrumentation channels

inoperable and/or with be suppression pool water temperature instrumentation division noperable, restore at least one !

- inoperable water level channel and one water temperature division to OPERA 8LE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following i^ 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

e. With the drywell-to-suppression chamber bypass leakage in excess of

t the limit. restore the bypass leakage to within the limit prior to increasing reactor coolant temperature above 200*F.

i SURVEILLANCE REQUIREMENTS 4.6.2.1 The suppression chamber shall be demonstrated OPERA 8LE:

~

a. By verifying the suppression chamber water volume to be within the i limits at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in OPERATIONAL CONDITION 1 or 2 by verifying the suppression chamber average water temperature to be less than or equal to 100*F, except:

I 1. At least once per 5 minutes during tasting which adds heat to j the suppression chamber, by verifying the suppression enamber average water tamperature less than or equal to 105'F.

At least once per 60 minutes when suppression chamber average 2.

water temperature is greater than 100*F, by verifying suppression

! chamber average water temperature less than or equal to 110*F

! and THERMAL POWER less than or equal to 1% of RATED THERMAL POWER.

l j i 3. At least onca per 30 minutes following a scram with suppression i

' chamber average water temperature greater than or equal to 100*F, by verifying suppression chamber average water temperature less than or equal to 120*F.

. =-==;===. ,=: = = = = = == = : = = = : = = == : =

8 ef eeW etie- ef ;-! bum te emture r muu-d by ee d'"*:*e- te ;ee!

! b@ *-mture as-. measured by ce-th-d4*hbm. 5: !:: ;f .;;n: tat's and -

- "mj ne .etnemm4emsm =m t~mwo-

.wi thi n.90-days.of. demonstration /-

j LA SALLE - UNIT 1 3/4 6-17 ,

1 s

e e

. CONTAINMU R SYST9t5 .

glLANCE REQUIiiEigcNia (Continued)

c. By verifying at least two suppressiongamber water level instru-mentation channels and at least fm_w suppression pool water .

f

- tamperature instrumentation channels, y in each of two divisions,

?

, OPERABLE by performance of a: -

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

2. CHANNEL FUNCTIONAL TEST at least once per 31 days, and

3. CHANNEL CALIBRATION at least once per 18 months, with the water level and tamperature alarm setpoint for:

ii ,

L Nigh water level i 2F -

2. Law water level 3,2g447 ._

3. High temperature i 100* T

d. By conducting drywell-to-suppression chamber bypass leak tests and verifying that the A/4 calculated from the seasured leakage is within the specified limit when drywell-to-suppression chamber

bypass leak taats are conducted

[

1. At least once per 18 months at an initial differential pressure l of L5 psi, and

2. At the first refueling outage and then on the schedule required for Type A Overall Integrated Containment Leakage Rate tests by Speci-fication 4.6.1.2.aj at an initial differential pressure of 5 psi, l l except that, if the first two 5 psi leak tests performed up to that ties result in:

i 1. A calculated A/ 4 within the specified Itait, and

- 2. The A/ 4 calculated from the leak tests at 1.5 psi is 1 20% of

! the epecified limit, then the leak tests at 5 psi may be discontinued.

i !

l r- P \

! Mru wee-ammT..--Ffnel-setpoint-to-be-deter;'ned during-startup-te;; preg .

l --

^-" mt:d 2 ;;; i: tti; e.tpeia^. 5.'.e'! L.

w L si.i.o Lu Um C s.. iwi.

,j . - MtMa*^ dy' ef teet r ,-?;ti:n. ,

.i I

LA SALLE - UNIT 1 3/4 6-18 l

I ee- em * *me>g=* -*** "******** " * * " " * * * * * * * " " * * * * "

e-m *=--e e.e ,,neegma ese g S' = s_myw ww e e' - --wew--*wumpe- w ww --+ ---- --o-w - - - - - r v-, - - - -- * ---

I 2

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) '

f

- If any 1.5 psi or 5 psi leak test results in: -

,p .i . .

1. A calculated A/ 8 greater than the specified limit, or

}-

~

2. A calculated A/4 from a 1.5 psi leak test > 20% of the p; ,

specified limit, then the test schedule for subsequent tests shall be reviewed by the Commission.

If two consecutive 1.5 psi leak tests result in a calculated A//k greater than the specified limit, then:

1. A 1.5 psi leak test shall be performed at least once per 9 months until two consecutive 1.5 psi Ic:h te:ts result in the calculated A/8 within the specified limits, and

2. A 5 psi leak test, performed with the second consecutive successful 1.5psileaktest,resultsinacalculatedA/8 within the specified limit, after which the above schedule for only 1.5 r;si leak tests may be resumed.

If two consecutive 5 psi leak. tests result in a calculated A/.6 greater thanthespecifiedlimit,thena5psileaktestshallbeperformedat least once per7Xmonths until two consecutive 5 psi leak tests result in a calculated A/d within the specified limit, after which the above k schedule for only 1.5 psi leak tests may be resumed.

1 1

l 1

LA SALLE - UNIT 1 3/4 6-19

. ... . . - . x=......__ : - - .. em -

. I CONTAINNENT SYSTEMS l dm. ~'

' SUPPRESSION POOL SPRAY

. ~-

.I

. LIN! TING CGNOITION FOR OPERATION '

i 3.6.2.2 The suppression pool spray mode of the residual heat resoval (RHR) system shall be OPERA 8.LE with two independent loops, each loop consisting of:

a

a. One OPERA 8LE RHR pump, and

!' b. An OPERA 8LE flow path capable of recirculating water from the

, suppression chamber.

APPLICA8ILITY: OPERATIONAL CONDITIONS 1, 2, and 3. ,

ACTION: . .

a. With one suppression pool spray loop inoperable, restore the inoperable

- loop to OPERABLE status within 7 days or be in at least HOT SHUTDOWN

/ within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following

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

i . b. With both suppression pool spray loops inoperable, restore at least

^

one loop to 0PERA8LE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN" within the

,/ following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . .

SURVEILLANCE REQUIRENENTS

}

4. 6. 2. 2 The suppression pool spray mode of the RHR system shall be demonstrated I

OPERA 8LE:

I

a. At least once per 31 days by verifying that each valve manual, power-operated j or automati$ in the flow path that is not locked, j sealedj or otherwise secured in position, is in its correct position.

b. By verifying that each of the required RHR pumps develops a flow of l ; at least 450 gpa on recirculation flow tnrough the suppression pool 1 spray sparger when tasted pursuant to Specification 4.0.5.

I Whenever Dotn RHR subsystems are inoperable, if unable to attain COLD SHUTDOWN

as required by this ACTION, maintain reacter coolant temperature as low as 3

practical by use of alternate heat re: oval methods.

I

fm

! \ .) '

