ML20127L472

Amends 90 & 83 to Licenses DPR-19 & DPR-25,respectively, Adding Limiting Conditions for Operation & Surveillance Requirements to Tech Specs for Mods Required by Generic Ltr 83-36

ML20127L472
Person / Time
Site:	Dresden   (DPR-19-A-090, DPR-19-A-90, DPR-25-A-083, DPR-25-A-83)
Issue date:	06/24/1985
From:	Zwolinski J Office of Nuclear Reactor Regulation
To:
Shared Package
ML20127L476	List: ML20127L472
References
GL-83-36, NUDOCS 8506280029
Download: ML20127L472 (61)

Text

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UNITED STATES NUCLEAR REGULATORY COMMISSION o

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t WASHINGTON, D. C. 20065 s.....)

COMMONWEALTH EDISON COMPANY DOCKET NO. 50-237 DRESDEN NL' CLEAR POWER STATION, UNIT NO. 2 AMENDMENT TO PROVISIONAL OPERATING LICENSE Amendment No. 90 License No. OPR-19 1.

The Nuclear Regulatory Comission (the Comission) has found that:

A.

The application for amendment by the Commonwealth Edison Company (the licensee) dated October 10, 1984 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter It B.

The facility will operate in confonnity with the appitcation, the provisions of the Act and the rules and regulations of the Comission; C.

There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (11) that such activities will be conducted in compliance with the Comission's regulations; D.

The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the pubitc; and

. E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have l

been satisfied.

i 8506200029 850624 I

ADOCKOS00g7 DR

. 2.

Accordingly, the Itcense is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 3.8 of Provisional Operating License No. DPR-19 is hereby amended to read as follows:

B.

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 90, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of its issuance.

FOR T E NUCLEAR REGUl AT0if COMMISSIO n _1 4 John. Zwolinski, Chief Opera ing Reactors Branch #5 Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: June 24, 1985.

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A7ACHMFNT TO LICENSE AMENDMENT NO. 90 PROVISIONAL OPERATING LICENSE DPR-19 DOCKET NO. 50-237 Revise Appendix A Technical Specifications by removing the pages identified below and insertina the attached pages.

The revised pages are identified by

+he captioned amendment number and contain marginal lines indicattre the cree of change.

REMOVE INSERT 11 11 vii vif viii viii 3/4.2-3 3/4.2-3 3/4.2-17 3/4.2-17 3/4.2-18 3/4.2-18 3/4.2-19 3/4.2-19*

3/4.2-20 3/4.2 20*

3/4.2-21 3/4.2-21*

3/4.2-22 3/4.2-22*

3/4.2-23 3/4.2-23*

B 3/4.2-24 3/4.2-24*

B 3/4.2-25 3/4.2-25*

B 3/4.2-26 3/4.2-26 8 3/4.2-27 3/4.2-27 B 3/4.2-28 B 3/4.2 28*

B 3/4.2-29 B 3/4.2-29*

B 3/4.2-30 B 3/4.2-30*

B 3/4.2-31 B 3/4.2-31*

B 3/4.2-32 8 3/4.2-32*

B 3/A.2-33 8 3/4.2-33 B 3/4.2-34 8 3/4.2-34*

B 3/4.2-35*

B 3/4.2-36*

B 3/4.2-37 8 3/4.2-38*

• Pagination change only

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DRESDEN II DPR-19 AmendmentNo.pe,p690 Table of Contents g

1.0 Definitions 1.0- 1 1.1 safety Limits - Fuel Cladding Integrity 1/2.1-1 Safety Limit Bases B 1/2.1-6 1.2 Safety Limits - Reactor Coolant System Safety Limit Bases 1/2.2-1 3 1/2.2-2 2.1 Limiting Safety System Settings - Fuel Cladding Integelty 1/2.1-1 Limiting Safety system settings Bases B 1/2.1-10 2.2 Limiting Safety System Settings - Reactor Coolant System 1/2.2-1 Limiting Safety System Settings Bases B 1/2.2-4 3.0 LIMITING CONDITION FOR OPERATION Limiting Condition for Operation Bases 3.0- 1 B 3.0- 3 3.1 Reactor Protection System 3/4.1-1 Limiting Conditions for Operation Bases (3.1) 1 3/4.1-9 Surveillance Requirement Bases (4.1) 3.2 Protective Instrumentation B 3/4.1-15 3/4.2-1 3.2.A Primary Containment Isolatlon Functions 3/4.2-1 3.2.B Core and Containment Cooling systems - Initiation and Control 3.2.C Control Rod Block Actuation 3/4.2-1 3 / 4. 2J"2 3.2.D Refueling Floor Radiation Monitors 3.2.E Pos' Accident Instrumentation 3/4.2 2 3/4.2-3 3.2.F Red etive Liquid Effluent Instrumentation 3/4.2-4 3.2.0 Radioactive caseous Effluent Instrumentation 3/4.2-5 Limiting Conditions for Operation Bases (3.2)

B 3/4.2-28 Surveillance Requirement Bases (4.2) 3.3 Reactivity Control B 3/4.2-34 3.3.A Reactivity Limitations 3/4.3-1 3.3.8 Control Rods 3/4.3-1 3.3.C Scram Insertion Times 3/4.3-4 3/4.3-10 3.3 D Control Rod Accumulators 3.3.E Reactivity Anomalles 3/4.3-11 3/4.3-12 3.3.0 Economic Generation Control System

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3/4.3-13 Limiting Condit1ons for Operation Bases (3.3)

B 3/4.3-14 Surveillance Requirement Bases (4.3)

B 3/4.3-22 3.4 Standby Liguld control System 3/4.4-1 3.4.A Normal Operation 3/4.4-1 3.4.B operation with Inoperable Components 3.4.C Liquid Polson Tank 3/4.4-2 3/4.4-3 3.4.D Reactor Shutdown Requirement 3/4.4-3 Limiting Conditions for Operation Bases (3.4)

B 3/4.4-6 surveillance Requirement Bases (4.4)

B 3/4.4-7

• 5 Core and containment Cooling systems 3/4.5-1 3.5.A Core Spray and LPCI subsystems 3/4.5-1 3.5.5 Containment cooling subsystem 3/4.5-5 11 3959a 3843A

DRESDEN II DPR-19 a

Amendment No. S/, pf, g6 9 0 T

List of Tables EA11 Table 3.1.1 Reactor Protection System (Scram) 3/4.1 - 5 Instrumentation Requirements Table 4.1.1 Scram Instrumentation Functional Tests 3/4.1 - 8 Table 4.1.2 Scram Instrumentation Calibration 3/4.1 -10 Table 3.2.1 Instrumentation that Initiates Primary Containment Isolation Functions 3/4.2 - 8 Table 3.2.2 Instrumentation that Initiates or controls the Core and Containment Cooling System 3/4.2 -10 Table 3.2.3 Instrumentatlon that Initlates Rod Block l

Table 3.2.4 Radlosctive Liquid Effluent 3/4.2 -12 l

Monitoring Instrumentation 3/4.2 -14 Table 3.2.5 Radioactive Caseous Effluent l

Monitoring Instrumentation l

Table 3.2.6 Post Accident Monitoring Instrumentation 3/4.2 -15 Regulroments 3/4.2 -17 Table 4.2.1 Minimum Test and Calibration Frequency for Core and Containment Cooling Systems Instrumentation, Rod Blocks, and Isolations 3/4.2 -19 Table 4.2.2 Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements 3/4.2 -f2 Table 4.2.3 Radloactive Caseous Effluent Nonitoring Instrumentation Surveillance Requirements 3/4.2 -24 Table 4.2.4 Post Accident Monitoring Instrumentation i

Surveillance Requirements 3/4.2 -26 Table 4.6.2 Neutron Flus and Sample Withdrawal B 3/4.6-26 Table 3.7.1 Primary Containment Isolation 3/4.7 -31 Table 4.8.1 Radioactive Caseous Waste Sampilns and Analysis Program 3/4.8-22 Table 4.8.2 Maximum Permissible Concentration of Dissolved 4

or Entrained Noble Cases Released From the Site to Unrestricted Areas in Liquid Weste 3/4.8-24 Table 4.8.3 Radioactive Liquid Weste Sampilns and Analysis Program 3/4.8-25 Table 4.8.4 P.adiological Invironmental Monitoring Program 3/4.8-27 Table 4.8.5 Reporting Levels for Radioactivity Concentrations in Environmental Samples 3/4.8-28 Table 4.8.6 Practical Lower Limits of Detection (LLD) for Standard Radiological Environmental Monitoring Program 3/4.8-29 Table 4.11-1 Surveillance Requirements for High Energy Piping Outside Containment 3/4.11-3 Table 3.12-1 Fire Detection Instruments S 3/4.12-17 Table 3.12-2 Sprinkler Systems 3 3/4.12-18

• Table 3.12-3 CO2 Systems B 3/4.12-19 Table 3.12-4 Fire Mose St4tions

- 3 3/4.12-20 & 21 Table 6.1.1 Minimum Shift Manning Chart y-6-4 Table 6.6.1 Special Reports 6-23 vil 3959a 3843A i

DRSSDRN 11 DFR-19 Ameldme:t No. %, p, $4y4 List of Flaures

i. alt.

Figure 2.1-3 AFRM Blas Scram Relationship to Normal Operating conditions B 1/2.1-17 Figure 4.1.1 Graphical Aid la the Selection of an Adequate Interval Between Tests B 3/4.1-18 Figure 4.2.2 Test Interval vs. Systee Unavailability B 3/4.2-38 Figure 3.4.1 Standby Liquid Control Solution Requirements 3/4.4-4 Figure 3.4.2 Sodium Fontaborate solution Temperature Requirements 3/4.4-5 Figure 3.5-1 Nazimum Average Flaner LNOR 3/4.5-17 (consisting of eight fuel type curves) thru 24 Figure 3.5-2 Core Flow %

3/4.5-2,7 & 28 Figure 3.6.1 Minimus Temperature Requirements r-per Appendla C of 10 CFR 50 3/4.6-20 Figure 4.6.1 Ninimum teactor Pressurisation Temperature 3 3/4.6-25 Figure 4.6.2 Chloride Stress Corrosion Test Results at 500'F S 3/4.6-27 Figure 4.8.1 Owner Controlled / Unrestricted Area Boundary 3 3/4.8-38 Figure 4.8.2 Detail of Central Comples 3 3/4.8-39 Figure 6.1-1 Corporate and statloa Organisation 6-3 t

v111 Stita 8401D i

s DRESDEN II DPR-19 Amendment No. %. % 9(

3.2 LIMITING CONDITION POR OPERATION 4.2 SURVEILLANCE. REQUIREMENTS (CONT'D)

(CONT'D) in the fuel storage pool treatment, system initiation and during refueling or shall be performed at least fuel movement operations, each operating cycle.

2.

One of the two re-fueling floor radiation monitors may be inoper-able for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. If the inoperable monitor is not restored to service in this time, the reactor building ventilation system shall be isolated and the standby gas treet-ment operated until repairs are complete.

3.

The trip setting for the refueling floor radiation monitors shall be set at a less than or equal to 100me/hr.

4 Upon loss of both re-fueling floor radiation monitors while in use, the reactor building ventilation system shall be isolated and the standby gas treet-ment operated.

E.

Post Accident E.

Post Accident Instrumentation Instrumentation The limiting conditions for operation for the post accident instrumen-instrumentation, which tation,shall be function-is read out in the con-ally tested and calibrated as indicated in Table 4.2.4.

trol room, required for post accident monitoring are given in Table 3.2.6.

3/4.2-3 3960s 3843A

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DRESDEN II DPR-19 Amendment No. $ 0 Table 3.f.61 Post Accident Monitorina Instrunertation Reautrements li i Minimum Ins trument Number of Readout Operable Location Number Instrument ihannels (1)

Parameter Unit 2 reovided ganze l

1 Reactor Pressure 902-5 1

0-1500 psig 2

0-1200 psig 1

Reactor Water Level 902-3 2

-360 to +60 inches 1

Torus Water Temperature 902-4 2

0-200*F 2 (3)

Torus Water Level 902-3 1

-25 to +25 inches l

Indicator 902-3 1

-7 t) +3' inches (narrow range) 902-2 2

0-?.0 ft.

