ML20113G333

Amend 84 to License DPR-19,changing Tech Specs to Support Cycle 10 Operation & Authorizing Use of Hafnium in Control Rod Blades

ML20113G333
Person / Time
Site:	Dresden
Issue date:	01/17/1985
From:	Zwolinski J Office of Nuclear Reactor Regulation
To:
Shared Package
ML20113G336	List: ML20113G333 ML20113G360
References
NUDOCS 8501240320
Download: ML20113G333 (51)

Text

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/

UNITED STATES 8

NUCLEAR REGULATORY COMMISSION o

'Eg

4.,I WASHINGTON, D. C. 20555

%p**C#W /

COMMONWEALTH EDIS0N COMPANY DRESDEN NUCLEAR POWER STATION, UNIT NO. 2 DOCKET NO. 50-237 AMENDMENT TO PROVISIONAL OPERATING LICENSE Amendment No. 84 License No. DPR-19

1., The Nuclear Regulatory Comission (the Commi,ssion) has found that:

A.

The applications for amendment by the Commonwealth Edison Company (the licensee) dated September 11, 27 and 28, 1984 and October 2, 1984 comply with the' standards and requirements of the Atomic 2

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 confomity with the applications, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance:

(1) that 'the activities authorized by this amendment can be conducted without endangering the health

< s and safety of the public, and (ii) that such activities will be

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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.

1 s

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2.

Accordingly, the licen'se is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 3.B 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. 84, 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 THE UCLEAR REGULATORY 0 SSION 1

C John

. Zwolinski, Chief Operat ng Reactors Branch #5 Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: January 17, 1985

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ATTACHMENT TO LICENSE AMENDMENT NO. 84 PROVISIONAL OPERATING LICENSE NO. DPR-19 DOCKET NO. 50-237 Revise the Technical Specifications by replacing the following pages, which reflect the pagination of Amendment 83, with the attached pages. The revised pages contain the captioned amendment number and marginal lines to reflect-the area of change. Some of the changes are on Radiological Effluent Technical Specifications pages which are not effective until March 15, 1985. These pages have a previous captioned amendment number of 83. The changes marked a

by marginal lines on these pages are effective as of the date of this amendment.

REMOVE INSERT iii (1) 114-(2) v (1) v (2

vii (1) vii (2 viii (1) viii (2 s

3/4.1-5 3/4.1-5 3/4.1-8 through 3/4.1-10 3/4.1-8 through 3/4.1-10

'B"3/4.1-12

'B 3/4.1-12

.l B 3/4.1-19 and B 3/4.1-20 B 3/4.1-19 and B 3/4.1-20 3/4.2-12 and 3/4.2-13 (1) 3/4.2-12and3/4.2-13(2) 3/4.2-17(1) 3/4.2-17(2) 3/4.2-19(1) 3/4.2-19 12) _.

'B 3/4.2-32 and B 3/4.2-33 (1)

B 3/4.2-32 and B 3/4.2-34.(2) 3/4.5-15 3/4.5-15 is 3/4.5-17 through 3/4.5-27 3/4.5-17 throuuh 3.4.5-30 l

B 3/4.5-28 through B 3/4.5-41 B 3/4.5-31 through B 3/4.5-44 5.1 5.1

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(1) Pages which are from Amendment 83.

(2) Pages which contain information from Amendment 83 but also have changes l

approved in this amendment which are immediately effective.

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DRESDEN II.

DPR-19 AmendmentNs.pd,p8,84 (Table of Contents. Cont'd.)

f.818 3.5.C HPCI-Subsystem 3/4.5 - 6 3.5.D Automatic Pressure Relief Subsystems 3/4.5 - 8 3.5.E Isolation Condenser System 3/4.5 - 9 3.5.F Minimum Core and Containment Cooling System Availability 3/4.5 -11 3.5.G Deleted 3.5.H Maintentace of Filled Discharge Pipe 3/4.5 -13 3.5.I Average Planar LHGR 3/4.5 -15 3.5.J Local LHGR 3/4.5 -15 3.5.K Minimum Critical Power Ratio 3/4.5 -25 3.5.L Condensate Pump Room Flood Protection 3/4.5 -26 Limiting Conditions for Operation Bases (3.5)

B 3/4.5 -31 Surveillance Requirement Bases (4.5)

