ML20054F671

Proposed Tech Spec Changes Re Fire Detector Tables Reflecting Addl Detectors

ML20054F671
Person / Time
Site:	LaSalle
Issue date:	06/14/1982
From:	COMMONWEALTH EDISON CO.
To:
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ML20054F667	List: ML20054F666 ML20054F671
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NUDOCS 8206170205
Download: ML20054F671 (15)

Text

..

TABLE 4.6.1.5-1 TENDON SURVIELLANCE TENDON NUhBERS Tears After Initial Structural 1

3 5

10 15 Integrity Test Type of Inspection H

V H

V H

V H

V H

V Visual Inspection 48AC 15C 48AC 15C 48AC 15C 48AC 15C 48AC 15C of End Anchorages 56CB 15A 2CB 6C 3BA 28A 4BA 30B SOCB 19A Adjacent Concrete:

12CB 20A 14AC 17A 12BA 23A 41CB 22A 53BA 13B Surface and Pre-70B 47C 24BA 32C 21CB SB 50AC 57AC stress Monitor-20CB 29A 37CB 42C 23BA 31C ing Tests 1CB 47CB 38CB 12AC 57CB 49AC 56BA 60B 68B 21AC s

Detensioning and 20CB 47C 2CB 42C 23BA 31C 4BA 22A 50CB 19A Material Tests TENDON NUMBERS Tears After Initial Structural 20 25 30 35 40 Integrity Test Type of Inspection H

V H

V H

V H

V H

V Visual Inspection 48AC 15C 48AC 15C 48AC 15C 48AC 15C 48AC 15C of End Anchorages 39CB 25B IBA 3B 48CB 7B 49CB 25A 36CB 13A Adjacent. Concrete 49BA llA 47AC 12A 51AC 18A SlBA 18B 48BA 27B Surface and Pre-71D 57BA 58BA 59D 69D stress Monitor-ing Tests Detensioning and 48BA llA 47AC 3B 48CB 18A SlBA 18B 36CB 13A Material Tests 6

e e

P LA SALLE - UNIT 1 3/4 6-11

TABLE 4.6.1.5-2 TENDON LIFT-OFF FORCE V TENDONS Tendon First Year

• Number Ends Maximum'(kips)

Minimum (kips)

X1 X2 V15C A

N/A

._,,,654.95

,,_, N/A 4.369 B

' ~'N/A

'N/A N/A N/A V28A A

N/A 650.53 N/A 4.270 B

N/A N/A N/A N/A V23A A

N/A 650.53 N/A 4.270 B

N/A N/A N/A N/A V5B A

N/A 653.27 N/A 4.270 K/A N/A N/A N/A B

V31C A

N/A 651.90 N/A 4.369 B

N/A N/A N/A N/A V30B A

N/A 655.74 N/A 4.270 B

N/A N/A N/A N/A V22A A

N/A 655.74 N/A 4.270 B

N/A N/A N/A N/A Vl9A A

N/A 650.53 N/A 4.270 B

N/A N/A N/A N/A V138 A

N/A 653.27 N/A 4.270 B

N/A N/A N/A N/A V25B A

N/A 653.23 N/A 4.263 B

N/A N/A N/A N/A VilA A

N/A 650.49 N/A 4.263 B

N/A N/A N/A N/A V3B A

N/A 653.27 N/A 4.270 B

N/A N/A N/A N/A__

V12A A

N/A 650.53 N/A 4.270 B

N/A N/A N/A N/A V7B

~A N/A 653.27 N/A 4.270 B

N/A N/A

'N/A N/A V18A A

N/A 650.53 N/A 4.270 l,

~

B N/A N/A N/A N/A V25A A

N/A 650.53 N/A 4.270 B

N/A N/A N/A N/A V18B A

N/A 653.27 N/A 4.270 B

N/A N/A N/A N/A j

A N/A 643.28 N/A 4.263 i

V13A B

N/A N/A N/A N/A

I V27B A

N/A 653.23 N/A 4.253 l

B N/A N/A N/A N.t I

• First Inspection LA SALLE - UNIT 1 3/4 6-12

TABLE 4.6.1.5-2 (CONT.)