(wids range) i Torus Water Local 1

Sight Class 18 inch range r-(narrow range) 1 (4)

Torus Pressure 902-5 1

-2,45-5 psig 2

Drywell Pressure 902-5 1

0-5 psig 902-3 1

C-75 psig 902-3

-2 0-250 psig i

2 Drywell Temperature 902-21 6

0-600*F 2

Neutron Monitoring

'932-5 4

0.1-106 CPS s

1 (4)

Torus to Drywell 902-3 2

0-3 psid Differential Pressure 1

Drywell Radiation Mon! tor 902-55,56 2

1 to 108 R/hr 2/ valve (2)

Main Steam RV Position, 902-21 1 per valve N/A Acoustic Monitor Main Steam RV Position.

902-21 1 per valve 0-600*F Temperature Monitor 2/ valve (2)

Main Steam SV Position, 90?-21 2 per valve N/A Acoustic Monitor Main Steam SV Position,

902-21 1 per valve 0-500*F s

Temperature Monitor 1 (5)

Drywell Hydrogen 902-55 2

0 10t Concentration 902'35 s

Notes:

(See Next Page)

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3/4.*t-17 3960a 3843A h

DRESDgN II DPR-19 Amendment No. $ f Table 3.2.6 lhL.if.

1.

From and after the date that a parameter is reduced to the minimum number of channels, continued operation is not permissible beyond thirty (30) days unless such instrumentation is sooner made operable.

In the' event that all indications of a parameter is disabled and such indication cannot be restored in six (6) hours, an orderly shutdown shall be inittsted and the reactor shall be in a cold shutdown condition in twenty-four (24) hours. See notes 2, 3, 4 and 5 for exceptions to this requirement.

2.

If the number of position indicators is reduced to one indication on one or more valves, continued operation is permissible; however, if the reactor is in a cold shutdown condition for longer than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, it may not be started up until all position indication is restored.

In the event that all position indication is lost on one or more valves and such indication cannot be restored in thirty (30) days, an orderly shutdown shall be initiated, and the reactor shall be depressurized to less than 90 pais in twenty-four (24) hours.

3.

From and after the date that this parameter is reduced to either one narrow-range indication or one wide-range indication, continued reactor-operation is not permissible beyond thirty (30) days unless such instrument is sooner made operable. In the event that either all narrow-range indication or all wide-range indication is disabled, continued reactor operation is not permissible beyond seven (7) days unless such instruments are sooner made operable.

In the event that all indication for this parameter is disabled, and such indication cannot be restored in six (6) hours, an orderly shutdown shall be intitiated and the reactor shall be in a cold shutdown condition in twenty-four (24) hours.

4 From and af ter the date that one of these parameters becomes inoperable, continued operation is not permissible beyond thirty (30) days unless such instrumentation is sooner made operable.

In the event that all indication of these parameters is disabled and such indication cannot be restored in six (6) hours, an orderly shutdown shall be intitiated and the reactor shall be in cold shutdown in twenty-four (24) hours.

5.

From and after the date that one of the drywell hydrogen monitors becomes inoperable, continued reactor operation is permissible, If both drywell hydrogen monitors are inoperable.-continued reactor a.

operation is permissible for up to 30 days provided that during this time the HRSS hydrogen monitoring capability for the drywell is operable.

b.

Ifalldrywellhydrogenmonitoringcapabilityislost,[ continued j

reactor operations is permissible for up to 7 days.

3/4.2-18 3960s 3843A y

Tabla 4.2.1 DRESDEN II DPR-19 *

• MINIful TEST AND CALIBRATION FREQUENCY FOR CDRE AND CONTAINENT CEOLING SYSTEMS INSTRtMNTATION, R00 BLOCKS, AND ISOLATIO i

Instrument Channel Instrument Instrunent Functional Test Calibration Check ECCS Instrunentation 1.

Reactor Low. Low Water Level (1)

Once/3 Months once/ Day 2.

Drywell High Pressure (1)

Once/3 Months None 3.

Reactor Low Pressure (1)

Once/3 Months None 4.

Contaltunent Spray Interlock a.

2/3 Core Height (1) (13)

(13)

None b.

Contaltunent High Pressure (1)

Once/3 Months None 5.

Low Pressure Core Cooling Purp (1)

Once/3 Months None Discharge 6.

Undervoltage Emergency Bus Refueling Outage Refuel Outage once/3 months 7.

Sustained High Reactor Pressure (1)

Once/3 Months None 8.

Degraded voltage Energency Bus Refueling Outage (10)

Refuel Outage Monthly Rod Blocks 1.

APRM Downscale (1) (3)

Once/3 Months None 2.

APRM Flow Variable (1) (3)

Refuel Outage None 3.

APRM t$ scale (Startup/ Hot Staney)

(2) (3)

(2) (3)

(2) 4.

IRM15 scale (2) (3)

(2) (3)

(2) 5.

IRM Downscale (2) (3)

(2) (3)

(2) 6.

IRM Detector Not Fully Inserted (2)

N/A None in the core 7.

ROM Upscale (1) (3)

Refuel Outage None r-8.

ROM Downstale (1) (3)

Once/3 Months None 9.

SRM Upscale (2) (3)

(2) (3)

(2)

10. SRM Detector Not in Startup Position (2) (3)

(2) (3)

(2)

11. Scran Instrument Volume Level High Once/3 Months (9)

None Ihne Containment Monitorina 1.

Pressure Minus 5 in. Hg to plus 5 psig None once/3 Months Once/ Day a.

Indicator b.

O to 75 psig Indicator None once/3 Months None 2.

Temperature None Refuel Outage Once/ Day 3.

Drywell-Torus Differential None Once/6 Months (Two None Pressure (5) (6)

Channels Operable)

(0-3 psid)

Once/ Month (One Channel Operable) 4.

Torus Water Level (5) (6)

None Once/6 Months Plus or minus 25 in. Wide Range a.

Indicator b.

18 in. Sight Glass safetv/ Relief Valve Monitorino 1.

Safety / Relief Valve (7)

None Once Per Position Indicator (Acoustic Monitor) (8) 31 Days

/ 2.

Safety / Relief Valve Position None Once overy Once Per Indicator (Taiperature 18 months -

31 Days Monitor (8) 3.

Safety Valve Pxition Indicator (7)

None E Once Per (Acoustic Monitor) (8) 31 Days 4.

Safety valve Position Indicator None once every Once Per (Temperature Monitor) (8) 18 months 31 Days (Table cont'd next page) 3/4.2-19 l

8401D

s Tabla 4.2.1 DRELDEN II DPR-19 MINIftfl TEST AND CALIBRATION FREQUENCY FOR CORE AND Amendment No. d.

CONTAlifENT COOLING SYSTEMS INSTRUPENTATION, R00 BLOCKS, AND ISOLATION 5 t

Instrsment Channel Instrument Instrisnent Functional Test Calibration Check Main Steam Line Isolation 1.

Steen Tunnel High Tapperature Refueling Outage Refuel Outage None 2.

Steen Line High Flow (1)

Once/3 Mon'ths once/ Day 3.

Steam Line Low Pressure (1)

Once/3 Months None 4.

Steam Line High Radiation (1) (3)

Once/3 Months (4)

Once/ Day t

Isolation Condenser Isolation 1.

Steam Line High Flow (1)

Once/3 Months None 2.

Condensate Line High Flow (1)

Once/3 Months None IPCI Isolation 1.

Steen Line High Flow (1) (11) (12)

(11) (12)

None 2.

Steen Line Area High Tamperature Refueling Outage Refuel Outage None 3.

Low Reactor Pressure (1) (13)

(13)

None Reactor Buildina Vent Isolation and 50GTS Initiation 1.

Refueling Floor Radiation Monitors (1)

Once/3 Months Once/ Day NOTES: (For Table 4.2.1)

F 1.

Initially once per month until exposure hours (N as defined on Figure 5

4.1.1) is 2.0 x 10 ; thereafter, according to Figure 4.1.1 with an interval not less than one month nor more than three months.The compilation of instrument failure rate data may include data obtained from other Boiling Water Reactors for which the same design instrument operates in an environment siellar to that of Dresden Unit 2.

2.

Function test calibrations and instrument checks are not required when these instruments are not required to be operable or are tripped.

Functional tests shall be performed before each startup with a required frequency not to exceed once per week. Calibrations shall be performed during each startup or during controlled shutdowns with a required frequency not to exceed once per week.

Instrument checks shall be performed at least once per week.

Instrument checks shall be performed at least once per day during those periods when the instruments are required to be operable.

3.

This instrumentation is excepted from the functional test definition.

The functional test will consist of injecting a simulated electrical signal into the measurement channel.

See Note 4.

4.

These instrument channels will be calibrated using simulated electrical signals once every three months.

In addition, calibration including the sensors will be performed during each refue14ag outage.

5.

A minimum of two channels is required.

P (Cont'd. next page) 3/4.2-20 l

8401D 1

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i DRESDgN II DPR-19 Amendment No. f), $6, g4 $0 EgIgg:

(For Table 4.2.1) (Cont'd.)

i 3

6.

From and after the date that one of these parameters (...nither drywell-torus differential pressure or torus water level indication) s is reduced to one indication, continued operation is not permissible

]

beyond thirty days, unless such instrumentation is sooner made operable.

j In the event that all indications of these parameters

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(...either drywell-torus differential pressure or torus water level) is disabled and such indication cannot be restored in six (6) hours, 1

an orderly shutdown shall be initiated and the reactor shall be in a cold shutdown condition in twenty four hours.

i j

7.

i Functional tests will be conducted before startup at the end of each refueling outage or after maintenance is performed on a particular i

Safety / Relief Valve.

8.

If the number of position indicators is reduced to one indication on one or more valves, continued operation is permissible; however, if the reactor is in a cold shutdown condition for more than seventy-two 3

j hours, it may not be started up until all position indication is restored.

In the event that all position indication is lost on one or more valves and such indication cannot be restored in thirty days, anr-orderly shutdown shall be initiated, and the reactor shall be depressurized to less than 90 psig in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

9.

The functional test of the Scram Discharge Volume thermal switches is not applicable; i.e., the switch is either on or off.

Further, these switches are mounted solidly to the device and have a very low probability of moving; e.g., the thermal switches in the scram discharge volume tank.

Based on the above, no calibration is required for these instrument channels.

10.

Functional test shall include verification of the second level undervoltage (degraded voltage) timer bypass and shall verify i

operation of the degraded voltage 5-minute timer and inherent 7-second timer.

11.

Verification of time delay setting between 3 and 9 seconds shall be performed during each refueling outage.

12.

Trip units are functionally tested monthly (staggered one channel out of four every week). A calibration of the trip units is to be performed concurrent with the functional testing.

13.

Trip units are functionally testod monthly (staggered one division out of two every two weeks).

s A calibration of the trip units is to be performed concurrent with the functional t'esting.

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3920s

$401D

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DRESDEN II DPR-19 Ameidment ND. p, f gp TABLE 4.2.2 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS Instrument Calibration Functional Source Instrument Check (1)(7)

(1)(7)(3)(4)

Test (1)(2)(7) Check (1)

Liquid Radweste D

R Q (6)

(5)

Effluent Cross Activity Monitor Service Water D

R Q (6)

R Effluent Cross Activity Monitor Tank Level Indicating Device

a. A Waste Sample Tank D

R Q

N/A

b. B Waste Sample Tank D

R Q

N/A

c. C Waste Sample Tank D

R Q

N/A r-

d. A Floor Drain Sample D

R Q

N/A Tank

e. B Floor Drain Sample D

R Q

N/A Tank

f. Weste Surge Tank D

R Q

N/A Notes:

(See Next Page)

I c

3/4.2-22 I

l 3960s 3843A

e DRESDEN II DPR-19

/endmentNo.14.P6 90 TABLE 4.2.2 (Notes) 1.

D = Once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> M = Once per 31 days Q = Once per 92 days R = Once per refueling outage S = Once per 6 months 2.

The Instrument Functional Test shall also demonstrate that control room alarm annunciation occurs, if any of the following conditions exist, where applicable.

Instrument indicates levels above the alarm setpoint.

s.

b.

Circuit Failure.

Instrument indicates a downscale failure.

c.

d.

Instrument controls not set in OPERATE mode.

3.

Calibration shall include performance of a functional test.