O-j LA SALLE - UNIT 1 3/4 6-20 ,

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

[

~ ~ ~

~~~~~

.. . v .

. (.

k '

CONTAD01ENT SYSTDt5

' ~

f

- i SUPPRESSION POOL COOLING -

- a

! gr w _

i . Lm u anu CON 0! TION FOR OPERATION h

3.6.2.3 The suppression pool cooling mode of the residual heat removal (RHR) -

system shall be OPERA 8LE with two independent loops, each loop consisting of:

a. One OPERABLE RHR pump; and

b. An OPERA 8LE flow path capable of recirculating water from that suppression chamber through an RHRSW heat exchanger.

APPLICA8ILITf: OPERATIONAL CONDITIONS 1, 2, and 3.

ACTION:

a. With one suppression pool cooling loop inoperable; restore the inoperable loop to OPERA 8LE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least

, HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within l the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . .

( .

i' b. With both suppression pool cooling loops inoperable, be in at.least.

h. HOT SHUTDOWN v't.hin 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN" within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE REQUIREMENTS .

i .

t 3 l 4.6.2.3 The suppression pool cooling mode of the RHR system shall be l

demonstrated OPERA 8LE

j a. At least once per 31 days by verifying that each valveksanual,

. powerwoperated or automatid in the flow path that is not locked, f

sealed or otherwise secured in position, is in its correct position, i

l a 7

. b. By verifying that each of the required RHR pumps develops a flow of 1

at least 7200 gpa on recirculation flow through the RHR heat exchanger and the suppression pool when tested pursuant to

! Specification 4.0.5.

. ~

I l

e I

'l u "Whenever Doth RHR subsystems are inoperable, if unable to attain COLD SHUTDOWN as required by this ACTION, maintain reactor coolant temperature as lcw as practical by use of altarnate hea removal :nethods.

I

(' s

, o LA SALLE - UNIT 1 3/4 6-21 l

. . - _ . . . . . - . - - --- - - - '--"-~~~- -"* ~ ~~ ~ ~~ ' '~

TABLE 3.6.3-1 m PRIMARY CONTAIMENT ISOLATION VALVES l;; MAXIMLM .

ISOLATION TIME h VALVE FUNCTION AND NUMBER VALVE GROUPI ") (Seconds) 3 H a. Automatic Isclation Valves s

1. Main Steam Isolation Valvas* 1 5* I 1821-F022A, B, C, D I) 1821-F028A, B, C, DI)

2. Nain Steam Line Drain Valves M 1 1821-F016 1821-F019 1 15 {

1821-F067A, B, C, D(b) -< 15

3. 1 23 R Reactor Coolant S stem Sample I Line Valves (c 4 1833-F019 3 15

IB33-F020 6

4. Drywell Equipment Drain Valves ?

1RE024 5 20 1RE025 10E026 1riE029

5. Drywell Floor Drain Valves 2 1RF012 1 20 1RF013

6. Reactor Water Cleanup Suction Valves 5 < 30

~

1G33-F001(d) 1G33-F004

7. RCIC Steam Line Valves 8 1E51-F008(*) < 20 1E51-T063 I 15 IE51-F064 I#) 15 IE51-F076 -

I 15

, e TABLE 3.6.3-1 (Continued) ,

9 .

PRIMARY CONTAINMENT ISOLATION VALVES E MAXIMUM

. ISOLATION TIME c- VALVE FUNCTION AND NUMBER VALVE GROUP (a) (Seconds)

?.

Automatic Isolation Valves (Continued)

8. Containment Vent and-Purge Valves # 4

< 4iHF10 #

IVQO26 IVQ027 I.EMPfO**

IVQO29 I 44HP-/O**

IVQO30 I JiHrlod*

IVQO31 7.&30"/O""

1VQO32 7 -16 S' #

7 aiHP 'O w IVQO34 h IVQO35 75 cn IVQO36 7.EHr/O "

4 IVQ040 7 430-/ #

IVQ042 iM/*

IVQ043 2 !Mr/O**

IVQ047 7.3fP 5' 7 .3iF-S" IVQ048 1VQ050 IMS" 7.33-U" 1VQ051 IVQ068

<5

9. RCIC Turbine Exhaust Vacuum Breaker 9 M N.O .

Line Valves 1E51-F080 1E51-F086

10. LPCS, HPCS, RCIC, RHR Injection Testable Check Bypass Valves I9) 2 - NA-M. A . l IE21-F333 1E22-F354 1E12-F327A, B, C 1E51-F354 IE51-F355

}

I i -

I / k J l

YABLE 3.6.3-1 (Continued) 9 PRINARY CCNTAlletENT ISOLATION VALVES' . .

un !

l N HAMilG20 C ISOLATION TlfEE

! s VALVE fullCTION AND NUh8ER VALVE GRottI *) (Seconds) , , :

I.

l E

q Automatic Isolation Valves (Continued) 1 ' '

i 11. Contalement Monitoring Valves 2 <5 3 Cit 317A,8 M 1Cl10 t hA,BF '

IC119 t'JA,BW '

ICliO20A,8 # .-

l 10 10218 } . .

1C11022

11 8

! N

fi164 A

! 10s:oy?,A'IO i ICI;02Lil INI . '

! S 10.1o?/ s

10. :,2.> . .*

l *

ICi .fii9 ,

1C: 330 t l ICt.in s t ,*

! ICita't2 .

ICIEll3 -

ICfin'l4 .

i; 12. Dr;.si:ll Pneumatic Valves t lith,ulA and 8 g < 40 '

1111017 7 30 '.

11:10 /-l

I 30 1 11n0/?# 7 30 11 h0 i l* e/ 7~

1 111o 111 l .

13. R ,intdown Coollag Mode Valves 6 ,

IEI1-1008

< 41 1E!? 1009 7 41 1EI? 1023 7 90 1E 1/-l %3 A and 8 - 7 29 3' 8) 1E I7-1 sl99A , 30 .

8 f

eMe g e

. t, ,

Q,)

(' N a .

TABLE 3.6.3-1 (Continued)

[i

^

i 9 -

a PRIMARY CONTAlletENT ISOLATION VALViES .,

V .

j;; MAXIMIN ISOLATIGH TISE i c- VALVE FUllCTION AND NUMBER VALVE GROUP I ") (Seconds)

! 5 d

l

_Aulc.m.itic Isolation Valves (Continued) g ,

Tip Cuide Tube Valve 7 --MHl .O. ~l 1

) 14.

Ball W 1ve >

l 1C31 .1004 *

15. Reactor Building Closed Cooling Water .

Sys, tem Valves 2 1 30 ,

1W11029

) IWl:040 t* 1 Wit i19 i

IWl:Ihu i i' 16. Primary Containment Chilled , ,

m

" U.:8. r. Inlet Valves

2 l l IVIil.s A and B -

< 90 -r*

Ivi*hf>3 A and 8 .

340

17. Prim.iry Containment Chilled uatur Outlet Valves

2 l

, IVIu5.1 A and 8 < 40 i IVi'l14 A and B ~ 3 90 i 18. Re.irt. ilydraulic Flow Control l Lii.e W ives IU) 2 l 55 , ,

j IB n l338 A and S .

! 10 rs -1139 A and 8 IB is-1140 A and 8 IB s1 1 141 A .and 6

! 1B'13-1342 A and 8 . .

I IB'l3-i '143 A and B ,

i 111&1-I:144 A and 8 -

g 10 s's 1 :145 A and 8

19. Fe..t.s..ter Testable Check Valves 2 -NA- /I.A -

I k )

IB?l-l .132 A and B ' V.

f me -

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

(

C<' \

1 /. - ;

1

, .]-

I TABLE 3.6.3-1 (Continued)

I

E .

m PRIMARY CONTATNMENT ISOLATION VALVES r-

};; WRIleM .

ISOLATION TIIE c VALVE FUNCTION AND NUNBER VALVE GROUP (a) (Seconde) iE .

H b. Manual Isolation Valves @ .

. >4

1. 1FC086 hA/.A.

2. 1FCI13 . mar A/.4-

3. IFCil4 4W3/.A. '

j 4. IFCll5 . N##.A.

4

5. 1HCO27(1) Nard.4-

6. 1HCO33(1) -NRAlG. s '

7. 15A042(1) NArA/.A-

8. I'.A046(1) .NA A/.d.

i w '

s .

5 m

8 -

g 9 M

=

1 g

l i i t

o t

l ',

g.*._

. I I

l.

.?

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

l I

-' J . .

TABLE 3.6.3-1 (Continued) , l g i 1 m PRIMARY CONTAll#4ENT 150LAT10N VALVES 2- ! .

E VALVE FilllCTION AND NUMBER j i

$ d. Ottier Isolation Valves ,

z ; ;

E 1. MSIV Leakage Control Systee j I y

.I IE32-F001A, E. J. N(b) ..l'

'. ~

2. Reactor Feedwater and RWCU System Return j 1821-F010A, 8 ,

i 1821-F065A, 8 3g ,

IG33-I040 .

y 3. N.lihialHeatRemoval/LowPressureCoolantInjectionSystes.

[ IE12 lu42A, 8. C .

4 N

1Elz-lul6A, 8 1E12-1017A, 8 IS ,'i IE12-lug 4A, 8 1E12-l'u27A, B Ilg ) -

I3) '

1E12-1024p)38 1E12-1021 .

1E12-1lu2 III -

IS) 1E12-lu64A, 1E12 IullA, B 8Ilj8 (

1Elz-lt.88A, 8, C IS) ' .

IS)

IE12-1025py)8, C g IE12-1030 1E12-lug 5 I3)

Il}y IE12-lH13A, IE12-tu74A, B BI3p ,

IElz-luSSA, 8 EI) 3Cl2-fu36A, B Il ~

IE12-t311A, B IS) !-

1E12-1041A, 8 IE12-1050A, 8 pg ) IN Q

)

k

. y -

?.

6i

i . a. . .4 . e

' . - ~- -- -- - - -

. . . . . . . . . . . . . . ~ . .. ..-. -...-... - .

.s , ,

i f I. ' ,

n j l

~ -

i TABLE 3.6.3-1 (Continued) '

t C .

. I' J

v. PRIMARY CONTAllMENT ISOLATI001 VALVES I

l' i ;;;

. VALVE FtmCIION AND NUMBER c-2 f.

-4 Other Isolation Va'Ives (Contfnued) .

" 4. Low Pressure Core Spray Systes ;

l j IE21-F005 . ,

1E21-f001 III i

IE21-f012 III .

IE21-f0ll III IE21-f018 Il s ._ ,'.-

IE21-ID31 Il III l 1[21-1006 i t' i

5. .Il.ioli Pressure Core Spray System J. 1E22-I004 -

'd 1E22-F015 III 1E22-F023 III IE22-IU12 III '

1[22-IG14 III 1E22-1n05 IkI i 6. Re.ir.ts.r Core Isolation Coolina Systes

['

IE *.1 -11113 .

i gE*,1-luo6 I LIE S l-l'080 .; ,l IESI-1069 i IESI-1028 i 1E S1-1 u68 !.

IE51-1040 r IESI-f031 III J .

I -

j-1E51-1019(N 1E51-l'065 k) ,

IESI-1066 II (.

I 47

. t.

5 ef*e e e e

i s .

I s .) -

I V

1

'~

TABLE 3.6.3-1 (Continued) -

iE PRIMAR'1 CONTAINMENT ISOLATION VALVES i

vi

?

! E ,

l . . VAIVE FUNCTION AND NUNBER , '.

l c l-i y other Isolation Valves (Continued) ,

" 7. --.0CA Post 1 Hydrogen Control

'c IIK.001A, 8 .-

I ,

i; l Elicou2A, B . .,

lilcule',A, 8 .I IllG006A, 8

8. Stanilhy Liquid Control Systee c., IC41-12004A, 8 '!

1 ) 3C41-1007 9.

l .', Reactne Recirculation Seal Injection i ,

i 18:s: 1.ll3A, B ISI

i

' 1Bys-1017A,8(J} . r l 10. Drywell h mona W <_ Ty h . ( '

j, IIN oi s ,}.

- u But > 3 seconds.

l I i # TheIIrovisionsofSpecification3.0.4arenotapplicable. -

. . 'i (a) See *p.cification 3.3.2 Table 3.3.2-1, for isolation signal (s) that operates each valve groep, :

i (b) Not is.i.luded in total sum of Type 8 and C tests. 5

, (c) May leis opened est an intermittent basis under administrative control. _-

(d) Not ..Insed by SLCS actuation. I.

l (e) Not closed by Trip functions Sa, b or c. Specification 3.3.2, Table 3.3.2-1.

c : ,

(f) Not a.luded by Trip functions 4a, c, d, e or f of Specification 3.3.2 Table 3.3.2-1. '

(g) Not f.iihject to Type C leakage test.

(h) Open. on an isolation signal. Valves will be open during Type A test. No Type C test required, k{.t I (i) Alt.o closed by drywell pressure-high signal. ,

A' (j) Hyii mile leak test at 43.6 psig. .?

I (k) Not ai.bject to Type C leakage test - leakage rate tested per Specification 4.4.3.2.2. *) .

(1) The .c l.enetrations are provided with removable spr.ols outboard of the outboard isolation 14 valsc. During operation, these lines will be blind flanged using a double 0-ring and a I ';

tyg . 11 leak test. In addition, the packing of these isolation valves will be soap-bubble .

l}'

ter. .il to ensure insignificant or no leakage at the containment test pressure each refuel- ,

' .M* s es Sk All h AC A h*AKiMUM INl*hd*' bE"* *$ O '

P !# 'N '

j rdm/iba m4me. .-

= .......: ~ - '-"

M W O S -:. C : .:; :' ' % ~ ? ~ + 6 <

. , . a..

. CONTAINNENT SYSTENS

.I 3/4.6.4 VACUtM RELIEF

! .4 t '

I c LIMITING CONDITION FOR OPERATION

]-

t 3.6.4 All suppression chanber - drywell vacuum breakers shall be OPERA 8LE and

' closed.

~

gPLICA8ILITY,: OPERATIONAL CONDITIONS 1, 2 j and 3.

ACTION:

a. With one suppression chamber - drywell vacuum breaker inoperable and/or open, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months clo;,e the manual isolation valves on both sides of the inocerable and/or open vacuum breaker. Restore

?' the inoperable and/or open vacuum breaker to CPERA8LE and closed -

status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTOOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b. With one position indicator of any OPERA 8LE suppression chamber -

l drywell vacuum breaker inoperable, restore the inoperable position l- indicator to OPERA 8LE status within 14 days or visually verify the .

a vacuum breaker to be clos ~ed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Otherwise,

v. declare the vacuum breaker inoperable.

SURVEILLANCE REQUIREMENTS 4.6.4.1 Each suppression chamber - drywell vacuum breaker shall oe:

a. Verified closed at least once per 7 days.

i b. Demonstrated OPERABLE: ,

3 I 1. At least once per 31 days and within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after any discharge of steam to the suppressica chamber from the safety-relief valves, by cycling each vacuum breaker through at least one

. complete cycle of full travel.

l l 2. At least once per 31 days by verifying both position indicators I

OPERA 8LE by performance of a CHANNEL FUNCTIONAL TEST.

l

3. At least once per 18 months by; a) Verifying the force required to open the vacuum breaker, from the closed positica, to be less than or equal to 0.5 psid, and i

] b) Verifying both cosition indicators OPE. A3LE R by performance of a O CHANNEL CALIBRATION.

O' ' .

I LA SALLE - UNIT 1 3/4 6-35

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CONTAINNENT SYSTE45 Q*

i SURVEILLANCE REQUIRi!MENTS (Continued) l i

, i 4.6.4.2 The manual isolation valves on both sides of an inoperable and/or l t open suppression chamber-drywell vacuum

breaker shall be verified to be closed at least once per 7 days.

4.6.4.3 Vacuum breaker header fTanges which have been broken shall Ce feak tested after et-saking by S "h et et . r....... ;' i 5 ;:f;,-

Typ B TAAt a131.b gs; ,

pc 5' c;44iw 4,b. ,2 d ,;

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'M CONTAINMENT SYSTDIS j 3/4.6.5 SECONDARY CONfAINNENT ,

' l,

$ECONDARY CONTADetENT INTEGRITY i LDt! TING CONDITION FOR OPERATIGI i

l 1 3.6.5.1 SECONDARY CONTAINNENT INTEGRITY shall be maintained.

APPLICA8ILITY: OPERATIONAL CWWITIONS 1, 2, 3 and *. l j

ACTION:

Without SECONDARY CONTAIP9 TENT INTEGRITY:

a. In OPERATIONAL CONDITION 1, 2 or 3, restore SECONDARY CONTAINNENT l

. INTEGRITY within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT SHUTOCWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLO 5 HUT 00WN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

I

b. In Operational Condition *, suspend handling of irradiated fuel in ~

! the secondary containment, CORE ALTERATIONS and operations with a l potential for draining the reactor vessel. The provisions of i : Specification 3.0.3 are not applicable. -

SURVEILLANCE REQUIRENENTS l O .

!' . .m 4.6.5.1 SECONDARY CONTAIMENT INTEGRITY shall be demonstrated by:

! '7 a. Verifying at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months that the pressure within the l l secondary containment is less than or equal to 0.25 incher of l

' vacuum water gauge.# .

b. Verifying at least once per 31 dayr, that:

I 1. At least one door in each access to tha secondary containment .

is closed.

,i . 2. All secondary containment penetrations not capable of being

' closed by OPERABLE secondary containment automatic isolation j

despers and required to be closed during accident conditions I are closed by valves, blind flanges, or deactivated automatic dampers secured in position.

c. At least once per is months:

1. Verifying that one standby gas treatment subsystem will draw down the secondary containment to greater than or equal to 0.25 inches of vacuum water gauge in less than or equal to '(

300 seconds, and

2. Operating one standby gas treatment subsystem for one hour and maintaining greater than or equal to 0.25 inches of vacuus l I water gauge in the secondary containment at a flow rata not exceedi.g 4000 CFN t 10%.

l r "When tresciatea tual is beiniJ har.::let in e HLindary c .'.t 21- := .c anc 3 r'.'c CORE ALTERATIONS and operations witn a potential for craining tne reactor vessei.

{l # SECONDARY CONTAINMENT INTEGRITY is maintained when secondary containment vacuum O', # is less than required for up to. pat hour solely due to Reactor Building ventilation system failure. .i l

l j t

i LA SALLE - UNIT 1 3/4 6-37 I

1

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Dd,'

CONTAIlWEENT Sh ium

,i i SECD M ARY CONTAINNENT AUTOMATIC ISOLATION DANPERS lI I .

j Lammum COMITION FOR OPERATION

\* t l

' ;! 3. 6. 5.2 The secondary containment ventilation system automatic isolation j dampers shown in Table 3.6.5.2-1 shall be OPERA 8LE with isolation times equal to or less than shown in Table 3.6.5.2-1.

' APPLICA8ILITY: OPERATIONAL CON 0!TIONS 1, 2, 3, and ". l i j ACTION:

1 With one or more of the secondary containment ventilation system automatic isolation dampers shown in Table 3.6.5.2-1 inoperable:

f

a. Maintain at least one isolation damper OPERA 8LE in each affected .

penetration that is open and within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , either:

- - .- L Resto e the inoperable damper to OPERA 8LE status, or -

l 2. Iso 1a'te each affected penetration by use of at least one deactivated automatic damper secured in the Tsolation position, or j 3. Isolata each affected penetration by use of at least one cicsed .

b. Otherwi in 0 C0 ON 1, 2 or 3, be in at least HOT l SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in, COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

l- c. Otherwise, in Operational Condition 8, suspend handling of irradiated I fuel in the secondary containment, CORE ALTERATIONS and operations I

with a potential for draining the reactor vessel. The provisions of .

Specification 3.0.3 are not applicable.

'+ .

i i SURVEILIANCE REQUIREMENTS i

e 4.6.5.2 Each secondary containment ventilation system automatic isolation j damper shown in Table 3.6.5.2-1 shall be demonstrated OPERA 3LE:

j a. Prior to returning the damper to service after maintenance, repair

or replacement work is performed on the damper or its associated

} actuator, control or power circuit by cycling the damper through at i

least one cssplete cycle of full travel and verifying the specified isolation time.

i b. During COLD SHUTDOWN or REFUELING at east once per 18 months by verifying that on a containment isolation test signal each isolation l damper actuates to its isolation position.

. c. By verifying the isolation time to be within the limit when tested

! m pursuant to Specification 4.0.5.

! ( '

o

' "When irradiated fuel is being handled in the secondary containment and during O'

t. ORE ALTERATIONS and operations with a potential for draining the reactor

., vessel.

i LA SALLE - UNIT 1 3/4 6-38 P

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CONTAINEENT SYSTEMS .

.v t STANDBY GAS TREATNENT SYSTEN .e I

l i -

LIMITING CONDITION FOR OPERATION e

l 'l 3.6.5.3 Two independent standby gas treatment subsystems shall be OPERA 8LE.#

APPLICA8ILITY: OPERATIONAL CONDITIONS 1, 2, j3 and 8 g i i ACTION:

l i i a. With one standby gas treatment subsystem inoperable, restore the

! inoperable subsystes to OPERA 8LE status within 7 days, or:

1 1 1. In OPERA 8LE CONDITION 1,32 er 3, be in at least HOT SHUTCOWN l -

l within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and.in COLD SHUTDOWN within the

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

[ 2. In Operational Condition *, suspend handling of irradiated .

fuel in the secondary cratainment, CORE ALTERATIONS and opera- l l 1 tions with a potential for draining the reactor vessel. The . j

.. provisions of 5 ecification 3.0.3 are not applicable. j d b. With both standby gas treatment subsystems inoperable in Operational !

Condition *, suspend handling of irradiated fuel in the secondary j containment, CORE ALTERATIONS and operations with a potential for draining the reactor vessel. The provisions of Specification 3.0.3 are not applicable. ; g i

SURVEILLANCE REQUIREMENTS j l 4.6.5.3 Each standby gas treatment subsystes shall be demonstrated OPERA 8LE:

I

a. At least once per 31 days by initiating, frcm the control room, flow throug* the HEPA filters and charcoal adsorbers and verifying that the l subsystem operates for at least 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months with the heaters OPERABLE.

"When irradiated fuel is being handled in the secondary containment and during ,

CORE ALTERATIONS and operations with a potential for draining the reactor vessel.

A The r.crmal or emergency power sour:a '

.~.ay be inoperable in Operational

,/ Condition *.

t 1

l : LA SALLE - UNIT 1 3/4 6-40

~~

~

i, CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

\- At least once per 18 months or (1) after any structural maintenance b.

(f on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or l j chemical release in any went11ation zone communicating with the subsystem by:

1. Verifying that the subsystem satisfies the in place testing

,' acceptance criteria and uses the test procedures of Regulatory f Positions C.5.a. C.5.c and C.5.d of Regulatory Guide 1.52,

Revision 2, March 1978, and the system flow rate is 4000 cfm 2 105.

' 2. Verifying within 31 days after removal that a laboratory analysis f

of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52. Revision 2,

(

March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978.

L

3. Verifying a subsystem flow rate of 4000 cfm + 10% during system operation when tested in accordance with ANSI N510-1975.

c.* After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operation by verifying -

within 31 days after removal that a laboratory analysis of a

=

representative carbon sample obtained in accordance with Regulatory l

Position C.6.b of Regulatory Guide 1.52 Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52. Revision 2, March 1978.

~

)

d. At least once per 18 months by:

q

1. Verifying that the pressure drop across the combined HEPA filters

'.4 ' and charcoal adsorber banks is less than or equal to 8 inches l Water Gauge while operating the filter train at a flow rate of 4000 cfm 2 105.

2. Verifying that the filter train starts and isolation dampers open on each of the following test signals:

I

a. Reactor Building exhaust plenum radiation - high,

b. Drywell pressure - high,

-c. Reactor vessel water level - low low, level 2, and

d. Fuel pool vent exhaust radiation - high.

3. Verifying that the heaters dissipate 20 1 2.0 kw when tested in accordance with ANSI N510-1975.

I l

O 3/4 6-41 LA SALLE - UNIT 1

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i PLANT SYSTEMS

,% - SURVEILLANCE REQUIRE 4ENTS

c. At least once per 18 months by:

l 1. Performing a system functional test which includes simulated automatic actuation and verifying that each automatic valve in 2.--..- _

~ ~ ~ ~ ~ ~ ~ ~

r - -- -

~

the flow path actuates to its correct position, but may exclude N1 injection of coolant into the reactor vessel.

2. Verifying that the system is capable of providing a flow of greater than or equal to 600 gpa to the reactor vessel when

- steam is supplied to the turbine at a pressure of 150.i 15 psig using the test flow path.# l

3. Performing a CHANNEL CALIBRATION of the discharge line " keep filled" pressure alare instrumentation and verifying the low pressure setpoint to be > 62 psig.

d. By demonstrating MCC-121y and the 250-volt batter 9 9 "and charge # l OPERABLE:

1. At least once per 7 days by verifying that:

a) MCC-121y is energized, and has correct breaker alignment, indicated power availability from the charger and battery, and voltage on the panel with an overall voltage of greater than or equal to 250 volts.

b) The electrolyte level of each pillt cell is above the plates, c) The pilot cell specific gravity, corrected to 77*F, is O,- d) areater taan or eauai to 1.200. aad The overall battery voltage is greater than or equal to 250 volts.

(

2. At least once per 92 days by verifying that:

a) The voltage of each connected battery is greater than or _

equal to 250 volts under float charge and has not decreased more than 12 volts from the value observed during the original test, b) The specific gravity, corrected to 77'F, of each connected cell is greater than or equal to 1.195 and has not decreased more than 0.05 from the value observed during the previous test, and c) The electrolyte level of each connected call is above the plates.

3. At least once per 18 months by verifying that:

a) The battery shows no visual indication of physical damage or abnormal deterioration, and b) Battery terminal connections are clean, tight, free of corrosion and coated with anti-corrosion material.

"The provisions of Specification 4.0.4 are not applicably provided the surveillance is performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure is j dequate to perform the tests.

2: : :e *

. '45

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LA SALLE - UNIT 1 3/4 7-8 Amendment 5

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l PUNT SYSTEMS

!,. SURVEILLANCE REQUIREMENT 3 4.7.5.1.1 The fire suppression water systes shall be demonstrated OPERA 8LE:

i a. At least once per 31 days by verifying that each valve, manual, power operated or automatic, in the flow path is in its correct position.

b. At least once per 12 months by cycling each testable valve in the flow path through at least one complete cycle of full travel.

c. At least once per 18 months by performing a system functional test which includes simulated automatic actuation of the system throughout its operating sequence, and:

1. Verifying that each automatic valve in the flow path actuates to its correct position, ,

2. Verifying that each fire suppression pump develops at least gpaatasystemheadofgfeet,

3. Cycling each valve in the flow path that is not testable during -

plant operation through at least one complete cycle of full travel, and s.

~

4. - Verifying that each fire suppression pump starts sequentially ,

to maintain the fire suppression water systes pressure greater l

than or equal to)libDt psig.

1

d. At least once per 3 years by performing a flow test of the system in -

accordance with Chapter 5, Section 11 of the Fire Protection Handbook,

14th Edition, published by the National Fire Protection Association.

4.7.5.1.2 Each diesel driven fire suppression pump shall be demonstrated OPERA 8LE:

I a. At least once per 31 days by:

1

1. Verifying the fuel day tank contains at least 130 gallons of fuel,

2. Starting:

f ,

a) The fuel transfer pump and transferring fuel from the

} storage tank to the day tank.

b) The diesel driven pump from ambient conditions and operating

.l 8 for at least 30 minutes on recirculation flow.

I l i . ,,

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. J .

1 i

i LA SALLE - UNIT 1 3/4 7-12 9

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... , 'vNTSYSTens OELUGE AN0/0R SPRINKLER SYSTEMS .

i t ,

! LIMITING CONDITION FOR OPERATION i

I 3.7.5.2 The deluge and sprinkler systems of Unit 1 and Unit 2 shown in

I Table 3.7.5.2-1 shall be OPERABLE.*

A l

APPLICA8ILITY: Whenever equipment protected by the deluge / sprinkler systems l

l are required to be OPERA 8LE. '

i t ACTION:

l j a. With one or more of the deluge and/or Sprinkler systems shown in l4 .

Table 3.7.5.2-1 inoperable, within hour establish a continuous l

! fire watch with backup fire suppression equipment for those areas in -

j which redundant systems or components could be damaged; for other i areas, establish an hourly fire watch patrol. Restore the system to

OPERABLE status within 14 days or, in lieu of any-other report j raquired by Specification 6.6.8, prepare and submit a Special Report j to the Commission pursuant to Specification 6.6.C within the next 30 days outlining the action taken, the cause of the inocerabil.ity and .

C- the plans and schedule for restoring the system to OPERA 8LE status.

V b. The provisions of Specification 3.0.3 and 3.0.4 are not appitcable.

i 1 -

SURVEILLANCE REQUIREMENTS 1 -

4.7.5.2 Each of the above required deluge and sprinkler systems shown in

. Table 3.7.5.2-1 shall be demonstrated CPERAGLE:

a. At least once per' 31 days by verifying that each valveksanual, power-operated)or automatirfy'in the flow path is in its correct position. i

b. At least once per 12 months by cycling each testable valve in the flow path through at least one complete cycle of full travel.

c. At least once per 18 months: ,

1. By performing a system functional test which includes simulated automatic actuation of the system, and:

I. a) Verifying that the automatic valves in the flow path

! actuate to their correct positions on a test signal, and

~

b) Cycling each vase in :. e ticw cath th.1t is not testsMa during plant cperation .r.rcuIn at 1.ust er.e ccm:lete ejcle 3

of full travel.

((y ,, -

"The normal or emergency powter source may be inoperable in OPERATIONAL CONDITION 4 or 5 or when defueled.

i LA SALLE - UNIT 1 3/4 7-14

~

i PLANT SYSTEMS CO, SYSTEMS 1

LIMITING CONDITION FOR OPERATION

- - 3.7.5.3 The following low pressure C0 systems of Unit 1 and Unit 2 shall be i OPERA 8LE.*

a. Division 1 diesel generator 0 room.

b. Division 2 diesel ponerator IA room.

^

c. Division 3 diesel generator 18 room. ,

i d. Unit 2 Division 2 diesel generator 2A room.

APPLICA8ILITY: Whenever equipment protected by the low pressure C02 systems is required to be OPERA 8LE.

ACTION:

I I

a. With one or more of the above required low pressure CO systems Saoperable, within. hour establish a continuous fire watch with I backup fire suppression equipment for those areas in which redundant systems or components could be damaged; for other areas, establish O. - an hourly fire watch patrol. Restore the system to OPERA 8LE status

( -

within 14 days or, in lieu of any other report required by Specification 6.6.8, prepare and submit a Special Report to the I Commission pursuant to Specification 6.6.C within the next 30 days outlining the action taken, the cause of the inoperability and the f plans and schedule for restoring the system to OPERA 8LE status. .

b. The provisions of Specification 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS i

l l 4.7.5.3 Each of the above requirsa low pressure CO 2 systems shall be demonstrated OPERA 8LE:

I

a. At least once per 7 days by verifying CO 2 storage tank level to be greater than 505 full and_ pressure to be greater than 290 psig, and

b. Atleastonceper31daysbyverifyingthateachvalvee(manual, power-

} operated, or automatief)in the flow path is in the correct position.

i

c. At least once per 18 months by verifying:

1. The system valves and associated motor operated ventilation dampers actuate, manually and automatically, upon receipt of .

a simulated actuation signal, and

- 2. Flow from each nozzle during a " Puff Test."

The normal or emergency power source may be inoperable in OPERATIONAL l

l CONDITION 4 or 5 or when defueled.

3/4 7-17 Amendment No.13

. LA SALLE - UNIT.1

i i

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i

'). ,u,,r sv5 FIRE HOSE STATIONS 5

.4

r. ,

l Linunu CONDITION FOR OPERATION n n _______,.__

l 3. 7.-5.4 The fire hose stations of Unit'l and Unit 2 shown in Table 3.7.5.4-1 l shall be OPERA 8 M.

t-APPLICA8ILITY: Whenever equipment in the areas protected by the fire hose stations is required to be OPERA 8LE.

ACTION:

a. With one or more of the fire hose stations shown in Table 3.7.5.4-1 inoperable, route an additional fire hose of equal or greater diameter to the unprotected area (s)/ zone (s) from an OPERABLE hose station I within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months if the inoperable f, ire hose is the primary means of fire suppression; othentise, route the additional hose within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Restore the inoperable fire hose station (s) to GPERABLE status within 14 days or, in lieu of any other report required by Specification 6.6.8, prepare and submit a Special Report to the Commission pursuant to Specification 6.6.C within the next 30 days outlining the action taken, the cause of the inoperabi.11ty and the plans and schedule -

O. ,

for restoring the system to OPERA 2LE status.

.- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

~

I l SURVEILLANCE REQUIREMENTS _

l + 4.7.5.4 Each of the above required fire hose stations shown in Table 3.7.5.4-1 shall be demonstrated OPERA 8LE:

l a. At least once per 31 days by a visual inspection of the fire hose l

l t stations accessible during plant operation to assure all required

! equipment is at the station.

b. At least once per 18 months by:

1. Visual inspection of the fire hose stations not accessible during j plant operation to assure all required equipment is at the station.

l 2. Removingthehoseforinspectionandredeking,and l

3. Inspecting all gaskets and replacing any degraded gaskets in l

i

{. the couplings.

c. At least once per 3 years by partially cpening each hose station valve l

! to verify valve OPERASIL*TY and no f1:w bic cags.

d. Within 5 years and betseen 5 ana 8 years after purcnase cata ano at least every 2 year . thereafter by conducting a hose hydrostatic test at

! a pressure of 150 psig or at least 50 psig above the maximum fire main operating pressure, whichever is greater.

LA SALLE - UNIT 1 3/4 7-18

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f) PLANT SYSTEMS 3/4.7.6 FIRE RATED ASSSELIES .

I LIMITING CONDITION FOR OPERATION

addiag.