B 3/4.5 -39 3.6 Primary System Boundary 3/4.6 e 1 3.6.A Thermal Limitations 3.6.B Pressurization Temperature 3/4.6 - 1 3.6.C Coolant Chemistry 3/4.6 - 2 3/4.6 - 3 3.6.D Coolant Leakage 3/4.6 - 5 3.6.R Safety and Relief Valves 3/4.6 - 6 3.6.F Structural Integrity 3/4.6 - 7 3.6.C Jet Pumps 3/4.6 -10 3.6.H Recirculation Pump Flow Mismatch 3/4.6 -11 3.6.I shock Suppressors (Snubbers) 3/4.6 -12 Limiting Conditions for Operation Bases (3.6)

B 3/4.6 -25 Surveillance Requirement Bases (4.6)

B 3/4.6 -38 3.7 Containment Systems

'-'3/4. 7 - 1

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3.7.A Primary Containment 3/4.7

,1 3.7.B Standby Gas Treatment System 3/4.7 -19 3.7.C Secondary Containment 3/4.7 -25 l

3.7.D Primary Containment Isolation Valves 3/4.7 -27 Limiting Conditi'ons for Operation Bases (3.7)

B 3/4.7 -33 Surveillance Requirement Bases (4.7)

B 3/4.7 -40 3.8 Radioactive Materials 3/4.8 - 1 3.8.A Airborne Effluents 3/4.8 - 1 3.8.B Mechanical Yacuum Pump 3/4.8 - 9 3.8.C Liquid Effluents 3/4.8 -10 l

3.8.D Radioactive Wa.ste Storage 3/4.8 -11 l

3.8.R General Information 3/4.8 -12 3.8.F Miscellaneous Radioactive Materials Sources 3/4.8 -14 Limiting Conditions for Operation Bases (3.8)

B 3/4.8 -18 Surveillance Requirement Bases (4.8)

B 3/4.8 -22 3.9 Auxilary Electrical Systems 3/4.9 - 1 l

3.9.A Requirements 3/4.9 - 1 3.9.B Availability of Electric Power 3/4.9 - 2 111 3959a 3843A

F DRESDEN II DPR-19 Amendment No. p(, $8, 84 (Table of Contents. Cont'd.)

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P_agg, 4.3.D Control Rod Accumulators 3/4.3 -11 4.3.E Reactivity Anomalies 3/4.3 -12 4.3.F (N/A) 4.3.G Automatic Generation Control System 3/4.3 -13 4.4 Standby Liquid Control System 3/4.1 - 1 4.4.A Normal Operation 3/4.4 - 1 4.4.B Surveillance With Inoperable Components

" 3/4.4 - 3 3/4.4 - 2 4.4.C Boron Solution 4.5 Core and Containment Cooling Systems 3/4.5 - 1 4.5.A Core Spray and LPCI Subsystems 3/4.5 - 1 4.5.B Containment Cooling Subsystem 3/4.5 - 5 4.5.C HPCI Subsystem 3/4.5 - 6 4*.5.D Automatic Pressure Relief Subsystems 3/4.5 - 8 4.5.E Isolation Condenser System 3/4.5 - 9 4,.5.F Core and Containment Cooling System 4.5.G (Deleted) 3/4.5 -11 4.5.H Maintenance of Filled Discharge Pipe 3/4.5 -13 4.5.I Average Planar Linear Heat Generation Rate 3/4.5 -15 4.5.J Linear Heat Generation Rate 3/4.5 -15 4.5.K<~-Minimum Critical Power Ratio 3/4.5 -25 4.5.L Condensate Pump Room Flood Protection 3/4.5 -26 4.6 Primary System Boundary 3/4.6 - 1 4.6.A Thermal Limitations 3/4.6 - 1 4.6.B Pressurization Temperature 3/4.6 - 2 4.6.C Coolant Chemistry

-3/4.6 - 3 4.6.D Coolant Leakage 3/4.6 - 5 4.6.E Safety and Relief Valves 3/4.6

'6 4.6.F Structural Integrity 3/4.6 - 7 3/4.6 -10 4.6.C Jet Pumps 4.6.H Recirculation Pump Flow Mismatch 3/4.6 -11 4.6.I Snubbers (Shock Suppressors) 3/4.6 -12 4.7 Containment System 3/4.7 - 1 4.7.A Primary Containment 3/4.7 - 1 4.J.B Standby Gas Treatment System 3/4.7 -19 4.7.C Secondary containment 3/4.7 -25 4.7.D Primary containment Isolation Valves 3/4.7 -27 4.8 Radioactive Materials 3/4.8 - 1 4.8.A Airborne Effluents 3/4.8 - 1

)

'4.8.B Mechanical Vacuum Pump 3/4.8 --9 4.8.C Liquid Effluents

~3/4.8 -10 4.8.D Radioactive Waste Storage 3/4.8 -11 4.8.E General 3/4.8 -12

' 4.8.F Miscellaneous Radioactive Materials Sources

-3/4.8 -14 V

3959a

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DRESDEN II DPR-19 Amendment No. gl, Q& 84 List of Tabies i

P9.1.9.