4.

Calibration shall include performance of a source check.

5.

Source check shall consist of observing instrument response during a discharge.

r-6.

Functional test may be performed by using trip check and test circuitry associated with the monitor chassis.

7.

Function test calibrations and instrument checks are not required when these instruments are not required to be operable or are tripped.

Calibration shall be performed once per refueling outage and not more than once every 18 months. Instrument checks shall be performed at least once a day during those periods when the instruments are required to be operable.

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3/4.2-23

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3960s 3843A

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DRESDEN II DPR-19 Amendment ND. %, p 3; TABLE 4.2.3 RADIDACTIVE CASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS Instrument Calibration Function Test Source Instrument cheek (1)(6)

(1)(6)(3)

(1)(4)(2)(6)

Check (1)

SJAE Radiation D

R Activity Monitor Q

R Reactor Blds Vent D (4)

N/A N/A N/A Particulate and Iodine Sampler Reactor Blds Vent D

R Exhaust Duct Q

Q Radiation Monitor Reactor Blds Vent D

R SPING Noble Cas Q

M Monitor Lo, Mid, High Range Main Chimney Noble D

R Cas Activity Monitor Q

M Main Chimney SPING D

R Noble Gas Monitor Q

M Lo, Mid, High Range Main Chimney D (4)

N/A N/A N/A Particulate and Iodine Sampler Main Chimney Flow D

R Q

N/A Rate Monitor Main Chimney Sampler D

R Q (5)

N/A Flow Rate Monitor Reactor Bldg Vent D

R Flow Rate Monitor Q

N/A Reactor Bldg Sampler D

R Flow Rate Monitor Q (5)

N/A P.

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3843A i

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DRESDEN II DPR-19 i

Amendment No. 9. M.9 0 TABLE 4,2.3 (Notes) 1.

D = Cnce per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> M = Once per 31 days Q = Once per 92 days R = Once per refueling outage 2.

The Instrument Functional Test shall also demonstrate that c alarm annunciation occurs, if any of the following conditions exist m applicable,

, where Instrument indicates levels above the alarm setpoint.

n.

b.

Circuit Failure.

Instrument indicates a downscale failure.

c.

d.

Instrument controls not set in OPERATE mode.

3.

Calibration shall include performance of a functional test 4.

place and functioning properly. Instrument check to verify operabili n

5.

Function Test shall be performed on local switches providing low flow alarm.

6.

r-Function test calibrations and instrument checks are not required wh these instruments are not required to be operable or are tripped en Calibration shall be performed once per refueling outage and not more th once every 18 months.

per day during those periods when the instruments are required an operable.

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3/4.2-25 l

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3960s 3843A

DRESDEN II DPR-19 g

Amendment ND. 9 0 Table 4.2.4 i

Post Accident Monitorina Instrumentation Surveillance Reauirements 4

Minimum Instrument Number of Readout Operable Location Instrument Channels Parameter Unit 2 Calibration Check 1

Reactor Pressure 902-5 Once Every Once'Per Day 6 Nonths 1

Reactor Water Level 902-3 Once Every Once Per Day 1

6 Months 1

Torus Water Temperature 902-4 Once Ryery Once Per Day 12 Months Torus Water Level 902-3 Once Every Once Per Day Indicator (Narrow Range) 6 Noaths (Sight Class)

N/A None (Wide Range) 902-2 Once Every Once Per 31 12 Nonths Days 1

Torus Pressure 902-3,5 Once Ryery Once Per Day 3 Months 1

Torus to Drywell 902-3 Differential Pressure Once Every Once Per Def 6 Months 2

Drywell Pressure (0-5 psig) 902-5 Once Ryery Once Per Day 3 Nonths (0-75 psis) 902-3 once Every Once Per 31 3 Months Days (0-250 psis) 902-3 Once Every once Per 31 i

Refuel Days s.

2 Drywell Temperature 902-21 Once Ryery once Per Day Refuel 2

Neutron Monitoring 902-5 Once Every Once Per Day 3 Month:

1 Drywell Radiation Monitor 902-55,56 Once Every Once Per 31 Refuel (2)

Days Main Steam NY Position, 902-21 2/ Valve Temperature Monitor Once Every Once Per 31 Refuel Days Main Steam NY Position.

(1)

Acoustic Monitor Main Steam SV Position, 902-21 Temperature Monitor Once Every Once Per 31 Refuel

.' Days 2/Yalve Main Steam SV Position.

Acoustic Monitor (1)

I0ece Per 31

- Days 1

Drywell Mydrogen 902-55 Once avery Once Per 31 Concentration 902-56 3 Months Days Notes:

(See Next Page) 3960s 3/4.2-26 3843A

Dresd n II DPg-19 j

Amendment No. J)

Table 4.2.4 (Notes)

Notes 1.

Calibration of Acoustic Monitors shall consist of verifying the instrument threshold levels, and will be performed monthly.,

Functional tests will be conducted before startup at the end of each refueling outage or after maintenance is performed on a particular safety or relief valve.

2.

Calibration shall consist of an electronic calibration of the channel, not including the detector, for range decades above 10 R/hr; and a one-point calibration check-of the detector below 10 R/hr with an installed or portable samma source.

r-m 3/4.2-27 3960a 3843A

DRESDEN II DPR-19 g

Amendment No. N. @ 9 0 3.2 LIMITING CONDITION FOR OPERATION BASES In addition to reactor protection instrumentation which initiates a reactor scram, protective instrumentation has been provided which initiates action to mitigate the consequences of accidents which are beyond the operator's ability to control, or terminates operator errors before they result in serious consequdhees.

This set of specifications provides the limiting conditions of operation for the primary system isolation function, initiation of the emergency core cooling system, control rod block and standby gas treatment systems.

The objectives of the specifications are

1) to assure the effectiveness of the protective instrumentation when required by preserving its capability to tolerate a single failure of any component of such systems even during periods when portions of such systems are out of service for maintenance, and

2) to prescribe the trip settings required to assure adequate perfoemance.

When necessary, one channel may be made inoperable for brief intervals to conduct required functional tests and calibrations.

Some of the settings on the instrumentation that' initiates or control core and containment cooling have tolerances explicitly stated where the high and low values are both critical and may have a substantial effect on safety.

It should be noted that the r-setpoints of other instrumentation, where only the high or low end of the setting has a direct bearing on safety, are chosen at a level away from the normal operating range to prevent inadvertent actuation of the safety system involved and exposure to abnormal situations.

Isolation valves are installed in those lines that penetrate the primary containment and must be isolated during a loss-of-coolant accident so that the radiation dose limits are not exceeded during an accident condition.

Actuation of these valves is initiated by protective instrumentation which serves the condition for which isolation is required (this instrumentation is shown in Table 3.2.1).

Such instrumentation must be available whenever primary containment integrity is required.

The objective is to isolate the primary containment so that the guidelines of 10 CFR 100 are i

i not exceeded during an accident.

The instrumentation which initiates primary system isolation is connected in a dual bus arrangement. Thus the discussion given in the bases for Specification 3.1 is applicable here.

The low-reactor level instrumentation is set to trip at greater than 8 inches on the level instrument (top of active fuel is defined to be 360 inches above vessel zero) and after allowing for the full power pressure drop across the steam dryer the lo'w level trip is at 504 inches above vessel zero, or 144 inches,_above the B 3/4.2-28 l

3960a 3843A E

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DRESDEN II DPR'-19 g

Amendment No. p, p 9 0 3.2 LIMITING CONDITION FOR OPERATION BASES (Cont'd.)

top of active fuel.

Retrofit 8 I 8 fuel has an active fuel length l

1.24 inches longer than earlier fuel designs. However, present trip setpoints were used in the LOCA analyses. This trip i

initiates closure of Group 2 and 3 primary containmenb isolation vsTves but does not trip the recirculation pumps (reference SAR.

S6ction 7.7.2).

For a trip setting of 504 inches above vessel sero (144 inches above top of active fuel) and a 60-second valve closure time, the valves will be closed before perforation of the cladding occurs even for the maximum break; the setting is therefore adequate.

The low low reactor level instrumentation is set to trip when reactor water level is 444 inches above vessel zero (with top of active fuel defined as 360 inches above vessel zero, - 59 inches is 84 inches above the top of active fuel). This trip initiates closure of Group I primary containment isolation valves (reference SAR Section 7.7.2.2) and also activates the ECC subsystems, starts the emergancy diesel generator, and trips the recirculation This trip setting level was chosen to be high enough to pumps.

i prevent spurious operation but low enough to initiate ECCS

~

operation and primary system isolation so that no melting of the fuel cladding will occur and so that post accident cooling can be c accomplished and the guidelines of 10 CFR 100 will not be exceeded. For the complete circumferential break of a 28-inch recirculation line and with the trip setting given above, ECCS initiation and primary isolation are initiated and in time to meet the above criteria. The instrumentation also covers the full spectrum of breaks and meets the above criteria.

The high-drywell pressure instrumentation is a backup to the water level instrumentation and, in addition to initiating ECCS, it causes isolation of Group 2 isolation valves. For the breaks discussed above, this instrumentation will initiate ECCS operation at abcut the same time as the low low water level instrumentation; thus the results given above are applicable here, also Group 2 isolation valves include the drywell vent, purge and sump isolation valves.

High-drywell pressure activates only these valves because high drywell pressure could occur as the result of non-safety-related causes such as not purging the drywell air during startup. Total system isolation is not desirable for these conditions, and only the valves in Group 2 are required to close.

The low low water level instrumentation initiates protection for the full spectrum of loss-of-coolant accidents and causes a trip cf Group 1 primary system isolation valves.

P.

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3960s 3843A i

i DRESDgN II DPR-19 d

AmendmentNo,pt,9890 I

i 3.2 LIMITING CONDITION FOR OPERATION BASES (Cont'd.)

Venturis are provided in the main steamlines as a means of measuring steam flow and also limiting the loss of mass inventory from the vessel during a steamline break accident.

In addition to monitoring steam flow instrumentation is provided which causes a trip of Group 1 isolation valves.

The primary functidh of the instrumentation is to detect a break in the main steamline outside the drywell, thus only Group 1 valves are closed.

For the worst case accident, main steamline break outside the drywell, this trip setting of 120% of rated steam flow in conjunction with the flow limiters and main steamline valve closure, limit the mass inventory loss such that fuel is not uncovered, fuel temperatures remain less than 1500 degrees F and release of radioactivity to the environs is well below 10 CFR 100 guidelines.

(Ref. Sections 14.2.3.9 and 14.2.3.10 SAR)

Temperature monitoring instrumentation is provided in the main steamline tunnel to detect leaks in this area.

Trips are provided to this instrumentation and when exceeded cause closure of Croup 1

^

isolation valves.

Its setting of 200*F is low enough to detect leaks of the order of 5 to 10 spm; thus, it is capable of covering the entire spectrum of breaks.

For large breaks, it is a back-up to high steam flow instrumentation discussed above, and for small r-l breaks with the resultant small release of radioactivity, gives isolation before the guidelines of 10 CFR are exceeded.

High radiation monitors in the main steamline tunnel have been provided to detect gross fuel failure.

This instrumentation causes closure of Group 1 valves, the only valves required to close for this accident.

With the established setting of 3 times full power background for all conditions except for greater than 20% power with hydrogen being injected during which the Main Steamline trip setting is less than or equal to 3 times full power j

background with hydrogen addition, and main steamline isolation valve closure, fission product release is limited so that 10 CFR 100 guidelines are not exceeded for this accident.

(Ref. Section 14.2.1.7 SAR)

The performance of the process radiation monitoring system relative to detecting fuel leakage shall be evaluated during the first five years of operation.

evaluation will be reported to the NRC.

The conclusions of this Pressure instrumentation is provided which trips when main steam-line pressure drops below 850 psig. A trip of this instrumenta-tion results in closure of Group 1 isolation valves.

In the

" Refuel" and "Startup/ Hot Standby" mode this trip function is bypassed.

This function is provided to provide protection against

/

a pressure regulator malfunction which would cause'the_ control P.

B 3/4.2-30 3960a 3843A i

DRESDgN II DP Amendment No pt, $8 g19 t

i 3.2 LIMITINC CONDITION FOR OPERATION BASES (Cont'd.)

and/or bypass valves to open.