3.7.6 All fire rated assemb11esM11s, floor / ceilings, cable tray enclosures j and other fire barriergyseparating safety related fire areas or separating i portions of redundant systems important to safe shutdown within a fire area, and all sealing devices in fire rated assembly penetrations (fire doors, fire

windows, fire dampers, cable and piping penetration seals and ventilation

; seals) shall be 0.PERA8LE.

! APPLICA8ILITY: At all times.

!, $ .. ACTION;_ _

l i a. With one or more of the above required fire rated assembifes and/or sealing devices inoperable, within one hour either establish a con-tinuous fire watch on at least one side of the affected assembly (s)

- and/or device (s) or verify the OPERA 8ILITY of fire detectors on at i least one side of the inoperable assembly (s) and/or sealing device (s) -

and establish an hourly- fire watch patrol. Restore the inoperable fire rated assembly (s) and/or sealing device (s) to OPERA 8LE status

- qa.T within 7 days or, in lieu of any other report required by Specifica-tion 6.6.8, prepare and submit a Special Report to the Commission

' pursuant to Specification 6.6.C within the next 30 days outlining

' the action taken, the cause of the inops.rable fire rated assembly (s) and/or sealing device (s) and plans and schedule for restoring the t fire rated assembly (s) and/or sealing device (s) to OPERABLE status.

f !

' " b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS k

I

' 4.7.6.1 Each of the above required fire rated assembites and sealing devices shall be verified to be OPERA 8LE at least once per 18 months by performing a

visual inspection of: ,

a. The exposed surfaces of each fire rated assembifes.

b. Each fire window / fire damper and associated hardware.

c. At least 10 percent of each type, of sealed penetration. If apparent changes in ar.ce.arance or aancre ' -= ;n:2-icc3 are fcuri, a vis.ai inspecti.:n ci an additionai *0 . . --O ;f cs:

rc:e cf setiec penetration s.uil oe mace. Tais inspec. ion process snail contint.a O ' >N : -

until a 10 percent sample with no apparent changes in appearance or abnormal degradation is f.ound.

i a .

i LA SALLE - UNIT 1 3/4 7-22

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i 3f4,7.7 AREA TEMPERAW RE M

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ll l LIMIHNG CONDITION FOR OPERATION '

~

h n in Table 3.7.7-1 l'

\

l The temperature of each area of Unit 1 and Unft 2 s owd in Tab 3.7.7 is required to be shall be maintained Whenever the equipment in an affected area within the limits indicate APPLICABILITY:

l CPERA8LE.

M: Itait(s) shown in Table 3.7.7-1:

Specift- l-With one or more areas exceeding the temperaturehou

a. For cationmore than 6.6.8, prepare and submiti a Special Report 30 days p time the temperature pursuant to Specification 6.6.C within l isthe next to demonstrate .

record of the amount by which i ment. and the cumula 3 the continued CPERASILITY of the affected t required equabove, p

, J b.

By more than 30'F, in addition to theinoperable. Special Reporithin i within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months either restore the area to wlimit or 1 .

l SURVEILLANCE REQUIREMENTS

[ ,

uired areas shown in Table 3.7.

i 4.7.7 The temperature in each of the above reqleast once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

l shall be determined to be within its limit at I

J i.

I l

\ (-

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. - - - - . --- . ._ . _ . _ _ . . _ . _ _ . ._..3/4.7-24 _ ,,

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O l

!, TA8LE 3.7.7-1

. ! AREA TEMPERATURE MONITORING

,t j . .:, . . _ _ . . _

\ .

l: e TEMPERATURE LIMIT (*F) l:

tMVICED EQUIPMtNi AREA ,

NOT OPERATING OPERATING A. Unit 1 Area Temperature Monitorina

- SC

,6tf-104

1. Control Roos < 104

2. Auxiliary Electric Equipment Room h104 < 104 i

3. Diesel Generator Room 122 < 122

4. Switchgear Room M'104 < 104

~~

5. HPC5, LPC5, RHR & RCIC Rooms ,

150 < 150

6. Primary Containment

a. Orywell 150 < 150 ,

% 50

b. Beneath Reactor Pressure Vessel M185 < 185

. 1

8. Unit 2 Area Temeerature Monitorina Recuired For Unit 1 _l

~

1. Auxiliary Electric Equipment Room 104 < 104 t 2. Ofesel Generator 2A Room -

-122 < 122 l 3. Division 1 and 2 Switchgear Rooms -104 < 104-l l

l -

1

\

I

+

l

I l i 1

l l l

l I' l

i

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1 i

f LA SALLE - UNIT 1 3/4 7-25

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. 3/4.7.9 SNU88ERS l:

l

-.i ,

1 i LIMITI!M CONDITION FOR OPERATION 7f

~

o gg; Se packu; I i 3.7.9 All, snubbers *-' M ** N -- ? '. 5 1 d 0.7.3-2 shall be OPERA 8LE. l t

\ ITIONS 1, 2 and 3, and OPERATIONAL CONDITIONS 4 l lt APPLICA8ILITY: OPERATIONAL l ! and 5 for snubbers located on stems required OPERA 8LE in thoss OPERATIONAL

', CONDITIONS.

! ACTION:

With one or more' snubbers inoperahle, within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months replace or restore the

, inoperable snubber (s) to OPERA 8LE tatus and perform an engineering evaluation ~

' ! perSpecification4.7.94onthes ported component or dsclare the supported l l- system inoperable and follow the ap priare ACTION statement for that system.

} SURVEILLANCE REQUIREMENTS

. 4.7.9 Each snubber shall be demonstrat OPERA 8LE by performance of the following augmented inservice inspection program and the requirements of Specification 4.0.5.

a. Visual Inspection The first inservice visual inspec$1on of snubbers shall be performed

' after 4 months but within 10 months of commencing POWER OPERATION _

l and shall include all snubbers listed in Tables 3.7.9-1 and 3.7.9-2.

. If less than two snubbers are foun inoperable during the first i inservice visual inspection, the se nd inservice visual inspection i shall be performed 12 months 2 25% fbom the date of the first

! inspection. Othenvise, subsequent vi ual inspections shall be i performed in accordance with the foll ing schedule:

No. Inoperable Snubbers ubsequent Visual per Inspection Period nspection Period *#

I 0 ~

18 months t 25%

1 months 2 25%

2 months 2 25%

4 3, 4 12 days 2 25%

5,6,7 6 days 2 25%

8 or more 31 ays 2 25%

i The snubbers may be categori:ed i. to tue grcup : These accessib!a 1 - and thcu irs:cassible duri g n;; cr coentic.. Ea:r. grecc uy :e inspected inoepenoently in accorcance witn ene ove scneoule.

t "The inspection interval shall not be lengthened more than ne step at a time.

I t #The provisions of Specification 4.0.2 are not applicable.

LA SALLE - UNIT 1 3/4 7-27

n . . - , . - - , - , - - - - ,- - - _ - - - ,,,n-- - - . .,,,,ve-----,--- r -,---.,~~c. -

i-

'zescer ATrpenac P%E ]

O _1 ' PLANT SYSTEMS

~

. ! SURVEILLANCE REQUIR S(Continued)

.i

!1 b. Visual Inspection Ahenkance Criteria ,

! Visual inspections s 11 verify (1) that there are no visible indications of damage or impai that attachments to the foundation i or supporting structu OPERA are secure 8ILITY,-(2)d an (3) in those locations where

- snubber movement can b manually in,duced without disconnecting the snubber, ,

that the snubber has f edom of movement and is not frozen up. Snubbers H which appear inoperable a result of these visual inspections aay be determined OPERABLE for e purpose of establishing the next visual inspection interval pro!dingthat(1)thecauseoftherejectionis clearly established,and ri died for that particular snubber and for other snubbers that say be gene allysusceptible,and(2)theaffectedsnubber is functionally tested in is as found condition and determined OPERABLE per Surveillance Requiremen 4.7.9.d and 4.7.9.e as applicable. However, when a fluid part of a hydr lic snubber is found,to be_.uncavaced, the

' rable and cannot be determined OPERABLE by

, snubber shall be declared i -

functional testing for the p sose of 1stablishing the next visual inspection interval. All sn ers connected to an inoperable common hydraulic fluid reservoir shal be counted as inoperable snubbers.

c. Functional Tests -

~

f Ouring the first refueling shutd and at least once per 18 months t O . _,: therearter durias shutdo a. re9 sentative s pie of saubbers shati be functionally tested.

2 For hydraulic snubbers, a represen tive sample of at least 10% of the

total of hydraulic snubbers listed n Table 3.7.9-1 shall be functionally

- tested either in place or in a bene test. For each hydraulic snubber that does not meet the functional te t acceptance criteria of Surveil- -

lance Raquirement 4.7.9.d or 4.7.9.e, an additional 10% of the hydraulic i -

. snubbers shall be functionally tested.

! For mechanical snubbers, a representat ve sample of that number of mec anical snubbers listed in Table 3.7.9-2 which where c* is the allowable number of sec allowstheexpression35(1+nical snubbers I

I acceptance criteria selected by the oper tor shall be functionally l tested either in place or in a bench test foreachnumberofmechanical

snubbers above c which does not meet the nctional test acceptance j criteria of Specifications 4.7.9.d. or 4.7 9.e, an additional sample selectedaccordingtotheexpression35(1 {)(c J.) (a-c)shall where a is the total cumber of mechanical be functionally snubbers tested,le during the functio 1 testing of the found inoperab I representative saeple.

l' j Functionaltestingshallcontinueaccordingtotheexpression

!' b ( 35 (1 + j) (g)2] where b is the number of Techanical snubbers found inoperable in the previous re-sa cle until no 1 sitional iwparible l

'_~ .nachanical snucters are icW.a i;ain,_a sacpla cc r.Lil all cachanical snubbers in Table 3.7.9-1 and 3.7.9-2 have been f ctionally tested.

C=2 l

LA SALLE - UNIT 1 3/47-28

. . ... . .:. ~ .- . .

. .:._ _ . v .

! i I D. '

~ ## ##

PUNT SYSTEMS SUINEILLANCE REQUI 5 (Continued) .

l '-

'., ~ ~ ~ ~ ~ Functional Tests (conti ) ,

lhe representati sample selected for fucctional testing shall R include the vario configurations, operating environments and the )

range of size and apacity of draulic and mechanical snubbers. i At least 25% of snubbers i the representative samples shall include snubbers f the following three categories:

L The first her away from each reactor vessel nozzle.

2. Each snubber within 5 feet of heavy equipment (valve, !

,. pump, turbine actor, etc.). l

~

3. Each snubber w thin 10 feet of tfie discharge from a safety relief valve. _

- snubbers which failed the previous i In addition test functional to the regular shall be reample,during sted the next test period. If.a .

' spara snubber has been insta led in place of a failed snubber, then ,

both the failed snubber, if i is repaired and installed in another 1 O n. position, and the spara snubb shall be retested. Test results of for the re-sampling.

j these snubbers say not be incl )

l l

If any snubber selected for f ional testing either fails to j l -

lockup or fails to move, i.e., f zen in place, the cause will be ,

1 1 evaluated and if caused by manufa turer or design deficienc' all snubbers.ofthesamedesignsubje to the same defect shal be functionally tested. This testing equiremant shall be independent 1 of the requirements stated above fo snubbers not meeting the j

functional test acceptaitce criteria. .

! For any snubber (s) found incperable, engineering evaluation shall l be performed on the components which a supportedbythesnubber(s).

The purpose of this engineering evaluat n shall be to deter:nine if

, the components supported by the snubber ( ) were adversely affected by the inoperability of snubber (s) in or r to ensure that the supported component remains capable of me ing tne designed service.

d. Hydraulic Snubbers Functional Test Acceatan e Critaria i

! The hydraulic snubber functional test shall rify that:

l i . 1. Activation (restraining action) is achiev d witnin the specified range of velocity cr se:eWition in bet 5 ansi:n ird c 0.as5 : .

l

2. Snubber blaec, or raiessa cta, ..ncre es;.n ad, is .::.* :..2 or snubbers specifi-specified range in compression or tension.

Q' call required to not displace under continuo s load, the abil ty of the snubber to withstand load with t displacement shall be verified.

LA SALLE - UNIT 1 3/4 7-29

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! ,* SURVEILLANCEREQUIREMENTS(Continued (

! e. Mechanical Snubbers Functionak Test Acceptance Criteria The mechanical snubber functio I test shall verify that:

L The force that initiates f movement of the snubber rod in

.Ii either tension or compressio is less than the specified maximum breakaway friction drag force Breakaway friction drag force shall not have increased more than 5 since the last surveillance test.

2. Activation (restrainingaction) s achieved within the specified range of velocity or acceleratto in bsth tension and compression.

3. Snubber release rate, where requi , is within the specified range

} in c ression or tension. For sn bers specifically required not .

to dis lace under continuousr load, e ability of the snubber to withs nd load without displacesant hall be_ verified.

j f. Snubber Service Life Monitoring A record of the se nice life of each snubber the date at which the m designated service life commences and the ins llation and maintenance

j records on which the designated service life i based shall be

, saintained as required by Specification 6.5.8.. .

Concurrent with the first inservice visual inspe tion and at least once per 18 months thereafter, the installation a d maintenance -

records for each snubber listed in Tables 3.7.9-1 nd 3.7.9-2 i shall be reviewed to verify that the indicated serv *ce life has not l

been exceeded or will not be exceeded prior to the t scheduled i snubber service life review. If the indicated servi life will be i exceeded prior to the next scheduled snubber service ife re/iew l the snubber service life shall be reevaluated or the s ubber shall i

be replaced or reconditioned so as to extend its servic life beyond i the date of the next scheduled service life review. Thi reevaluation, replacement or reconditioning shall be indicated in the cords.

l s

LA SALLE - UNIT 1 3/4 7-30

_.'....-,-y.-,.,_-,-,.

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

i . ___ .

.. . -- - --. . - : . . . . . . ~ . . . . . . - - . . - - . . - - - ---

i O. .

.. - 3 O..

,l TABLE 3.7.9-1

! 5 SAFETY RELATED HYORAULIC SNUB 8ERS*

C E HIGi RADIATION i . SNUBBER SYSTEM SNUB 8ER INSTALLED ACCESSIBLE OR ZONE ESPECIALLY DIFFICULT '

c- NO. ON, LOCATION AND ELEVATION INACCESSIBLE DURING SiluT00WN** 10 REMOVE H

(A or I) (Yes or No) (Yes or No)

None ,

i I

M.

Y t .

l

?! .

t .

t '

' 5niEhers n may be added to safet ryte' e systems without prior License Amendment to Table 3.7.9-1

, provided that a revision to T lt 3.7.9-1 is included with the next License Amendment request. .

l ** Modification to t yhi6 unn due to cha'nges in high radiation areas may be made without prior .

License Amen at provided that a revision to Table 3 7.9-1 is in91uded with the next License Amende nt, re uest. ,

l

.I i

. [

i.

. e

.1 PLANT SYSTEMS 7.

l. ". 3/4.7.9 SNUBBERS . -

LIMITING CONDITION FOR OPERATION

! 3.7.9 All hydraulic and mechanical snubbers shall be OPERABLE.

i APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3. OPERATIONAL CONDITIONS 4 and 5 for snubbers located on systems required OPERABLE in those OPERATIONAL CONDITIONS.

ACTION:

With one or more snubbers inoperable on any system, within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months replace or restore the inoperable snubber (s) to OPERABLE status and perform an engineering evaluation per Specification 4.7.99. on the attached component or declare the attached ' system inoperable and follow the appropriate ACTION statement for that system.

SURVEILLANCE REQUIREMENTS 4.7.9 Each snubber shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program and the requirements of Specification 4.0.5.

a. Inspection Types l

l As used in this specification, type of snubber shall mean snubbers of the same design and manu~facturer, irrespective of capacity.

' h V' fC b. Visual Inspections Snubbers are categorized as inaccessible or accessible during reactor operation. Each of these groups (inaccessible and accessible) may be

inspected independently according to the schedule below. The first l

inservice visual inspection of each type of snubber shall be performed l

after 4 months but within 10 months of commencing POWER OPERATION and shall include all hydraulic and mechanical snubbers. If all snubbers

l. of each type on any system are found OPERABLE during the 4efes inser-vice visual inspection, the second inservice visual inspectioh of thatg system shall be performed at the first refueling outage. Otherwise, subsequent visual inspections of a given system shall be performed in accordance with the following schedule:

l No. Inoperable Snubbers of Each Type Subsequent Visual On Anv System oer Insoection Period Insoection Period" #

0 18 montns 2 25%

1 12 months 25%

2 6 months 2 25%

3,4 124 days 25%

5,6,7 62 days 25%

6 or more 31 days ! 25%

l l ,

"The inspection interval for each type of snubber on a given system shall not l be lengthened more than one step at a time unless a generic problem has been

/I identified and corrected; in that event the inspection interval may be N lengthened one step the first time and two steps thereafter if no inoperable l

U' snubbers of that type are found on that system.

The provisions of Specification 4.0.2 are not applicable.

LA SALLE - UNITA [ 3/4 7,,2ff 91

t l PLANT SYSTEMS

( SURVEILLANCE REQUIREMENTS (Continued) ,

c. Visual Inspection Acceptance Criteria

,; Visual inspections shall verify that: there are no visible indica-

i tions of damage or impairso OPERABILITY and (2) attachments to the

foundation or supporting structure are secure, and (3) fasteners for attachment of the snubber to the componert and to the snubber anchorage are secure. Snubbers which appear inoperable as a result

> of visual inspections may be determined OPERABLE for the purpose gf

} establishing the next visual inspection interval, provided that: ,"

j (1) the cruse of the rejection is clearly established and remedied for that particular snubber and for other snubbers irrespective of type on that system that may be generically susceptible; and (2) the I affected snubber is functionally tested in the as-found condition and detenmined OPERABLE per Specification 4.7.9f. All snubbers connected to an inoperable common hydraulic fluid reservoir shall be counted as inoperable snubbers. For those snubbers common to more than one system, the OPERABILITY of such snubbers shall be considered in !

assessing the surveillance schedule for each of the related systems. J I

d. Transient Event Inspection

- Aninspectionshallbeperformedchallhydraulicandmechanical '

snubbers attached to sections of systems that have experienced I l

l une2pected, potentially d.usaging transients as determined from a !

t review of operational data and a visual inspection of the systems I

[g within 6 months following such an event. In addition to satisfy-ing the visual inspection acceptance s criteria, freedom-of-motion of l

mechanical snubbers shall be verified using at least one of the 1 following: (1) manually induced snubber movement; or (2) evaluation l

of in place snubber piston setting; or (3) stroking the mechanical :

snubber through its full range of travel.

e. Functional Tests -l During the first refueling shutdown and at least once per 18 months thereafter during shutdown, a representative sample of snubbers shall be tested using one of the following sample plans. The sample plan ,

l shall be selected prior to the test period and cannot be changad during the test period. The NRC Regional Administrator shall be notified in writing of the sample plan selected prior to the test period or the sample plan used in the prior test period shall be implemented: ;

1) At least 10% of the total of each type of snubber shall be !

functionally tested either in place or in a bench test. For I each snubber of a type that does not meet the fuactional test acceptance criteria of Specification 4.7.9f. , aa additional 10%

of that type of snubber shall be functionally tested until no i

l more failures are found or until all snubbers of that type have been functionally tested; or A representative sample of each type of snubber shall be func-f 2) tionally tested, in accordance with Figure 4.7-1. "C" is the i

LASALLE-UNITg{ 5 3/47-)98 ,

1 1

l

5

, PLANT SYSTEMS D~

d SURVEILLANCE REQUIREMENTS (Continued) 3

e. Functional Tests (Continued) total number of snubbers of a type found not meeting the accep- -

tance requirements of Specification 4.7.9f. The cumulative number of snubbers of a type tested is denoted by "N". At the i and of each day's testing, the new values of "N" and "C" (pre-vious day's total plus current day's increments) shall be plotted on Figure 4.7-1. If at any time the point plotted

. falls in the " Reject" region, all snubbers c.f that type may be functionally tested. If at any time the point plotted falls in the " Accept" region, testing of snubbers of that type may be teminated. When the point plotted lies in the " Continue Testing" region, additional snubbers of that type shall be tested until the point falls in the " Accept" region or the

" Reject" region, or all the snubbers of that type have been tested. Testing equipment failure during functional testing may invalidate that day's testing and allow that day's testing to resume anew at a later time provided all snubbers tested with the failed equipment during the day of equipment failure are ratested; or

3) An initial representative sample of 55 snubbers shall be func-tionally tested. For each snubber type which does not meet the l

functional test acceptance criteria, another sample of at least one-half the size of the initial sample shall be tested until h.' the total number tested is equal to the initial sample size multiplied by the factor,1 + C/2, where "C" is the number of snubbers found which do not meet the functional test acceptance criteria. The results from this sample plan shall be plotted using an " Accept" line which follows the equation N = 55(1 +

C/2). Each snubber point should be plotted as soon as the snubber is tested. If the point plotted falls on or below the

" Accept" line, testing of that type of snubber may be terminated.

If the point plotted falls above the " Accept" line, testing must continue until the point falls in the " Accept" region or all the l snubbers of that type have been tested.

l l The representative sample selected for the functional test sample plans shall be randomly selected from the s1ubbers of each type and reviewed before beginning the testing. The review shall ensure, as far as practicable, that they are representative of the various con-figurations, operating environments, range of size, and capacity of snubbers of each type. Snubbers placed in the same 1ccation as i

snubbers which failed the previous functional test shall be ratested

! at the time of the next functional test but shall not be included in the sample plan. If during the functional testing, additional sampling is required due to failure of only one type of snubber, the func-tional test result!. shall be reviewed at that time to determine if additional samples should be limited to the type of snubber which has failed the functional testi.ng. .

l O-LASALLE-UNIT /l 3/4 7-

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

k

, . . PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) ,

f. Functional Test Acceptance Criteria

'; The snubber functional test shall verify that:

1) Activation (restraining action) is achieved within the specified range in both tension and compression;

2) Snubber bleed, or release rate whers required, is present in both tension and compression, within the specified range; l

l' 3) Where required, the force required to initiate or maintain

! - motion of the snubber is within the specified range in both directions of travel; and

! 4) For snubbers specifically required not to displace under continueus load, the ability of the snubber to withstand load

'4 without displacement.

$ Testing methods may be used to measure parameters indirectly or l parameters other than those specified if those results can be correlated to the specified parameters through established methods.

g. Functional Test Failure Analysis An engineering evaluation shall be made of each failure to meet the functional test acceptance criteria to determine the cause of the h, _ - failure. The results of this evaluation shall be used, if applicable, U in selecting snubbers to be tested in an effort to determine the

.i OPERABILITY of other snubbers irrespective of type which may be subject to the same failure mode.

For the snubbers found inoperable, an engineering evaluation shall -

be performed on the components to which the inoperable snubbers are -

attached. The purpose of this engineering evaluation shall be to t determine if the components to which the inoperable snubbers are attached were adversely affected by the inoperability of the snubbers in crder to ensure that the component remains capable of meeting the l designed service. l If any snubber selected for functional testing either fails to lock up or fails to move, i.e., frozen in place, the cause will be evaluated and, if caused by manufacturer or design deficiency, all snubbers of the same type subject to the same defect shall be func-tionally tested. This testing requirement shall be independent of the requirements stated in Specification 4.7.9e. for snubbers not meeting the functional test acceptance criteria.

]

h. Functional Testino of Repaired and Replaced Snubbers Snubbers which fail the visual inspection or the functional test acceptance criteria shall be repaired or replaced. Replacement snubbers and snubbers which have repairs which might affect the functional test results shall be tested to meet the functional test LASALLE-UNITyl 3/4 7-Jr 30

,v y m--< , --w,,.. ,.,,,m-m--,,

...._.,..m..___.__ . .. . . . . . . . -. . . . _ , . ..

\

.i k

(, ,_ PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

,' h. Functional Testino of Repaired and Replaced Snubbers (Continued) -

4

, criteria before installation in the unit. Mechanical ~ snubbers shall

., have met the acceptance criteria subsequent to their most recent ser-

, vice, and the freedom-of-motion test must have been performed within l 12 months before being installed in the unit.

i. Snubber Service Life Program l/

The service life of hydraulic and mechanical snubbers shall be monitored to ensure that the service life is not exceeded between surveillance inspections. The maximum expected service life for various seals, springs, and other critical parts shall be deter-

~ ~,

mined and established based on engineering information and shall be extended or shortened based on monitored test results and fail- '

ure history. Critical parts shall be replaced so that the maximum service life will not be exceeded during a period when the snubber is required to be OPERABLE. The parts replacements shall be docu- ,

mented and the documentation shall be retained in accordance with Specification 6.58.

[ -

l LASALLE-UNITgg' 3/4 7,X si

. . . . . . . _ . . . . ._s.__,___. . u.. - _ _ . . .

l a

(.

t .

l 5

.m ?-

I, s .: - -

. s.-

l 10 -

ll 9 8

7 Y

l REJECT i ,

/

k S.s' 1'

f b' 4 g

b 6 CONTINUE 3

2 TESTING , [

2

/ jt M7 ~

ACCEPT 1 f l

J 0 10 20 30 40 50 60 70 80 90 100 l

l N !

l FIGURE 4.7-1 i SAMPLING PLAN FOR SNUBBER FUNCTIONAL TEST l LASALLE-UNITf[ 3/4 7 ,25' 32

a e

Table 3.7.9-2 Safety Related Mechanical Snubbers ,

j t~

Snubber System snubber installed on Snubber Systaa sr.ubber installed No. ocation and elev. No. on location and elev.

DG15-10055 DG Reactor 704 FWO2-n735 FW Containment 779

,! DG18-00115 DG Diesel Generator 746 N02- H745 FW Containment 779 l DG13-00145 OG Diesel Generator 735 FWO2- n755 FW Containment 792

! DG34-10025 DG Diesel Generator 734 FWO2- n765 FW Auxiliary 714

, DG34-10035 DG iesel Generator 734 FW u-10035 FW Auxiliary 753

'~

DG34-10045 DG D esel Generator 745 N H-19045 FW Auxiliary 753 DG34-1005S DG Di el Generator 745 FWu-10055 FW Auxiliary 754 i, DG34-18005 DG Die 1 Generator 740 FWu-1006S FW Auxiliary 754 FWO2-10375 FW Conta neent 792 NG01-10035 HG Reactor 799

, FWO2-10435 N Contai nt 792 HG01-10055 HG Reactor 797 N02-10445 FW Contai t 787 HG01-10065 HG Reactor 797 FW2-10455 FW Contai t 779 HG01-10095 HG Reactor 797 FWO2-10465 FW Containes 779 HG01-10105 HG Reactor 797 l' FWO2-10475 FW Containment 779 HG01-10465 HG Contaiceent 803 FWO2-10505 FW Containment 792 NG02-20165 HG Reactor 798 FWO2-10555 'W containment 755 NG02-20225 HG Reactor 798 FWO2-10605 FW Auxiliary 744 HG02-20275 HG Rsactor 798 FWO2-10715 FW Containment 778 NG02-20315 HG Reactor 798 FWO2-11335 FW Auxiliary 743 HG05-10165 HG Reactor 756 i FWO2-n435 FW Containment HG05-10255 HG Reactor 772

! FW Containment 7 HG05-10295 HG Reactor 782 Q<'WO2-11455 102- n465 FW Containment 7 HG05-1041S HG Reactor 734 rWO2-H485 FW Containment 784 HG06-1046S HG Reactor 757 FW02- n495 FW Containment 778 -10915 HG Reactor 726 FWO2-11505 FW Containment 778 -10985 HG Reactor 737 FWO2- n525 FW Containment 785 2105S HG Reactor 772 FWO2-n535 FW Containment 785 21095 HG Reactor 769 f Reactor 675 I FWO2-n545 FW Containment 780 HP0 075 HP I FWO2-n555 FW Containment 780 HPol- 175 HP Reactor 675 FW02-n565 N Containment 783 HP02-1 HP Reactor 685 FWO2-n575 FW Containment 777 HP02-10 HP Reactcr 764 FWO2-11595 FW Containment 758 HP02-1019 HP Reactor 735 FWO2-n695 FW Containment 780 HP02-10205 HP Reactor 690 FW02-n70S FW Containment 754 HP02-10215 HP Reactor 685 FWO2-n715 FW Containment 791 HP02-10255 HP Containment 774 FWO2- n725 FW Containment 792 HP02-1026$ HP Containment 789 HP02-10275 HP Containment 774 LC01-1024S LC Auxiliary 690 HP02-15005 HP Containment 789 LC01-10265 Auxiliary 690 HP02-15025 HP Reactor 753 LC01-10315 L Auxiliary 690 HP02-15035 HP Reactor 753 LC01-10475 LC Auxiliary 682 HP02-1507S HP Containment 774 LC01-10505 LC Auxiliary 681 HP02-15085 HP Containment 774 LC01-10515 LC Auxiliary 681 HP02-15105 HP Containment 774 LC01-10525 LC Auxiliary 681 i

HP02-15135 HP Reactor 690 LC01-10535 LC uxiliary 682 HP02-15195 HP Reactor 768 LC01-10545 LC A iliary 682 HP02-15205 hP Reactor 769 LC01-10555 LC Au *11ary 681 802-1523S HP Turbine 708 LC01-10565 LC Auxi fary 682 LA Salle - UNIT 1 3/4 7-32 Ng Amendment % I i

- .~ .

Table 3.7.9-2 (Continuad) i 0;

\ \ .

- Snubber System sadber installed on Snumber System snubber installed No. location and elev. No. on location and elev.

a

~HP02-15245 Containment 690 LC01-10575 LC Auxiliary 681 i HP32-18005~ Reactor 686 LC01-10585 LC Auxiliary 681 HP08-10095 Reactor 704 LC01-10595 LC Auxiliary 706 HP08-10245 HP Reactor 699 LC01-10655 LC Auxiliary 722

{2 -- J1R08-10265_.____ HP Reactor 700 LC01-10665 LC Auxiliary 706 HP08-10285 HP Reactor . 699 LC01-10685 LC Auxiliary 722 HP08-10295 HP ntainment 703 LC01-10695 LC Auxiliary 722 HP09-10135 HP actor 688 LC01-10705 LC Auxiliary 721 732 p -NP09-10365 ---HP R r 693 LC01-10895 LC Auxiliary HP14-10035 HP Re r 681 LC01-10965 LC Auxiliary 733 HP22-10035 HP Reac r 768 LC01-11065 LC Auxiliary 731 HP75-10035 HP Oiesel nerator 735 LC01-11205 LC Auxiliary 732 HP75-10055 HP Diesel nerator 746 LC01-11215 LC Auxiliary 732 l; 732 LC01-1002S LC Auxilia 707 LC01-11235 LC Auxiliary ll Auxiliary 707 LC01-11265 LC Auxiliary 732 p LC01-1003S LC LC01-10045 LC Auxiliary 706 LC01-18075 LC Reactor 680 LC01-10055 LC Auxiliary 706 LC01-18085 LC Reactor 680 LC01-10075 LC Auxiliary 707 LC01-18095 LC Reactor 680 LC01-10085 LC Auxiliary 707 LC01-18105 LC Reactor 680 LC01-10135 LC Auxiliary 706 LC03-18175 LC Reactor 680 LC01-10145 LC Auxiliary 06 LC03-18205 LC Reactor 680 LC01-10155 LC Auxiliary LC03-18235 LC Reactor 680 Auxiliary 7 LC03-18265 LC Reactor 680 0LC01-10195 LC Auxiliary 681 LC09-10015 LC Reactor 681 mC01-10215 LC Auxiliary 706 LC09-10135 LC Auxiliary 679 LC01-10225 LC Auxil'iary 681 00-10155 MS Containment 780 LC09-10185 LC Auxiliary 702 M500-10165 MS Containment 780 l LC09-10315 LC Auxiliary 702 MS 10175 MS Containment 775