TENDON LIFT-OFF FORCE BOOP TENDONS Tendon First Year

• Number Ends Maximum (kips)

Minimum (kips)

Y1 Y2 48AC A

N/A 647.00 N/A 4.500 B

N/A 647.00 N/A 4.500 3BA A

N/A 656.46 N/A 4.226 B

N/A 656.46 N/A 4.226 12BA A

N/A 637.48 N/A 6.173 B

N/A 637.48 N/A 6.173 N/A 637.48 N/A 6.173 21CB A

B N/A 637.48 N/A 6.173 23BA A

N/A 629.71 N/A 6.173 B

N/A 629.71 N/A 6.173

~

38CB A

N/A 631.76 N/A 5.437 B

N/A 631.76 N/A 5.437 49AC A

N/A 647.00 N/A 4.500 B

N/A 647.00 N/A 4.500 68B A

N/A 655.39 N/A 4.332 B

N/A 655.39 N/A 4.332 4BA A

N/A 651.16 N/A 4.226 B

N/A 651.16 N/A 4.226 41CB A

N/A 644.51 N/A 4.975 B

N/A 644.51 N/A 4.975 50AC A

N/A 650.35 N/A 4.500

~B N/A 650.35 N/A 4.500 50CB A

N/A 650.35 N/A 4.500 B

N/A 650.35 N/A 4.500

' 53BA A

N/A 649.82 N/A 4.538 B

N/A 649.82 N/A 4.538 57AC A

N/A 650.14 N/A 4.862 B

N/A 650.14 N/A 4.862 39CB A

N/A 644.69 N/A 5.437 B

N/A 644.69 N/A 5.437 49BA A

N/A 647.00 N/A 4.500 B

N/A 647.00 N/A 4.500_

71D A

N/A 645.20 N/A 4.332 B

N/A 645.20 N/A

__ 4.332 IBA A

N/A 655.82 N/A 1914 B

N/A 655.82 N/A 3. 914___

47AC A

N/A 644.51 N/A 4.975 1

B N/A 644. 5.L N/A 4, 9 '_

57BA A

N/A 650,1C N/A

/

B N/A 650.1F N/A M2 48CB A

N/A 64 f>. E N/A 592 B

N/A 646.48 N/A-607 51AC A

N/A 653.75 N/A B

N/A 653.75 N/A 4.507,

• First Inspection LA SALLE - UNIT 1 3/4 6-13 i

L

_ TABLE 4.6.1.5-2 (CONT.)

TENDON LIFT-OFF FORCE

_ HOOP TDIDONS_

Tendon First Year

• Number Ends Maximum (kips) -

Minimum (kip::)

Y1 Y2 58BA A

N/A 640.84 N/A 4.912 B

N/A

'640.84 N/A 4.912 49CB A

N/A 639.80 N/A 4.500 B

N/A 639.80 N/A 4.500 51BA A

N/A 653.76 N/A 4.500 B

N/A 653.76 N/A 4.500 59D A

N/A 638.18 N/A 4.906 B

N/A 638.18 N/A 4.906 36CB A

N/A 644.69 N/A 5.437 B

N/A 644.69 N/A 5.437 48BA A

N/A 653.76 N/A 4.500 B

'N/A 653.76 N/A 4.500 69D A

N/A 642.31 N/A 4.332 B

N/A 642.31 N/A 4.332

• First Inspection LA SALLE - UNIT 1 3/4 6-14 l

l

LASALLE COUNTY STATION UNIT 1 TECH SPEC CHANGE REQUEST NPF-ll/82-6

Subject:

Revise Fire Detector Tables Reference (a):

License NPF-ll, Condition 2.C. (25). (b)

Background:

Reference (a) mandates that a general sprinkler system be installed in the diesel-generator corridor prior to initial criticality.

Discussion:

This change incorporates the additional sprinkler system associated fire detectors into the Technical Specifications.

==

Conclusion:==

Commonwealth Edison finds no unreviewed safety question.

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TABLE 3.3.7.9-1

/~'\\

FIRE DETECTION INSTRUMENTATION INSTRUMENT LOCATION MINIMUM INSTRUMENTS OPERABLE 8 HEAT Fi.n.;E SMOKE A.