With the trip set at 850 psis, inventory loss is limited so that fuel is not uncovered and peak 4

clad temperatures are much less than 1500 degrees F; thus, there are no fission products available for releaso other than those in the reactor water.

(Ref. Section 11.2.3 SAR) t Two sensors on the isolation condenser supply and return lines are provided to detect the failure of isolation condenser line and i

actuate isolation action. The sensors on the supply and return sides are arranged in a 1 out of 2 logic and, to meet tie single failure criteria, all sensors and instrumentation are required to be operable.

The trip settings of 20 psig and 32 inches of water and valve closure time are such as to prevent uncovering the core or exceeding site limits.

The sensors will actuate due to high flow in either direction.

The HPCI high flow and temperature instrumentation are provided to detect a break in the HPCI piping. Tripping of this instrumenta-tion results in actuation of HPCI isolation valves, i.e., Group 4 valves.

Tripping logic for this function is the same as that for th,e isolation condenser and thus all sensors are required to be

~

operable to meet the single failure of design flow and valve closure time are such that core uncovery is prevented and fission r-product release is within limits.

The instrumentation which initiates ECCS action is arranged in a dual bus system.

As for other vital instrumentation arranged in this fashion the Specification preserves the effectiveness of the system even during periods when maintenance or testing is being performed.

The control rod block functions are provided to prevent excessive control rod withdrawal so that NCPR does not go below the MCPR fuel cladding integrity safety limit. JThe trip logic for this function is 1 out of n, e.g., any trip on one of the six APRN's, 8 IRM's, or 4 SRM's will result in a rod block.

The minimum instrument channel requirements assure sufficient instrumentation to assure the single failure criteria are met.

The minimum instrument channel requirements for the RRN may'be reduced by one j

for a short period of time to allow for maintenance, testing or calibration.

This time period is only approximately 3% of the operating time in a month and does not significantly increase the risk of preventing an inadvertent control rod withdrawal.

The APRM rod block function is flow biased and prevents a significant reduction in NCPR, especially during operation at j

1 B 3/4.2-31 3960s 3843A 6

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DRESDEN II DPR-19 AmendmontNo.pi,9690 8

I 3.2 LIMITING CONDITION FOR OPERATION BASES (Cont'd.)

reduced flow. The APRM provides gross core protection, i.e.,

limits the gross withdrawal of control rods in the normal withdrawal sequence.

In the refuel and startup/ hot standby modes, the APRN rod block function is set at 12% of rated power. This control rod block i

Provides the same type of protection in the Refuel and Startup/ Hot Standby mode as the APRM flow-blased rod block does in the Run mode, i.e., prevents control rod withdrawal before a scram is reached.

The RBM rod block function provides local protection of the core, i.e., the prevention of transition boiling in a local region of the core for a single rod withdrawal error from a limiting control rod pattern. The trip point is flow biased. The worst-case single control rod withdrawal error is analyzed for each' reload to assure that, with the specific trip settings, rod withdrawal is blocked before the MCPR reaches the NCPR fuel cladding integrity safety limit.

)

Be' low 30% power, the worst-case withdrawal of a single control rod _

i

{

without rod block action will not violate the NCPR fuel cladding integrity safety limit. Thus, the RBM rod block function is not required below this power level.

i The IRN block function provides local as well as gross core i

protection. The scaling arrangement is such that the trip setting is less than a factor of 10 above the indicated level. Analysis of the worst-case accident results in rod block action before MCPR approaches the MCPR fuel cladding integrity safety limit.

A downscale indication on an APRM or IRN is an indication the instrument has failed or is not sensitive enough. In either case the instrument will not respond to changes in control rod motion and the control rod motion is prevented. The downscale trips are set at 5/125 of full scale.

9 l

The rod block which occurs when the IRN detectors are not fully 1

inserted in the core for the refuel and startup/ hot standby position of the mode switch has been provided to assure that these detectors are in the core during reactor startup. This, therefore, assures that these instruments are in proper position j

to provide protection during reactor startup. The IRM's primarily provide protection assinst local reactivity effects in the source i

and intermediate neutron range.

j

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3960s 3843A

DRESDgN II DPR-19 Amendment No. h, N 9 0 4

f I

3.2 LIMITINC CONDITION FOR OPERATION BASES (Cont'd.)

t' For effective emergency core cooling for small pipe breaks, the J

HPCI system must function since reactor pressure does not decrease rapidly enough to allow either core spray or LPCI to operate in time.

l The automatic pressure relief function is provided as a backup to the HPCI in the event the HPCI does not oper' ate.

The arrangement of the tripping contacts is such as to provide this function when necessary and minimize spurious operation.

The trip sett11gs given in the specification are adequate to assure the above criteria are met (Ref. SAR Section 6.2.6.3).

The specification preserves the effectiveness of the system during periods of maintenance, testing or calibration and also minimizes the risk of inadverten.t operation; i.e., only one instrument channel out of service.

Two radiation monitors are provided on the refueling floor which initiate isolation of the reactor building and operation of the i

standby gas treatment systems. The trip logic is one out of two.

Trip settings of less than or equal to 100 mR/hr for the monitors on the refue?.ing floor are based upon initiating normal ventilation isolation and standby gas treatment system operation so that none of the activity released during the refueling accident leaves the reactor building via the normal ventilation r-stack but that all the activity is processed by the standby gas treatment system.

The instrumentation which is provided to monitor the post accident condition is listed in Table 3.2.6.

The instrumentation listed and the limiting conditions for operation on these systems ensure adequate monitoring of the containment following a loss-of-coolant accident.

Information from this instrumentation will provide the operator with a detailed knowledge of the conditions rosulting from the accident.

Based on this information he can make logical decisions regarding post accident recovery.

The specifications allow for post accident instrumentation to be out of service for a period of 30 days. This period is based on the fact that several diverse instruments are available for guiding the operator should an accident occur, on the low probability of an instrument being out of service and an accident occurring in the 30-day period, and on engineering judgemont.

l The radioactive 11guld and gaseous effluent instrumentation is provided to monitor the release of radioactive materials in 11guld and gaseous effluents during releases. The alarm setpoints for the instruments are provided to ensure that the alarms will occur prior to exceeding the limits of 10 CFR 20.

h B 3/4.2-33 3960s 3843A I

J

DRESDEN II DPR-19 Amendment No. f, @ 9 0 a

8 4.2 SURVEILLANCE REOUIREMENT BASES The instrumentation listed in Table 4.2.1 will be functionally tested and calibrated at regularly scheduled intervals. Although this instrumentation is not generally considered to be as important to plant safety as the Reactor Protection System, the same design reliability goal of 0.99999 is generally a*pplied for all applications of (1 out of 2) I (2) logic. Therefore, on-off sensors are tested once/3 months, and bi-stable trips associated with analog sensors and amplifiers are tested once/ week.

i Those instruments which, when tripped, result in a rod block have their contacts arranged in a 1 out of n logic, and all are capable of being bypassed. For such a tripping arrangement with bypass capability provided, there is an optimum test interval that should be maintained in order to maximize the reliability of a given channel (see note 7).

This takes account of the fact that testing degrades reliability and the optimum interval between tests is approximately given by:

i = (2t/c)1/2 Where:

i = optimum interval between tests r-t = the time the trip contacts are disabled from performing their function while the test is in progrecs e = The expected failure rate of the relays To test the trip relays requires that the channel be bypassed, the test made, and the system returned to its initial state. It is assumed this task requires an estimated 30 minutes to complete in a thorough and workmanlike manner and that the relays have a failure rate of 10-6 failures per hour. Using this data and the above operation, the optimum test interval is:

1 = [2(0.5)/10-6)1/2, 1 g 103 hours0.00119 days <br />0.0286 hours <br />1.703042e-4 weeks <br />3.91915e-5 months <br />

= approximately 40 days For additional marzin a test interval of once per month will be used initially.

Note:

(7) UCRL-50451, Improving Availability and Readiness of Field Equipment

~

Through Periodic Inspection, Benjamin Epstein, Albert Shiff, July 16, 1968, page 10, Equation-(24), Lawrence Radiation Laboratory.. -

B 3/4.2-34 l

3960s 3843A i

DRESDEN II DpR-19 Amendment No. pd, p$ 90 4.2 SURVEILLANCE REOUIREMENT BASES (Cont'd.)

The sensors and electronic apparatus have not been included here as these are analog devices with readouts in the control room and the sensors and electronic apparatus can be checked by comparison with other like instruments.

The checks which are made on a daily basis are adequate to assure operability of the sensors and electronic apparatus, and the test interval given above provides for optimum testing of the relay circuits.

The above calculated test interval optimizes each individual channel, considering it to be independent of all others.

As an example, assume that there are two channels with an individual technician assigned to each. Each technician tests his channel at the optimum frequency, but the two technicians are not allowed to communicate so that one can advise the other that his channel is under test.

Under these conditions, it is possible for both channels to be under test simultaneously. Now, assume that the technicians are required to communicate and that two channels are never tested at the same time. Forbidding simultaneous testing improves the availability of the system over that which would be achieved by testing each channel independently. These one out of

~

n trip systems will be tested one at a time in order to take advantage of this inherent improvement in availability.

r-Optimizing each channel independently may not truly optimize the system considering the overall rules of system operation.

However, true system optimization is a complex problem. The optimums are broad, not sharp, and optimizing the individual channels is generally adequate for the system.

4 The formula given above minimizes the unavailability of a single channel which must be bypassed during testing. The minimization of the unavailability is illustrated by curve No. 1 of Figure 4.2.2 which assumes that a channel has a failure rate of 0.1 x 10-6/ hour and that 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> is required to test it.

The unavailability is a minimum at a test interval 1, of 3.16 x 103 hours0.00119 days <br />0.0286 hours <br />1.703042e-4 weeks <br />3.91915e-5 months <br />.

If two similar channels are used in a 1 out of 2 configuration, the test interval for minimum unavailability changes as a function of the rules for testing. The simplest case is to test each one independent of the other. In this case, there is assumed to be a finite probability that both may be bypassed at one time.

This case is shown by Curve No. 2.

Note that the unavailability is lower as expected for a redundant system and the minimum occurs at the same test interval. Thus, if the two channels are tested g

independently, the equation above yields the test inteeval for minimum unavailability.

B 3/4.2-35 3960s 3843A w

n-y

DRESDEN II DPR-19 i

Amendment No. p, P 9 0 I

4.2 SURVEILLANCE REOUIRe nT BASES (Cont'd.)

A more usual case is that the testing is not done independently.

If both channels are bypassed and tested at the same tLae, the result is shown in Curve No. 3.

Note that the minimum occurs at about 40,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />, much longer than for cases 1 and 2.

Also, the minimum is not nearly as low as case 2 which indicates that this i

method of testing does not take full advantage of the redundant channel. Bypassing both channels for simultaneous testing should be avoided.

The most likely case would be to stipulate that one channel be bypassed, tested and restored, and then immediately following, the j

second channel be bypassed, tested and restored. This is shown by Curve No. 4.

Note that there is no true minimua. The curve does have a definite knee and very little reduction in system unavailability is achieved by testing at a shorter interval than

{

computed by the equation for a single channel, j

The best test procedure of all those examined is 'o perfectly sta'gger the tests. That is, if the test interval is four motths, test one or the other channel every two months. This is shown in Curve No. 5.

The difference between Cases 4 and 5 is negligible.

There may be other arguments, however, that more strongly support the perfectly staggered tests, including reductions in human eeror.

The conclusions to be drawn are these:

1.

A 1 out of a system may be treated the same as a single channel in terms of choosing a test interval; and 2.

More than one channel should not be bypassed for testin^g at any one time.

The analog trip system consists of an analog sensor (transmitter) and a master / slave trip unit setup which ultimately drives a trip relay. The frequency of calibration and functional testing for instrument loops of the analog system, including reactor low water j

l 1evel, has been established in Licensing Topical Report i

NEDO-21617-A (December, 1978).

For instruments 2(3)-2389A, B, C, D, the one-of-two-taken-twice logic exists, and NEDO-21617-A states that each trip unit be subjected to a calibration / test frequency (staggered one channel out of four per week) of one month. An adequate calibration /

surveillance test, interval for the transmitter is once per-i operating cycle.