l. LC09-10365 LC LC09-10495 LC Auxiliary 681 MS00 0195 MS Containment 763

l. Containment 743 i LC09-1050S LC Auxiliary 681 MS00- 0205 MS Auxiliary 681 MS00-10215 MS Containment 743 LC09-10525 LC Auxiliary 737 M500-10225 MS Containment 743 LC09-10535 LC Reactor 681 MS00-1023 MS Containment 769 LP01-10145 LP LP02-10135 LP Containment 778 M500-10245 MS Containment 786 LP02-10155 LP Containment 774 MS00-1025S MS Containment 787 Containment 774 MS00-10265 MS Containment 780 LP02-10165 LP Containment 774 MS00-10295 MS Containment 743 LP02-10175 LP LP02-10185 LP Containment 774 M500-1030S MS Containment 772 Containment 774 MS00-10315 MS Containment 785 LP02-10195 LP Reactor 777 MS00-10325 M Containment 790 LP02-10205 LP Reacter 703 M500-10335 MS Containment 790 LP02-10255 LP Reactor 777 MS00-10345 MS Containment 780 LP02-10545 LP 777 M500-1038S MS Containment 743 LP02-10555 LP Reactor Reactor 777 H500-10395 MS Containment 743 LP02-10575 LP Reactor 777 M500-10405 MS Cqntainment 772 LP02-10595 LP Reactor 689 MS00-10415 MS Cogtainment 786 LP02-1062S LP 790 LP02-1067S LP Containment 788 MS00-10423 MS Containment L LP Reactor 763 MS00-10435 MS Cont nment 790 Reactor 733 MS00-10445 MS Contal ment 780 h .P19-1011S P20-1025S LF 3/4 7-33 Amendment 1 LA SALLE - UNIT 1

l _ _ . . . . _

i Table 3.7.9-2 (Continued) .

lI .

System snubber installed ~

l

' (( ). nubber System snubber installed on location and elev.

Snubber No. on location and elev.

t4 785 LP20-10285 Reactor 728 MS00-10455 MS Containment l 143

LP20-1029S Reactor 727 MS00-10495 MS Containment Reactor 726 M500-10505 MS Containment 737 LP20-10305 LP 769 LP20-10355 LP Reactor 767 MS00-10515 MS Containment Containment 775 M500-10525 MS Containment 775 MS00-10065 MS Containment 780 MS00-10535 MS Containment 787

. MS00-10075 MS 748 MS00-10545 MS Containment 786 i M500-10095 MS tainment inment 774 M500-10555 MS Containment 780 MS00-10105 MS inment 780 M501-12735 MS Auxiliary 691 MS00-10115 MS 760 MS00-1012S MS Con nment 780 MSO4-n355 MS Containment 780 MS04- u375 MS Containment 771

+

M500-10135 MS Conta nt 78; M500-10145 MS Contal nt 774 MS04- n385 MS Containment 781 MSO4- n925 MS Containment 760 MSO4-11395 MS Contai t 758 MSO4- n935 MS Containment 751 M504-n415 MS Contai 758 MS04- n945 MS Containment 766 MS04-11425 MS Containment 775 MS04-n955 MS Containment 779 M504-n435 MS Containment 779 MS04- n965 MS Containment. 779 M504- n455 MS Containment 779 MSO4-n975 MS Containment 778 M504-11475 MS Containment 748 M504-12015 MS Containment 753 MSO4- n505 MS Containment 770 MSO4-12025 MS Containment 755 M504-n515 MS Containment 0 M504-12035 HS Containment 760 MSO4-11525 MS Containment M504-12045 MS Containment 781 M504-u535 MS Containment KSO4-1205S MS Containment 783 MS Containment OMSO4-n545 781 MS34-12075 MS Containment 783

'SO4-n55S MS Containment Containment 760 -12105 MS Containment 783

'.SO4-u575 MS 760 l

MSO4- n585 MS Containment 773 04-12nS MS Containment l

MS Containment 773 -12125 MS Containment 757p MSO4-11595 758 MSO4-11615 MS Containment 781 MSC -12135 MS Containment i

748 M504 12155 MS Containment 748 1 l MSC4-n635 MS Containment 762 M504-12165 MS Containment 756 MSO4- n655 MS Containment 763 M504-12175 MS Containment 748 MSO4-n665 MS Containment 749 M504-1218S MS Containment 764 MSO4-n695 MS Containment 751 MS04-12195\ MS Containment 756 M504-n705 MS Containment 766 MSO4-12205 \ MS Containment 757 M504-H71S MS Containment 766 MSO4-12215 \\ MS Containment 764 M504- u725 MS Containment 774 MSO4-12225 MS Containment 772 MSO4-n735 MS Containment 772 MSO4-12235 \MS Containment 751 MS04-1174S MS Containment 774 M504-12255 'MS Containment 750 ;

M504-n755 MS Containment 758 MSO4-12295 MS Containment 779 MSO4-n77S MS Containment 779 Containment 761 MSO4-12305 MS\ Containment M504-n785 MS MSO4-n795 MS Containment 761 MSO4-12325 MS \ Containment 746 746 MSO4-11805 MS Containment 778 MSO4-12335 MS \ containment 770 Containment 749 MSO4-12345 MS MSO4-u83S MS NContainment 770 MS Containment 772 MSO4-1235S MS Containment MSO4- n845 C'entainment 780 Containment 773 M504-12375 MS MS04-n855 MS MS Containment 779 :

MSO4-n865 MS Containment 772 MSO4-12385 781 M504-12395 MS Containment 747 MSO4-n895 MS Containment 747 Containment 781 MSO4-12405 MS Conta'inment MSO4- n905 MS Contai'neent 744 Cont 11nnent 747 MSO4-12955 MS MSO4-1241S MS LA Salle - UNIT 1 3/4 7-34 Amendment i I l

l _ . _ -

> e g

(

Table 3.7.9-2 (Continued) g

~

ubber System snubber installed on Snubber System snubber installed -

jj '-

on location and elev.

location and elev. No.

Containment 780 MS04-12985 MS Containment 761

[ M504 M504- 465 455 MS MS Containment 779 MS04-12995 MS Containment 770 MS04-12

MS Con.tainment 752 MS04-13005 MS Containment 772

.'li M504-124 MS Containment 752 'M504-13015 MS Containment 772 781

, MSO4-12505 MS Containment 771 MSO4-13025 MS Containment Containment 771 MS04-1?04S MS Containment 781 M504-12515 MS Containment 771 M504-13055 MS Containment 774

M504-12525 MS Containment 778 M504-13075 MS Containment 771

M504-12535 MS Containment 766 M504-13085 MS Containment 773

'MSO4-12565 Containment 761 MSO4-1320S MS Containment 746 M504-12585 Containment 780 M504-13215 MS Containment 768 M504-12595 MS Containment 774 MSO4-13225 MS Containeent 766

,i MSO4-12615 MS Containment 748 MSO4-1323S MS Containment 766 MSO4-12625 MS

( ,; MSO4-12655 MS ntainment 745 MSO4-1324S MS MS Containment Containment 749 766 h

MS04-12665 MS tainment 764 MSO4-1325S Co inment 767 MSO4-13285 MS Containment 751 MSO4-12675 MS Con inment 780 MSO4-13295 MS Containment 772 M504-12695 MS Conta nt 781 MSO4-13305 MS Containment 772 l MS04-12705 MS Contai nt 747 M504-13315 MS Containment 774 MSO4-12715 MS 747 M504-1332S MS Containmen?. 745 MSO4-12725 MS Contai t Containee 769 MSO4-13335 MS Containment 748 MSO4-12745 MS 769 HS04-13345 MS Containment 774 MS Containment 772 MSO4-13375 MS' Containment 749 Containment OMS04-12755 4504-12765

- MS04-12775 MS MS Containment 773 MSO4-13415 MS Containment 748 782 MSO4-12785 MS Containment 773 MSO4-13445 MS Containment 780 MSO4-13455 MS Containment 743 MSO4-12795 MS Containment 3 M504-13465 MS Containment 743 M504-12815 MS Containment MSO4-13475 MS Containment 757 M504-12825 MS Containment -

75 MSO4-13495 MS Containment 742 M504-12835 MS Containment 770 MSO4-1350S MS Containment 759 M504-12885 MS Containment 773 MSO4-13515 MS Containment 758 M504-12895 MS Containment 772 ; 04-1353S MS Containment 749 M504-12905 MS Containment 780 MS -13545 MS Containment 749 MS04-12915 MS Containment 781 MSO 13555 MS Containment 759 M504-12935 MS Containment 744 MSO4- 565 MS Containment 766 M504-12945 MS Containment 775 MSO4-1 S MS Containment 760 M504-13585 MS Containment 780 M504-154 MS Containment 760 M504-13595 MS Containment 760 Containment 751 MSO4-1550 MS Containment MSO4-1362S MS 781 MS Containment 759 MSO4-15515 MS Containment MSO4-1363S 748 MS Containment 759 M504-1553S MS Containment MSO4-1364S 769 MS Containment 757 MSO4-16835 MS Containment MSO4-13665 751

'~

Containment 775 MSO4-17055 MS Containment MSO4-13675 MS 810 Containment 743 M505-10025 S Containment M504-1368S MS Containment 804 Containment 779 MS05-10045 M M504-13695 MS MS Centainment 804 MSO4-13715 MS Containment 783 M505-10055 M505-10085 MS Containment 304 MSO4-1372S MS Containment 749 ,

MS05-10105 MS Containment 804 M504-1373S MS Containment 766 M506-10065 MS ontainment 737 MSO4-13765 MS Containment 745 MS06-10165 MS ntainment 776 A504-13785 MS Containment 761 i

M506-10175 MS Co tainment 779 MSO4-13795 MS Containment 760 3/4 7-15 Amen ent 1 LA SALLE - UNIT 1 i

_. -- . ~_ . -

l

,i Table 3.7.9-2 (Continued) (-

h' Snubber

~

j System snubber installed on Snubber System snubber installed ~

j No. location and elev. No. on beation and elev.

M504-13845 MS Containment 749 MS10-10025 MS Containment 761 MSO4-13905 MS Containment 783 M510-10075 MS Containment 761 M504-13915 MS Containment 783 M510-10135 MS Containment 759

) Containment 738

j M504-13925 Containment 783 MS14-10305 MS

!i MS04-13935 Containment 783 MS14-10315 MS Containment 738 1 MSO4-13945 MS Containment 751 MS14-10345 MS Containment 738

,j M504-13955 MS Containment 751 MS14-10375 MS Containment 739

!; MSO4-13965 MS Ccntainment 747 M514-10385 MS Containment 738 MSO4-13975 MS ontainment 744 M514-10395 MS Containment 739 MSO4-13985 MS ntainment 758 M514-10445 MS Containment 739 MSO4-13995 MS Co inment 760 M514-10475 MS Containment 740 M504-15025 MS inment 760 M514-10485 MS Containment 741 M504-1503S MS Conta nt 763 MS14-1050S MS Containment 739 l- MSO4-15045 MS Contai nt 771 MS14-10515 MS Containment 740 MSO4-15055 MS Contai nt 770 M514-10525 MS Containment 741 Containee 758 MS14-10545 MS Containment 741 MS04-15065 MS Containeen 748 MS14-10555 MS Containment 741 MSO4-15085 MS

-i MS Containment 756 M514-10565 MS Containmen* 741 MS04-15095 MS Containment 749 MS14-10585 MS Containment 741 i MSO4-15105 Containment 760 MS14-10595 MS Containment 741 ft MS04-15115 MS Containment 761 MS14-10635 MS Containment 739 i

. MSO4-15135 MS '

Containment 741 N815-10025 NB Containment 830 MS14-10665 MS N815-10055 N8 Containment' 828 Q., A525-10235 MS Auxiliary NB15-10085 N816-10025 N8 N8 Containment Containment 827 810 i

l MS25-10615 MS Auxiliary 68 MS Containment 787 N816-10055 M8 Containment 810 l t MS88-1005S Containment 787 16-1006S N8 Containment 809 MS88-10065 MS Reactor 740 3-10035 M8 Containment 808 -

MS88-10115 MS -

Reactor 740 NB -1002S N8 Containment 808 MS88-1012S MS Reactor 740 PC01 0145 PC Reactor 741 l MS88-10135 MS MS Reactor 740 PC01- 155 PC Reactor 741 l MS88-1015S MS Reactor 740 PC01-1 65 PC Reactor 741 MS88-10205 MS Containment 761 PC01-101 PC Reactor 741 l MSC6-10035 747 1

MS Reactor 761 PC01-1019 PC Reactor MSC6-10045 747 Containment 760 PC01-1020S PC Reactor MSC6-10055 MS 751 MS Reactor 761 PC01-18005 PC Reactor MSC6-1006S 795 761 RG10-0014S RG Auxiliary MSC6-10095 MS Containment MS Containment 760 RG21-00125 RG Auxiliary 797 MSC6-10135 675 MS Containment 760 RH01-10055 H Reactor MSC6-10155 675 MS Containment 761 RH01-10065 R Reactor MSC6-10165 681 MS Containment 760 RH01-10085 RH Reactor MSCS-1018S 681 MS Containment 760 RH01-10135 RH Reactor MSC6-1021S 681 MS Containment 761 RH01-10175 RH Reactor MSC6-1024S 681 MS Containment 787 RH01-10185 RH Reactor l MS01-10015 actor 678 Containment 785 RH01-10255 RH 1 MSF9-10025 MS 705 Containment 784 RH02-1010S RH R ctor MSF9-10045 MS MS Containment 787 RH02-1012S RH Rea tor 636 MSF9-10065 h .

3/4 7-36 Amendment 1 LA SALLE - UNIT 1 l

^

.*s .*

Table 3.7.9-2 (Continued) i i'l' O.'$nubber \ System snutter installed on Snubber System snubber installed

~~

' No. gl ocation and elev. No. on location and elev.

[ ,

HSF9-10075 Containment 788 RH02-10175 RH Reactor 732 l* Reactor 686 NB11-10035 M Containment 808 RH02-10185 RH i N813-10015 NB Containment ' 832 RH02-1019S RH Reactor 686 NB13-10025 W Containment 832 RH02-10245 RH Reactor 735 W Containment 829 RH02-10255 RH Re&ctor 695 213-10045 M813-1006S 2 inment 828 RH02-10265 RH Reactor 703 m Reactor 695

,i NB13-10255 inment 813 RH02-10275 RH Reactor 735

' M813-10275 NB Co inment 811 RH02-10435 RH Con inment 814 RH02-10475 RH Reactor 725 i N813-10285 NB M Conta nt 811 RH02-10485 RH Reactor 722 NB13-10315 Reacto 732 RH03-15175 RH Containment 741 RH02-10515 RH Reactor 712 RH03-15245 RH Reactor 733 RMC2-10525 RH 710 RH03-15255 RH Reactor 736 RH02-10565 RH Reactor

- RH02-10575 RH Reactor 696 RH03-15265 RH Reactor 736 700 RH02-10585 RH Reactor 686 RH03-1527S RH Reactor Reactor, 710 RH03-15285 RH Reactor 730 RH02-10605 RH Peactor 715 RH03-15305 RH Reactor 736 RH02-10625 RH Reactor 727 RH03-15325 RH Reactor 736 RH02-10635 RH Reactor 696 RH03-15335 RH Reactor 715 RH02-10645 RH RH03-15345 RH Reactor 736 RH02-10655 RH Reactor RH03-15375 RH Reactor 734 RH02-10675 RH Reactor Reactor 7 RH03-15405 RH Reactor 718 RH03-10345 RH 733 RH03-15415 RH Reactor 700 aH Reactor O aH03-10355 RH03-1036S RH Reactor 719 RH03-15445 RH containment 738 717 RH04-10205 RH Reactor 682 RH03-10375 RH Reactor 718 RH04-10215 RH Reactor 681 RH03-10385 RH Reactor 734 RH04-10225 RH Reactor 700 RH03-10445 RH Reactor 703 RH04-10235 RH Reactor 703 RH03-10465 RH Reactor -

736 RH04-10245 RH Reactor 703 RH03-10475 RH Reactor' 700 RH04-10255 RH Reactor 703 RH03-10495 RH Reactor 729 RH04-10275 RH Reactor 703 RH03-10515 RH Reactor 738 RH04-10285 RH Reactor 703 RH03-15005 RH Containment 738 RH04-10295\ RH Reactor 703 RH03-15025 RH Containment 738 RH04-10315 i RH Reactor 703 RH03-15035 RH Containment 738 RH04-10325 \ RH Reactor 703 RH03-15045 RH Containment 740 RH04-10335 \ RH Reactor 703 RH03-15055 RH Containment 742 RH04-10355 RH Reactor 700 RH03-15065 RH Containment RH Containment 738 RH04-10365 RH Reactor 692 RH03-15075 688 Containment 738 RH04-10385 RH Reactor RH03 15085 RH 688 Containment 738 RH04-10395 RH' Reactor RH03-15095 RH RH03-15115 RH Containment 738 RH04-10405 RH \ Reactor 688 738 RH04-10445 RH \ Reactor 682 RH03-15125 RH Containment 682 RH03-15135 RH Containment 738 RH04-10455 RH \ Reactor 738 RH04-10465 RH Reactor 688 RH03-15145 RH Containment RH Containment 741 RH04-1051S AH Reactor GS2 RH03-15155 738 Containment 741 RH04-1416S RH Containment RH03-15165 RH 688 Containment 738 RH12-10715 RH Reactor RH04-141SS RH Containment 738 RH12-1072S RH Reactkr 697 RH04-14205 RH 582 Containment 738 RH12-10773 RH Reacto' l, RH04-15005 RH 3/4 7-37 Amendment 1 l LA SALLE - UNIT 1 l

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

S Table 3.7.9-2 (Crntinued) ,

I*

stem snubber installed on Snubber System snubber installed Ij h" ubber ' -

t. . 1 ation and elev. No. on location and elev.

Containment 738 RH13-10095 RH Reactor 720 d - RH04-15015 RH RH04-1502S RM Containment 738 RH13-10nS RH Reactor 719 j

i RH04-15035 RM Containment 738 RH13-10415 RH Reactor 702 inment 754 RH13-10855 RH Reactor 687 i RH04-15065 RH Reactor 682

':i RH04-15085 RH inment 744 RH13-1086S RH RH Reactor 698 RH04-15135 RH r 703 RH13-10885 RH04-15155 RH 2e r 688 RH13-10895 RH Reactor 698

- Reactor 686

RH04-15165 RH 682 RH13-n005 RH

' RH04-15185 RH Reacto 734 RH13-n015 RH Reactor 695 RH04-15195 RH Reector 734 RH13-n105 RH Reactor 690 Reactor 703 RH13-u nS RH Reactor 702 RH04-15205 RH

. RH04-1521S RH Reactor 703 RH13-n125 RH Reactor 703

- RH04-15225 RH Reactor 688 RH13-n13S RH Reactor 683

-RH05-10105 RH Reactor 679 RH13-n145 RH Reactor 683 Reactor 679 RH13-in95 RH Reactor 689

, RH05-10125 RH i RH05-10485 RH Reactor 683 RH13-n205 RH Reactor 705 Reactor 705

l. RH05-n105 RH Reactor 683 RH13-11225 RH Reactor 705

!, RH05- n135 RH Reactor 683 RH13-D235 RH 701 RH13-D265 RH Reactor l'

l RH05- n145 RH05-n205 RH RH Reactor Reactor 703 RH13-11275 RH Reactor 701 RH Reactor 701 RH05-n225 RH Reactor 69A RH13- u315 Reactor 687 \ RH13-n325 RH Reactor 701 RH12-10075 RH Reactor 697 \RH13-H345 RH Reactor 706 RH12-10095 RH 706 12-10105 RH Reactor 697 RH13- n355 RH Reactor 12-10HS RH Reactor 697 RH13-11375 RH Reactor 705 697 RH13-H385 RH Reactor 708 i RH12-10125 RH Reactor Reactor 695 RH13\n395 RH Reactor 707 RH12-10135 RH 708 RH12-10145 RH Reactor 695 RH13-1140S RH Reactor RH12-10185 RH 682 RH13-11415 RH Reactor 712 Reactor ~

RH12-1060S RH Reactor 690 RH13-n425 RH Reactor 713 RH12-10675 RH Reactor 692 RH13-n541 RH Reactor 719 695 RH13-uS5S RH Reactor 705

! RH12-10695 RH Reactor 695 RH13-n565 RH Reactor 705 RH12-10705 RH Reactor 7n RH23-10395 RH Reactor 706 RH13-n615 RH Reactor 720 RH23-10405 RH Reactor 704 RH13-n655 RH Reactor 719 RH23-10415 RH Reactor 705 RH13-n665 RH Reactor Reactor 724 RH23-10425 RH Reactor 705 i RH14-1012S RH 720 RH Reactor 724 RH23-10435 Reactor RH14-10135 721 Reactor 724 RH23-10445 R Reactor RH14-10145 RH 720 Reactor 723 RH23-10455 RH Reactor RH14-10165 RH 720 Reactor 707 RH23-10465 RH Reactor RH14-10185 RH 707 Reactor 717 RH23-10475 RH Reactor RH14-10465 RH 707 Reactor 724 RH23-10485 RH . Reactor

, RH14-10475 RH 717 Reactor 708 RH23-10525 RH Reactor RH14-10505 RH 720 RH14-10515 RH Reactor 723 RH23-10535 RH R4ctor 703 RH14-10525 RH Reactor 718 RH26-10075 RH Reactor RH Reactor 734 RH28-1001S RH Reactor 720 RH14-10555 696 Reactor 703 RH29-18045 RH Reacth ,

RH15-10135 RH D- 3/4 7-38 LA Salle - UNIT 1 AmendmentI I l

i, 7

Table 3.7.9-2 (Continued) jj ,

.t Snubber System snubber installed on Snubber . System snubber installed

, s No. location and elev. No. on location and elev.

RH18-10075 Reactor 703 RH33-10535 RH Reactor

746 o RH18-10095 Reactor 709 RH33-10545 RH Reactor 747

, RH18-1011S RH Reactor 715 RH33-10555 RH Reactor 746

' RH18-10155 RH Reactor 705 RH34-10325 RH Reactor 740 1

RH19-10095 RH . Reactor 718 RH34-10323 RH Reactor 740 RH19-10105 RH Reactor 715 RH39-10255 \ RH Reactor 732 RH19-10135 RH actor 718 RH39-10265 RH Reactor 732 RH19-10145 RH r 719 RH39-10275 RH Reactor 732 RH19-10153 RH Re tor 719 RH40-10285 RH Reactor 733 RH19-10165 RH Read $or 718 RH40-10295 RH Reactor 734 RH19-10335 RH Reactor 723 RH40-10305 RH Containment 794 '

RH19-10345 RH Reacto 722 RH40-10325 RH Containment 792

. RH19-10365 RH Reactor 717 RH40-10335 RH Containment 782 i-

'; RH19-1037S RH Reactor 721 RH40-10345 RH Containment 778

RH19-1041S RH Reactor 717 RH40-10365 RH Reactor 757 RH19-10425 RH Reactor 716 RH40-10375 RH Reactor 729 RH19-10435 RH Reactor 717 RH40-10395 RH Reacter 734 RH19-10445 RH Reactor 717 RH40-10405 RH Reactor 734 RH19-10455 RH Reactor 715 RH40-10415 fH Reactor 734 RH23-10335 RH Reactor 712 RH40-10425 RH Reactor 733

RH23-10365 RH Reactor 721 RH40-10435 RH Reactor 729 l RH40-10445 RH Reactor . 30 RH50-10025 RH Reactor 675 RH40-10475 RH Reactor 734 RH50-10045' RH Reactor 675

~9H40-10485 RH Reactor 750 RH50-10175 RH Reactor 679 d440-10495 RH Reactor 73 RH53-10165 RH Reactor 695 RH40-14015 RH Reactor 729 RH53-10185 RH Reactor 718 RH40-15005 RH Containment 799 RH53-10205 RH Reactor 722 i

RH40-15015 RH Containment 799 RH53-10215 RH Reactor 725 RH40-15045 RH Containment 774 Rl453-10225 RH Reactor 725 -

RH40-15055 RH Containment 774 RH5 -10245 RH Reactor 731 RH40-15065 RH Containment 784 RH53 0255 RH Reactor 763 RH40-15225 RH Containment 797 RH53- 265 RH Reactor 780 RH40-15235 RM Containment 797 RH53-1 75 RH Reactor 775 RH40-15245 RH Containment 793 JtH53-10 S RH Reactor 780 RH40-15265 RH Containment 797 RH53-1029 RH Reactor 800

, RH40-15395 RH Reactor 734 RH53-10305 RH Reactor 780

'i RH40-1540S RH Reactor 734 RH53-10315 RH Reactor 777 t

R'440-15415 RH Reactor 734 RH53-10325 RH Reactor 776 RH40-15435 RH Reactor 733 RH53-10695 RH Reactor 725 RH40-15445 RH Reactor 734 RH53-15505 RH Containment 796 RH40-15485 RH Reactor 734 RH53-15515 Containment 7%

RH40-15495 RH Reactor 735 RH53-15535 Containment 798 l RH40-15505 RH Reactor 757 RH53-15545 RH Containment 796

, RH43- 4 515 RH Auxiliary 732 RH53-15575 RH Containment 794 RH40-1"?S RH Reactor 747 RH53-15615 RH Containment 783 RH40-15b35 RH Reactor 750 RH53-15625 RH Containment 779l RH40-15545 RH Reactor 757 RH53-15653 RH Reactor 776 RH G 15575 RH Reactor 764 RH53-15685 RH ntainment 793 ,

RH40-15595 RH Containment 771 RH53-15735 RH R ctor 770 l LA Salle - UNIT 1 3/4 7-33 y

~___ .__ _ _ _ _ _ _ _ _ _ . _ _ _ . ____;____.____ E_________

Table 3.7.9-2 (Continued) .

(hhSnubber stem snubber installed on Snubber System snubber installed ~

No. 1 ation and elev. No. on location and elev.

RH40-15605 RH Containment 794 KH53-15745 RH Reactor 771 i RH40-15615 RH Containment 799 RH53-1575S RH Reactor 771

RH40-15725 RH Reactor 734 RH56-10035. RH Reactor 688 RH40-15735 RH eactor 728 RH56-10075 RH Reactor 688 RH41-10915 RH R ctor 708 RH59-10305 RH Reactor 762 j RH42-10325 RH Re tor 705 RH59-10315 RH Reactor 762 .

RH42-10335 RH Re or 706 RH59-10485 RH Reactor 758 RH42-10375 RH Reac r 735 RH59-10495 RH Reactor 738 RH59-10525 RH Reacto 758 RI01-10725 RI Reactor 702 RH59-10565 RH - Reactor 754 RI01-10735 RI Reactor 701 RH32-10305 RH Reactor 708 RIO1-10745 RI Reactor 701 RH82-10375 RH Reactor 687 RIO1-1076S RI Reactor 692 RH82-1038S RH Reactor 686 R101-10775 RI Reactor 693 RH82-1040S RH Reactor 686 RIO1-10805 RI Reactor 685 RH82-10415 RH Reactor 687 RI01-1081S RI Reactor 685 RH82-1046S RH Reactor 695 RI01-10835 RI Reactor 676 RH82-1068S RH Auxiliary 698 RI01-10845 RI Reactor 678 .

RH82-10745 RH Auxiliary 698 R101-10855 RI Reactor 678 RH83-10115 RH Reactor 689 RIO1-10885 RI Reactor 679 RH83-10145 RH ' Reactor 689 RIO1-10895 RI Reactor 683 RH83-10155 RH Reactor 68' RIO1-10905 RI Reactor 683 RH83-1016S RH Reactor ~

684 RI01-1091S RI Reactor 683 RH83-1017S RH Reactor 684 R101-10925 RI Reactor 677 RH Reactor 684 4I01-10935 RI Reactor 678

({}JRH83-1018S RH83-10375 RH Reactor 689 R101-11015 RI Containment 718 t RHA6-1003S RH Reactor 715 RIO1-11025 RI Containment 718 RHB4-10025 RH Containment 738 RI0h11035 RI Containment 750 RHB4-1005S RH Containment 738 RIO1-11065 RI Containment 776 RHB4-10075 RH Containment 742 RIO1-11085 RI Containment 779 '

RHB4-10085 RH Containment 742 RIO2-10075 RI Reactor 680 RH64-10115 RH Containment 745 RIO2-10095 RI Reactor 687 RIO1-1006S RI Containment 753 RIO9-1005 RI containment 742 RIO1-10075 RI Containment 771 RIO9-10075 RI Containment 742 RI01-10085 RI Containment 769 RIO9-1008S RI Containment 7a7 R101-10095 RI Containment 769 RIO9-10095 RI Containment 747 RIO1-10105 RI Containment 775 RIO9-10115 RI Containment 743 RIO1-10115 RI Containment 775 RIO9-10165 RI Containment 743 RIO1-10125 RI Containment 774 RIO9-1021S I Containment 743 RI01-10635 RI Reactor 743 RIO9-10245 R! Containment 747 RIO9-10255 \ Containment 746 RIO1-10645 RI Reactor 743 RI \

RI01-10655 RI Rea; tor 744 RIO9-10265 RI \ Containment 747 RI01-10675 RI P mctor 740 RIO9-10275 RI 743 736 RI16-1016S RI \ Containment gReactor 688 RIO1-10695 RI 4 actor RIO1-10705 RI Reactor 736 RI16-10225 RI Reactor 583 RI16-10235 RI Reactor 683 RI41-1067S RI Reactor 703 RI16-10255 RI Reactor 687 RI41-10685 RI Rehetor 703 RI24-1015S RI Containment 793 RI41-10895 RI Rea'ctor 703 RI24-10165 RI Containment 793 RI41-1092S RI React r 711 LA SALLE - UNIT 1 3/4 7-40 Amendment H

I Tablo 3.7.9-2 (C:ntinued) -

t(

, \

System snubber installed on Snubber System snubter installed '

i.

QSnubber No. location and elev. No. on location and elev.

h

'~

RI24-10175 containment 779 RI42-10105 RI Reactor 742

,e RI24-10185 RL Containment 780 RR00-10015 RR Containment 743 l '; RI24-10195 RI Containment 759 RR00-1002S RR Containment 739 RI24-1020S RI Containment 760 RR00-3M35 RR Containment 743 i RI24-10215 RI Containment 754 RRo0-iv045 RR Containment 759

,; RI24-10225 RI Containment 754 RR00-10055 RR Containment 743 1: RI24-10635 RI r 741 RR00-10065 RR Containment 739 i RI24-10705 RI actor 736 RR00-1007S RR Containment 739

RI24-10715 RI R tor 726 RR00-10085 RR Containment 739 RI24-10725 RI r 726 RR00-10095 RR Containment 743 i

RI24-1080S RI r 742 RR00-10105 RR Containment 748

, RI24-11205 RI Conta t 821 RR00-10115 RR Containment 748 l, RI24-11215 RI Contal nt 821 RR00-10125 RR Containment 759 l i4 RI24-11225 RI Contai nt 821 RR00-10135 RR Containment 758

' Contai 826 RR00-10145 RR Centainment 759 RI24-11245 RI RI24-11305 RI Containes 810 RR00-10155 RR Containment 743 j RI24-11315 RI Containment 809 RR00-1016S RR Containment 753 RI24-11325 RI Containment 810 RR00-10175 RR Centainment 753 i' RI24-15115 RI Containment 810 RR00-10185 RR Containment 743 RI24-15125 RI Containment 810 RR00-10195 RR Containment 765 RI24-15135 RI containment 785 RR00-10205 RR Containment 765 RI24-15145 RI Containment 764 RR00-10215 RR Containment 761 RI24-15155 RI Containment 749 RR00-10225 RR Containment 761 M I41-10565 RI Reactor 7,40 RR00-10235 RR Containment 758

, V RI41-10575 RI Reactor 712 RR00-10245 RR Containment 757 RI41-10585 RI Reactor 7 RR00-10255 RR Containment 743

. RI41-10595 RI Reactor 708 RR00-10265 RR Containment 762 RI41-10605 RI Reactor 703 RR00-10275 RR Containment 762 RI41-10615 RI Resctor 705 RR00-1028S RR Containment 765 RI41-10645 RI Reactor 705 RR00-10295 RR Containment 765 - l RI41-10655 RI Reactor 703 RR00-10305 RR C-mtainment 743

, RI41-10665 RI Reactor 703 RR00-10315 RR Containment 746

. RR00-10325 RR Containment 746 RR28s1016S RR Containment 750 l' RR00-10415 RR Containment 748 RR28-10175 RR Containment 750 RR00-16425 RR Containment 750 RT01-10855 RT Containment 748

, RR00-10435 RR Containment 757 RT01-1087S RT Containment 739 RR00-10445 RR Containment 759 RT01-10885 RT Containment 738 RR00-10455 RR Containment 743 RT01-10895 RT Containment 738 l' RR00-10465 RR Containment 753 RT01-10915 RT Containment 738 RR00-10475 RR Containment 754 RT01-10935 RT Containment 738 RR00-10485 RR Containment 743 RT01-10945 RT Containment 738 RR00-10495 RR Containment 768 RT01-109GS RT Containment 738 ,

RR00-10505 RR Containment 765 RT01-10975 RT Containment 738 RR00-1051S RR Containment 764 RT01-11005 RT Containment 738 RR00-10525 RR Containment 764 RT01-11295 RT Containment 738 RR00-1053R RR Containment 743 RT01-11305 RT Containment 739 RR00-10545 RR Containnent 762 RT17-10025 RT Containment 748 RR00-1055S RR Contairveent 762 RT17-10085 Rf Containment 746

, RR00-10565 RR Containment 765 SC02-10155 SC Reactor 814 ,

i LA Salle - UNIT 1 3/4 7-41 AmendmenthI b[

Table 3.7.9-2 (Continued) (

Snubber System snubber installed on Snueber System snubber installed'

_ . . _ No. location and elev. No. on location and elev.

RR00-10575 Containment 765 SC02-10275 SC Reactor 781 RR00-10585 R Containment 739 SC02-10365 SC Containment 774 4 RR00-10595 RR Containment 743 SC02-10385 SC . Containment 770 RR00-10605 RR Containment 746 SC02-10475 SC Containment 757 RR00-10615 RR Containment 746 $C02-10555 SC Containment 750 RR00-10625 RR Containment' 748 VG01-00015 VG Reactor 831 RR01-10325 RR tainment 743 VG01-00055 VG Reactor 831 RR07-14325 2R tainment 738 VG01-00065 VG Reactor 831 RR17-10015 RR Co tainment 738 VG01-00085 VG Reactor 831 RR17-10025 RR Con inset $ 737 Vr,01-00105 VG Reacter 831 RR17-10035 RR Con inment 737 VG02-10045 VG Reactor 794 RR17-10045 RR Conta nt 737 VG02-30055 VG Reactor 794 RR17-10055 RR Contai t 737 VG02-18005 VG Reactor 802 RR17-10065 RR Contai nt 737 VG04-10035 VG Reactor 821 RR17-10075 RR Containee t 737 VG04-1005$ VG Reactor 809 RR17-10085 RR Containeen 737 VG04-10115 VG Reactor 794 RR28-10075 RR Containment 756 V604-10145 VG Reactor 794 RR28-10125 RR Containment 750 VG04-10155 VG Reactor 798 RR28-1015S RR Containment 741 VG04-10195 VG Reactor 798 VQ02-10355 VQ Reactor 805 LP-25-H045 LP

, VQ02-10405 VQ Reactor 810 N.P. 30A RR-59 VQ05-1001S VQ Reactor 733 RR68-H-4 RR VQ05-10095 VQ Reactor 38 RR68-N-6 RR .

N.P. 83 RR-59 N.P. 85 RR-59 RR69-H-4 RR RR-59 kN.P.164 '

FHP1204-H02 HP N.P. 170 RR-59 -

FHP1204-H03 HP M P. 175 RR-59 FHP1204-H02 HP Nh.180 RR-59 FHP1204-H03 HP N.PK 240 RR-59 N.P. 288 MS-51A HG-08L-H035 HG N. P. 38 MS-51A HG-08-H085 HG N.P. 41 MS-51A HG-08-H075 HG N.P. 48A MS-51A N.P. 114\ . HG-61 N.P. 2968' MS-51A N.P. 115 \ HG-61 N.P. 3368 MS-51A N.P.116\ HG-61 N.P. 3768 MS-51A HG21-H04 N HG N.P. 855 MS-51A FRH1213-H15 \ RH N.P. C43 MS-51A FRH1213-H14 \ RH M8-125-H075 RH FRH1213-H12 N.P. 465 VG-03 FRH1213-H11 \.RH

.RH N.P. 475-X VG-03 FRH1213-H09 RH M.P. 475-Z VG-03 FRH1213-H07 R

, M.P. 308 RN-75A FRH1213-H08 RH

, N.P. 30A RH-76 FRH1213-H06 RH N.P. 308 RH-76 FP.H1213-H03 RH N.P. 1358 RH-76 FRH1213-H02 RH 1.P. 112 HG-61 FRH1213-H01 RH LA SALLE - UNIT 1 3/4 7-42 Am ndment 1 1

i

Tabl6 3.7.9-2 (continued) i

, I ' n _ a '4.H Snubber Syst snubber installed en snubbeP $Nsidisnubberinstalled

, _ No. locati n and elev. -

No.. m.. ; BH Incation and elev.

HP- u-H025 HP PMgm , I ,

i MP-12-H045 HP PRWif0 ' '

, N.P. 125 RR-59 PRM18tj(a g* i i N.P. 135 RR-59 RM209 0 '

+ 0s-H125 Ha 26

,gggg,

'! N. P. 52 RH-68 .I 1 FRH 3)= .

N.P. 63 RN-68 FRH 07a N . I ,

N.P. 95 RH-68 FRN $f= MS I!

RH52-H07 RH FR11215*40 I RH52-H08 RH FR21209- "O I PH52-H09 RH FRHitsia <

l RH52-H06 RH FNH1fil- l RH52-H05 RH F 1232-H06 I; RH52-H02 RH $ a 23 . -f:,

l' '

RH52-H03 RH N4 : Aid 1 71 N.P. 133 RN-68 FAHif06-Nil IN FRH1211-H01 RH FAH120 H18 L. ' "

l- FRH1211-H04 RH W. P. , j , .la FRH1211-H03 RH N,P ':, '

X' kt-FRH42-1037 RH N'. Pi .l,8*1 546 i H-4RM-87 RH \ N. P: f3 i LC-

! . H-2RH-88 RH \ N.P. 23-1 LC 85 N.P. 120 RH-69 X C-SS

\N.P.

)RI120723*2Hdi.::ll '

N.P. 135 RN-69 FRH1214-H-125 CS 1RIO78-2al-H08 J FRH1230-H07 RH 1RIO78-2 1-H04 R:

FRH1230-H06 RH 1RIQ78-2-1-H05 RI FRH1230-H05 RH 1RI07,8*2 1-H16 RI a -

FRH1230-H03 RH N.P. 53 RH-ti

l. FRH1230-H02 RH N. P. 94 . RH-21 FRH1230-H01 RH H.P.-100 RH-21 FRH1206-H03 RH N.P.125\' RH-21 FRH1233-H05 RI MS-52-H065 MS FRH1233-H04 RI MS-53-Hes M5 FRH42-1031 RH MS-53-H055 M4 FRH42-1032 RH MS-50-H025 MS FRH42-1033 RH MS-50-H045 \ MS N.P. 135 RH-16 MS-51-H055 \ HS FRH1231-H09 RI RT-33-H085 NHs FRH1231-H06 RI RT-33-H095 tS FRH1231-H04 RI RT-33-H105 P6 RH25-H01 RH N.P. $6 RSN RH25-H02 RH N.P. 52 . RH 3 RH25-H04 RH M-1302-2I-169 4 RH25-H03 RH M-130f*22-110 D l RH25-H05 RH M-1302-22-112 RN RH25-H06 RH M-1302-22-113 RN ;

HG-04-H04S HG M-1302-23-96 HS HG-61 N.P. 60A RH-C3 h N.P. 117 1

LA SALLE - UNIT 1 3/4 7-43 Amendment 1 i

i

e Table 3.7.9 ,2 (C:ntinued) .

n ..r-0 5a"* - Sr **" *a"*6*" 'a" 'd 'a 5"66*" '5 daub 6d"ia

4. I tion and elev. ,. No t , , , ,

,g 8,facitionande'e**"'d l v. ~

N.P. 118 NG-6 M=1J03-2461 R.I N.P. 119 NG-61 5180Pt44 i i N.P. 15 N8-83 M4408-24*% F ll t M.P. 558-Y RN-41 M=1362*84480 l' l- N.P. 558-Z RH-G1 W130taf49%81 II :

', N.P. 40 HP-Al *24'Hf Ill M-1302-14-48 M8 Ma @303 2>01 P1 '

2 N.P. 2558 RN-01 M=M82*28*8 , >

j N.P. 50 M-C1 ' PD0p

, 8, W

. N.P. 65 M-C1 M 180a*

i tL M-1302-15-3 RH ,tii13024

110 4' M-1302-15-56 RH M-1302?28-134 R' I

M-1302-15-72 RH N.P..08-16 R'-Cs W 1302-16-35 NB M-11Deff*II$ Al M-1302-16-26 M M4402d2,12) R

, M-1302-16-56 N8 PUOPt 485 lR l' '

M-1302-16-55 N8 M 1302*2 428 i

M-1302-16-23 RH M-130247= 9 R l W1302-16-51 RH M-130P27* l}

M-1302-16-52 RH M 1302df74 9 Rt N.P. 1708 HP 28t?4 L <

W 1302-20-205 MS M-1302=,28578 M-1302-M-1302-20-208 MS M-1302-28-84 M-1302-20-209 MS M-1302-30-52 RH -

MS M-1302-30-54 RN

~Q^S-1302-20-212

.r1302-20-145 MS .

M-1302-30-41 RH I'

M-1302-20-131 N8 $1302-35-41 LC N.P. 375A-Y HP Al Ms1302-35-43 LC N.P. 375A-Z HP-Al M-1302-36-45 R1 _

M-1302-36-138 RI M-1302-36-131 RI M-1302-36-141 RI M-1302-23-95 MS M-1302-36-142 RI W 1302-23-140 s

MS M-1302-36-148 RI M-1302-23-143 MS M-1302-36-152 RI M-I302-23-124 MS l M-1302-36-154 RI l M-1302-36-169 RI M-1302-36-94 RI l M-1302-36-167 RI M-1302-9-1 NB M-1302-36-133 RI M-1302-36-146 RI M-1302-36-147 RI M-1302-36-150 RI M-1302-36-151 RI M-1302-36-125 RI M-1302-36-126 RI M-1302-36-170 RI M-1302-21-40 RR N.P. 45 RR-A7 O-LA Salle - UNIT 1 3/4 7-44 Amendment \l e-mb u ge

8 Tablo 3.7.9-2 (Ccntinued) .

(-

~

-05aar 5n nar i=u.4 -

loca on and elev.

= = ,

. No ., . . , , .,

unu 48mne. -

DM location and elev. -i

._..N.P. 118 NE-51 M-1302*24*k45 itk

N.P. 119 NG-61 P1302*f4*146 RR !

,: N.P. 15 NS-83 M-1302-244131 -

,i N.P. 558-Y RN-G1 M-1302?24488 I

. N.P. 558-Z RN-G1 M-1302=24?181 l N.P. 40 HP-Al M=1302.=24 152 I l

M-1302-14-48 NB M-1302*25-41 l N.P. 2558 RN-G1 M-1302-26-8 8.)i D )

i N.P. 50 NS-C1 M=130263640$

M-1302-264132

$ i N.P. 65 NS-C1 RH M-1302-15-3 RH M-1302-26-110 46

! M-1302-15-56 RH M-1302-26?134 Rt M-1302-15-72 RM N.P. C8-10.. . h-C8 M-1302-16-35 NB M-1302-27-113 RH i

! W1302-16-26 NB M 1302-27-121 RH M-1302-16-54 W M-1302-27-122 4R

. M-1302-16-55 NB M-1302-27-128 RH M-1302 16-23 RH M-1302-27-129 RN l M-1302-16-51 RH M-i302-27-83 RW

' \ M-1302-27-139 RH M-1302-16-52 RN N.P. 1708 HP M-1302-28-74 LP M-1302-20-205 MS M-1302-28-76 LP M-1302-20-208 MS W 1302-28-84 RH MS M-1302-30-52 RH M-1302-30-54 O~M-1302-20-209 61302-20-212 A-1302-20-145 MS MS sM-1302-30-41 RH RH l i

M-1302-20-131 NB M-1302-35-41 LC LC

(

l N.P. 375A-Y HP-Al $1302-35-43 )

N.P. 375A-Z HP-Al M-1302-36-45 RI 1 M-1302-36-138 RI M-1302-36-131 RI _