Unit 1 Fire Detection Instrumentation 1.

Cable Spreading Room 21 (Dry Pipe Sprinkler System) 2.

Diesel Generator Corridor GP (Dry Pipe Sprinkler System) 3.

Unit 0 Cables Over Lab 32 (Dry Pipe Sprinkler System) 4.

Diesel Generator (ODG01K) Room 2

(CO2 Flooding System) 5.

Diesel Generator (1DG01K) Room 2

(CO Flooding System) 2 6.

HPCS Diesel Generator Room 2

(CO Flooding System) 2 7.

Control Room Ventilation (VC) Return Air Monitor System A 1

System B 1

8.

Control Room Ventilation (VC) Outside Air Monitor System A 1

System B 1

9.

Auxiliary Electric Equipment Room Ventilation (VE)

Return Air Monitor System A 1

System B 1

10.

SGTS Equipment Train (1VG015) 1 (2VG015) 1 11.

Control Room Emergency Make-up Air Filter Unit (0VC01SA) 1 Filter Unit (0VC01SB) 1 12.

Control Room HVAC Supply Air Filter Unit (0VC01FA) 1 Filter Unit (OVC01FB) 1

13. Auxiliary Electric Equipment Room HVAC Supply Air Filter Unit (OVE01FA) 1 HVAC Supply Air Filter Unit (OVE01FB) 1 O

LA SALLE - UNIT 1 3/4 3-76

~

p TABLE 3.3.7.9-1 (Continued)

FIRE DETECTION INSTRUMENTATION INSTRUMENT LOCATION MINIMUM INSTRUMENTS OPERABLE

• HEAT FLAME SMOKE Unit 1 Fire Detection Instrumentation (Continueo)

17. Off Gas Building a.

Off Gas Building, Zone 1-15P 3

El. 710'6", Fire Hazard Zone 10A1 b.

Off Gas Building, Zone 1-15 6

E1. 690', Fire Hazard Zone 1081 B.

Unit 2 Fire Detection Instrumentation Required For Unit 1 1.

Cable Spreading Room 15 (Dry Pipe Sprinkler System) 2.

Diesel Generator Corridor 9

(Dry Pipe Sprinkler System) 3.

Diesel Generator (20601K) Room 2

(Co2 Flooding System)

• p 4.

Auxiliary Building / Turbine Bldg.

a.

Aux. Bldg. Vent Floor, Zone 2-1 5

El. 815', Fire Hazard Zone 4A b.

Aux. Bldg. Vent Floor, Zone 2-2 9

E1. 786'6", Fire Hazard Zone 4B c.

Control Room, Zone 2-5 17 El. 768', Fire Hazard Zone 4C1 d.

Record Room, Zone 2-6 3

El. 768', Fire Hazard Zone 4C5 e.

Reactor Prot. M-G Set Room, Zone 2-12 12 El. 749', Fire Hazard Zone 404 f.

Cable Spreading Area, Zone 2-18 13 El 749', Fire Hazard Zone SA4 g.

Div. 2 SWGR Room, Zone 2-8 15 E1. 731', Fire Hazard Zone 4E4 h.

Aux. Electric Equipment Room, Zone 2-27 12 E1. 731', Fire Hazard Zone 4E2 1.

Aux. Bldg. Corridor, Zone 2-3 5

E1. 731', Fire Hazard Zone 5B13 j.

Aux. Bldg. Corridor, Zone 2-7 12 El. 731', Fire Hazard Zone 5B13 LA SALLE - UNIT 1 3/4 3-79 L

LASALLE COUNTY STATION UNIT 1 TECH SPEC CHANGE REQUEST NPF-ll/82-7

Subject:

Revise SRM Count Rate Reference (a): -Jay M. Pilant letter to Voss A. Moore dated April 2, 1974, " Proposed Technical Specificatons Waiver, Cooper Nuclear Station, AEC Docket No. 50-298, DPR-46."