P B 3/4.2-36 I

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DRESDEN II DPR-19' Amendment No M. %. f 90 4.2 SURVEILLANCE REQUIREMENT BASES (Cont'd.)

For instruments 2(3)-263-73A, 738 and 2(3)-2352, 2353, the logie downstremn of the output relay contacts exhibits a one-out-of-two logic and, by utilizing the Availability Criteria identified in NEDO-21617-A, each of these trip units should also be subjected to a calibration / test frequency (staggered one division out of two per two weeks) of one month.

An adequate calibration / surveillance test interval for the transmitter is once per operating cycle.

The radiation monitors in the ventilation duct and on the refueling floor which initiate building isolation and standby gas treatment operation are arranged in two 1 out of 2 logic systems.

The bases given above for the rod blocks applies here also and were used to arrive at the functional testing frequency.

Based on experience at Dresden Unit I with instruments of similar design, a testing interval of once every three months has been found to be adequate.

The automatic pressure relief instrumentation can be considered to -

be a 1 out of 2 logic system and the discussion above applies also.r The instrumentation which is required for the post accident condition will be tested aad calibrated at regularly scheduled intervals.

The basis for the calibration and testing of this instrumentation is the same as was discussed above for Protective Instrumentation in Table 4.2.4.

4*

B 3/4.2-37

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Q DRESDEN II DPR-19 A

a..nt no. g, p, p on e

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CuftvE 1

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Cunyt 3 1 i.-

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Figure 4.2.2 I

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TEST INTERVAL VS. SYSTEN UNAVAILABILITY '

B 3/4.2-38 3960s 3843A

4

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UNITED STATES NUCLEAR REGULATORY COMMISSION o

3 E

WASHINGTON, D. C. 20555

%.....)

COMMONWEALTH EDISON COMPANY DOCKET NO. 50-249 DRESDEN NUCLEAR POWER STATION, UNIT NO. 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 83 License No. OPR-25 1.

The Nuclear Regulatory Comission (the Comission) has found that:

A.

The application for amendment by the Comonwealth Edison Company (thelicensee)datedOctober 10, 1984 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.

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

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.

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

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

W e.

. 2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 3.B of Facility Operating License No. DPR-25 is hereby amended to read as follows:

B.

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 83, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of its issuance.

FOR T E NUCLEAR REGULAT RY OMMISSION

-[_us John A. Zwolinski, Chief Oper ting Reactors Branch #5 r-Divis'on of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: June 24, 1985.

r g

e W

I

ATTACHMENT TO LICENSE AMENDMENT NO. 83 FACILITY OPERATING LICENSE DPR-25 DOCKET NO. 50-249 Revise Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages. The revised pages are identified by the ceptioned amendment number and contain marginal lines indicating the area of change.

REMOVE INSERT 11 11 vii vii viii viii 3/4.2-3 3/4.2-3 3/4.2-17 3/4.2-17 3/4.2-18 3/4.2-18 3/4.2-19 3/4.2-19*

3/4.2-20 3/4.2-20*

3/4.2-21 3/4.2-21*

3/4.2-22 3/4.2-22*

3/4.2-23 3/4.2-23*

B 3/4.2-24 3/4.2-24*

B 3/4.2-25 3/4.2-25*

8 3/4.2-26 3/4.2-26 B 3/4.2-27 3/4.2-27 8 3/4.2-28 8 3/4.2-28*

P 3/4.2-29 B 3/4.2-29*

B 3/a.2-30 B 3/4.2-30*

B 3/4.2-31 B 3/4.2-31*

B 3/4.2-32 B 3/4.2-32*

B 3/4.2-33 B 3/4.2-33 B 3/4.2-34*

B 3/4.2-35*

B 3/4.2-36 8 3/4.2-37*

• Pagination change only

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DRESDEN III DPR-25 Amendment No. )$, )1 82 Table of Contents

Page, a

1.0 Definitions 1.0- 1 1.1 Safety Limits - Fuel Cladding Integrity 1/2.1-1 Safety Limit Bases B 1/2.1-6 1.2 Safety Limits - Reactor Coolant System 1/2.2-1 Safety Limit Bases B 1/2.2-2 2.1 Limiting Safety System Settings - Fuel Cladding Integrity 1/2.1-1 Limiting Safety System Settings Bases B 1/2.1-10 2.2 Limiting Safety System Settings - Reactor Coolant System 1/2.2-1 Limiting Safety System Settings Bases B 1/2.2-4 3.0 LIMITING CONDITION FOR OPERATION 3.0- 1 Limiting Condition for Operation Bases B 3.0- 3 3.1 Reactor Protection System 3/4.1-1 Limiting Conditions for Operation Bases (3.1) 3/4.1-9 surveillance Requirement Bases (4.1)

B 3/4.1-15 3.2 Protective Instrumentation 3/4.2-1 3.2.A Primary Containment Isolation Functions 3/4.2-1 3.2.B Cor's and Containment Cooling Systems - Initiation r-and Control 3/4.2-1 3.2.C Control Rod Block Actuation 3/4.2-2 3.2.D Refueling Floor Radiation Monitors 3/4.2-2 3.2.E Post Accident Instrumentation 3/4.2-3 3.2.F Radioactive Liquid Effluent Instrumentation 3/4.2-4 3.2.G Radioactive caseous Effluent Instrumentation 3/4.2-5 Limiting conditions for Operation Bases (3.2)

B 3/4.2-28 Surveillance Requirement Bases (4.2)

B 3/4.2-34 3.3 Reactivity Control 3/4.3-1 3.3.A Reactivity Limitations 3/4.3-1 3.3.B Control Rods 3/4.3-4 3.3.C Scram Insertion Times 3/4.3-10 3.3.D Control Rod Accumulators 3/4.3-11 3.3.E Reactivity Anomalies 3/4.3-12 3.3.G Economic Generation Control System 3/4.3-13 Limiting Conditions for Operation Bases (3.3)

B 3/4.3-28 Surveillance Requirement Bases (4.3)

B 3/4.3-34 3.4 Standby Liquid Control System 3/4.4-1 3.4.A Normal Operation 3/4.4-1 3.4.B operation with Inoperable Components 3/4.4-2 3.4.C Liquid Poison Tank 3/4.4-3 3.4.D Reactor Shutdown Requirement 3/4.4-3 Limiting Conditions for Operation Bases (3.4)

B 3/4 4-6 Surveillance Requirement Bases (4.4)

_ B 3/4.4-7 3.5 Core and Containment Cooling Systems 3/4.5-1 3.5.A Core Spray and LPCI Subsystems 3/4.5-1 3.5.B Containment Cooling Subsystem 3/4.5-5 11 3958a 0009A

DRESDEN III DPR-25 Amendment No. 7/, 7/,j 7/ $ $

.P.Agt Table 3.1.1 Reactor Protection System (Scram) 3/4.1 - 5 Instrumentation Requirements Table 4.1.1 Scram Instrumentation Functional Tests 3/4.1 - 8 Table 4.1.2 Scram Instrumentation Calibration 3/4.1 -10 Table 3.2.1 Instrumentation that Initiates Primary Containment Isolation Functions 3/4.2 - 8 Table 3.2.2 Instrumentation that Initiates or controls the core and containment Cooling System 3/4.2 -10 Table 3.2.3 Instrumentation that Initiates Rod Block 3/4.2 -12 Table 3.2.4 Radioactive Liquid Effluent Monitoring Instemmentation 3/4.2 -14 Table 3.2.5 Radioactive Gaseous Effluent Monitoring Instrumentation 3/4.2 -15 Table 3.2.6 Post Accident Monitoring Instrumentation Requirements 3/4.2 -17 Table 4.2.1 Minimum Test and Calibration Frequency for Core and Containment Cooling Systems Instrumentation, Rod Blocks, and Isolations 3/4.2 -19 Table 4.2.2

, Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements 3/4.2 Table 4.2.3 Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements 3/4.2 -24 Table 4.2.4 Post Accident Monitoring Instrumentation Surveillance Requirements 3/4.2 -26 l

Table 4.6.2 Neutron Flux and Sample Withdrawal B 3/4.6 -26 Table 3.7.1 Primary Containment Isolation 3/4.7 -31 Table 4.8.1 Radioactive Gaseous Waste Sampling and Analysis Program 3/4.8 -22 Table 4.8.2 Maximum Permissible Concentration of Dissolved or Entrained Noble Gases Released from the site to Unrestricted Areas in Liquid Waste 3/4.8 -24 Table 4.8.3 Radioactive Liquid Waste Sampling and Analysis Program 3/4.8 -25 Table 4.8.4 Radiological Environmental Monitoring Program 3/4.8 -27 Table 4.8.5 Reporting Levels for Radioactivity concentrations in Environmental Samples 3/4.8 -28 Table 4.8.6 Practical Lower Limits of Detection (LLD) for Standard Radiological Environmental Monitoring Program 3/4.8 -29 Table 4.11-1 Survalliance Requirements for High Energy Piping Outside containment 3/4.11-3 Table 3.12-1 Fire Detection Instruments 3 3/4.12-17 Table 3.12-2 Sprinkler Systems JB 3/4.12-18 Table 3.12-3 CO2 Systems B 3/4.12-19 7

Table 3.12-4 Fire Hose Stations 8 3/4.12-20 & 21 Table 6.1.1 Minimum Shift Manning Chart 6

-4 Table 6.6.1 Special Reports 6

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DERSDEN III DPR-25 I

Amendment No. )(, 74, f% I f

]

List of Flaures EA&t i.

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Figure 2.1-3 APRM Blas Scram Relationship to Normal i

Operating conditions B 1/2.1-17 Figure 4.1.1 Graphical Aid in the Selection of an Adequate Interval Between Tests B 3/4.1-18 l

Figure 4.2.2 Test Interval vs. System Unavailability B 3/4.2-37 Figure 3.4.1 Standby Liquid Control Solution Requirements 3/4.4-4 Fi.'

e 3.4.2 Sodium Pentaborate Solution Temperature Requirements 3/4.4-5 Figure 3.5-1 Bundle Average Exposure (MWD /NT)

(Sheet 1 of 5) 3/4.5-17 Figure 3.5-1 Planar Average Exposure (MWD /T)

(Sheet 2 of 5) 3/4.5-18 l

Figure 3.5-1 Planar Average Exposure (MWD /T)

(Sheet 3 of 5) 3/4.5-19 Figure 3.5-1 Maximum Average Planar Linear Heat r-Generation Rate (MAPLHGR)

(Sheet 4 of 5) 3/4.5-20 Figure 3.5-1 Planar Average Exposure (MWD /ST)

(Sheet 5 of 5) 3/4.5-21 Figure 3.5-2 Core Flow %

3/4.5-25 & 26 Figure 3.6.1 Minimum Temperature Requirements per Appendix G of 10 CFR 50 3/4.6-20 Figure 4.6.1 Minimum Reactor Pressurization Temperature B 3/4.6-25 Figure 4.6.2 Chloride Stress Corrosion Test Results at 500*F B 3/4.6-27 Figure 4.8-1 Owner Controlled / Unrestricted Area Boundary B 3/4.8-38 Figure 4.8-2 Detail of Central Complex B 3/4.8-39 Figure 6.1-1 Corporate and Station Organization 6-3 vili 3958a 0009A

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DRESDEN III _,

!)PR 5 }

Amendment,No. 7[ /

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3.2 LIMITING CONDITION FOR OPERATION 4.2 SURVEILLANCE REQUIRENENTS J

(CONT'D)

(COJT'D) in the fuel storage pool treatment system initiatien and during refueling or',

shall be performed at least c

' eash openating cycle.

fuel movement operations.

2.

One of the two re-fueling floor radiation monitors may be inoper-

,i, l

able for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. If

/

I the inoperable monitor

]

is not restored to j

I service in this time.

the reactor building s

ventilation system I

shall be isolated and j

l the standby gas treet-l ment operated until

'f.

j l

repairs are complete.

I s

s

\\

3.

The trip setting for the refueling floor j.

y radiation monitors j

shall be set at a less than or equal to 100mr/hr.

l 4.

Upon loss of both re-fueling floor radiation monitors while in use, s

the reactor building-ventilation system shall be isolated and the standby gas treat-ment operated.

E.

Post Accident E.