M-1302-36-141 RI M-1302-23-95 MS M-1302-36-142 RI M-1302.-23-140 MS M-1302-36-148 RI M-1302,23-143 MS M-1302-36-152 RI M-1302-23-124 MS ,

M-1302-36-154 RI M-1302-36-169 RI M-1302-36-94 RI M-1302-36-167 RI M-1302-9-1 NB M-1302-36-133 RI l M-1302-36-146 RI l l M-1302-36-147 RI -

, M-1302-36-150 RI l

M-1302-36-151 RI M-1302-36-125 RI W 1302-36-126 RI

! M-1302-36-170 RI

, M-1302-21-40 RR l' N.P. 45 RR-A7 O-LA Salle - UNIT 1 3/4 7-44 Amendmen 1 l -

l

i . .

Table 3.7.9-2 (Continued) k f

~

Snubber Sys snubber installed on Snubber System sn2ber installe.i

! No. locatio and elev. No. on location and elev.

, W1302-22-83 RR I

i M-1302-22-33 RR l M-1302-22-127 RR

.i M-1302-24-110 RR

, M-1302-24-111 RR

M-1302-24-106 RR i M-1302-24-107 RR M-1302-24-103 RR l M-1302-21-183 MS l

l- $ 1302-21-181 MS N.P. 858 MS-C1 l i

N.P. 65-Z MS-C2 N.P. 65-Y MS-C2 i M-1302-21-162 MS

3 M-1302-21-163 MS

, M-1302-21-74 MS

M-1302-21-189 MS HP75-20095 HP Diesel Generator 49 2HG-24-H02 2HG-62 HP75-20145 HP Diesel Generator 5 2HG-16-H01 2HG-62 HP75-20175 HP Diesel Generator 7 RH-27H-H035 2HG-61 HP75-28005 HP Diesel Generator 74 RH-27H-H065 2HG-61 RH83-20065 RH Reactor 684 \ 2HG-4H-H105 2HG-61 RH Reactor 689 \2HG-15H-H035 2HG-61 OTH83-20075 AH83-20095 RH Reactor 684 '2HG-15-H05S 2HG-61 RH83-20125 RH Reactor 689 2HG-15H-H015 2HG-61 RH83-20135 RH Reactor 684 2DG-66-H03 2CS-75 f 2HG-70 RH83-20165 RH Reactor 689 2HG23-H06 SC02-29675 SC Reactor 782 2HG23-H09 2HG-70 ~

VG01-00155 VG Reactor 831 2VQ-10-H02 2VG-03

^

Reactor 831 k 2YQ-10 N04 2VG-03 VG01-00165 VG

i VG01-20195 VG Reactor 830 2VQ-11-M1 2VG-03 VG01-20215 VG Reactor 803 2VQ-11-H03 2VG-03 VG02-00125 VG Reactor 831 2VQ-11-H06 2VG-03 VG02-20195 VG Reactor 803 2VQ-11-H04 2VG-03 VG02-20215 VG Reactor 831

\

VG02-20225 VG Reactor 794 VG02-2023S VG Reactor 826 \

Reactor 831 \

l2 VG02-20245 VG

\

VG03-20015 VG Reactor 794 VQ04-20195 VQ Reactor Reactor 738 733

\s VQ04-2020S VQ \

VQ05-20235 VQ Reactor 762 VQ05-20245 VQ Reactor 762 VQ05-20255 VQ Reactor 787 VQ05-20265 VQ Reactor . 732

. VQ05-20315 VQ Reactor 733 l VQ04-2800S VQ Reactor 733 VQ06-20025 VQ Auxiliary 810 ;

O 3/4 7-45 Ame ent 1 LA SALLE - UNIT 1 l

PLANT SYSTENS, 3/4.7.10 MAIN TUR8INE 8YPASS SYSTEM .

i LIMITING C M ITION FOR OPERATION a i

- 3.7.10 The main turtpine bypass system shall be OPERA 8LE.

APPLICA8ILITY: OPERATIONAL COMITION 1 w&w g, THrtm4. PowAR 15 <Jreak M er ep,M to as$ ef Ir.ATao TWieneAL howse, l ACTION: With the main turbine bypass system inoperable, withia hvo hours restore the system to OPERABLE status 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 REQUIRENENTS 4.7.10 The main turbine bypass system shall be demonstrated OPERA 8LE at least once per:

a. 7 days by cycling each turbine bypass valve through at least one complete cycle of full travel.

b. 18 months by:

1. Performing a system functional test which includes simulated automatic actuation and verifying that each automatic valve actuates to its correct position.

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4.2. Demonstrating TUR8INE BYPASS SYSTEM RESPONSE TIME to be less than or equal to 200 milliseconds 0; . .;h: ;;;;;tter,

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l LA SALLE - UNIT 1 3/4 7-f6' 39 Amendment 1 I

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! ~3 /4.8 ELECTRICAL POWER SYSTEMS .

'3/4.8.T A.C.' SOURCES ,,,.

i A.C. SOURCES - OPERATING , ,

f i j LIMITING CONDITION FOR OPERATION i .

3.8.1.1 As a minimum, the fo11owing A.C. electrical power sources shall be l '

OPERA 8LE:

a. Two physically independent circuita between the offsite transmission network and the onsite Class 1E distribution system, and

b. Separate and independent diesel generators 0, 1A, 2A and 18 with:

l 1. For diesel generator 0. IA and 2A:

c) A separate day fuel tank containing a minimum of

~

250 gallons of fuel. ~

! b) A separate fuel storage systes containing a minimum of 31,000 gallons of fuel.

2. For diesel generator 18, a separate fuel storage tank / day tank .

t '.

~

containing a minimum of 29,750 gallons of fuel.

2

3. A separate fuel transfer pump.

APPLICA8ILITY: OPERATIONAL CONDITIONS 1, 2, .snd 3. .

t j ACTION: -

i

a. With either one offsite circuit or diesel generator 0 or 1A of the above required A.C. electrical power sources inoperable, demonstrate .

l,

' the lanceOPERA Requirements 8ILITY of the remaining 4.8.1.1.1.a within A.C.%

hour, and sources by performing 4.8.1.1.2.a.4., i Surve I for one diesel generator at a time, within four hours, and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter; restore at least two offsite circuits and diesel generators 0 and IA to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or i be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLO SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , J

b. With one offsite circuit and diesel generator 0 or IA of the above i required A.C. electrical power sources inoperable, demonstrate the i OPERA 81LITY of the remaining A.C., sources by performing Surveillance ,

Requirements 4.8.1.1.1.a within.ck hour, and 4.8.1.1.2.a.4, for one j i i diesel generator at a time, within three hours, and at least once per . l 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter; restore at least one of the inoperable A.C. l i sour:es to OPERA 8LE status within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or be in at least HOT SHUTC0hN witnin tne next 12 haurs and in COLD SiiUTCN witnia the

! followinc .'; cours. Restare at least two offsita :f r uits vc diesel gEnernars 0 ana lA ta CFERABLE status wicnia 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months frc.3 l

!.'] s the time of initial loss or be in at least HOT SHUTDOWN within the I

w'. i next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUT 00WN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . I t

j ,

LA SALLE - UNIT 1 3/4 8-1

ELECTRICAL POWER SYSTEMS

L_INITING CONDITION FOR OPERATION (Continued)

ACTION (Continued)

~~ '

c. With both of the above required offsite circuits inoper:ble, \

'i demonstrate the OPERABILITY of the remaining A.C. sources by per-

' forming Surveillance Requirement 4.8.1.1.q[a.4, for one diesel I generator at a time, within four hours, and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter, unless the diesel generators are already operating; I restore at least one of the inoperable offsite circuits to OPERABLE j stat,us within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT SHUTDOWN within the i next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . With only one offsite circuit restored to OPERABLE l status, restore at least two offsite circuits to CPERABLE status wfthin 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months from time of initial loss or be in at least HOT 4 .

SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the l, following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

d. With diesel generators 0 and 1A of the above required A.C. electrical r -

power sources inoperable, demonstrate the OPERABILITY of the remaig,-

ing A.C I within M. sources hour andby performing for 4.8.1.1.2.a.4, Surveillance one diesel Requirements generator at a 4.8.1.1.1da

(

time, within two hours and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter;

- restore at least one of the inoperable diesel generators 0 and 1A to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Restore both diesel generators 0 and 1A to OPERABLE status within U

72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months from time of inftial loss or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following

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

a. With diesel generator 18 of the above required A.C. electrical power sources inoperable, demonstrate the OPERABILITY of the remaining l A.C. sources by performing Surveillance Requirements 4.8.1.1.Ma withinbne hour, and 4.8.1.1.2.a.4, for one diesel generator at a l time, within three hours, and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter; l

restore the inoperable diesel generator 18 to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or declare the HPCS system inoperable and take the ACTION required by Specification 3.5.1.

f. With diesel generator 2A of the above required A.C. electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.C. sources by performing Surveillance Requirements 4.8.1.1.Ma and I 4.8.1.1.2.a.4, for diesel generator 1A, within one hour, and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter; restore the inoperable diesel generator 2A to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or declare standby gas treatment system subsystem B, Unit 2 drywell and suppression

f. chamber hydrogen recombiner system, and control room and auxiliary l' electric equipment room emergency filtration system train B inoperable

~

and take the ACTION required by Specifications 3.6.5.3, 3.6.6.1.,it and 3.7.23 coo +;woed performance of Surveilleuce. Ibygewwts 4g. I.l. U*-

and 4.9.1.I.k.4 k dressi made M is miit refuld PmA*

M sys4eus are. daelma Me.W -Ma Acmu f Leic resfocke-v spech+;ows is hireu.

LA SALLE - UNIT 1 3/4 8-2

~

ELECTRICAL POWER SYSTEMS 7 SURVEILLANCE REQUIREMENTS (Continued)

d. At least once per 18 months during shutdown by: M.

1. Subjecting the diesel to an inspection in accordance with l procedures prepared in conjunction with its manufacturer's l

! recommendations for this class of standby service. l lh 2. Verifying the diesel generator capability to reject a load of greater than or equal to 1190 kw for diesel generator 0, greater than or equal to 638 kw for diesel generators IA and 2A, and

greater than or equal to 2381 kw for diesel generator 18 while

~

maintaining engine speed less than or equal to 75% of the b . difference between nominal speed and the overspeed trip setpoint or 15% above nominal, whichever is lass.

3 .' Verifying the diesel generator capability to reject a load of i 2600 kw without tripping. The generator voltage shall not exceed 5000 volts during and following the load rejection.

4. Simulating a loss of offsite power by itself, and: l a) For Divisions 1 and 2 and for Unit 2 Division 2:

,, 1) Verifying de-energization of the emergency busses and l load shedding from the emergency busses. j

2) Verifying the diesel generator starts on the auto-start l signal, energizes the emergency busses with permanently '

connected loads within 13 seconds, energizes the auto- l connected loads and operates for greater than or equal to 5 minutes"while its generator is so loaded. After v energization, the steady state voltage and frequency i; of the emergency busses shall be maintained at 4160 i 150 volts and 60 1 1.2 Hz during this test.

b) For Division 3:

1) Verifying de-energization of the emergency bus.

2) Verifying the diesel generator starts on the auto-start signal, energizes the emergency bus with its loads with-in 13 seconds and operates for greater than or equal to 5 minutes while its generator is so loaded. After energization, the steady state voltage and frequency of the emergency bus shall be maintained at 4160 2 150 volts and 60 i 1.2 Hz during this test.

5. Verifying that on an ECCS actuation test signal, without loss of offsite power, diesel generators 0, 1A and 18 start on the auto-start signal and operate on standby for greater than or equal to 5 minutes. The generator voltage and frequency shall be 4160 + 416, -150 volts and~ 60 + 3.0, -1.2 Hz within 13 seconds after the auto-start signal; the steady state generator voltage and 1

frequency shall be maintained within these limits during this test.

6. VeH fying that on a simulated loss of the diesel generator,

! with offsite power not available, the loads are shed from the i

% emergency busses and that subsequent loading of the diesel b generator is in accordance with design requirements. ,. -

LA SALLE - UNIT 1 3/4 8-4 l

l

ELECTRICAL POWER SYSTEMS f SURVEILLANCE REQUIREMENTS (Continued)

NO.- v

. #.y . Simul ding a loss of offsite power in conjunction with an ECCS l actuation test signal, and:

,- a) For Divisions 1 and 2:

1 1) Verifying de-energiz.ation of tha emergency busses and load shedding from the emergency busses. '

2) Verifying the diesel generator starts on the auto-start l

- signal, energizes the emergency busses with permanently connected loads within 13 seconds, energizes the auto-connected emergency loads through the load sequencer and operates for greater than or equal to 5 minutes while its generator is loaded with the amergency leads. After energization, the steady state voltage and frequency of the emergency busses shall be maintained at 4160 1 416 volts and 60 i 1.2 Hz during this test.

b) For Division 3:

1) Verifying de-energization of the emergency bus.

2) Verifying the diesel generator starts on ther auto-start signal, energizes the emergency bus with its loads

~

within 13 seconds and operates for greater.than or equal to 5 minutes while its generator is loaded with the. emergency loads. After energization, the steady i

state voltage and frequency of the emergency bus shall be maintained at 4160 t 416 volts and 6011.2 Hz during this test.

14. Verifying that all diesel generator 0, IA and 18 automatic trips l except the following are automatically bypassed on an ECCS actuation signal:

a) For Divisions 1 and 2 - engine overspeed, generator differential current, and emergency manual stop.

b) For Division 3 - engine overspeed, generator differential or overcurrent, and emergency manual stop.

O,

4. Verifying the diesel generator operates for at least 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . l During the first 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of this test, the diesel generator shall be loaded to greater than or equal to 2860 kw and during the remaining 22 hours2.546296e-4 days 0.00611 hours 3.637566e-5 weeks 8.371e-6 months of this test, the diesel generator shall be loaded to 2600 kw. The generator voltage and frequency shall be 4160 + 420, -150 volts and 60 + 3.0, -1.2sHz within 13 seconds after the start signal; the steady state generator voltage and frequency shall be maintained within these limits during this test. Within 5 minutes after completing this 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months test, (O a tra = Surv iii "c a auire at 4 8 1 1 2 4 4 -).23 and b>.23.-

LA SALLE - UNIT 1 3/4 8-5

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

l

.... w a -

A.. ~

ELECTRICAL POWER SYSTEMS,

$UltVEILLANCE REQUIREMENTS (Continued)

} .i .

! ; T ,W. Verifying thai, the auto-connected loads to each diesel generator' (

, donotexceedthe2000hourratingof2860ly.p j 1

la W. Verifying the diesel generator's capability to: l j a) Synchronize with the offsite power source while the ,

generator is loaded with its emergency loads upon a simulated .

' restoration of offsite power, b) Transfer its loads to the offsite power source, and c) 8e restored to its standby status.

jf. ,32: Verifying that with diesel generator 0, IA and 18 operating in a l g test mode and connected to its bus:

l a) For Divisions 1 and 2, that a simulated ECCS actuation i . signal overrides the test mode by returning the diesel -

generator to standby operation.

b) For Division 3, that a simulated trip al' the diesel i generator overcurrent relay trips the SAT feed breaker to

bus 143 and that the diesel generator continues to supply' p' . Verify ng t wi all diosal generator air start receivers 3 pressurized to less than or equal to the compressors auto-start l

I setpoint and the compresso.s secured, diesel generators 0, IA l l and 2A start at least 5 times and diesel generator 18 starts at

' i' least 3 times from ambielt conditions and accelerates to l 900 rps + 5%, ~25, in less than cr equal to 13 seconds. .

~'

1 1 34: Ver1rying r.nar the automatic load sequence timer is OPERABLE i

( with the interval between each load block within 210% of its l

design interval for diesel generators 0 and 1A.

l3,AS'. Verifying that the following diesel generator lockout features )

{ prevent diesel generator operation only when required:

l a) Generator underfrequency.

, I b) Low lube oil pressure.

c) High jacket cooling temperature d) Generator reverse power. -

e) Generator overcurrent.

f) Generator loss of field.

g) Engine cranking lockout.

l

/

l ^If surveillance Requirement 4.3.1.1.2.d.4.a)2) ar.d/or am ara not satisfactorily

.m com)1eted, it is not necessary to receat the pre ecia.; la nour test. Instead.

G .

the diesel generator may be operated at 2600 lp/ for 9x nour or until operating temperature has stabilized. y .f, 1

LA SALLE - UNIT 1

3/4 8-6 -

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ELECTRICAL POWER SYSTEMS l

1 l, SURVEILLANCE REQUIREMENTS (Continued)

e. At least once per 10 years or after any modifications which could l

- . affect diesel generator interdependence by starting diesel gener- I pt stars 0, IA and 18 simultaneously, during shutdown, and verifying i that all three diesel generators accelerate to 900 rps + 5, -2% in

, i less than or equal to 13 seconds.

e

f. At least once per 10 years by:

I I 1. Draining each fuel oil storage tank, removing the accumulated I: i sediment and cleaning the tank using a sodium hypochlorite or

equivalent solution, and

'. 2. Perfoming a pressure test of those portions of the diesel

! fuel oil system designed to Section III, subsection NO, of the f'

i ASME Code in accordance with ASME Code Section 11 Article .

IWD-5000. .

9

4.8.1.1.3 Reports - All diesel generator failures, valid or non-valid, shall

, be reported to the Commission pursuant to Specification 6.6.8. Reports of j diesel generator failures shall include the information recommended in Regula-tory Postion C.3.b of Regulatory Guide 1.108, Retision 1, August 1977. If the l.-

a r or < 51 r ia ta i

ioo iid t t=. a P r a ci r uait 6 t=.

O. .w. is greater than or equal to 7, the report shall be supplemented to include the additional information recommended in Regulatory Position c.3.5 of Regulatory

Guide 1.108, Revision 1, August 1977. .

$ TA8tf 4.8.1.1.2-1 ~

i

~

, DIESEi. GENERATOR TEST SCHEDULE

)

I Number of Failures in i last 100 Valid Tests * .

Test Frequency 1 '

.l

! <1 At least once per 31 days 2 At least once per 14 days 3 At least once per 7 days

. . i

> 4~ At least once per 3 days

" Criteria for astermining number of failures and number of valid tests shall be in accordance with Regulatory Position C.2.e of

{

Regulatory Guidi 1.1C3, Revisien 1, August 1977, eere the last ,

i . 100 tests are determined on a per nuciaar unit basis. I 1 )

Q ,.

./

( LA SALLE - UNIT 1 3/4 8-7

ELECTRICAL-POWER SYSTEMS A.C. SOURCES - SHUTOOWN k

v} ,

LIMITING CONDITION FOR" OPERATION 1

W .

ii 3.8.1.2 As a minimum, the following A.C. electrical power sources shall be

OPERABLE

. a. One circuit between the offsite transmission network and the onsite ljl Class IE distribution system, and

b. Diesel generator 0 eneMor 1A, and diesel generator 18 when the HPCS l system is required to be OPERABLE, and diesel generator 2A when the of'fsite power source for standby gas treatment system subsystem B or

i control room and auxiliary electric equipment room emergency filtra-

! .i tion system train B is inoperable and either or both systems are required to be OPERABLE, with each diesel generator having:

l

1. For diesel generator 0,1A and 2A:

a) A separate day fuel tank containing a minimum of 250 gallons of fuel.

b) A separate fuel storage system containing a minimum of 31,000 gallons of fuel.

'.O , 2. For diesel generator 18,s a separate fuel storage tank / day tank v l containing a minimum of 29,750 gallons of fuel.

l 3. A fuel transfer pump.

APPLICABILITY: OPEPATIOEL CONDITIONS 4, 5, and *.

5 ACTION:

l' With all offsite circuits inoperable and/or with diesel generators 0 l- ae

!i .end/or 1A inoperable, suspend CORE ALTERATIONS, handling of irradiated }

l fuel in the secondary containment and operations with a potential for draining the reactor vessel.

l

b. With diesel generator 1B inoperable, restore the inoperable diesel

, generator 18 to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or declare the HPCS system inoperable and take the ACTION required by Specification 3.5.2 and 3.5.3.

I i

% en handling irradiated fuel in the secondary containment.

r0 -_

LA SALLE - UNIT 1 3/4 8-8

ELECTRICAL POWER SYSTEMS

, LIMITING CONDITION FOR OPERATION (Continued) ~

~

,. v . ._ .

ff ACTION: (Continued)

p ,

c. With diesel generator 2A inoperable, declare standby gas treatment l

{ ,

system subsystem 8 and control room and auxiliary electric equipment l 3

l room emergency filtration system train B inoperable and take the

~f ACTION required by Specifications 3.6.5.3 and 3.7.2.

d. The provisions of Specification 3.0.3 are not applicable.

i SURVEILLANCE REQUIREMENTS j

- l

l .1 4.8.42 At least the above required A.C. electrical power sources shall be \ l

,j demonstrated OPERABLE per Surveillance Requirements 4.8.1.1.1; 4.8.1.1.2 and l 4.8.1.1.3, except for the requirement of 4.8.1.1.2.a.5. l

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so LA SALLE - UNIT 1 3/4 8-9

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(l i h3 ELECTRICAL, POWht SYSTUIS .

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.3/4.8 2 ONSITE POWER OISTRIBUTION SYSTEMS.

_'1_. . . .

t

s. A. C. DISTRIBUTION - OPERATING .

i !

i :

, ; LIMITING CLwn! TION FOR OPERATION ,

,i i

~i

! 3. 4. 2.1 The following A.C. distribution system electrical' divisions shall be l j OPERABLE and energized:

I a. Division 1, consisting of; -

i '

i 1. 4160-volt bus 141Y.

! 2. 480-volt buses 135X and 135Y.

I j 3. 480-volt MCC's 135X-1,135X-2,135X-3,135Y-1 and 135Y-2. I

!, ; 4. 120-volt A.C. distribution panels in 480 volt MCCs 135X-1, -

l 135X-2, 135X-3 and 135Y-1. .

,. i ,_

,. l b. Division 2, consisting of; *

\ i

! 1. 4160. volt bus 142Y. .

8

2. 480-volt buses 136X and 136Y.

480-volt MCC's 136X-1,136X-2,136X-3,136Y-1 and 136Y-2. I-3.

O. #

4. 120 volt A.C. distributiert panels.in 480 volt MCCs 136X-1, 136X-2, 136X-3 and 133Y-2.

c. Division 3, consisting of;

1. 4160-volt bus 143.' -

2. 480-volt MCC 143-1.

3. 120-volt A.C. distribution panels in 480 volt MCC 143-1.

d. Unit 2 Division 1, consisting of;

1. 4160-volt bus 241Y.

2. M Breaker 3H4 OPGeA6L.E or closh..

s =:t =C- : ? =-2 x: 2:5M-3.-

1. '*^ relt ^.C. diet.1,4 ...

,,e. ele "- i?? Yelt TC': 2 5 ; end lI 236N4:

e. Unit 2 Division 2, consisting of; .

1. 4160-volt bus 242Y.

2. 480-volt buses 236X and 236Y. l

3. 480-volt MCC's 236X-1, 236X-2, 236X-3, 236Y-1, and# 236Y-2.

. 4. 120 volt A.C. distribution panels in 480 volt MCC s 236X-1, 1 225X-2, 236X-3, and 225Y-2.~

- APPLIC.13I'.ITY: GPERATIO:tAL CCNDITIO:13 1, r.d 3.

~

LA SALLE - UNIT 1 3/4 8-10

l

+ ELECTRICAL POWER SYSTEMS c

h A.C. DISTRIBUTION - SHUTDOWN t LIMITING CONDITION FOR.0PERATION

!j. !

3.8.2.2 As a minimum, Division 1-end/or Division 2, and Division 3 when the j

!~

HPCS system is required to be OPERABLE, and Unit 2 Division 2 when the standby gas treatment system and/or the control room and auxiliary electric equipment i

,i room amergency filtration system are required to be OPERABLE, of the A.C. j l

l ,

distribution system shall be OPERAELE and energized with ,

l a. Division 1, consisting of; j 4160 volt bus 141Y.

1. l

2. 480 volt buses 135X and 135Y. I

3. 480 volt MCC's 135X-1, 135X-2, 135X-3, 135Y-1 and 135Y-2.

4. 120 volt A.C. distribution panels in 480 volt MCCs 135X-1,

,i 135X-2, 135X-3 and 135Y-1.

b. Division 2, consisting of;

1. 4150 volt bus 142Y.

2. 480 volt buses 136X and 136Y. I

3. 480 volt MCC*i 136X-1,136X-2,136X-3,136Y-1 and 136Y-2.

4. 120 volt A.C. distribution panels in 480 volt MCCs 136 X-1, 136X-2, 136X-3 and 136Y-2.

c. Division 3, consisting of;

. 1. 4160 volt bus 143

2. 480 volt MCC 143-1.

3. 120 voit A.C. distribution panels in 480 volt MCC 143-1.

j d. Unit 2 Division 2, consisting of; f 1. 4160 volt bus 242Y.

2. 480 volt buses 236X and 236Y. I

3. 480 volt MCC8 s 236X-1, 236X-2, 236X-3, and 236Y-1.

120 volt A.C. distribution panels in 480 volt MCCegi 236X-1, I 4.

236X-2, and 236X-3.

APPLICABILITY: OPERATIONAL CONDITIONS 4, 5,and *.

l^

l

'" O LA SALLE - UNIT 1 3/4 8-12

lt

!i jl'

< ELECTRICAL POWER SYSTEMS li

, D.C. DISTRIBUTION - OPERATING

.(r~5d 3

LIMITING CONDITION FOR OPERATION d i'

!l

! 3.8.2.3 The following D.C. distribution system electrical divisions shall be j jj. OPERABLE and energized:

I# _ _ L Division 1, consisting of;

~ ~ ~ ~ 1. 125 volt battery 1A.

! 2. 125 volt full capacity charger.

3. 125 volt distribution panel 111Y.

b. Division 2, consisting of;

. 1. 125 volt battery 18.

2. 125 volt full capacity charger.

!- 3. 125 volt distribution panel 112Y. '

c. Division 3, consisting of;

'2 1. 125 volt battery 1C.

l4 2. 125 volt full capacity charger. ,

I' 3. 125 volt distribution panel 113. t (C Unit 2 D b ision 1, consisting of N O - '

l 1. 125 volt battery 2A.

i I

- 2. 125 volt full capacity charger.

3 125 volt distributi n panel 211Y.

V

! -c. Unit 2 Division 2, consisting of;

1. 125 volt battery 28.

2. 125 volt full capacity charger.

l 3. 125 volt distribution panel 212Y.

l APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3.

ACTION:

l .'

a. With either Division 1 distrit thn p;n;1 111Y or Division 2 }

@ +- *>t % r -M 1114 inoperable or r.at energized, restore the

' inoperable division dktributka p;n;1 to OPERABLE and energized l

status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in at least HOT SHUTDOWN within the next 1 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b. With Division 3 d h tM h tfr p:n;l li> inoperable or not energized, declars the HPCS system inoperable and take the ACTION required by Specification 3.5.1.

With one division battery and/or battery charger inoperable, either:

1. Operation may continue provided the unit tie breakers for the

! affected division are OPERABLE and aligned to supply power to the affected distribution panel from the associated Unit 2 m

LA SALLE - UNIT 1 3/4 8-14

l' ELECTRICAL POWER SYSTEMS n

s'd v

LIMITING CONDITION FOR OPERATION (Continued)

J ~ ACTION: (Continued) y l

125t voit DC distribution panel", restore the inoperable battery !

f' and/or battery charger to OPERABLE status within .72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be l in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or

2. Restore the inoperable battery and/or battery charger to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Q.

d:'- With dit.er '.l ,'t 2 2hhh 1 dht:5tuc, ,. ..1 211Y cr Unit 2 i.

l

Division 2 sh i ivuuivi, n n.1 2127 inoperable or not energized, restore the inoperable division dhtriLtha ;;; .;l to OPERABLE and energized status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 7

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

V both Unit 2 Division 1 distribution panel 211Y and Unit 2 With

e.

Division 2 distribution panel 212Y inoperable or not energized, restore at least one of the inoperable distribution panels to OPERABLE j status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in at least HOT SHUTDOWN within the next l 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

'Os - s i

SURVEILLANCE REQUIREMENTS 4.8.2.3.1 Each of the above required D.C. distribution system electrical divisions shall be determined OPERABLE and enargized at least once per 7 days by verifying correct breaker alignment, indicated power availability from the charger and battery, and voltage on the panel with an overall voltage of greater than or equal to 125 volts.

4.8.2.3.2 Each 125-volt battery and charger shall be demonstrated OPERABLE:

a. At least once per 7 days by verifying that:

1. The parameters in Table 4.8.2.3.2-1 meet the Category A limits, and

2. Total battery terminal voltage is greater than or equal to 128 volts on float charge.

4'

" Unit 2 equipment in service to supply Unit 1 shall be demonstrated OPERABLE ACTIONS "a" and "b" shall be revised per Unit 1 Technical Specifications.

and ACTION "c" and this footnote shall be deleted upon issuance of an l

I pd perating License for Unit 2.

LA SALLE - UNIT 1 3/4 8-15

.c ... ,, ... u c. . _,

~

Q ' 1 ELECTRICAL POWER SYSTDes l y, l M ILLANCE REQUIREMENTS (Continued) 1 1 .

l b. At least once pr.:e 92 days and within 7 days after a battery discharge 1 i with battery voltage below 110 volts, or battery overcharge with l i battery terminal voltage above 150 volts, by verifying that:

g

. j 1. The parameters in Table 4.8.2.3.2-1 seet the Category 8 limits, i 2. There is no visible corrosion at either terminals or connectors, or the connection resistance of these items is less than 150 x 10 8 obspi,and l l

! 3. The average electrolyte temperature of at least 10 connected tt . cells is above 60*F.

l4 j i c. At least once per 18 months by verifying that:

i 1. The cells, cell riates and battery racks show no visual

.. indication of physical damage or abnormal deterioration, i

1 2. The cell-to-cell and terminal connections are clean, tight, free of corrosion.and coated with anti-corrosion material,-

l'~ 3. The resistance of each cell and terminal connection is less

-l than or equal to 150 x 10.e oy,and l

4. The battery charger will supply at least 200 amperes for division 1, 75 amperes for division 2 and 50 amperes for division 3 at a minimum of 130 volts for at least 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

1 d. At least once per 18 months, during shutdown, by verifying that 1 either:

. 1. The battery capacity is adequate to cupply and maintain in j OPERA 8LE status all of the actual emergency loads for the design cycle when the battery is subjected to a battery service

< test, or .

2. The battery capacity is adequate to suppi a dummy load, which i is verified to be greater than the actual emergency load, of the l following profile while maintaining the battery terminal voltage l

' greater than or equal to 105 volts. l l

a) Division 1, greater than or equal to: l

1) 483.4 amperes for the first 50 sec:nes, !

.s 2) 251.2 amperes for the nex '.4 minutas, j 227.7 amr.eres for tne next 15 .ainur.as,

' ,_. 3) )

4) 151.7 amperes for the next 30 minutes, and '

5) 83.7 amperes for the last 180 minutes. -l I LA SALLE - UNIT 1 3/4 8-16 l

. . . . .._ -. ... . . *. : f.'-- ..

i- - -

=ma9.w 9- -r - -

=, - - - .m. .-.--ye-- - - - - - - - - - - - - - -__ __ - - - - = * -

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

. ELECTRICAL POWER SYSTEMS i

SURVEILLANCE REQUIREMENTS (Continued) . _

li~

b) Division 2 greater than or equal to:

l 1) 488.5 amperes for the first 60 seconds,

2) 237.6 amperes for the next 14 minutes, ll 177.6 amperes for the next 15 minutes, and

.! 3)

4) 141.6 amperes for the next 30 minutes, and

?

5) 54.4 amperes for the last 180 minutes.

c) Division 3, greater than or equal to:

1) 58.4 amperes for the first 60 seconds, '

2) 11.1 amperes for the next 239 minutes.

Unit 2Okvision1,greaterthanorequalto: O ,

ll d)

1) 483.4 amperes for the first 60 seconds,

2) 251.2 amperes for the next 14 minutes, s

- 3) 227.7 amperes for the next 15 minutes,

4) 151.7 amperes for the next 30 minutes, and

5) 83.7 amperes for the last 180 minutes.

, d) p Unit 2 Division 2, greater than or equal to:

li 1) 488.5 amperes for the first 60 seconds,

2) 237.6 amperes for the next 14 minutes, l! 177.6 amperes for the next 15 minutes, 3) l2

' 4) 141.6 amperes for the next 30 minutes, and

5) 54.4 amperes for the last 180 minutes. .

Ii

e. At least once per 60 months, auring shutdown, by verifying that the

battery capacity is at least 80% of the manufacturers rating when ;

i subjected to a performance discharge test. Once per 60 month interval, this performance discharge test may be performed in lieu of the battery service test.