Background:

The startup/ operational sources are the Antimony-Beryllium type which lose activity due to the 60.2 day half-life o f Antimony-124, unless regenerated at approximately 25% power.. Based on the current SRM count rate it is possible that the source strength will be insuffi-cient to maintain the required 3 cps until 25% power is reached.

Resourcing is possible (by either new source installation or re-irradiation of the current antimony pins in a test reactor) out would significantly delay startup.

i Similar license amendments uave been recently discussed with the Core Performance Branch of NRR by other utilities (e.g.

Pennsylvania Power and Light) in anticipation of startup delays and waivers have been received in the past at Cooper and other GE BWRs.

It is also noted that Reg. Guide 1.68 allows count rate as low as 1/2 cps.

Discussion:

The proposed change reduces the minimum required count rate on the LaSalle 1 Source Range Monitors (SRM's) from 3 cps to 0.7 cps for modes other than shutdown.

Tnis is conservative with respect to USNRC Regulatory Guide 1.68 Rev. 2 " Initial Tes t Programs for Water-Cooled Nuclear Power Plants" which states:

" A neutron count rate at least 1/2 count per second should register on the startup channels before startup begins, and the signal-to-noise ratio should be known to be greater than two."

Startup Mode Since a Level 1 Acceptance Criterion for Startup Test Procedure #6 requires an SRM signal-to-noise ratio greater than 2 (consistent with Regulatory Guide 1.68) the proposed minimum count rate of 0.7 cps is acceptable for monitoring initial criticality.

Th e allowable rod block setpoint o f 1/2 cps is consistent with Reg. Guide 1.68 while the nominal setpoint is specified at the same level as the operability surveillance value o f 0.7 cps.

.___-m

,.,y,

. The Source Range Monitoring System provides the operator with continuous indication of reactivity changes in the Refuel Mode and the Startup Mode until the Intermediate Range Monitoring system is well on scale (at least Range 3).

Although an upscale rod block occurs at 105 cps and the SRM shorting links are removed to activate a non-coincident scram at 5x105 cps under some circumstances, the SRM system is not credited with performing any protective or mitigating functions in any of the transients or accidents analyzed in the LaSalle FSAR.

Only the 120% scram from the APRM system is credited in the FSAR analysis of the Rod Drop Accident (RDA) which is the limiting fault at low power conditions.

The SRM and IRM systems, while providing additional monitoring and protection, are not needed for mitigation of the design basis RDA.

The e f fects o f the Dappler reactivity coefficient alone are sufficient to turn the neutron flux transient and the 120% APRM scram is sufficient to shutdown the core.

No credit is therefore taken for scrams from high flux on the SRMs, IRMs, c r the 15% APRM scram in the startup mode.

As such, the only important consideration in a modification of the minimum count rate requirement is that sufficient monitoring capability is maintained to detect positive reactivity insertions from the initial subcritical condition in a smooth and continuous f ashion.

That is, subcritical multiplication should be observed sufficiently in advance of achieving criticality that the operator is prepared and capable of placing the reactor on a stable and managable positive period to reach the heat-up range.

The use of neutron sources to produce on-scale SRM count rates for new cores fulfills this function and avoids any need to use the tedious " pull and wait" technique that might otherwise be necessary.

Since the bottom of scale on both the SRM meter and recorder is 0.1 cps, the proposed value o f 0.7 cps is well on-scale and will maintain adequate monitoring capability.

It is believed that the current 3 cps value is basically historical in nature having initially been chosen somewhat arbitrarily based on the SRM scale in an earlier BWR design.

A more technical basis was, however, presented by General l

Electric in April 1974 in support o f a similar amendment for Nebraska Public Power District's Cooper Station (Reference (a)).

According to i

General Electric, Cooper is one o f several BWRs which have previously received Tech. Spec. waivers to allow less than 3 cps during initial startup due to source decay.

Cooper was the first BWR to utilize antimony-beryllium (Sb-Be) sources for both loading and startup instead of the previous practice i

of using americium-beryllium sources for core loading with subsequent changeover to Sb-Be as the startup/ operational sources.

As at LaSalle 1, delays in Cooper's startup and the resultant decay of the antimony then necessitated a Tech. Spec. waiver on the SRM count rate.