Po/h Accident Instrumentation Instiumentation l

The lisilting conditions, <

Post accident instrumen-4 l

fcr operation for the tation shall be function-in trumentation, which ally tested and calibrated is read out in the con-es indicated in Table 4.2.4.

trol room, required for post accident monitoring are given in Table 3.2.6.

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-7 DRRSDRN III DPR-25 Amendment No. $$

Table 3.2.6 Post Accident Monitorina Instrumentation Requirements Minimum Instrument Number of Readout Operable Location Number Instrument channels (1)

Parameter Unit 3 Provided

,Ranae 1

Reactor Pressure 903-5 1

0-1500 psig 2

0-1200 psig 1

Reactor Water Level 903-3 2

-340 to +60 inches 1

Torus Water Temperature 903-4 2

0-200*F 2 (3)

Torus Water Level 903-3 1

-25 to +25 inches Indicator 903-3 1

-7 to +3 inches (narrow range) 903-2 2

0-30 ft (wide range) l

~

Torus Water Local 1

18 inch rasse Sight Glass (narrow range) 1 (4)

Torus Pressure 903-5 1

-2.45-5 psig 2

Drywell Pressure 903-5 1

0-5 psig 903-3 1

0-75 psig 903-3 2

0-250 psig 2

Drywell Temperature 903-21 6

0-600*F 2

Neutron Monitoring 903-5 4

0.1-106 CPS 1 (4)

Torus to Drywell 903-3 2

0-3 psid Differential Pressure 1

Drywell Radiation Monitor 903-55,56 2

1 to 108 R/hr 2/ valve (2)

Main Steam RV Position, 903-21 1 per valve N/A Acoustic Monitor Main Steam RV Position, 903-21 1 per valve 0-600*F Temperature Monitor l

2/ valve (2)

Main Steam SV Position, 903-21 1 per valve N/A lt Acoustic Monitor g

Main Steam SV Position, 903-21 1 per valve

~0-600*F 4

Temperature Monitor

~

r

/

1 (5)

Drywell Hydrogen 903-55 2 10%

/

Concentration 903-56 Notes:

(See Next Page) 3/4.2-17 3840a 3845A

DRESDEN III DFR-25 Amendment No. 83 Table 3.2.6 Notes 1.

From and after the date that a parameter is reduced to the minimum number of channels, continued operation is not permissible beyond thirty (30) days unless such instrumentation is sooner made operable. In the event that all indications of a parameter is disabled and such indication cannot be restored in six (6) hours, an orderly shutdown shall be initiated and the reactor shall be in a cold shutdown condition in twenty-four (24) hours. See notes 2, 3, 4 and 5 for exceptions to this requirement.

2.

If the number of position indicators is reduced to one indication on one or more valves, continued operation is permissible; however, if the reactor is in a cold shutdown condition for longer than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, it may not be started up until all position indication is restored. In the event that all position indication is lost on one or more valves and such indication cannot be restored in thirty (30) days, an orderly shutdown shall be initiated, and the reactor shall be depressurized to less than 90 psig in twenty-four (24) hours.

3.

From and after the date that this parameter is reduced to either one narrow-range indication or one wide-range indication, continued reactor operation is not permissible beyond thirty (30) days unless such instrument is sooner made operable. In the event that either all narrow-range indication or all wide-range indication is disabled, continued reactor operation is not permissible beyond seven (7) days unless such instruments are sooner made operable. In the event that all indication for this parameter is disabled, and such indication cannot be restored in six (6) hours, an orderly shutdown shall be intitiated and the reactor shall be in a cold shutdown condition in twenty-four (24) hours.

4.

From and after the date that one of these parameters becomes inoperable.

continued operation is not permissible beyond thirty (30) days unless such instrumentation is sooner made operable. In the event that all indication of these parameters is disabled and such indication cannot be restored in six (6) hours, an orderly shutdown shall be intitiated and the reactor shall be in cold shutdown in twenty-four (24) hours.

5.

From and after the date that one of the drywell hydrogen monitors becomes inoperable, continued reactor operation is permissible, a.

If both drywell hydrogen monitors are inoperable, continued reactor operation is permissible for up to 30 days provided that during this time the HRSS hydrogen monitoring capability for the drywell is operable.

b.

If all drywell hydrogen monitoring capability is lost, tontinued reactor operation is permissible for up to 7 days.

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3/4.2-18 3840a 3845A

Tabla 4.2.1 DRESDEN III DPR-25 MINImm TEST AND CALIERATION FREQUENCY FOR CORE AND Mendment No. )6 /) g (X)NTAINENT CX)0 LING SYSTEMS INSTRtPENTATION, R00 BLOCKS, AND 150LATIONS I

Instrunent Channel Instrunent Instrument Functional Test Calibration Check t

ECCS Instrumentation 1.

Reactor Low-Low Water Level (1)

Once/3 Months Once/ Day 2.

Drywell High Pressure (1)

Once/3 Months None 3.

Reactor Low Pressure (1)

Once/3 Months hone 4.

Contaltunent Spray Interlock a.

2/3 Core Height (1)

None b.

Contaltunent High Pressure (1)

Once/3 Months None 5.

Low Pressure Core Cooling Pug (1)

Once/3 Months None Discharge 6.

Undervoltage Emergency Bus Refueling Outage Refuel Outage once/3 months 7.

Sustained High Reactor Pressure (1)

Once/3 Months None 8.

Degraded Voltage Emergency Bus Refueling Outage (10)

Refuel Outage Monthly Rod 8tocks 1.

APRM Downscale (1) (3)

Once/3 Months None 2.

APM Flow variable (1) (3)

Refuel Outage None 3.

APRM Upscale (Startup/ Hot Staney)

(2) (3)

(2) (3)

(2) 4.

I M Upscale (2) (3)

(2) (3)

(2) 5.

IM Downscale (2) (3)

(2) (3)

(2) 6.

IM Detector Not Fully Inserted (2)

N/A None in the Core 7.

ROM Upscale (1) (3)

Refuel Outage NorE 8.

RBM Downscale (1) (3)

Once/3 Months None 9.

SRM Upscale (2) (3)

(2) (3)

(2)

10. SM Detector Not in Startup Position (2) (3)

(2) (3)

(2)

11. Scram Instrument Volume Level High Once/3 Months (9)

None None Contairunent Monitorina 1.

Pressure a.

Minus 5 in. Hg to plus 5 psig None Once/3 Months Once/ Day Indicator b.

O to 75 psig Indicator None once/3 Months None 2.

Temperature None Refuel Outage Once/ Day 3.

Drywell-Torus Differential None Once/6 Months (Two None Pressure (5) (6)

Channels Operable)

(0.-3 psid)

Once/ Month (One Channel Operable) 4.

Torus Water Level (5) (6)

None once/6 Months a.

Plus or minus 25 in. Wide Range Indicator b.

18 in. Sight Glass safetv/ Relief Valve Monitoring 1.

Safety / Relief Valve (7)

None Once Per Position Indicator 31 Days (Acoustic Monitor) (8) 2.

Safety / Relief Valve Position None Once every

'Once Per Indicator (Temperature 18 months.

31 Days Monitor (8) 3.

Safety valve Position Indicator (7)

Nine P Once Per (Acoustic Monitor) (8) 31 Days 4.

Safety Valve Position Indicator None once every Once Per (Temperature Monitor) (8) 18 months 31 Days (Table cont'd next page) l 3/4.2-19 3840a 3845A

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l Table 4.2.1 (Cont'd)

DRESDEN III DPR-25 MINIltfl TEST AND CALIBRATION FREQUENCY FOR (X)RE AND Amendment No.

4 CONTAlifENT COOLING SYSTEMS INSilRNENTATION, ROD BLOCKS, AND ISO j

d)

Instrument Channel Instrtsment InstrLanent Functional Test Calibration Check

}

Msin Steam Line Isolation 1.

Steam Tunnel High Temperature Refueling Outage RefuelOutage None

^

2.

Steam Line High Flow (1)

Once/3 Months Once/ Day 3.

Steam Line Low Pressure (1)

Once/3 Months None 4.

Steam Line High Radiation (1) (3)

Once/3 Months (4)

Once/ Day 1

Isolation Condenser Isolation 1.

Steam Line High Flow (1)

Once/3 Months None 2.

Condensate Line High Flow (1)

Once/3 Months None IFCI Isolation 1.

Steam Line High Flow (1) (11)

(11)

None i

2.

Steam Line Area High Temperature Refueling Outage Refuel Outage None 3.

Low Reactor Pressura (1)

None Reactor Buildine Vent Isolation and SBGTS Initiation 1.

Refueling Floor Radiation Monitors (1)

Once/3 Months Once/ Day E2ns.:

(For Table 4.2.1) 1.

Initially once per month until exposure hours (M as defined on Figure 5

i 4.1.1) is 2.0 x 10 ; thereafter, according to Figure 4.1.1 with an interval not less than one month nor more than three months. The i

compilation of instrument failure rate data may include data obtained 1

from other Boiling Water Reactors for which the same design instrument operates in an environment similar to that of Dresden Unit 3.

2.

Function test calibrations and instrument checks are not required when these instruments are not required to be operable or are tripped.

Functional tests shall be performed before each startup with a required frequency not to exceed once per week. Calibrations shall be performed during each startup or during controlled shutdowns with a required frequency not to exceed once per week. Instrument checks shall be performed at least once per week. Instrument checks shall be performed at least once per day during those periods when the I

instruments are required to be operable.

3.

This instrumentation is excepted from the functional test definition.

The functional test will consist of injecting a simulated electrical l

signal into the measurement channel. See Note 4.

4.

These instrument channels will be calibrated using simulated electrical signals once every three months. In addition, calibration including the sensor's will be performed during each refueling outage.

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DRESDEN III DPR-25 i

Amendment No. /$, % r. 3 4

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(For Table 4.2.1) (Cont'd.)

5.

A minimum of two channels is required.

6.

From and after the date that one of these parameters (...elther drywell-torus differential pressure or torus water level indication) is reduced to one indication, continued operation is not permissible beyond thirty days, unless such instrumentation is sooner made operable. In the event that all indications of these parameters

(...either drywell-torus differential pressure or torus water level) is disabled and such indication cannot be restored in six (6) hours, an orderly shutdown shall be initiated and the reactor shall be in a cold shutdown condition in twenty four hours.

7.

Functional tests will be conducted before startup at the end of each refueling outage or after maintenance is performed on a particular Safety / Relief Valve.

8.

If the number of position indicators is reduced to one indication on one or more valves, continued operation is permissible; however, if the reactor is in a shutdown conditionfor more than seventy-two hours,-

~

it may not be started up until all position indication is esstored.

s.

In the event that all position indication is lost on one or more valves and such indication cannot be restored in thirty days, an orderly shutdown shall be initiated, and the reactor shall be depressurized to less than 90 psig in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

9.

The Functional Test of the Scram Discharge Volume float switch shall include actuation of the switch using a water column.

10.

Functional test shall include verification of the second level undervoltage (degraded voltage) timer bypass and shall verify operation of the degraded voltage 5+ minute timer and inherent 7-second timer.

j 11.

Verification of time delay setting between 3 and 9 seconds shall be performed during each refueling outage.

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3/4.2-21 i

3840a 3845A I

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DRESDEN III DPR-25 Amendment No. %, 7/ g3 TABLE 4.2.2 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS Instrument Calibration Functional Source Instrument Check (1)(7)

(1)(7)(3)(4)

Test (1)(2)(7) Check (1)

Liquid Radwaste D

R Q (6)

(5)

Effluent Gross Activity Monitor Service Water D

R Q (6)

R Effluent Cross Activity Monitor Tank Level Indicating Device

s. A Waste Sample Tank D

R Q

N/A

b. B Waste Sample Tank D

R Q

N/A r-

c. C Waste Sample Tank D

R Q

N/A

d. A Floor Drain Sample D

R Q

N/A Tank

e. B Floor Drain Sample D

R Q

N/A Tank

f. Waste Surge Tank D

R Q

N/A 4

Notes:

(See Next Page) 4*

1 3/4.2-22 l

1 3840s 3845A

DRESDEN III DpR-25 d

Amendment No. f,

3 TABLE 4.2.2 (Notes) 1.

D = Once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> M = Once per 31 days Q = Once per 92 days R = Once per refueling outage S = Once per 6 months 2.