i f. Annual performance discharge tests of battery capacity shall be given to any battery that shows signs of degradation or has reached 85% of the service life expected for the application. Degradation is indicated when the battery capacity drops more than 10% of rated

- capacity from its average on previous performance tests, or is below 90% of the manufacturer's rating.

i' l

LA SALLE - UNIT 1 3/4 8-17

i

\

l ELECTRICAL POWER SYSTEMS D.C. DISTRIBUTION - SHUTDOWN l

v

, LIMITING CONDITION FOR OPERATION l

. 3.8.2.4 As a minimum, Division 1.and/or Division 2, and Division 3 when the l 'l '

HPCS system is required to be OPERABLE, and Unit 2 Division 2 when the standby

gas treatment system and/or the control room and auxiliary electric equipment l

room emergency filtration system are required to be OPERABLE, of the D.C. l distribution system shall be OPERABLE and energized with:

.; a. Division 1, consisting of;

1. 125 volt battery 1A.

2. 125 volt full capacity charger.

3. 125 volt distribution panel 111Y.

b. Division 2, consisting of;

1. 125 volt battery 18.

2. 125 volt full capacity charger.

. 3. 125 volt distribution panel 112Y.

c. Division 3, consisting of;

1. 125 volt battery 1C.

l ' (n) 2. 125 volt full capacity charger.

L 3. 125 voit distribution panel 113.

d. Unit 2 Division 2, consisting of; l

l

1. 125 volt battery 28.

I 2. 125 volt full capacity charger.

3. 125 volt distribution panel 212Y.

I APPLICABILITY: OPERATIONAL CONDITIONS 4, 5,and *. l ACTION:

a. With both Division 1 distribution panel 111Y and Division 2 distribution panel 112Y of the above required D.C. distribution system inoperable or not energized, suspend CORE ALTERATIONS, handling of irradiated fuel cask in the secondary containment and operations 2

with a potential for draining the reactor vessel.

l; b. With Division 3 distribution panel 113 of the above required D.C.

distribution system inoperable or not energized, declare the HPCS

' system inoperable and take the ACTION required by Specific.ations 3.5.2 and 3.5.3.

When handling irradiated fuel in the secondary containment.

l I

LA SALLE - UNIT 1 3/4 8-19

/ . . ... .

. . .:. i

. .. .. a . . . .

- 6 t -

j

~

. ELECTRICAL POWER SYSTEMS

4. . . . .. . . . . ..

3/4.8.3ELECTRICALEQUIPMENTPROTECTIVEDEVIClp -

A.C. CIRCUITS INSIDE PRIMARY CDt(TAINNENT ..

j '

t 8

LIMITING CONDITION FOR OPERATION ,

6

i i II t 3.8.3.1 At least the following A.C. circuits inside primary containment shall <

)

he de-energized *: i I

a. Installed welding grid systems 1A arid 18, and

b. All drywell lighting circuits. i

. c. All drywe.ll bi 43 W tra uss e.le w R s .

i APPLICA81LITY: OPERATIONAL ColeITIONS h-2 7 ard-3.---

j ACTION: ,

ifith any of the above required circuits energized, trip the associated circuit breaker (s) in the specified panel (s) within I hour. .

0W SURVE!LLANCE REQUIREMENTS i

{

' 4.8.3.1 Each of the above required A.C. circuits shall be determined to be de-energized at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ** by verifying that the associated

--l circuit breakers are in the tripped cor.dition.

l l

l

i l l 1

i I i l j "Except during entry inta the drywell. ,

1 - . )

- Exceot at least once per 31 days if 1cck:3, sealed or cthersise securse in the tripped cor.dition. - l

{,,

l 1

i 1 LA SALLE - UNIT 1 3/4 S-21

. .- . u O' .

v

}

v.

! . TABLE 3.8.3.2-1 1

.t j PRIMARY ,0NTAINMENT PENETRATION CON 00CTOR i ; QVERCURRENT PROTECTIVE DEVICE 5 t

TRIP RESPONSE SYSTEM /

' TIME DEVICE NUMBER SETPOINT ANO LOCATION (Amperes) (Milliseconds / Cycles)(a)POWEREO COMPONENT

a. 6.9 KV Circuit Breakers

1. Swgr. 151 (Campt. 4) 840 CC) 83.3/5 RR Pump 1A

~

2. Swgr. 152 (Compt. 4) 840(c) g3,375 gg pu,, gg

. i 3. Swgr.151-1 (8kr. 2A) 720(b) 83.3/5 RR Pump _1_A, _,

low speed .

4. Swgr.152-1 (8kr. 28) 720(b) 83.3/5 RR Pump 18, low speed i b. 480 VAC Circuit Sceakers

1. .%gr.136Y (Compt.. 160(C) 50/3 VP/Pri. Cont.

403C) Vent Supply Fan 18 k ~~ 2. Swgr. 135Y (Compt. 160(C) 50/3 VP/Pri. Cont.

l

203A) Vent Supply Fan IA~

I j c. 480 VAC (Molded Case) Circuit Breakers ~

i - P j

t Type X-M Cat # NZ Mi-160/ZM6 l a) MCC 134Y-3 36 N.A. HC/ Hoist Motor j j (Compt. A10)

HC/Eqpt. Handling

.I 25 N.A.

b) MCC 134X-2 (Compt. A4) Platform c) MCC 134X-2 60 N.A. HC/ Crane &

Compt. 03) Trolley Power

.I

1 I g. Type K-M Cat # NZ Mi-160/IM6C .

4 a) MCC 136Y-2 174 N.A. RR/MOV 1833-F0678 l~ .(Comet. C4)

.i b) MCC 136Y-2 72 N.A. RR/MOV 1833-F0238 (Compt. A3) l ,

! c) MCC 134X-1 10 N.A. N9/MOV1 1921-F001

. !. (Compt. B3) -

10, N.A. NBiMOV 1821-F002

[d

. d) MCC 134X-1

/ (Compt. 84)

~

LA SALLE - UNIT 1 3/4 8-24 e ,.

mm i..

. 2 _ ._ _ . ..__.. . . . . _ . . _ . _ _ . _ .

i.

h.

ELECTRICAL POWER SYSTDt5

!)

MOT 0S OPERATED VALVES THERMAL OVERL0A0 PROTECTION y

Limimu CONDITION FOR OPERA 1 ION l

y --:_ r .-- ,

i i

3.8.3.3 The thermal overload protection of each valve shown in Table 3.8.4.2-1

~

shall be bypassed continuously or under accident conditions, as appitcable, by

- an CPERABLE bypass device integral with the motor starter.

APPLICA8ILITY: Whenever the motor operated valve is required to be OPERA 8LE.

ACTION:

I au With the thermal overload protection for one or more of the above required valvas ;

not bypassed continuously or under accident conditions, as applicable, by an i OPERA 8LE integral bypass device, take administrative action to continuously

~

bypass the thermal overload within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or declare the affected valve (s) inoperable and apply the appropriate ACTION statement (s) for the affected -

system (s). ,

b. Tle. provisious f etC ca4to a 8 0 4 49e. *

Aff!<Ah/8-. l i i SURVEILLANCE REQUIREMENTJ 0:(.'.

4.8.3.3.1 The thermal overload protection for the above required valves shall be j.. verified to be bypassed continuously or under accident conditions, as applicable, ,

! by an OPERABLE integral bypass device by the performance of a CHANNEL FUNCTIONAL TEST of the bypass circuitry for those thermal overloads which are normally in -

' force during plant operation and bypassed under accident conditions and by i verifying that the thermal overload protection is bypassed for those thermal I overloads which are continuously bypassed and temporarily placed in force only ,

when the valve motors are undergoing periodic or maintenance tasting:

a. At least once per 18 months, and i '

b. Following maintenance on the motor starter.

! 4.8.3.3.2 The thermal overload protection for the above required valves which are continuously bypassed shall be verified to be bypassed following testing f during which the thermal overload protection was temporarily placed in force.

i t .

l l,

LA SALLE - UNIT 1 3/4 8-26

l l

w.u

. o .t. *

.6 j .

. TA8LE 3.8.3.3-1 i ~

MITOR OPERATED VALVES THERMAL OVERLOAD PROTECTION .

! 8YPA55 DEVICE SYSTEM (5)

! VALVE NUMBER (Continuous)(Accident Conditions) AFFECTED

'a. IVG001 Accident Conditions 58GTS 1VG003 - Accident Conditions 2VG001 Accident Conditions .

2VG003 Accident Condition.s

b. IVP113A Accident Conditions Primary containment IVP1138 Accident Conditions chilled water coolers IVP114A Accident Conditions *

. IVP1148 Accident Conditions IVP053A Accident Conditions IVP0538 Accident Conditions --

' IVP063A Accident Conditions IVP0638 Accidont Conditions

c. IVQ040 Accident Conditions Primary containment

. IVQO36 Accident Conditions vent and purge system Accident Conditions

.G .IVQ026 IVQ029 Accident Conditions '

IVQO38 Accident Conditions i '

1VQ0e ^ -

M t C . m " .... 1 _

4 IVQO31 Accident Conditions ;

' Accident Conditions IVQO32

, IVQ034 Accident Conditions

IVQ035 Accident Conditions

! IVQ027 Accident Conditions .

! IVQ042 Accident Conditions

} IVQ043 Accident Conditions l IVQ047 A'ccident Conditions IVQ048 Accident Conditioris

, IVQ050 Accident Condition 1

IVQ051 Accident Conditions i IVQ068 Accident Conditions IVQO30 Accident Conditions -

IVQ037 Accident Conditions

d. 1WR179 Accident Conditions R8CCW system l 1WR180 Accident Conditions 1 1WR040 Accident Conditions 1WR029 Accident Conditions

e. 1821 - F057A Accident Conditions Main steam systes

! 1821 - FC67B Acetdent C:nditiens

' . 1821 - FC57C Accicent Conditions J 1821 - F067D Accident Conditions g 1821 - F019 Accident Conditions

, 1821 - F016 Accident Conditions l

LA SALLS - UNIT 1 3/4 8-27 ee

l . --

e

~ ~ **

~

i -

EO

~

ELECTRICAL POWER SYST9t5

( ;

REACTOR PROTECTION SYSTEM ELECTRICAL POWER,NONITORINE U '4 LIMITING CONDITION FOR OPERATION l- 3.8.3.4 Two RPS electric power monitoring assemblies for each inservice RPS

'. l MG set or alternate power supply shall be OPERABLE.

APPLICA81LITY: At all times.

ACTION:

s.

a. With one RP5 electric power monitoring assembly for an inservice RPS MG set or alternate power supply inoperable, restore the inoperable power monitoring assembly to OPERA 8LE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or remove the associated RPS MG set or alternate power supply froe

.1

, service. .

4 With both RPS electric power monitoring assemblies for an inservice b.

RPS MG set or alternate power supply inoperable, restore at least one electric power monitcring assembly to OPERA 8LE status within l 30 minutes or remove the associated RPS MG .et or alternate power ~

suppiy from service.

~ ~

SURVEILLANCE REQUIREMENTS v

1 l t 4.8.3.4 The above specified RPS eisctric power monitoring assemblies shall be -

determined OPERA 8LE:

B

a. rC ;.r.;,#y I -At TEST, ? end 70 de

d;4ime es. glewf

Ls /#performance COLD of a CHANNEL FUNCT l -kaa M hows upless s perlernelin 4ks pesvious (,nouAs.

b. At least once per 1) months by demonstrating the OPERA 8It.ITY of '

i overtvoltage, undertvoltage, and under2 frequency protective .I instrumentation by performance of a CHANNEL CALIBRATION including simuTated automatic actuation of the protective relays, tripping logic and output circuit breakers and verifying the following i

] setpoints.

.j 1. Oveholtage < 132 VAC, ,

2. Unde (voltage 1108VAC,

3. L,dedfrequency357Hz.

O I

t LA SALLE

UNIT I 3/4 8-31

- e- -~

s 5

REFUELIhG OPERATIONS

~

SURVEILLANCE REQUIREMENTS (Continued)

b. Performance of a CHANNEL FUNCTIONAL TEST:

1. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to the start of CORE ALTERATIONS, and 1

2. At least once per 7 days.

c. Verifying that the channel count rate is at least 0.7 cps #1 l

1. Prior to control rod withdrawal,

~

2. Prior to and at least onca per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during CORE ALTERATIONS, '

and

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

d. Verifying that the RPS circuitry "thorting links" have been removed within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior to and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during: ,

1. The time any control rod is withdrawn," or

2. Shutdown margin demonstrations.

O.

s

" Not required for control rods removed per Specification 3.9.10.1 or 3.9.10.2.

O- $ i'rottded sayal-to-natse t d'ho 1s 2 2. Othernhs e, 3 cps. l LA SALLE - UNIT 1 3/4 9-4 Amendment No. 2

V :--a

^

. - - +: .a. . . . , - , .-: .

)' .

(. 4

~ . .

< ,s REFUELING OPERATIONS ,

.i . ... 3

3/4.9.11 RESIDUAL HEAT RfNOVAL AND COOLANT CIRCULATION

. 1

. HIGH WATER LEVEL

~.

l.

j LIMITING CON 0! TION FOR OPERATION .

i

! 3.9.11.1 At least one shutdowg cooling mode loop of the residual heat removal j (RNR) system sha11 he OPERA 8LE" and in operation

with at least:

I a. Ona OPERA 8LE RHR pump, and 1,

b. One OPERA 8LE RHR heat exchanger.

i i APPLICA8ILITY: OPERATIONAL CONDITION 5, when irradiated fuel is in the reactor vessel and the water level is greater than or equal to 22 feet above the top j

t of the reactor pressure vessel. flange. ,

t ACTION:

. .L

a. With no RHR shutdown cooling mode loop OPERA 8LE, within.oner hour and at l .

least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter, demonstrate the operability of at i ~ . . . .. least one alternate method capable of decay heat removal. Othenvise,

, suspend all operations involving an increase in the reactor decay heat

. d load and establish SECON0ARY CONTAINMENT INTEGRITY within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

j

b. With no RHR shutdown cooling mode loop in operation, within hour [

l establish reactor coolant circulation by an alternate method and monitor

reactor coolant temperature at least once per hour.

i I

SURVEILLAN'CE REQUIREMENTS l

?

. ! 4.9.11.1 At least one shutdown cooling mode loop of the residual heat removal system or alternate method shall be verified to be in operation and circulating

- reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

'l 1

i . .

'I "The snute:wn ::c'ing :=p be re: coved fr: ::srs .io.. f:r r: to 2 .:urs :er S-noor period.

,) The normal or emergency power source may be inoperable.

a LA SALLE - UNIT 1 3/4 9-16

~"' ..

. . : . .. . . . , . . v .. .

a b =

I.

)f '

Oq REFUELING OPERATIONS

~...'

' LOW WATER LEVEL . .

i i

~

I LIMITING COMOITION FOR OPERATION i !

3.9.*u.2 Two shutdown eBoling moda loops ~of the residual heat removal (RHR)

! system shall be OPERA 8LE and at least one loop shall be in operation," with-i ser.h loop consisting of at least:

'. 1

} a. One OPERA 8LE RHR pump, and

b. One OPERA 8LE RHR heat exchanger.

'l APPLICA8ILITY: OPERATIONAL ColWIT10N 5, when irradiated fuel is in the reactor I vessel and the water level is less than 22 feet above the top of the reactor j~ pressure vessel" f1ange.--- -

t 1

ACTION:

a. With less than the above required shutdown cooling moda loops of the RHR system OPERA 8LE, within h hour and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter, l i demonstrate the operability of at least one alternate method capable of a decay heat removal for each inoperable RHR shutdown cooling mode loop.

O' b. With no RHR shutdown cooling mode loop in operation, within 45e hour establish reactor c.colant circulation by an alternate method and monitor g

reactor coolant temperature at least once per hour.

I i SURVEILLANCE REQUIREMENTS I

4.9. u.2 At least one shutdown cooling mode loop of the residual heat removal l system or alternate method shall be verified to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . ,

>l 4

"The sautccwn cooHng etnp may be ca.:cvec :'m 0:2racter. 'cr ec o 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months t

,#'- per 3-hour period.

The normal or emergency power source may be inoperable for each loop.

!. ~

LA SALLE - UNIT 1 3/4 9-17

a -.4 .. h

___-4 --

)

. ._.. \

, e -- -- -

. i.

h., .

~

V' - TABLE 4. u.1-1

'fj - RADI0 ACTIVE LIQUID WASTE SAMPLING MD ANALYSIS PROGRAM p _

E k -

Minimum Type of Lower Limit Liquid Release Sampling Analysis Activity of Detection Type Frequency Frequency Analysis (LLD) l (pCf/al),

A. Batch Waste P P Principal Gamma 5x10-7

'i Release Each Batch Each Batch Esitters#

d Tanks

. 1-131, 1x10 ,

P M Ofssolved and 1x10-5

One Batch /M Entrained Gaies (Gammaemitters)

-5 .

P M H-3 1x10 b

Each Batch Composita l - ..

Gross Aloha 1x10

( gH

~8 P Q Sr-89, Sr-90 5x10 __

b -

i Each Batch Composite -

~5 i, i Fe-55 1x10 i

~7 B. Continuous W Principal Gamma 5x10 C

Releasas' Continuous" Composite Emitters #

~5 I-131 1x10 4

~5 i M M Dissolved and 1x10 i Grah Sample Entrained Gases (Gamma Esitters) ,

~5 M H-3 1x10 C

I -

Continuous' Composite

~

, Gross Alpha _ 1x10"I

(

m Q ;Sr-89, ir-90 l 5x10 ' l

-6

. Continuous C

Composite C Fe-55 1x10 LA SALLE - UNIT 1 3/4 11-3

,--,-w- an--- ,- _, , - - - - . - - , , . , - , - . - - , . , , , , . . ,. - -- , . - -

~

.- . . . . . .z.

a . ..e .

(

~

Q. JK l, l - ./ RADI0 ACTIVE EFFLUEKr5 4

3/4.11.2 GASEQUS EFFLUENTS .i

, j ..; . <

m-~- 005E RATE

, ; LIMITING CONDITION FOR OPERATION 4

3.11.2.1 The dose rate due to radioactive materials released in gaseous effluents from the site (see Figure 5.1.1-1) shall be limited to the following:

l a. For noble gases: Less than or equal to 500 ares /yr to the total I: body and less than or equal to 3000 ares /yr to the skin, and

, b. For all radiciodines and for all radioactive materials in particulate

form and radionuclides (other than noble gases) with half lives greater than 8 days: Less than or equal to 1500 ares* (yr to any l -

organ vaa he mhalation pathway. I APPLICA8ILITY: At all times. -

_A,CTION

~

liith the dose rate (s) exceeding the above limits, immediately decrease the .

%, release rata to within the above Ifmit(s).

O'Q .

, ! SURVEILLANCE REQUIREMENTS I ~

i l

4.11.2.1.1 The dose rate due to noble gases in gaseous effluents shall be determined to be within the above limits in accordance with the methods and procedures of the 00CM.

4.11.2.1.2 The dose rate due to radioactive materials, other than noble gases, in gaseous affluents shall be determined to be within the above limits in accordance with the methods and procedures of the 00CM by obtaining represen-l tative samples and performing analyses in accordance with sampling and analysis program specified in Table 4.11.2-1.

t I

os] b l

. 1 LA SALLE - UNIT 1 3/4 11-9 f

._._..s_. . . . _ _ . . . _ . . . _.

c-

.c. .~ . ,.

I-h - <

-]'"

TA8LE 4.11.2-1 (Continued)

TABLE NOTATION .

j

[. b. Analyses shall also be performed following shutdown, startup, or a '

M THERMAL POWER change exceeding 15.p y eerit of the RATED THERMAL POWER

~ ~~

i within a, par hour period. 7e

.i.

'. Whenever there is flow through the 58GTS.

c.

~ ^

d. Samples shall be changed at least once per 7 days and analyses shall be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after changing /or after removal from sampler). Sampling shall also be performed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for at least 7 days following each shutdown, startup or THERMAL POWER change exceeding 15 percent of RATED THERMAL POWER in hour and l l :. analyses completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months of changing. When samples collected for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months are analyzed, the corresponding LLD's any be . .

increased by a factor of 10. ,

e. Tritium grab samples shall be taken at least once per 7 days from the plant vent to detemine tritium releases in the ventilation exhaust from the spent fuel pool area whenever spent fuel is in the -

! spent fuel pool.

O- :

f. The ratio or the soie n- rate to the sa. pied stre- fi- rate shall be known for the time period covered by each dose or dose rate

' calculation made in accordance with Specifications 3.11.2.1, 3.11.2.2 and 3~.11.2.3.

I g. The principal gassa emitters for which the LLD specification appifes -

! midade.4-ene4eebi,-n 2; following radionuclides: Kr-87. Kr-88, Xe-133, g Xe-133s, Xe-135, and Xe-138 for gaseous emissions and Mn-54, Fe-59, Co-58, Co-60, In-65, Mo-99, Cs-134. Cs-137, Ce-141 and Ce-144 for particulate emissions. This list does not mean that only these nuclides are to be detected and reported. Other peaks which are measurable and identifiable, at the 95% confidence level, together i

j j with the above nuclides, shall also be identified and reported.

i i

. This$quirementdoesnotapplyif(1)analysisshow's that the DOSE EQUIVALENT I-131 concentration in the primary coolant -

has not increased more than a factor of 3; and (2) the noble gas I I monitor shows that affluent activ,ity has not increased more than a factor of 3.

/

e i

LA SALLE - UNIT 1 3/4 11-12

_ce..,-.--,,m..- , _ - - , - . , . - _ ,,.,,_-m,- ,,,, ,- ,__. ,._.,, ,__ ,.- -,.-_, - - .- , - - - - , - . -,..___,__,._,---__--.-y,.,

~ * * ~ ~

~

a~ - -a :. .s *~ ~~-.. --~- * --

_. . . . .. .- -- - .- - ---. n. - -

(.

'.Q. % >.

I " RADI0 ACTIVE EFFLUENTS 00SE - NOBLE GASES .[ -

e LIMITING CONDITION FOR OPERATION _ __

~ ~ - ~ ~ ~

y...-....-.- -

3.11.2.2 The air dose due to noble gases released in gaseous effluents, from each reactor unit, from the site (see Figure 5.1.1-1) shall be limited to the following:

a. During any calendar quarter: Less than or equal to 5 erad for gamma radiation and less than or equal to 10 mrad for beta radiation, and

b. During any calendar year: Less than or equal to 10 mrad for gamma

' radiation and less than or equal to 20 arad for beta radiation. .

i

~~

APPLICA8ILITi: At all times.

. ACTION: ,

a. With the calculated air dose from radioactive noble gases in gaseous Av/ effluents exceedtng any of the above limits, in lieu of any other report required by Specification 6.6.A or 6.6.B. prepara and submit to the Commission within 30 days, pursuant to Specification 6.6.C, a Special Report which identifies the cause(s) for exceeding the limit (s) and defines the corrective actions to be taken to reduce the releases M c:dk:;th; n;th ;;;;: ia a~ eece ;ffh:..;5 dsring

~

i i

2: - : thd;r :f 2: cc ---t cale-d=- a"=*** =ad de- %g th 7 sch :;;;at t.'.rt: cah d-- ;ut-*:r 50 M it the c" ulethe dete **

! af th4 E--1G e. .d Tv. ; - : --d h t ka : d 2a - !d fer he t 7; diet ha.-

.l b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

'I t

! SURVEILLANCE REQUIREMENTS I .

I 4.11.2.2 Dose calculations Cumulative dose contributions for the current l

calendar quarter and current calendar year shall be determined in accordance with the 00CM at least once per 31 days.

I to as rum

! Ano TE PaoPoss concocrws Acnovs

%, TO ASSO Qe TlMT ,506SeQuedi" (eLeASc3 UlLLL BG-OH lA) CBM9Li A^1LG wsTH TH5 A60/5 LeN LTS.

1 LA SALLE - UNIT 1 3/4 11-13

_ _.:. a .-

ksS

..~ .7 ~ RADI0 ACTIVE EFFLUENTS 1 00SE - RADIGICDINES, RADI0 ACTIVE MATERIALS IN PARTICULATE FORM, AND RADIONUCLIDES OTHER THAN NOBLE GA5E5 -

); - LIMITING CONDITION FOR OPERATION 3.11.2.3 The dose to an individual from radiofodines and radioactive materials in particulate form, and radionuclides, other than noble gases, with half-lives greater than 8 days in gaseous effluents released, from each reactor unit, frun the site (see Figure 5.1.1-1) shall be limited to the following:

a. During any calendar quarter: Less than or equal to 7.5 are(to any }

organ, and 3

b. During any calendar year: Less than or equal to 15 mres,to any (

. : organ. .-

APPLICABILITY: At all times.

ACTION: '

O- a. With the calculated dose from the release of radiofodines, radioactive materials in particulate form, or radionuclides (other than noble gases) with half lives greater than 8 days, in gaseous effluents exceeding any of the above limits, in lieu of any other report required by Specification 6.6.A or 6.6.B. prepare and sabmit to the

-- Commission within 30 days, pursuant to Specification 6.6.C. a A have.beu b, Special Report and defines the corrective which identifies actions t- the t; t;h:2 cause(s)

M -d"en for the re1=M exceeding the limi 4 r:df:i:dian end cedie 4! " g g niaams "f::=lidn, c'tr t';n adi;; g;;n, dth

%... m ur.... . . . , , . . . . . . . . . . . . . , . . . . -

heif-1!;n ,. ;;ter th

=l bdd A ('l8#y

d:p ia ;;;;;;; ;ff1=nt: durin; the . inu.. of the ;;cr:n'-

gg gjians 4o cp1-a5 ;nn:r ud during th; 3 ;;;;;;;at thc;; nind:r q=n:r:,

f lAN' . d"*" '- i ""* " ' * "" I '"' A ' ""* I "

[_ M bea do AES 0' is=iuun u . ."'."."*w*

1_5. .$. 2....

.. .ni- e. gen.

M soleydg i

~- i g,,,s wM k j b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

}ipcoqUAMS iwl1L 4'M M 4 ;y,*43 SURVEILLANCE REQUIREMENTS

.. L i

1 4.11.2.3 Dese Calculations Cumulative dose contributices for the current calendar qua ter anc current calei:ar year snall be cetarmined in actor ance

=- - '

with the COC:1 at least once per 31 Jn s.

V ,

LA SALLE - UNIT 1 3/4 11-14

- -i .

_ ~. . .s ~ ~ * .

M n k.\~Tw .- =.. - :-. . s

a. . . . _ _ .

~

i O. g

(,, RADI0 ACTIVE EFFLUENTS VENTING OR PURGING ,

LIMITING CONDITION FOR OPERATION 3.11.2.8 VENTING or PURGING of the containment drywell shall be through the Primary Containsect Vent and Purge System or the 5,tandby Gas Treatment System.

- APPLICA8Ij.ITY: Whenever the drywell is vented or purged.

ACTION:

. a. With the requirements of the above specification not satisfied, suspend all VENTING and PURGING of the drywell. ,

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS

- y.

D

4. .8 The containment drywe shall be determined to be aligned for VENTING l or PU % through the Primary C tainment Vent and Purge System or the Standby Gas Trea t System within 4 hou prior to start of and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months dur VENTING or PURGING f the drywell. .

4.11.2.8.1 The containment drywell shall be determined to be aligned for VENTING

or PURGING through the Primary Containment Vent and Purge System or the Standby j

Gas Treatment System within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to start of and at least once per

. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during VENTING or PURGING of the drywell.

.4.11.2.8.2 Prior to use of the Purge System through the Standby Gas Treatment .~

System in OPERATIONAL CONDITION 1, 2 or 3 assure that:

l a. Both Standby Gas Treatment System trains are OPERABLE, and

!, i 1 j b. Only one of the Standby Gas Treatment System trains is used for PURGING.

e l .LA SALLE - UNIT 1 3/4 11-19 l . ..-

~. .. . . . . v. - .h . = . .a. .. .= ..... .... . . . .. .: ..:.

1. =. . ,,

i

h. .

3/4.12 RADIOLOGICAL ENVIRONMENTAL NONITORING i 3/4.12.1 NOMITORING PROGRAM . .

6

~

LIMITING CONDITION FOR OPERATION li 3.12.1 The radiological environmental abnitoring program shall be conducted

, as specified in Table 3.12.1-1.

l- APPLICA8ILITY: At all times.

ACTION:

a. With the radiological environmental monitoring program not being i

conducted as specified in Table 3.12.1-1, in lieu of any othery (,,g,p op (o.6,8 i report required by Specification J#1,~ prepare and submit to the j

! Commission, in the Annual Radiological Operating Report, a descrip-tion of the reasons for not conducting the program as required and .

the plans for preventing a recurrence.

. b. With the level of radioactivity in an environmenial sampling medium i exceeding the reporting levels in Table 3.12.1-2 when averaged over l any calendar quarter, in If eu of any other report required by b 'W spectrication%Srt; prepare and subarit to the Commission within 30 Os' <<e a r <r th < < th <<

a i ra a a et >=r at to Specification 6.9.1.13. When more than one of the radionuclides 3" ~- in Table 3.12.1-2 are detected in the sampling medium, this report

shall be submitted 17:

t concentration (1) , concent ation (2) + *** > 1.0

limit level (1) limit level (2) -

I* When radionuclides other than those in Table 3.12.1-2 are detected

- and are the result of plant effluents, this report shall be submitted if the potential annual dose to an individual is equal to or greater than the calendar year limits of Specifications 3.11.1.2, 3.11.2.2 1 and 3.11.2.3. This report is not required if the measured level of

} radioactivity was not the result of plant effluents; however, in

! such an event, the condition shall be reported and described in the Annual Radiological Environmental Operating Report.

c. With milk or fresh leafy vegetable samples unavailable froa one or more of the sample locations required by Table 3.12.1-1, in lieu of (,,6.A v any other report required by SpecificationJw971, prepare and submit Q,g,g -

i to the Commission within 30 days, pursuant to Specification 6,4-2 Special Report which identiff as the cause of the unavailability oh, a g*g '

l i samples and identifies loca'ans fcr obtaining replacement samples.

j The locations from which samples were unavailable may then be deleted

frem those required by hb!e 3.12.1-1, ;: revidad t' e !ccations frem t wnica the replacement sar:los were cbtainec are ac:ed to tne environmental monitoring ;regram ss replaca ent Icesticas.

Ov ;

d. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

LA SALLE - UNIT 1 3/4 12-1 .

m, n y- www--e - - , - - - + -m-.s , ee--,,ww ---mm,.w-,- m,,mww,wwe-rw-

II Q- g

- Q ..:.- . -..

.- l TABLE 3.12.1-1 ,

C RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAN :

W ; .

Number of Samples a Exposure Pathway and Sampilng and Type and Frequency g and/or Sample Sample Locations 8 Co11 action Frequency of Analysis ;

~

" 1. AIR (10RNE 6 lot Ahous * -

R.utiolodine and ('erett: : 1 3) Continuous operation of Radiolodine canister. -l 1 P wticulates saapier with sample col- Analyze at least once . .

1ection as required by g.ar 7 days for I-131. .

dust loading but at least once per 7 days. Particulate sampler.

Analyze for gross beta ;

radioactivity > 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months !.

following filter change. i, Perform gamma isotopic l-analysis on each sample i when gross beta activity :

i' R*

is > 10 times the yearly mean of control samples.

U * -

Perform gamma isotopic g O -

analysis on composite ,.

(by location) sample ;'

at least once per j 92 days.

3 8 locA4;our ,.

2. DInfCT RADIATION (L:::ttrr: 0- 05)-

> 2 dosimeters or > 1 At least once per 31 daye.

or' Gamma dose. At least once per 31 days.

[ .

p Tnstrument for con- or .

tinuously measuring At least once per 92 days. Gamma dose. At least and recording dose (Read-out frequencies are once per 92 days. .'.

rate at each determined by type of dosin-location. eters selected.) l

- LZ,fil Metie..e sie gi.ee en the figur ...e table-in-the 00CMT i

/- . .....T

r a \ :.

-) .. . i .,..

j 3 TABLE 3.12.1-1 (Continued) ;i G RADIOLOGICAL ENVIRONNENTAL MONITORING PROGRAM i

u. -

? !

E [

, Minlaum c Number of Samples '

and Sampljne and Type and frequency }

M b posure Pathway- Collection Frequency of Analysis H and/or Sample Sample Locations * ,.

s:*

3. WAIERBORl:E. ',

2 locations . Composite sample collected Gassa isotopic analysis ,

a. Surface of each composite sample, l

over a period of < 31 days, Tritium analysis ~of com-

- posite sample at least i once per 92 days. !

5 locations At least once per 92 days. Gamma isotopic and .

b. Ground trittua analyses of M '

each sample.

+ .

Gamma isotopic analysis

\

c. Sedleent free 1 location At least once per 184 days.

of each sample.

Shoreline h !.

Yample-4 cat +en -*e: &~- e e: f!:;re f.. the-cocM-I -

. l.

l ..

Ir i t 4

e

^

l l

l 8

t r "m .

\ _s j 3/4.0 APPLICA8ILITY e

l;

SASES t

i ,

The specifications of this section provide the general requirements applicable to each of the Lietting Conditions for Operation and Surveillance Requirements within Section 3/4.

p; sert. A -*'

3.0.1 This specification states the applicanility of each specification in tems of defined OPERATIONAL CONDITION or other specified appifcability condition ard is provided to delineate specifically when each specification is appifcable.

3.0.2 This specification defines those conditions necessary to constitute compliance with the tems of an individual Limiting Conditten for Operation and associated ACTION requirement. .

3.0.3 This specification delineates the asasures to be taken for circumstances not directly provided for in the ACTION statements and whose occurrence would violata the intent of the specification. For example, Specification 3.7.2 requires two control room and auxiliary electric equipment O' .,

U r= ' a=v 'n* *4 a *r='a= ==

aasaaa'5 a* ar 'a requirements if one train is inoperable. Under the requirements ofg5pecifica-tiotr 3.0.3, if both of the required trains are inoperable, within one hour a"c'* ac7to" i

esasuresr must be initiated to place the unit in at least STARTUP within the l

next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in l ,

COLD SHUTDOWN within the subsequent 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . As a further example, Specification 3.6.6.1 requires two primary containment hydrogen recombiner i systems to be OPERA 8LE and provides explicit ACTION requirements if one recombiner system is inoperable. Under the requirementsgof Specification 3.0.3, I if both of the required systems are inoperable, within.ent hour seas Jres must i be initiated to place the unit in at least STARTUP within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and M 6 in at least NOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