L

. In the supporting information for the Cooper amendment General Electric related the 3 cps minimum count rate to the assumed initial power for the RDA and addressed the sensitivity of the RDA rod worths and peak fuel enthalpy to potentially lower initial power levels.

The results o f the analysis clearly demonstrate that lower initial conditions do not produce significantly dif ferent RD A results.

This would also be true for LaSalle where the calculated RDA roo worths are less than 0.5%

K ( Table 15.4.9-1 o f the FSAR).

It should also be noted that the Cooper evaluation assumes that the RDA could be initiated from a condition near critical (kef f

.99) and yet have less than 3 cps on the SHMs.

Since subcritical multiplication would be expected to build the SRM count rate significantly before approaching criticality, the assumption of a shutdown countrate o f 3 cps is extremely conservative when applied to a condition near critical.

Re fueling Mode The SRMs are classified in the LaSalle FSAR as a " system not required for safety" and serve no protective or mitigating function for transients and accidents analyzed therein.

In the Refuel Mode the events of interest are the Fuel Handling Accident (Section 15.7.4.1) and the Control Rod Removal Error Du ring Re fueling (Section 15.4.1.1).

The Fuel Handling Accident involves the potential release of fission products from mechanical damage to fuel rods, which occurs if an assembly is dropped from the fuel grapple onto other assemblies in the core.

The event does not involve a nuclear transient and the SRMs are therefore not required.

The Control Rod Removal Error During Refueling is prevented by refueling interlocks on both rod withdrawal and movement of the refueling bridge.

Since all control rods must be inserted when fuel is being loaded into the core, loading fuel into an uncontrolled cell is precluded.

In addition to procedural / administrative requirements, this is physically assured by interlocks which a) prevent the refueling bridge from moving over the core if the grapple is loaded and a control rod is withdrawn; b) apply a rod block if the bridge is over the core and the grapple is loaded.

Since the core is typically more than 5%

K subcritical with all-rods-in, an inadvertant criticality due to fuel movement alone is extremely unlikely.

The potential for inadvertant criticality due to rod motion alone is minimized by the combination o f Tech. Spec. requirements on Shutdown Margin and the interlock which prevents more than one rod from being withdrawn when the mode switch is in " REFUEL".

Upward removal of the control rod is impossible without prior removal o f the four adjacent assemblies due to the design o f the velocity limiter.

_4_

The SRMs are therefore only needed to provide information to the reactor operator while core alterations are underway.

Maintaining an on-scale count rate during in-core fuel handling is therefore-prudent and desirable.

The proposed specification requiring 0.7 cps accomplishes this by assuring an indicatior which is~7 times the bottom of scale count rate of 10-1 cps.

Since the initial loading of LaSalle Unit 1 has been completed and verified (with adequate SDM demonstrated at several intermediate core configurations) a reduction in the required count rate does not present any new safety concerns.

In the unlikely event that core alterations should be required before the neutron sources are regenerated from power operation, dunking chambers could be utilized in lieu of the SRMs to achieve higher count rates from the better monitoring geometry which is possible from portable detectors.

==

Conclusions:==

Commonwealth Edison has determined that no new safety questions have been raised and the change does not involve an increase in consequences of any previously analyzed event or the creation of any new transient or accident scenarios.

i 1

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4

i TABLE 3.3.6-2 l

9 CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION SETPOINTS m

r-p TRIP FUNCTION TRIP SETPOINT ALLOWABLE VALUE 1.

ROD BLOCK MONITOR a.

Upscale 5 0.66 W + 40%

5 0.66 W + 43%

H b.

Inoperative NA NA H

c.

Downscale

> 5% of RATED THERMAL POWER

> 3% of RATED THERMAL POWER 2.

APRM i

a.

Flow Biased Simulated l

Thermal Power-Upscale 5 0.66 W + 42%*

5 0.66 W + 45%

• b.

Inoperative NA NA c.

Downscale

> 5% of RATED THERMAL POWER

> 3% oT RATED THERMAL POWER d.

Neutron Flux-High 312%ofRATEDTHERMALPOWER 314%ofRATEDTHERMALPOWER 3.

SOURCE RANGE MONITORS w

a.