The Instrument Functional Test shall also demonstrate that control room alarm annunciation occurs, if any of the following conditions exist, where applicable.

a.

Instrument indicates levels above the alarm setpoint.

b.

Circuit Failure.

c.

Instrument indicates a downscale failure, d.

Instrument controls not set in OPERATE mode.

3.

Calibration shall include performance of a functional test.

4.

Calibration shall include performance of a source check.

5.

Source check shall consist of observing instrument response during a F

discharge.

6.

Functional test may be performed by using trip check and test circuitry associated with the monitor chassis.

7.

Function test calibrations and instrument checks are not required when these instruments are not required to be operable or are tripped.

Calibration shall be performed once per refueling outage and not more than once every 18 months. Instrument checks shall be performed at least once a day during those periods when the instruments are required to be operable.

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DRESDEN III DPR-25 Amendment No. %, p 6 3 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS Instrument Calibration Function Test Source Instrument Check (1)(6)

(1)(6)(3)

(1)(4)(2)(6)

Check (1)

SJAE Radiation D

R Q

R Activity Monitor Reactor Blds Vent D (4)

N/A N/A N/A Particulats and Iodine Sampler Reactor Blds Vent D

R Q

Q Rxhaust Duct Radiation Monitor Reactor Blds Vent D

R Q

M SPING Noble Gas Monitor Lo, Mid, High Range

~

Main Chimney Noble D

R Q

M r-Cas Activity Monitor Main Chimney SPING D

R Q

M Noble Gas Monitor Lo, Mid, High Range Main Chimney D (4)

N/A N/A N/A Particulate and Iodine Sampler Main Chimney Flow D

R Q

N/A Rate Monitor Main Chimney Sampler D

R Q (5)

N/A Flow Rate Monitor Reactor Blds Vent D

R Q

N/A Flow Rate Monitor Reactor Bldg sampler D

R Q (5)

N/A Flow Rate Monitor

..~

V.

Notes:

(See Next Page) 3/4.2-24 3840a 3845A

e DRESDEN III DPR-25 Amendment No. %. P 8 3 TABLE 4.2.3 (Notes) 1.

D = Once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> M = once per 31 days Q = Once per 92 days R = Once per refueling outage 2.

The Instrument Functional Test shall also demonstrate that control room alarm annunciation occurs, if any of the following conditions exist, where applicable.

1 Instrument indicates levels above the alarm setpoint.

a.

b.

Circuit Failure.

c.

Instrument indicates a downscale failure, d.

Instrument controls not set in OPERATE mode.

3.

Calibration shall include performance of a functional test.

4.

Instrument check to verify operability of sampler; that the sampler is in place and functioning properly.

5.

Function Test shall be performed on local switches providing 10W fl0W r

alarm.

6.

Function test calibrations and instrument checks are not required when these instruments are not required to be operable or are tripped.

Calibration shall be performed once per refueling outage end not more than once every 18 months. Instrument checks shall be performed at least once per day during those periods when the instruments are required to be operable.

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i DRESDEN III DPR-25 Amendment No.g3 y

Table 4.2.4 Post Accident Monitorina Instrumentation Surveillance Recuirements Minia'im Instrument Number of Readout Operable Location Instrument Channels Parameter Unit 3 Calibration Check 1

Reactor Pressure 903-5 Once Every once Per Day 6 Months 1

Reactor Water Level 903-3 Once Every once Per Day 6 Monthe 1

Torus Water Temperature 903-4 Once Every Once Per Day 12 Months Torus Water Level 903-3 Once Every Once Per Day Indicator (Narrow Range) 6 Months 2

(Sight Class)

N/A None (Wide Range) 903-2 Once Every once Per 31 12 Months Days 1

Torus Pressure 903-3,5 Once Every Once Per Day 3 Months 1

Torus to Drywell 903-3 Once Every Once Pir Day Differential Pressure 6 Months 2

Drywell Pressure (0-5 psig) 903-5 Once Every once Per Day 3 Months (0-75 psig) 903-3 Once Every Once Per 31 3 Months Days (0-250 psis) 903-3 Once Every once Per 31 Refuel Days 2

Drywell Temperature 903-21 Once Every Once Per Day Refuel 2

Neutron Monitoring 903-5 Once Every once Per Day 3 Months 1

Drywell Radiation Monitor 903-55,56 Once Every Once Per 31 Refuel (2)

Days Main steam IV Position, 903-21 Once Every once Per 31 2/ Valve Temperature Monitor Refuel Days Main Steam IV Position, (1)

Acoustic Monitor Main Steam EY Position, 903-21 Once Every once Per 31 Temperature. Monitor Refuel

, - Days 2/ Valve Main steam EY Position, Acoustic Monitor (1) 5~

Once Per 31 Days 1

Drywell Hydrogen 903-55 Once Every Once Per 31 Concentration 903-56 3 Months Days Notes:

(See Next Page) 3/4.2-26 3840a 3845A

Drosd:n III DPR-25 Amendment No. 83 Table 4.2.4 (Notes)

Notes 1.

Calibration of Acoustic Monitors shall consist of verifying the instrument threshold levels, and will be performed monthly."

Functional tests will be conducted before startup at the end of each refueling outage or after maintenance is performed on a particular safety or relief valve.

2.

Calibration shall consist of an electronic calibration of the channel, not including the detector, for range decades above 10 R/hr; and a one-point calibration check of the detector below 10 R/hr with an installed or portable gamma source.

r-s-

3/4.2-27 3840a 3845A

DRESDEN III DpR-25 AmendmentNo.7/,7/gg 3.2 LIMITING CONDITION FOR OPERATION BASES In addition to reactor protection instrumentation which initiates a reactor scram, protective instrumentation has been provided which initiates action to mitigate the consequences of accidents which are beyond the operator's ability to control, or terminates operator errors before they result in serious conseque'nces. This set of specifications provides the limiting conditions of operation for the primary system isolation function, initiation of the emergency core cooling system, control rod block and standby gas treatment systems. The objectives of the specifications are

1) to assure the effectiveness of the protective instrumentation when required by preserving its capability to tolerate a single failure of any component of such systems even during periods when portions of such systems are out of service for maintenance, and

2) to prescribe the trip settings required to assure adequate performance. When necessary, one channel may be made inoperable for brief intervals to conduct required functional tests and

, calibrations.

Some of the settings on the instrumentation that initiates or control core and containment cooling have tolerances explicitly stated where the high and low values are both critical and may have a substantial effect on safety. It should be noted that the r setpoints of other instrumentation, where only the high or low end of the setting has a direct bearing on safety, are chosen at a level away from the normal operating range to prevent inadvertent actuation of the safety system involved and exposure to abnormal situations.

Isolation valves are installed in those lines that penetrate the primary containment and must be isolated during a loss-of-coolant accident so that the radiation dose limits are not exceeded during an accident condition. Actuation of these valves is initiated by protective instrumentation which serves the condition for which isolation is required (this instrumentation is shown in Table 3.2.1).

Such instrumentation must be available whenever primary containment integrity is required. The objective is to isolate the primary containment so that the guidelines of 10 CFR 100 are not exceeded during an accident.

The instrumentation which initiates primary system isolation is connected in a dual bus arrangement. Thus the discussion given in the bases for specification 3.1 is applicable here.

The low-reactor level instrumentation is set to trip at greater than 8 inches on the level instrument (top of active fuel is defined to be 360 inches above vessel sero) and after pilowing for the full power pressure drop across the steam dryer the l'ow level trip is at 504 inches above vessel zero, or 144 incheg_above the B 3/4.2-28 3840a 3845A

DRESDEN III' DpR-25

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Amendment No. }4, [ gg 4

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l 3.2 LIMITING CONDITION'FOR OPERATION BASES (Cont'd.)

I top of active fuel. Retrofit 8 I 8 fuel has an active fuel length i

1.24 inches longer than earlier fuel designs. However, present trip setpoints were used in the LOCA analyses. This trip initiates closure of Group 2 and 3 primary containment

• isolation valves but does not trip the recirculation pumps (reference SAR gection 7.7.2).

For a trip setting of 504 inches above vessel zero (144 inches above top of active fuel) and a 60-second valve closure time, the valves will be closed before perforation of the cladding occurs even for the maximum break; the setting is therefore adequate.

The low low reactor level instrumentation is set to trip when reactor water level is 444 inches above vessel sero (with top of active fuel defined as 360 inches above vessel zero - 59 inches is 84 inches above the top of active fuel). This trip initiates closure of Group I primary containment isolation valves (reference SAR Section 7.7.2.2) and also activates the ECC subsystems, starts the emergency diesel generator, and trips the recirculation pumps. This trip setting level was chosen to be high enough to prevent spurious operation but low enough to initiate ICCS operation and primary system isolation so that no melting of the r-fuel cladding will occur and so that post accident cooling can be accomplished and the guidelines of 10 CFR 100 will not be exceeded. For the complete circumferential break of a 28-inch recirculation line and with the trip setting given above, ECCS initiation and primary isolation are initiated and in time to meet the above criteria. The instrumentation also covers the full spectrum of breaks and meets the above criteria.

The high-drywell pressure instrumentation is a backup to the water-level instrumentation and, in addition to initiating ECCS, it

. causes isolation of Group 2 1 solation valves. For the breaks discussed above, this instrumentation will initiate ECCS operation at about the same time as the low low water level instrumentation; thus the results given above are applicable here, also Group 2 1 solation valves include the drywell vent, purge and sump isolation valves. High-drywell pressure activates onl~ thase valves because high drywell pressure could occur as tL tasult of non-safety-related causes such as not purging the drywell air during startup. Total system isolation is not desirable for these conditions, and only the valves in Group 2 are required to close.

l The low low water level instrumentation initiates protection for the full spectrum of loss-of-coolant accidents and causes a trip of Group 1 primary system isolation valves.

P.

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DRESDEN III DPR-25 AmendmentNo./$,7/ 83 3.2 LIMITING CONDITION FOR OPgRATION MASES (Cont'd.)

Venturis are provided in the mair. steam 11nes as a means of measuring steam flow and also limiting the loss of mass inventory from the vessel during a steamline break accident. In addition to monitoring steam flow instrumentation is provided whic'h causes a trip of Group 1 isolation valves. The primary function of the instrumentation is to detect a break in the main steamline outside the drywell, thus only Group 1 valves are closed. For the worst case accident, main steamline break outside the drywell, this trip setting of 120% of rated steam flow in conjunction with the flow limiters and main steamline valve closure, limit the mass inventory loss such that fuel is not uncovered, fuel temperatures remain less than 1500 degrees F and release of radioactivity to i

the environs is well below 10 CFR 100 guidelines.

(Ref. Sections 14.2.3.9 and 14.2.3.10 SAR)

Temperature monitoring instrumentation is provided in the main i

steamline tunnel to detect leaks in this area. Trips are provided 1

to this instrumentation and when exceeded cause closure of Group 1 isolation valves. Its setting of 200*F is low enough to detect

)

leaks of the order of 5 to 10 spm; thus, it is capable of covering' the entire spectrum of breaks. For large breaks, it is a back-up r i

to high steam flow instrumentation discussed above, and for small breaks with the resultant small release of radioactivity, gives isolation before the guidelines of 10 CFR are exceeded.

High radiation monitors in the main steamline tunnel have been provided to detect gross fuel failure. This instrumentation causes closure of Group 1 valves, the only valves required to closu for this accident. With the established setting of 3 times full power background for all conditions except for greater than 201 power with hydrogen being injected during which the Main Steam 11ne trip setting is less than or equal to 3 times full power background with hydrogen addition, and main steamline isolation 4

valve closure, fission product release is limited so that 10 CFR 100 guidelines are not exceeded for this accident.

(Ref. Section 14.2.1.7 SAR) The performance of the process radiation monitoring i

system relative to detecting fuel leakage shall be evaluated during the first five years of operation. The conclusions of this evaluation will be reported to the NRC.

1 Pressure instrumentation is provided which trips when main steam-line pressure drops below 850 psig. A trip of this instrumenta-tion results in closure of Group 1 isolation valves. In the "Refucid and "Startup/ Hot Standby" mode this trip function is bypassed. This function is provided to provide protection against

~

a pressure regulator salfunction which would cause the. control B 3/4.2-30 3840a 3845A

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DRESDEN III DPR-25 o

AmendmentNo.[,'//gg 3.2 LIMITING CONDITION FOR OPERATION BASES (Cont'd.)

and/or bypass valves to open. With the trip set at 850 psig, inventory loss is limited so that fuel is not uncovered and peak clad temperatures are much less than 1500 degrees F; thus, there are no fission products available for~ release other th'an those in the reactor water.