l l 3.0.4 This specification provides that entry into an OPERATIONAL CONDITION l

sust be made with (a) the full complement of required systems, equipment or

!' components OP3RA8LE and (b) all other parameters as specified in the Limiting l , Conditions for Operation being met without regard for allowable deviations and out of service provisions contained in the ACTION statements.

i The intent of this provision is to ensure that unit operation is not l

, initiated with either required equipment or systems inoperable or other limits being exceeded.

I Exceptions to this provision have been provided for a limited number of

!' specifications when startup with inoperable ecuipment would not affect plant safety. These exceptions are stated in the ACTION statements of the appropriate j i specifications.

l l

l LA SALLE - UNIT 1 8 3/4 0-1

s pseus roe m se B 9/+ *-l l

Q . .

l j %ntsasT e i

In the event of a disagreement between i the requirements stated in these Technical Specifications and those stated

in an applicable Regulation or Act, the requirements stated in the applicable Regulation or Act, shall take precedence and shall be met.

i O

suster B

. It is acceptable to initiate and complete a reduction in OPERATIONAL CONDITIONS in a shorte: time

} interval than required in the ACTION statement and to and the unused portion -

of this allowable out-of-service time to that provided for operation in sub-sequent lower OPERATIONAL CONDITION (S). Stated allowable out-of-service times

are applicable regardless of the OPERATIONAL CONDITION (S) in which the inopera-bility is discovered but the times provided for achieving a CONDITION reduction are not applicable if the inoperability is discovered in a CONDI-TION lower than the applicable CONDITION.

I t

O 1

) EACTIVITY CONTROL SYSTEMS

: 3/4.1 REACTIVITY CONTROL SYSTEMS k-Ii

' ~BASES- .

'i 3/4.1. 1 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that(1) the reactor can be made I

subcritical from all operating conditions,(2) the reactivity transients 8 associated with postulated accident conditions are controllable within acceptable limits, anaQ) the reactor will be maintained sufficiently I suberitical to preclude inadvertent criticality in the shutdown condition.

Since core reactivity values will vary through core life as a function of

~

funi depletion and poison burnup, the demonstration of SHUT 00WN MARGIN will be performed in the cold, xenon-free condition and shall show the core to be subcritical by at least R + 0.38K delta K or R + 0.28K delta K, as appropriate.

The value of R in units of % delta K is the difference between the calculated value of maximus core reactivity during the operating cycle and the calculated beginning-of-life core reactivity. The value of R aust be positive or zero and sust be detemined for each fuel loading cycle.

oj s Two different values are suppliM in the Limiting Condition for Operation to provide for the different methods of demonstration of the SHUTDOWN MARGIN.

v The highest worth rod any be determined analytically or by test. The SHUTDOWN MARGIN is demonstrated by an insequence 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 indicates that the required margin could be reduced as a function of exposure. Observation of suberiticality in this i condition assures subcriticality with the most reactive control rod fully j withdrawn.

t

! This reactivity characteristic has been a basic assumption in the i analysis of plant performance and can be best demonstrated at the time of fuel

'. I loading, but the margin must' also be determined anytime a control rod is incapable of insertion.

3/4.1.2 REACTIVITY ANOMALIES

\

Since the SHUTDOWN MARGIN requirement for the reactor is small, a careful check on actual conditions to the predicted conditions is necessary, and the changes in reactivity can be inferred from these comparisons of rod patterns.

l . Since the comparisons are easily done, frequent checks are not an imposition l on normal operations. A 1% change is larger than is expected for normal

- j 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 postu: ated transients.

LA SALLE - UNIT 1 8 3/4 1-1

I

l l

i . . .

l !!

i h

M j REACTIVITY CONTROL SYST995 BASES

{

. i.

3/4.1.3 CONTROL R005 l! The specificationsof this section ensure that (1) the minimus SHUTDOWN I

', i i MARGIN is maintained (2) the control rod insertien times are consistent with h, those used in the accident analysis, and (3)-Matt the potential effects o.f l the red dron accidenta The ACTION statements permit variations from the basfc

.g requirements but at the same time impose more esstrictive criteria for continued operation. A limitation on inoperable rods is set such that the resultant

'l'; .. gE,.g effect on total rod worth and scras shape will be kept to a afnisus. The requirements for the various scras time seasurements ensure that any indication

} of systematic problems with rod drives will be investigated on a timely basis.

!' Oamage within the control rod drive mechanism could be a generic probles, therefore with a control rod famovable because of excessive friction or sechanical interfennce, opention of the reactor is 11mitad to a time period which is l- .

reasonable to determine the cause of the inoperability and at the sans time i

prevent operation with a large number of inoperable control rods.

1 Control rods that are inoperable for other reasons are permitted to be l

O, j taa

ar rvic ar 44 4 ta t ta ia ta aaaruiiv-ia rt d aa itiaa r-consistent with the SHUTDOWN MARGIM requirements. '

The number of control rods permitted to be inoperable cogid be more than

- the eight allowed by the specification, but the occurrence of eight inoperable i-rods could be indicative of a generic probles and the reactor must be shutdown -

j for investigation and resolution of the probles.

l The control rod system is desi to brin the reactor subcritical at a~

M rata fast enough to prevent the MCP rosbeconfnglessthanh06%rtngr.no dAddWf l

limiting power transient analyzed in Section 15.0 of the FSAR. This analysis sof44

- i*

shows that the negative reactivity rates resulting from the scras with the p* 'Q

~j: average response of all the drives as given in the specifications, provide the I required protection and MCPR remains greater than-h06#Tne occu....w of i

scras times longer then those specified should be viewed as an indication of a systemic problem with the red drives and therefore the surveillanca interval is reduced in order to prevent operation of the reactor for long periods of

i. time with a potenti&11y serious probles.

t i

The scras dischargo volume is required to be OPERA 8LE so that it will be available when needed to accept discharge water from the control rods during a i reactor scras and will isolata the reactor coolant system from the environment j

when required.

Control rods with inoperable accumulators are declared inoperable and

-! Specification 3.1.3.1 then acpifes. This prevents a pattern of inoperable accumulators that would result in less reactivity insertion on a scras than i has been analyzed even though control rods with inoperable accumulators may

.m O ,

,_/

still be inserted with normal drive water pressure. Cperabflity of the accumu-lator ensures that. there is a means available to insert the control rods even under the most unfavorable depressuri:ation of the reactors.

I LA SALLE - UNIT 1 8 3/4 1-2

.a _ _ _ _ _ . _ . . _ . .

j

. Os .

h REACTIVITY CONTROL SYST BS

- 8ASES 1

i-- - . - . . . . . .

t

j, 3/4.1.6 ECONOMIC GDERATION CONTROL SYSTEM Operation with the economic generation coritrol (EGC) system, automatic flow control, is limited to the range of 65% to 100% of rated core flow. In this flow range and with THERMAL POWER > 20% of RATED THERMAL POWER, the reactor '

could safely tolerata a rata of change of load of 8 MWe/pc (reference FSAR Section PISr4).

67.4

- Lieits within the EGC and the flow control systas prevent rates of change greater than approximately 4 MWe/tec. When EGC is in operation, this fact l will be indicated on the main control room console.

O, .: .

3 I

i f .:

i LA SALLE - UNIT 1 B 3/4 1-5

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

l' l

[

t q _.

j t j ; 3/4.2 POWER OISTRIBUTION LIMITS

! BASES The specifications of this section assure that the peak cladding temperature

- following the postulated design basis loss-of-coolant accidant will not exceed the 2200'F limit specified in 10 CFR 50.46.

i; 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE This specification assures that the peak cladding temperature 'following

the postulated design basis loss-of-coolant accident will not exceed the limit specified in 10 CFR 50.46.

The peak cladding temperature (PCT) following a postulated loss-of-coolant 3

accident is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is dependent only secondarily on the rod to rod power distribution within an assembly. The peak clad temperature is calculated assuming a LHGR for the highest powered rod

.O a4ca 4= a= i *= ar ' ta a ta d 'ea 'aaa carr =* d rar e a=4ric=*4 a-

% This LHGR times 1.02 is used in the heatup coda along with the exposure d 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 red divided by its local peaking factor. The limiting value for APLHGR is shown in Figure 3.2.1-lf 4.c two '

W oper*+ ion.

i The calculational procedure used to establish the APLNGR shown on Figure 3.2.1-1 is based on a loss-of-coolant accident analysis. The analysis was

! performed using General Electric (GE) calculational models which are consistent with the requirements of Appendix K to 10 CFR 50. A complete discussion of each code employed in the analysis is presented in Reference 1.

c' Differences in this analysis compared to previous analyses performed with 1 Reference 1 are: (1) the analysis assumes a fuel assembly planar power consistant with 102% of the MAPLHGR shown in Figure 3.2.1-1, (2) fission l

product decay is computed assuming an energy release rate of 200 MEV/ fission; l

(3) pool boiling is assumed after nucleate boiling is lost during the flow

og stagnation period lV(4) the effects of core spray er.trainment and counter-j current flow limitation as described in Reference 2, are included in the reflooding calculations.

j A list 'of the significant plant input parameters to the loss-of-coolant I accident analyJis is presented in Bases Table B 3.2.1-1. j se. udesa sliall be mulli(lied.

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l' j - BASES 3/4.2.2 APRM SET 90INTS f The fuel cladding integrity Safety Limits of Specification 2.1 were based j on a power distribution which would yield the design LHGR at RATED THERMAL pas.Ar I The flow biased sfaulated thermal power upscale scraa setting and control rod block functions of the APRM instrumentsraust be adjusted to ensure

. POWER. /gs.g I. -

that the MCPR does not become less thanJr# or that 11% plastic strain does pyr.

i g 4,el r not occur in tne Gegraded situation. The scraa settings and rod block settings i

g ;nd are adjusted in accordance with the fomula in this specification when the com- ;

bination of THERMAL POWER and MFLPO indicates a higher peaked power distribution <

5 to ensure that an LHGR transient would not be increased in the degraded condition.

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j 3/4.2.3 MINIMUM CRITICAL POWER RATIO ;

The required operating limit MCPRs at steady state operating conditions as specified in Specification 3.2.3 are derived from the established fuel I

! cladding integrity Safety Limit MCPR# 1.".C, and an analysis of abnormal l operational transients. For any abnormal operating transient analysis evalua- I W tion with the Initial condition of the reactor being at the steady-state l

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operating limit, it is reouf red that the resulting MCPR does not de

/-s ment trip setting given in Specification 2.2.

L To assure that the fuel cladCng integrity Safety Limit is notlimiting exceeded the luost

, i during any anticipated abnoneal operational transienttransients have b -

l

tion in CRITICAL POWER RATIO (CPR). The type of transients evaluated were loss l

of flow, increase in pressure and power, positive reactivity insertion, I an e

coolant temperature decrease.When added to the Safety Limit MCPR d ?.2, the required MCPR.

i limit MCPR of Specification 3.2 3 is obtained and presented in Figure 3.2.3-1.

The evaluation of a given transient begins with the systas initial parameters

shown in FSAR Table 15.0-1 that are input to a GE-core dynamic behavior transient The code used to evaluate pressurization events is described computer program. I l

in NED0-24154(3) andtheprogramusedinnon6ressurizationeventsisdescribed j! in NE00-10802(2) The outputs of this program along with the initial MCPR fons the input for further analyses of the thermally limiting bundle with the0) single channel transient thermal hydraulic TASC code described in NEDE-25149 The principal result of this evaluation is the reduction in MCPR caused by the transient.

The need to adjust the MCPR operating limit as a function of scraa time

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I arises from the statistical approach used in the implementation of the 00YNGeneric stI

,j computer code for analyzing rapid pressurization events. 1 analyses were performed for plant groupings of similar desi n which considered I the statistical variation in several parameters, These i.e. , init al power analyses, whichlevel, are l

CR0 scram insertion time, and model uncertainty.

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i FIRE DETECTION INSTRUMENTATION (Continued) l

'~ In the event that a portion of the fire detection instrumentation is inoperable, increasing the frequency of fire watch patrols in the affected areas in rwuired to provide detection capability until the inoperable instrumentation is restored to CPERA8ILITY.

3/3.3.7.10 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION The radioactive liquid effluent monitoring instrumentation is provided to t

I- monit..' and control, as applicable, the releases of .*adioactive materials in I liquid effluents during actual or potential releases of Ifquid effluents. The alam/ trip setpoints for these instrumenta shall be calculated in accordance with the procedures in the 00CM to ensure that the alans / trip will occur prior l' to exceeding the limits of 10 CFR Part 20. The OPERA 81LITY and use of this instrumentation is consistant with the requirements of General Design Criteria 60, 63 and 64 of Appendix A to 10 CFR Part 50.

3/4.3.7.11 RACICACTIVE GASEnUS EFFLUENT MONITORING INSTRUMENTATION l

O Ta radia ctiv s== triu at aitarias in tru at tiaa i= Provia a ta monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual o potential releases of gaseous effluents. The alam/ trip setpoints for these insruments shall be calculated in accordance with the procedures in the 00CM to ensure that the alam/ trip will occur prior to exceeding the limits of 10 CFR Part 20. This instrumentation also includes

! provisions for monitoring (and controlling) the concentrations of potentially j explosive gas mixtures in the wasta gas holdup system. The OPERA 8ILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63 and 64 of Appendix A to 10 CFR Part 50.

. i 3/4.3.7.12 LOCSE-PART DETECTION SYSTEM I The OPERA 8ILITY of the loose-part detection system ensures that sufficient l capability is available to detect loose metallic parts in the primary systen j and avoid or mitigate damage to primary systes components. The allowable out-

! ofaservice times and surveillance requirements are consistent with the recom-mandations of Regulatory Guide 1.133, " Loose-Part Detection Program for the j Primary System of Light-Water-Cooled Reactors."

! 3/4.3.3 FEEDWATER/ MAIN TUR8INE TRIP SYSTEM ACTUATION INSTRUMENTATION The feedwater/ main turbine trip system actuation instrumentation is l provided to initiate the feedwater system / main turbine trip system in the event of reactor vessel water level equal to or greater than the level 8 setpoint i

i associated with a feedwater controller failuref s gevw6 over@llias 4)c.,4 Q' cu nacjecvessal whicL ruot+ W Up wswe tjoid dischp. %9 ;

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3/4.4 REACTOR COOLANT SYSTEM BAMS 3/4.4.1 RECIRCULATION SYSTEM I

Operation with one reactor core coolant recirculation loop inoperable 4e-7:tibi*:d =ti' = :=!=ti= ;f = grfe. c.;; ;f t.',e "",:; a, f r.; e..; le--

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I 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 r.he capability of reflooding the core; thus, the requirement for shut 4cwn 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 adequata core flow O '

co stdo n fro eitaer recircuiation 100, foiio ini a LOCA. I In order to prevent undue stress on the vessel nozzles and bottaa head l 18sd region, the recirculation loop temperatures shall be within 50*F of each other (offswdf prior to startup of an idle loop. The loop temperature must also be within f pMe 50*F of the reactor pressure vessel coolant temperature to prevent thermal _

3 0 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 tho upper regions of the core, undue stress on the vessel would result if the temperature differenca was greater than 145'F.

I i 3/4.4.2 SAFETY / RELIEF VALVES i

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l The safety valve fum: tion of the safety-relief valves operate to prevent the reactor coolant system from being pressurized above the Safety Limit of 8 1325 psig in accordance with the ASME Code. A total of 18 OPERA 8LE safety /

relief valves is required to limit reactor pressure to within ASME III

allowable values for the worst case upset transient.

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i Demonstration of the safety-reifef valve lift settings will occur only

! during shutdown and will be performed in accordance with the provisions of

- Specification 4.0.5.

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O 't i Where the recir- i culation loop flow mismatch limits can not be maintained during the reci6

- culation loop operation, continued operation is permitted in the single i

recirculation loop operation mode.

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3/4.5.1 and 3/4.5.2 ECCS - OPERATING and SHITfDOWN

- ECCS Division 1 consists of the low pressure core spray systas, low pressure coolant injection subsystas "A" of the RHR system, and the automatic l'

l. depressurization system (A05) as actuated by A05 trip systas "A". ECCS

t Division 2 consists of low pressure coolant injection subsystems "B"and "C"

of the RHR systen and the automatic depressurization system as actuated by A05 l trip systes B".

The low pressure core spray (LPCS) systas is provided to assure that the ,

core is adequately cooled following a loss-of-coolant accident and provides adequate core cooling capacity for all break sizes up to and including the double-ended reactor recirculation line break, and for smaller breaks following l depressurization by the A05.

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' The LPCS is a primary source of emerge.acy core cooling after the reactor vessel is depressurized and a source for fisoding of the core in case of

' accidental draining.

The surveillance requirements provide adequata assurance that the LPCS 1 j

systan will be OPERA 8LE when required. Although all active components are testable and full flow can be demonstrated by recirculation through a test

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i loop during reactor operation, a completa functional test requires reactor - '

shutdown. The pump discharge piping is saintained full to prevent water

)menerdasegetopipingandtostartcoolingattheearliestsoment. '

The low pressure coolant injection (UCI) mode of the RHR system is i

provided to assure that the core is adequately cooled following a loss-of-coolant !

accident. Three subsystans, each with one pump, provide adequata core flooding r- ~'

- for all break sizes up to and including the double-ended reactor recirculation line break, and for small breaks following depressurization by the A05.

The surveillance requirements provide adequate assurance that the LPCI I p system will be OPERA 8LE when required. Atthough all active components are l l '

testable and full flow can be demonstrated by recirculation through a test I

loop during reactor operation, a completa functional test requires reac+.or shutdown. The pump discharge piping is saintained full to prevent water hasser damage to piping and to start cooling at the earliest soment.

ECCS Division 3 consists of the high pressure cars spray system. The j high pressure core spray (HPCS) systas is provided to assure that the reactor core is adequately cooled to limit fuel clad tamperature in the event of a small break in the reactor coolant systasi and loss of coolant which does not result in rapid depressurization of the reactor vessel. The HPCS systes I permits the reactor to be snut down while maintaining sufficient reactor I vessel water level inventory until the vessel is depres:;urized. The HPCS system operatas over a range of 1160 psid, differential pressure between reactor vessel and HPCS suction source, to O psid.

The capacity of the HPCS systas is selected to provide the required core i

p cooling. T e HPCS pump is designed to celiver greatar than or equalInitially, to v 516/1550/ . sgpa at differential pressures of 1160/1130/200 psid.

s water fros')the condensate storage tank is used instead of injecting water frem i L Gaco LA SALLE - UNIT 1 33/43-1

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i 3/a.6.1.5 PRIMARY CONTAINMENT STRUCTURAL INTEGRITY .

This limitation ensures that the structural integrity of the containment will be maintained comparable to the original design standards for the life of l the facility. Structural integrity is required to ensure that the containment will withstand the maximum pressure of 45 psig in the event of a LOCA. The measurement of containment tendon lift-off force, the ta'sile n tests of the

.l tendon wires or strands, the visual examination of tendons, anchorages and i

exposed interior and exterior surfaces of the containment, the chemical and visual examination of the sheathing filler grease, and the Type A leakage

-; ,j test are sufficient to demonstrate this capability.

i i - The surveillance requirements for demonstrating the primary containment's structural integrity and the method of predicting the pre-stress loses are in comoliance with the recommendations of Regulatory Guide 1.35.1, " Inservice yn, W c_ _ m i r of Ungrouted Tendons in Prestfessed Concreta Containment Structures,"

1 Aril 1979 with the following clarification: the tested lift-off force of

( individual tendon tension shall be greater than or equal to the initial l'

pre-strsss minus the loses, as predicted in the as-built design, which occur between the initial pre-operational structural integrity test and the time of h subsequent surveillance.

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3/4.4.1.6 ORWELL AND SUPPRESSION CHAM 8ER INTERNAL PRE 55URE

.: ge The limitations on drywell and suppression chammer internal pressure ensure that the containment peak pressure of 39.6 psig does not exceed the design pressure of 45 psig during LOCA conditions or that the external pres-sure differential does not exceed the design maximum external pressure differen-tial of 5 psid. The limit of 2.0 psig for initial positive primary containment pressure will limit tho' total pressure to 39.6 psig which is less than the design pressure and is consistant with the accident analysis.

3/a.6.1.7 ORWELL AVERAGE AIR TEMPERATURE The limitation on drywell average air temperature ensures that the 4 containment peak air temperature does not exceed the design temperature of 340*F during LOCA conditions and is consistant with the accident analysis.

3/a.6.1.8 ORWELL AND SUPPRESSION CHAMBER PURGE SYSTEM The drywell and suppression chamber purge supply and exhaust isolation valves are required to be closed during plant operation except as required

. for inerting, de-inerting and pressure control. Until these valves have been comonstrated capable of closing during a LOCA or steam line break accident, they shall be blocked so as not to open more than 50*.

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! The required Special Reports from any engineering evaluation or contain-1 ment abnormalities shall include a description of the tendon condition, the condition of the concrete (especially at tendon anchorages), the inspection

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procedure, the tolerances on cracking, the results of the enginatring evalua-tion, and the corrective action taken.

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, 3/4.7.6 FIRE RATED ASSEMBLIES i '

The OPERA 8ILITY of the fire barriers and barrier penetrations ensure t ;

that fire damage will bs limited. These design features minimize the l

possibility of a single fire involving more than ene fire area prior to

' detection and extinguishment. The fire barriers, fire barrier penetrations for conduits, cable trays and piping, fire windows, fire dampers, and fire doors are periodically inspected to verify their OPERASILITY.

l-P 3/4.7.7 AREA TEMPERATURE NONITORING

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l The area temperature limitations ensure that safety-related equipment will l* not be sub,iected to temperatures in excess of their environmental qualification temperatures. Exposure to excessive temperatures may degrade equipment and can

! cause loss of its CPERASILITY. The temperature limits include allowance for an

j. Instrument error of 2 7'F.

~f 3/a.7.8 STRUCTURAL INTEGRITY OF Ct. ASS 1 STRtCTURES .

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In order to assure that settlement does not exceed predicted and allowable -

settlement values, a program has been established to conduct a survey at the

! 3 site. The alloweble total differential settlement values are based on original i settlement predictions. In estabitshing these tabulated values, an assumption is made that pipe and conduit connection have been designed to safely withstand :

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.i the stresses which wuld develop due to total and differential settlement.

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{' All snubbers are required OPERA 8LE to ensure that the structural integrity of the Reactor Coolant System and all other safet*related systems is maintsined

, during and following a seismic or other event initiating dynamic loads. Snutbers

i excluded from this inspection program are those installed on nonsafety-related t

systemt and then only if their failure or failure of the system on which they '

are installed, would have no adverse effect on any safety related system.

i;.

Snubbers are clasfified and grouped by design and manufacturer but not by

. size. For example, mechanical snubbers utilizing the same design features of the 2-kip,10-kip, and 100-kip capacity manufactured by Company "A" are of the same type. The same design mechanical snubbers manufactured by Company "B" for the purpose of this Technical Specification would be of a different type, as woulif hydraulic snubbers from either manufacturer.

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3/a. 9 SNU88ERS

, A snubbges am required OPERA 8LE to ensure that the structural integrity

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' i of the actorScoolantS ay tem and all other saf'ety related systems is maintair.ed during an following a seismic or other event initiating dynamic loads. Snubbers excluded f this inspection program are those installed on nonsafety related systems and en only if their failure or failure of the system on which they y are inttalled, would have no adverse effect on eny safety related system.

The visual nspection frequency is bued upon maintaining a constant level p of snubber prot on to systems. Therefore, the required inspection interval varies inversely w h the observed snubber failuns and is determined by the numeer of inoperabl snubbers found during an inspection. Inspections performed before that interval s elapsed say be used as a new referenct point to detemine the next inspection. ver, the results 'of such early inspections performed before the original req ' red time interval has elapsed, nominal time less 25%, may

- not be used to lengthen required inspection intar~al. Any inspection whose results require a shorter pection interval will override the previous senedule.

^ When the cause of the re ion of a snubber is clearly established and J

remedied for that saubber and r'any'othef snubbers that may be generically

- susceptible, and verified by ins ice functional testing, that snubber may be exempted from being counted as i rable. Generically susceptible snub h rs j are tt.ose snumbers which are of a s ific make or model and have the same design features dircctly related to ection of the snubber by visual inspec-t

tion, or are similarly located or expo to the same environmental conditions -

such as temperature, radiation, and vib tion.

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When a snubber is found inoperable, a engineering evaluetion is performed, l in addition to the detemination of the sn er mode of failure, in order to j stem has been adversly affected j determine if any safety-related component or by the inoperability of the snubber. The engi ering evaluation shall determine 1

I whether or not the snubbar mode of failure has arted a significant effect or degradation on the supported component or syst .

l To provide assurance of snubber functional rol 111ty, a regnsentative sample of the installed snubbers shall be funcional tested during plant snut-downs at 18 month intervals. Selection of a represen tive sample according to the expression 35 (1 + (f) provides a confidence level f appriximately 95% that 90% to 100% of the snubbers in the plant will be OPERAELE ithin acceptance limits.

Cbserved failures of these sample snubbers will require fun tional testing of l

additional units.

i Hydraulic snutbers and mechanical snubbers cay each be tre tad as a j

different entity for the above surveillance programs.

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' 9 3/4 7-4 La S.ALLE - UMIT 1

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PLANT SYSTEMS

n BASES 4

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pMUBBERS (Continued)

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A list of individual snubbers with detailed information of snubbers loca-tion and size and of system affected shall be available at the plant in accord-

ance with Section b0.71(c) of 10 CFR Part 50. . The accessibility of each snubber shall be determined and approved by the Onsite Review and Investigative Function.

The determination shall bs based upon the existing radiation levels and the expected time to perform a visual inspection in each snubber location as well as other factors associated with accessibility during plant operations (e.g.,

temperature, atmosphere, location, etc.), and the recommendations of Regulatory Guide 8.8 and 8.10. The addition or deletion of any hydraulic or mechanical

> snubber shall be made in accordance with Section 50.59 of 10 CFR Part 50.