Detector not full in NA NA 5

5 D

b.

Upscale 5 2 x 10 cps 5 5 x 10 cp, w

c.

Inoperative N

NA>fcps h

d.

Downscale cps 4.

INTERMEDIATE RANGE MONITORS

.7 0.5

~

a.

Detector not full in NA NA b.

Upscale 5 108/125 of full scale 5 110/125 of full scale c.

Inoperative NA NA d.

Downscale

> 5/125of full scale

> 3/125 of full scale 5.

SCRAM DISCHARGE VOLUME a.

Water Level-High 5 765' % "

5 765' % "

b.

Scram Discharge Volume Switch in Bypass NA NA 6.

REACTOR COOLANT SYSTEM RECIRCULATION FLOW a.

Upscale 5 108/125 of full scale

~$ 111/125 of full scale l

b.

Inoperative NA NA c.

Comparator 5 10% flow deviation 1711% flow deviation

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

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

y

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INSTRUMENTATION

\\

SOURCE RANGE MONITORS Ang LIMITING CONDITION FOR OPERATION 3.3.7.6 At least three source range monitor channels shall be OPERABLE.

APPLICABILITY: OPERATIONAL CONDITIONS 2*, 3 and 4.

ACTION:

In OPERATIONAL CONDITION 2* with one of the above required source a.

range monitor channels inoperable, restore at least 3 source range monitor channels to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

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 one hour.

SURVEILLANCE REQUIREMENTS 1

4.3.7.6 Each of the above required source range monitor channels shall be O

demonstrated OPERABLE by:

a.

Performance of a:

1.

CHANNEL CHECK at least once per:

1) 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> in CONDITION 2*, and b) 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in CONDITION 3 or 4.

2.

CHANNEL CALIBRATION ** at least once per 18 months.

b.

Performance of a CHANNEL FUNCTIONAL TEST:

1.

Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to moving the reactor mode switch from the Shutdown position, if not performed within the previous 7 days, and 2.

At least once per 31 days.

l c.

Verifying, prior to withdrawal of control rods, that the SRM count rateisatleast$cpswiththedetectorfullyinserted.

0.7 l

"With IRM's on range 2 or below.

• • Neutron detectors may be excluded from CHANNEL CALIBRATION.

g LA SALLE - UNIT 1 3/4 3-72 u

X.

REFUELING OPERATIONS 9

SURVEILLANCE REQUIREMENTS (Continued) b.

Performance of a CHANNEL FUNCTIONAL TEST:

1.

Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to the start of CORE ALTERATIONS, and 2.

At least once per 7 days.

g Verifying that the channel count rate is at leasth cps:

c.

1.

Prior to control rod withdrawal,

~2.

Prior to and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during CORE ALTERATIONS, and 3.

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

d.

Verifying that the RPS circuitry " shorting links" have been removed within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> prior to and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during:

1.

The time any control rod is withdrawn,## or 2.

Shutdown margin demonstrations.

O Not required for control rods removed per Specification 3.9.10.1 or 3.9.10.2.

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

r-Status o f Tech Spec-Change Requests NRC Topic Submitted Actio n NPF-ll/82-1 Hydrogen Recombiner

-5/24/82 Temperature Controller -

Setpoint Deviation.

(Required Prior'to Criticality)

NPF-11/82-2 MSIV Closure Scram Setpoint 5/24/82 (Required Prior to Criticality).

NPF-11/82-3 Special Test Exception Change 6/01/82 6/03/82 verbal for Confirmatory Flow Induced Authorization Vibration Test (Required Prior 6/07/82 letter to Test which commenced 6/03/82).

NPF-ll/82-4 Revised Snubber List 6/07/82

( Licens e Condition 2.C. (5). (a)

Required Prior to Criticality.

NPF-l'/82-5 Tendon Tables 6/14/82 (License Condition 2.C. (7)

Required Prior to Full Power) j NPF-ll/82-6 Fire Detector Tables 6/14/82 (License Condition 2.C. (25)(b)

Required Prior to Criticality)

NPF-ll/82-7 SRM Countrate 6 /14/8 2 (Required prior to source decay below 3 cps).

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