(Ref. Section 11.2.3 SAR)

Two sensors on the isolation condenser supply and return lines are provided to detect the failure of isolation condenser line and actuate isolation action. The sensors on the supply and return sides are arranged in a 1 out of 2 logic and, to meet the single failure criteria, all sensors and instrumentation are required to be operable. The trip settings of 20 psig and 32 inches of water and valve closure time are such as to prevent uncovering the core or exceeding site limits. The sensors will actuate due to high flow in either direction.

The HPCI high flow and temperature instrumentation are provided to detect a break in the HPCI piping. Tripping of this instrumenta-tion results in actuation of HPCI isolation valves, i.e., Group 4 valves. Trippics logic for this function is the same as that for '

the isolation condenser and thus all sensors are required to be r

operable to meet the single failure of design flow and valve closure time are such that core uncovery is prevented and fission product release is within limits.

The instrumentation which initiates ECCS action is arranged in a dual bus system. As for other vital instrumentation arranged in this fashion the Specification preserves the effectiveness of the system even during periods when maintenance or testing is being performed.

The control rod block functions are provided to prevent excessive control rod withdrawal so that NCPR does not go below the MCPR fuel cladding integrity safety limit. The trip logic for this function is 1 out of n, e.g.,

any trip on one of the sin APRM's, 8 IRM's, or 4 SRM's will result in a rod block. The minimum instrument channel requirements assure sufficient instrumentation to assure the single failure criteria are not. The minimum instrument channel requirements for the RBM may be reduced by one for a short period of time to allow for maintenance, testing or calibration. This time period is only approximately 3% of the operating time in a month and does not significantly increase the risk of preventing an inadvertent control rod withdrawal.

The APRM rod block function is flow biased and prevents a significant reduction in NCPR, especially during operskio'n at 7

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DRESDEN III DPR-25 Amendment No. %, J// 33 4

3.2 LIMITING CONDITION FOR OPERATION BASES (Cont'd.)

reduced flow. The APRM provides gross core protection, i.e.,

limits the gross withdrawal of control rods in the normal withdrawal sequence.

In the refuel and startup/ hot standby modes, the APRM rod block function is set at 12% of cated power. This control rod block provides the same type of protection in the Refuel and Startup/ Hot Standby mode as the APRM flow-blased rod block does in the Run

mode, i.e., prevents control rod withdrawal before a scram is reached.

The RBM rod block function provides local protection of the core, i.e.,

the prevention of transition boiling in a local region of the core for a single rod withdrawal error from a limiting control rod pattern. The trip point is flow biased. The worst-case single control rod withdrawal error is analyzed for each reload to assure that, with the specific trip settings, rod withdrawal is blocked before the MCPR reaches the MCPR fuel cladding integrity safety limit.

Below 30% power, the worst-caso withdrawal of a single control rod-without rod block action will not violate the MCPR fuel cladding integrity safety limit. Thus, the RBM rod block function is not required below this power level.

The IRM block function provides local as well as gross core protection. The scaling arrangement is such that the trip setting is less than a factor of 10 above the indicated level. Analysis of the worst-case accident results in rod block action before MCPR approaches the MCPR fuel cladding integrity safety limit.

A downscale indication on an APRM or IRN is an indication the instrument has failed or is not senaltive enough. In either case the instrument will not respond to changes in control rod motion and the control rod motion is prevented. The downscale trips are set at 5/125 of full scale.

The rod block which occurs when the IRM detectors are not fully inserted in the core for the refuel and startup/ hot standby position of the mode switch has been provided to assure that these detectors are in the core during reactor startup. This, therefore, assures that these instruments are in proper position to provide protection during reactor startup. The IRM's primarily provide protection against loen1 reactivity effects in the source and intermediate neutron range.

Y.

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O DRESDEN III DPR-25 o

AmendmentNo.J/a, 7/ $ 3 3.2 LIMIf!NG CONDITION FOR OPERATION BASES (Cont'd.)

For effective emergency core cooling for small pipe breaks, the HPCI system must function since reactor pressure does not decrease rapidly enough to allow either core spray or LPCI to operate in time. The automatic pressure relief function is provi'ded as a backup to the HPCI in the event the HPCI does not operate. The arrangement of the tripping contacts is such as to provide thic function when necessary and minimize spurious operation. The trip settings given in the specification are adequate to assure the above criteria are met (Ref. SAR Section 6.2.6.3).

The specification preserves the effectiveness of the system during periods of maintenance, testing or calibration and also minimizes the risk of inadvertent operation; 1.e., only one instrument channel out of service.

Two radiation monitors are provided on the refueling floor which initiate isolation of the reactor building and operation of the standby gas treatment systems. The trip logic is one out of two.

Trip settings of less than or equal to 100 mR/hr for the monitors on the refueling floor are based upon initiating normal ventilation isolation and standby gas treatment system operation so that none of the activity released during the refueling F

accident leaves the reactor building via the normal ventilation stack but that all the activity is processed by the standby gas treatment system.

The instrumentation which is provided to monitor the post accident condition is listed in Table 3.2.6.

The instrumentation listed and the limiting conditions for operation on these systems ensure adequate monitoring of the containment following a loss-of-coolant accident. Information from this instrumentation will provide tho operator with a detailed knowledge of the conditions resulting from the accident. Based on this information he can make logical decisions regarding post accident recovery.

The specifications allow for post accident instrumentation to be out of service for a period of 30 days. This period is based on the fact that several diverse instruments are available for guiding the operator should an accident occur, on the low probability of an instrument being out of service and an accident occurring in the 30-day period, and on engineering judgement.

The radioactive 11guld and gaseous effluent instrumentation is provided to monitor the release of radioactive materials in 11guld and gaseous effluents during releases. The alarm setpoints for the instruments are provided to ensure that the alarms..will occur prior to exceeding the limits of 10 CFR 20.

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e DRESDEN III DPR-25 Amendment No. p$, 74 g,q r

4.2 SURVEILLANCE REQUIREMENT BASES i

The instrumentation listed in Table 4.2.1 will be functionally g

tested anu calibrated at regularly scheduled intervals. Although j

this instrumentation is not generally considered to be as important to plant safety as the Reactor Protection System, the j

same design reliability goal of 0.99999 is generally applied for l-all applications of (1 out of 2) X (2) logic. Therefore, on-off sensors are tested once/3 months, and bi-stable trips associated with analog sensors and amplifiers are tested once/ week.

Those instruments which, when tripped, result in a rod block have their contacts arranged in a 1 out of n logic, and all are cdpable of being bypassed. For such a tripping arrangement with bypass capability provided, there is an optimum test interval that should be maintained in order to maximize the reliability of a given channel (see note 7).

This takes account of the fact that testing degrades reliability and the optimum interval between tests is approximately given by:

1 = (2t/c)1/2 Where:

~

i = optimum interval between tests r-t = the time the trip contacts are disabled from performing their function while the test is in progress r = The expected failure rate of the relayr To test the trip relays requires that the channel be bypassed, the test made, and the system returned to its initial state. It is assumed this task requires an estimated 30 minutes to complete in a thorough and workmanlike manner and that the relays have a failure rate of 10-6 failures per hour. Using this data and the above operation, the optimum test interval is:

1 = [2(0.5)/10-6}1/2. 1 g 103 hours0.00119 days <br />0.0286 hours <br />1.703042e-4 weeks <br />3.91915e-5 months <br />

= approximately 40 days For additional marain a test interval of once per month will be used initially.

Note:

(7) UCRL-50451 Improving Availability and Readiness of Field Equipment Through Periodic Inspection, Benjamin Epstein Albert shiff, July 16, 1968, page 10 Iguation.(24), Lawrence Radiation Laboratory.," -

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i DRESDEN III DPR-25 Amendment No. }$ jd g3 4

4.2 SURVEILLANCE REQUIRENENT BASES (Cont'd.)

4 The sensors and electronic apparatus have not been included here j

as these are analog devices with readouts in the control room and j.

the sensors and electronic apparatus can be checked by comparison with other like instruments. The checks which are made on a daily basis are adequate to assure operability of the sensors and electronic apparatus, and the test interval given above provides j

for optimum testing of the relay circuits.

r j

4..s abeva calulated test interval optimitas each individual channel, considering it to be independent of all others. As an example, assume that there are two channels with an individual l

technician assigned to each. Each technician tests his channel at the optimum frequency, but the two technicians are not allowed to communicate so that one can advise the other that his channel is j

under test. Under these conditions, it is possible for both l

channels to be under test simultaneously. Now, assume that the technicians are required to cosusunicate and that two channels are never tested at the same time. Forbidding simultaneous testing 3

improves the availability of the system over that which would be achieved by testing each channel independently. These one out of 4

j n trip systems will be tested one at a time in order to take r

advantage of this inherent improvement in availability.

f optimizing each channel independently may not truly optimize the system considering the overall rules of system operation.

1 However, true system optimization is a complex problem. The

}

optimums are broad, not sharp, and optimizing the individual channels is generally adequate for the system.

The formula given above minimizes the unavailability of a single channel which must be bypassed during testing. The minimization of the unavailability is illustrated by curve No. 1 of Figure i

4.2.2 which assumes that a channel has a failure rate of 0.1 x 10-6/ hour and that 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> is required to test it.

The unavailability is a minimum at a test interval 1, of 3.16 x 103 hours0.00119 days <br />0.0286 hours <br />1.703042e-4 weeks <br />3.91915e-5 months <br />.

3 i

j If two similar channels are used in a 1 out of 2 configuration, the test interval for minimum unavailability changes as a function l

of the rules for testing. The simplest case is to test each one independent of the other. In this case, there is assumed to be a finite probability that both may be bypassed at one time. This case is snown by curve No. 2.

Note that the unavailability is lower as expected for a redundant system and the minimum occurs at the same test interval. Thus, if the two channels'are_ tested independently, the equation above yields the test interval for minimum unavailability.

,r l

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DRESDEN III DPR-23 o

Amendment No. J8, N 83

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4.2 SURVEILLANCE REQUIREMENT BASES (Cont'd.)

A more usual case is that the testing is done independently. If both channels are bypassed and tested at the same time, the result is shown in Curve No. 3.

Note that the minimum occurs at about 40,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />, much longer than for cases 1 and 2.

Als,o, the mini-num is not nearly as low as case 2 which indicates that this method of testing does not take full advantage of the redundant channel.

Bypassing both channels for simultaneous testing should be avoided.

The most likely case would be to stipulate that ote channel be by-passed tested and restored. Then immediately following, the second channel will be bypassed, tested, and restored. This is shown by Curve No. 4.

Note that there is no true minimum. The curve does have a definite knee and very little reduction in system unavailability is achieved by testing at a shorter interval than computed by the equation for a single channel.

The best test procedure of all those examined is to perfectly stagger the tests. That is, If the test interval is four months, test one or the other channel every two months. This is shown in Curve No. 5.

The difference between Cases 4 and 5 is negligible.

There may be other arguments, however, that more strongly support the perfectly staggered tests, including reductions in human erro/T The conclusions to be drawn are these:

1.

A 1 out of n system may be treated the same as a single channel in terms of choosing a test interval; and 2.

More than one channel should not be bypassed for testing at any one time.

The radiation monitors in the ventilation duct and on the refueling floor which initiate building isolation and standby gas treatment operation are arranged in two 1 out of 2 logic systems.

The bases given above for the rod blocks applies here also and were used to arrive at the functional testing frequency.

Based on experience at Dresden Unit I with instruments of similar design, a testing interval of once every three months has been found to be adequate.

The automatic pressure relief instrumentation can be considered to be a 1 out of 2 logic system and the discussion above applies also.

The instrumentation which is required for the post accident condition will be tested and calibrated at regularly scheduled intervals. The basis for the calibration and testing of this instrumentation is the same as was discussed above for. Protective Instrumentation in Table 4.2.4.

B 3/4.2-36 3840a 3845A

DRESDEN III DPR-25 AmendmentNo.%,[gg i

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