The visual inspection'frequer.cy is based upon maintaining a constant level of snubber protection to each saf 6ty-related system. Therefore, the required inspection interval varies inversely with the observed snubber failures on a given system and is determined by the number of inoperable snubbers found during an inspection of each system. ~ In order to establish the inspection frequency for each type of snubber on a safety-related system, it was assumed that the frequency of snubber failures and initiating events is constant with time and that the failure of any snubber on that system could cause the system to be r unprotected and to result in failure during an assumed initiating event.

Inspections performed before that iqterval has elapsed may be used as a new Q reference point to determine the next inspection. However, the results of such early inspections performed before the original raquired time interval has elapsed (nominal time less 25%) may not be used to lengthen the required inspection interval. Any inspection whose results require a shorter inspection interval will override the previous schedule.

The acceptance criteria are to be used in the visual inspection to determine OPERABILITY of the snubbers. For example, if a fluid port of a hydraulic snubber is found to be uncovered, the snubber shall be declared inoperable and shall not be determined OPERABLE via functional testing.

To orovide assurance of snubber functional reliability, one of three functional testing methods is used with stated acceptance criteria:

1. Functionally test 10% of a type of snubber with an additional 10%

tested for each functional testing failure, or

2. Functionally test a sample size and determine sample acceptance or rejection using Figure 4.7-1, or .

3. Functionally test a representative sample size and determine sample acceptance or rejection using the stated equation.

LASALLE-UNIT /i B 3/4 7-4

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The servics Iffe o a snubber is evaluated via manufacturer input and i in#ormation through consi ration of the snubber service conditions and

associated installation maintenance records (newly installed snubber, seal

- replaced, spring replaced, high radiation area, in high temperature area,

, etc.). The requirement to ao tar the snubber service life is included to ensure that the snubbers pari cally undergo a performance evaluation in view

.' , of their age and operating condi tons. These records will provide statistical

bases for future consideration of nubber service life. The requirements for the maintenance of records and the ubber service life review are not intended

' i to affect plant operation.

3/4.7.10 MAIN TURBINE SYPA$$ SYSTEM 4

' The main turbine bypass system is req red OPERA 8LE as assumed in the

feedwater controller failure analysis.

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SNUBBERS (Continued)

Figure 4.7-1 was developed using "Wald's Sequential Probability Ratio Plan" as described in " Quality Control and Industrial Statistics" by Acheson J. Duncan.

Permanent or other exemptions from the surveillance program for individual snubbers may be grantsd by the Commission if a justifiable basis for exemption is presented and, if applicable, snubber life destructive testing was performed '

to qualify the snubber for the applicable design conditions at either the com-plation of their fabrication or at a subsequent date. Snubbers so exempted shall be listed in the list of individual snubbers indicating the extent of the exemptions.

The service life of a snubber is established via manufacturer input and information through consideration of the snubber service conditions and associated installation and maintenance records (newly installed snubbers, seal replaced, spring replaced, in high radiation area, in high temperature area, etc.). The requirement to monitor the snubber service life is included to ensure that the snubbers periodically undergo a performance evaluation in view of their age and cperating conditions.- These records will provide statistical

,/, # bases for future consideration of snubber service life.

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3/4.7.10 MAIN TURBINE BYPASS SYSTBt The main turbine bypass system is required OPERABLE as assumed in the feedwater controller failure analysis. .

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! 3/4.12 RADIOLOGICAL OfVIRONMENTAL MONITORING l I l

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. 3/4.12.1 MONITORING PROGRAM l t

The radiological monitoring program required by this specification provides i seasurements of radiation and of radioactive materials in those exposure path-ways and for those radionuclides which lead to the highest potential radiation j

' exposures of individuals resulting from the station operation. Bis monitoring

- program thereby supplements the radiological effluent monitoring program by verifying that the esasurable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measure-ments and modeling of the environmental exposure pathways. The initially specified monitoring program will be effective for at least thx first three 3 l years of commercial operation, as defined i,n the 00CM. ;

The detection capabilities required by Table 4.12-1 are state-of-the-art for routine environmental measurements in industrial laboratories. It should be recognized that the LLD is defined as an "a priori" (before the fact) limit

- representing the capability of a measurement system and not as "a posteriori" (after the fact) limit for a particular measurement. Analyses shall be performed in such a manner that the stated LLDs will be achieved under routine Q V) conditions. Occasionally background fluctuations, unavoidahl small sample ,

sizes, the presence of interfering nuclides, or other uncont 11able circum- i

, stances may render these LLDs unachievable. In such cases, the contributing factors will be identified and described in the Annual Radiological Environmental Operating Report.

,f

3/4.12.2 LANO USE CENSUS .

l This specification is provided to ensure that changes in the use of J:

unrestricted areas are identified and that modifications to the monitoring k program are made if required by the results of this census. The best survey SJ l4 ,

aerial 4r consulting with local agricut- ,

l l information from snaii the door-to-doog,This census satisfies the requirements of I g, e tural authorities os usea.

l Section IV.S.3 of Appendix I to 10 CFR Part 50.

9 l

i i

. i

, i

! LA SALLE - UNIT 1 3 3/4 12-1

i ('

i 3/4.'11 RADI0 ACTIVE EFFLUFNTS I

8ASES 3/4.11.1 LIQUID EFFLUENTS 1

3/4.11.I.1 CONCENTRATION .

This specification is provided to ensure t. hat the concentration of radioactive materials released in liquid wasta effluents from the site will be less than the concentration levels specified in 10 CFR Part 20, Appendix S, Table II, Column 2. This limitation provides additional assurance that tne levels of radioactive materials in bodies of water outside the site will

' result in exposure within (1) the Section II.A design objectives of Appendix I, 10 CFR 50, to an individual, and (2) the limits of 10 CFR 20.106(e) to the population. The concentration limits for dissolved or entrained noble gases were detarsined by converting their MPC's ift air to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2.

3/4.11.1.2 005E os -

This specification is provided to taplement the requirements of Sections II. A III.A and IV.A of Appendix I,10 CFR Part 50. The Limiting Condition for Operation implements to guides set forth in Section II.A of Appendix I. The ACTION statements provide the required operating flexibility j

and at the same time implement the guides set forth in Section IV. A of -

Appendix I to assure that the releases of radioactive material in liquid effluents will be kept "as low as is reasonably achievable." Also, for fresh

! water sitas with drinking water supplies which can be potentially affected by

!- plant operations, there is reasonable assurance that the operation of the facility will not result in radionuclide concentrations in the finished drinking water that are in excess of the requirements of 40 CFR 141. The dose calcula-

' tions in the 00CM implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of an individual through appropriata pathways is unlikely to be substantially underestimated. The equations specified in the 00CM for calculating the doses due to the ac*ual release rates of radioactive materials in liquid effluents are consistent witn the methodology provided in Regulatory Guide 1.109, j

. " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Punose of Evaluating Compliance with 10 CFR Part 50, Appendix I,"

Revision 1, October 1977 and Regulatory Guide 1.113, " Estimating Aquatic Dispersior, of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I," April 1977. rv $coc.be.M A4"MS N This specification applies to the release ofaliquid effluents from each k reactor at the site. For units with shared radwasta treatment systams, the

' liquid effluents frca the shared systam are prcportioned among the units l

O. - =a r4=s ta t =vit -

LA SALLE - UNIT 1 8 3/4 11-1

--.-.,,_-.-..-,---.--w v- .,_ e- , .- 4,.-. ,- , y-,..y--r--. w---,,--

_ . d 1 -

l li I. '

n,. .,.,.

! ! N N. !

! 5ASES l [

1i

'I ! 005E RATE (Continued)

i t' infant via the cow-silk-infant pathway to less than or equal to 1500 mres/

l l year for the nearest cow to the plant. Asaus 4g4, I

MioAdi& id $

This specification applies to the release of geoeous affluents 3 from all (

j :*

reactors at the site. For units within shared raawaste treatment systems, the

'; gaseous effluents from the shared system are proportioned among the units

sharing that system.

i 3/4.11.2.2 005E - N08LE GASE5

- ThisspecificationisprovidedtoimpienenttherequirementsofSectionsII.8, t

III.A and IV.A of Appendix I, 10 CFR Part 50. The Limiting Conditions for *

!

Operation are the guides set forth in Section II.8 of Appendix I. The ACTION statements provide the required operating flexibility and at the same time

]

l 1 implement the guides set forth in Section IV. A of Appendix ! to assure that 0, the releases of radioactive material in gaseous affluents will be kept "as low as is reasonably achievable." The Surveillance Requirements implement the requirements in Section III.A of Appendix I that conformance with the guides

of Appendix I be shown by calculational procedures based on models and data j such that the actual exposure of an individual through appropriate pathways is i

unlikely to be substantially uriderestimated. The dose calculations estaelished -

6 in the 00CM for calculating the doses due to the actual release rates of 1- ~ radioactive noble gases in gaseous effluents are consistent with the methodology

+

provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from ,

i- Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance I with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.111. " Methods for Estinating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors,"

Revision 1, July 1977. The 00CM equations provided for determining the air f I' : doses at the site boundary are based upon the historical, average atmospheric conditions.

1.

l ! 3/4.11.2.3 005E - RADICIODINES. RADI0 ACTIVE MATERIALS IN PARTICULATE FORM l 1 AND RAOSONUCLIOE5 OTHER THAN NOBLE GASE5

'i l The specification is provided to implement the requirements of Sections II.C, III.A and IV.A of Appendix I, 10 CFR Part 50. The Limiting Conditions for

l j j Operation are the guides set forth in Section II.C of Appendix I. The ACTION l

statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV. A of Appendix I to assure that eq the releases of radioactive materials in gaseous effluents will be kaot "as

. ./

low as is reasonably achievable." The CCCM calculational methods specifiec in the Surveillance Requirements isolement the requirements in Section III.A of I

LA SALLE - UNIT 1 8 3/4 11-3 MW. " "*NN e -- - - - - - - , , - - , , , - - , _ ,e,- .-_n ,.,.e-- , - . , - , , - - . ,w,,, - - , . - - . , , , , - - - . , , y, ------m- w,, , , . . r -

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5. 0 OESIGN FEATURES i:

'i .

If i; 5.1 SITE fi EXCLUSION AREA ll

! 5.1.1 The exclusion area shall be as shown in Figure 5.1.1-1.

LOW POPULATION ZONE l' 5.1.2 The low population zone shall be as shown in Figure 5.1.2-1.

l1 1' SITE 800NOARY FOR GASEQUS EFFLUENTS 5.1.3 The site boundary for gaseous effluents shall be as shown in Figure 5.1.1-1.

{i SITE BOUNDARY FOR LIQUID EFFLUENTS 5.1. 4 The site boundary for liquid effluents shall be as shown in Figure 5.1.1-1.

5.2 CONTAINMENT CONFIGURATION ,

5.2.1 The prisan containment is a steel lined post-tensioned concrete

~

structure consisting of a drywell and suppression chamber. The drywell is a steel-lined post-stressed concreta vessel in the shape of a truncated cone closed by a steel done. The drywell is above a cylindrical steel-lined post-stressed

.; concrete suppression chamber and is attached to the suppression chamber through a series of downconer vents. The drywell has a minimum free air volume of -

The suppression chamber has an air region of 164,800 to [

tii;g33 - 2th168,100 MS cubic feet. cubic feet and a water region of 128,800 to 131,900 cubic feet.

I DESIGN TEMPERATURE AND PRESSURE i

- - 5.2.2 The primary containment is designed and shall be maintained for:

a. Maximum internal pressure 45 psig.

b. Maximum internal temperature: drywell 340*F. !

suppression chamber 275'F.

, ]

4 i c. Maximum external pressure 5 psig.

I l d. Maximus floor differential pressure: 25 psid, downward.

5 psid, upward.

[ ,

I i SECONDARY CCNTAINNENT ,

5.2.3 The secondary containment consists of the Reactor Suilding, the equipment access structure and a portion of the main steam tunnel and has a minimum free

] volume of 2,875,000 cubic faet.

LA SALLE - UNIT 1 5-1

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I l_.---..___ . _ __ _

i,l i

((

O anatarsTaartve co*Tiivis m

i Any deviatior, from the above guidelines shall be authorized by the (Station

! Superintendent or his deputy', or higher levels of management, in accordance with es tablished procedures and with documentation of the basis for granting the

~

l' deviation. Controls shall be included in the procedures such that individual overtime shall be reviewed monthly by the Station Superintendent or his designee lll to assure th t excessive hours have not been assigned. Routine deviation from the il above guidelines.is not authorizud.

I D. Qualifications of the station management and operating staff shall meet minimum acceptable levels as described in ANSI M18.1, " Selection and Training of Nuclear Power Plant Personnel," dated March 8, 1971. The

' Rad /Chen hpervisor shall meet the requirements of radiation protection

l manager of Rag;1 story Guide 1.8, September,1975. The ANSI N18.1-1971 ous11fication requirements for Rad /Chen Technician may also be met by either of-the following alternacives:

t 1. Individuals who have completed tho' Rad /Chen Technician training program and have accrued p W year of working experience in the (

i

specialty, or 3. I Individuals who have completed Rad /Chen Technician training l

' 2.

year of working experience l

].. program, but have not yet accrued

.in the specialty, who are supervised by on-shift health physics supervision who meet the requirements of ANSI NB.1-1971 Q Section 4.3.2, " Supervisor Not Requiring AEC Licenses," or Section 4.4.4, " Radiation Protection."

l -

,i E. Retraining and replacement training of Station personnel shall be in accordance with ANSI N18.1, " Selection and Training of Nuclear Power

,i Plant Personnel", dated March 8,1971 and Appendix "A" of 10 CFR Part 55, and shall include familiarization with relevant industry operational

.I experience identified by the ON5G.

l i.

5 F. Retraining shall be conducted at intervals not exceeding 2 years.

G. The Review and Investigative Function and the Audit Function of activities affecting quality during facility operations shall be constituted and have the responsibilities and authorities outlined below:

l

, 1. The Supervisor of the Offsite Review and Investigative Function shall be appointed by the Director, Nuclear Safety. The Audit Function shall bc the responsibility of the Manager of Quality

I Assurance and shall be independent of operations.

'l Offsite Review and Investigative Function

,l a.

The Supervisor of the Offsite Review and Investigative Function ;

'shall: (1) provide directions for the review and investigative i function and appoint a senior participant to provide appropriate j direction, (2) select each participant for this function, (3) :

select a complement of more than one participant who collectively 1 l

possess background and qualifications in the subject matter under '

!' review to provide comprehensive interdisciplinary review coverage 6-3 Amendment No.14 LA SALLE - UNIT 1

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

I k- g gg .N tr PY . g . Fiaure 6.1-3 MINI?1LM SNIPT CREW COMP 05m0N i

     'i-P P05 m 0N               \ NUMBER OF IMOIVIDUALS REQUIRED TO FILL P05 m 0N
                                                                                                              \ CON 0m0N1,2and3                             CON 0 m 0N 4 and 5
       !                                                                                          SE                          1                                     1 SF                          1                                    Mone M                           2                                     1 j

AQ 2 1 Non. [' ScRr \1 , i-1

-j.

1 I I.

9 or, whenever a SCRE (SRO/STA) . composition, the minimum shift not included in the shift crew composition shati be as l

 ,,      ;,                                                                                                                                                                                          l I                                                                             follous:                                   .

1-i i t

   +     'l ;  .
          .   .                                                                         POSITION                    Nt36EROFINDIVIDUALSREQUMEDTOFILLP05m0N s

l CONDITION 1, 2 and 3 l\ CON 0m0N4and5 1 1 SE 1 None l SF f 1 ! M 2 { m 2 1 STA 1 pne i o' O ' LA SALLE - UNIT 1 6- 13 l m.

T:. - . . . . : z. . - - : . . - . . . ..- = . . . - - . h Floure 6.1-3 MINIMUM SNIFT CREW COMPOSITION WITHUNITgINCONDITION1,2,OR3 POSITION NUPRER OF INDIVIDUALS REQUIRED TO FILL POSITION L, _ . CONDITIONS 1, 2 and 3 CONDITIONS 4 and 5 8 ll SE 1 1" 8 SF 1 None M 5 1 h M 5 1 a y,,, SCRE l

     ;                                                                  or, whenever a SCRE (SR0/STA) is not included in the shift crew composition, the minimum shift crew composition shall be as follows:

WITHUNIT4INCONDITION1,2,OR3 a

                     ' r'
                                    ,                                   POSITION                                                    NUMBER OF INDIVIOUAL5 REQUIRED TO FILL POSITION                                                                    !
                       .'v'                                                                                                         CONDITIONS 1, 2 and 3                              CONDITIONS 4 and 5                                              l SE                                                   l a                                                  ga                                                l l                                                                                                                                            8 SF                                                    1                                               None                                                .

b ; R0 2

                                 .                                                       M                                                     5                                                     1                                                 l a                                            None
      !                                                                                   STA ~                                                1 I

1 l WITH UNIT 4 IN CONDITION 4 OR 5 OR DEFUELED J

      -                                                           ~ POSITION                                                        NUMBER OF INDIVIDUALS REQUIRED TO FILL POSITION CONDITIONS 1, 2 and 3                                CONDITIONS 4 and 5                                            )

SE l' 1* SF 1 None R0 2 1 D A0 2 2

STA 1 None n .' W LA SALLE - UNIT fl 6-13 00T 4"E

 --"-------m__                          - *-.           --- _ _ _ , . , - . . . - _ _ + _ , , . , _ - , , , , , , , , , _ _ . ,                               , . ,

                                 ~                * ~ ~
                                                             .-..: "      * ~ ~
                                                                                                    ..e .

1 e 1 Fiaure 6.1-3 (Continued) I j MINIMLM SHIFT CREW COMPOSITION

   -                                                                                      NOTES

<~ ; -;, SE - Shift Supervisor (Shift Engineer) with a Senior Reactor Operators t License on Unit 1. . . ~ . , - ~5F - Shift Foreman with a Senior Reactor Operators License on Unit 1.

   !     !                               RO       -    Individual with a Reactor Operators License on Unit 1.

3: AO - Auxiliary Operator. SCRE - Station Control Room Engineer with a Senior Reactor Operators License. I i Except for the Shift Supervisor, the Shift Crew Composition may be one less

         ;                               than the minimum requirements of Figure 6.1-3 for a period of time not to l      t                              exceed 2 hours in order to arc h ata unexpec ad obsence of on duty shift
          !                              crew members provided immediate action is taken to restore the shift crew
          !                              composition to within the minimum requirements of Figure 6.1-3. This provision does not perisit any shift crew position to be unmanned upon shift change due to an oncoming shift crewman being late or absent.
           '                             While the unit is in OPERATIONAL COMOITION 1, 2, or 3, in individual with a valid SR0 license shall be designated to assume the Cor. trol Room direction function. While the unit is. in OPERATIONAL CONDITION 4 or 5, an individual-
                 'B}-  -                 with a valid SRO or R0 license shall be designated to assume the Control Roos i.

direction function. . i; . ja - 1. l 1

           .  ,t
    )          '                    - Individual may fill the same position on Unit d.
                                     -sfOne of the two required individuais may fill the same position on Unit 2-o v

O O' LA SALLE - UNIT 1 6-14 1 1 **-* e m enowe -e 4 e

~.

t i ADMINISTRATIVE CONTROLS t

  ,I               ,

i Pt. ANT 09ERATING RECORDS (Continued)

  , -; t-                               --

i 8. Records and/or logs relative to the following items shall be recorded in

                                                                ~ a manner convenient for review and shall be retained for the life of the

) l .j: - plant:

1. Substitution or replacement of principal itans of equipment pertain-

  - -i i-j                                                                        ing to nuclear safety; (i,' ;

2. Changes made to the plant as it is described in the SAR;

  .5                                                                 3.      Records of new and spent fuel inventory and assembly histories; I

4 4. Updated, corrected, and as-built drawings of the plant; ! l. l

               .j .                                                  5.      Records of plant radiation and contamination surveys; i                                              6.      Records of offsite environmental sonitoring surveys;

7. Records of radiation exposure for all plant personnel, includin all i contractors and visitors to the p.lant, in accordance with 10 CF 0; i

  ;                    I-
   ' Ii                                                              8.       Records of of radioactivity in ifquid and gasects wastes releas to b           0 -)                                       the environment;
                                 =

9. Records of transient or operational cycling fo:' those components that have been designed to operate safety for a limited number of transient j or operational cycles (identified in Table 5.7.1-1);

10. Records of individual staff members indicating qualifications, s - experience, training, and retraining; ~

11. Inservice inspections of the reactor coolant systas;

12. Minutes of meetings and results of reviews and audits performed by the offsite and onsite review and audit functions;

13. Records of reactor tests and experiments;

14. Records of Quality Assurance activities required by the QA Manual;

15. Recorcs of reviews perfonned for changes made to procedures on equip-annt or reviews of tests and experiments pursuant to 10 CFR 50.59; and 16.. Records of the service Ifves of all hydraulic and mechanical snubbers li. -

M M z Li 1.7c114 41.L112 including the data at which the l

                        .l. reguired                                      7se:'vice r     life commences and associated installation and maintenance
    '                                                                           records.
                            ! 5P60ticq sn                              17. Records of analyses required by the radiological environmental i
                                             .7 M l         -

aonitoring program. 1 l ,, l (.s LA SALLE - UNIT 1 6-20 f 1

                                                                             . N M-
       .r- , - - ..-                            - . . - -                                ,. . _ _ . ~ , _ , - , , , - , , - , _ , - , - - . , , , .             . . - , ,   -----w,- - - - ,          , - - -    --r

!i ! jN i A l- ,

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l ' i i AONINISTRATIVE CONTROL 5 l

 !    !-                                                                                                                                                                           l 1

Thirty-Cay Written Reports (Continued) l

 '                                                                                                                                                                                 l F                                             e.      An unplanned offsite release of 1) more than 1 curie of radio-l-                                                            active meterial in liquid effluents, 2) more than 150 curies of noble gas in gaseous effluents, or 3) more than 0.05 curies of
       ;                                                     radiciodine in gaseous effluents. The report.cf an unplanned
       !                                                     offsite release of radioactive material shall include the i                                                     following information:

1. A description of the event and equipment involved. l 1

2. Cause(s) for the unplanned release.

i 3. Actions taken to prevent recurrence. l

.t

4. Consequences of the unplanned release.

I f. Measured levels of radioactivity in an environmental sampling i mediuerdetermined to exceed the reporting level values of , i ., Tabla 3.12-2 when averaged over any calendar quarter sampling l l (}' C. Unique Reporting Requirements l l~ 1. Special Reports shall be submitted to the Director of the Office of Inspection and Enforcement (Region III) within the time period _ specified for each report. - g

  .                                 6. 7 PROCESS CONTROL PROGRAi4 (PCPW                                                                                                 l
          !                         6.7.1 The PCP shall be approved by the Commission prior to implementation.

I i:., 6. 7. 2 Licensee initiated changes to the PCP: ,

a. ShallbesubmittedtotheCommissioninthesemibnualRadioactive l ;

          ;                                           Effluent Release Report for the period in which the change (s) was j                                          sade. This subsittal shall contain:

L Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental i information; !

2. A determination that the change did not reduce the overall conformance of the solidified waste product to existing criteria j for solid wastes; and i

     *i                                                3. Documentation of the fact that the change has been reviewed.and I            i                                                  found acceptacle by the Onsite Review and Investigative Function.

m V b. Shall become effective upon review and acceptance by the Onsite Review 1 andmon Investigative Function. den r l u 4 L JL h l e. j (P.? com 1o La <rtis. h,? Z

             .                       l.A 5 LLE - UNIT 1                                     6-2S

j O: n ' i~U l ADMINISTRATIVE CONTROLS li .. 6.8 0FFSITE 00$E CALCULATION MANUAL (00CMW l $ 6.8.1 The 00CM shall be approved by the Commission prior to implementation. I l' 6.8.2 Licensee initiated changes to the 00CM:

     !                           a. Shall be submitted to the Comeission within 90 days of the date the change (s) was made effective. This submittal shall contain:

j 1. Sufficiently detailed information to totally support the rationale for the change without benefit of additional or j supplemental information. Information submitted should consist l i of a package of those pages of the 00CM to be changed with each l page numbered and provided with an approval and date box, j- , together with appropriate analyses or evaluations justifying the change (s);

      !                               2. A detarmination that the change will not reduce the accuracy or i                                     reifability of dose calculations or setpoint determinations; and

3. Documentation of the fact that the change has been reviewed and 3 found acceptable by the Onsite Review and Investigative Function.

s Ii.! e ! b. Shall become effective upon review and acceptance by the Onsite Review and Investigative Function. [ _ i l 6.9 MAJOR CHANGES TO RADI0 ACTIVE WASTE TREATU4ENT SYSTEMS

        !                  6.9.1 Licensee initiated major changes to the radioactive wasta systems (liquid, gaseous and solid):
        '                        a. Shall be reported to the Commission in the Monthly Operating Report l                            for the period in which the evaluation was reviewed by the Onsite Review and Investicative Function. The discussion of each change l,                           shall contain; i

I 1. A summary of the evaluation that led to the determination that

          !                                  the change could be made in accordance with 10 CFR 50.59;

2. Sufficient detailed information to totally support the reason l

for the change without benefit or additional or supplemental l I ; information; l 3. A detailed description of the equipment, components and prt: cesses I f involved and the interfaces witn other plant systems; t l r

h v. f (JDcH) C2mma to [4fde [)n: T' I al L.: irle d: 72.

LA SALLE - UNIT 1 6-29 1}}

ML20079H418

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SUBJECT:

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