Table 3.1.1 Reactor Protection System (Scram) 3/4.1 - 5 Instrumentation Requirements i

l 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 Functiont 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 Instrumentation 3/4.2 -14 Table 3.2.5 Radioactive Gaseous Effluent Monitoring Instrumentation 3/4.2 -15 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 -20 Table 4.2.3 Radioactive Gaseou.s Effluent Monitoring Instrumentation Surveillance Requirements 3/4.2 -22 Table 3.6.la safety Related Hydraulle Snubbers 3/4.6 -18 & 19 Table'3.6.lb Safety Related Mechanical Snubbers 3/4.6 -20 & 21 Table 4.6.2 Neutron Flux and Sample Withdrawal B 3/4.6-30 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 Disolved 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 Radioactive Ravironmental 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 B 3/4.12-17 Table 3.12-2 Sprinkler Systems B 3/4.12-18 Table 3.12-3 CO2 Systems B 3/4.12-19 Table 3.12-4 Fire Hose Stations B 3/4.12-20 & 21 Table 6.1.1 Minimum Shift Manning Chart 6-5 Table 6.6.1 Special Reports 6-26 vil 3959a 3843A

DRESDEN II DPR-19 Amendment No. %, $8, 84 I

e List of Figures f.11e.

Figure 2.1-3 APRM Bias Scram Relationship to Normal Operating Conditions B 1/2.1-17 Figure 4.1.1 Graphical Aid in the Selection of l

an Adequate Interval Between Tests B 3/4.1-18 i

Figure 4.2.2 Test Interval vs. System Unavailability B 3/4.2-34 Figure 3.4.1 Standby Liquid control Solution Requirements 3/4.4-4 Figure 3.4.2 Sodium Pentaborate Solution Temperature Requirements 3/4.4-5 Figure 3.5-1 Maximum Average Planar LHCR-3/4.5-17 (consisting of eight fuel type curves) thru 24 Figure 3.5-2 Core Flow %

3/4.5-27 & 28 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 N

Figuro 4.8.2 Detail of Central complex B 3/4.8-39 ~

Figure 6.1-1 Corporate organization 6-3 Figure 6.1-2 Station Organization 6-4 Vill 3918a 8401D

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1 DRESDEN II DPR-19 Amendment. No. %, 84

)

TABLE 3.1.1 j

REACTOR PROTECTION SYSTEM,(SCRAM) INSTRUMENTATION REQUIREMENTS iinisan Nurber Modes in,2ich Function

perabla Inst.

Must be coerable 3mnnels per Trip

.Startup/ Hot l

(1) System Trio Function Trio level Settino Refuel D)

'Stane y Ruq Action

• I 1

Mode Switch in Shutdown X

X X

A-1 Manual Scram X

X X

A IRN 3

High Flux (LT/E) 120/125 X

X X(5)

A of Full Scale 3

Inoperative X

X X(5)

A

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2 High Flux Specification 2.1.A.1 i

,X X(9)

X A or 8 2

Inoperative **

X X(9)

X A or B 2

Downscale (GTA) 5/125 X(12)

X(12)

X(13)

A or 8 of Ful-1. Scale 2

High Flux (15% Scram)

Specification 2.1.A.2 X

X X(14)

A,

2

[tigh Reactor Pres,sure

'(LT/E) 1060 psig X(ll)

X X

A 2

High Drywell Pressure (LT/E) 2 psig X(8), X(10)

X(8), (10)

X(10)

A 2

Reactor Low Water Level (GTA) 1 inch ***

X X

X A

2 High Water Level in (LT/E) 40 inches above X(2)

X X

A or D' (Per Bank)

Scram Discharge Volume bottcm of the Instrument (Thermal and dP Switch)

Volune

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2 Turbine Condenser Low (GTA) 23 in. Hg Vacuun X(3)

X(3)

X

. A or C Vacuum s

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2 Main Steam Line High (LT/E) 3 X X(3)

X(3)

X(15)

A or C Radiation Full Power

• Background 4(6)

Main Steam Line (LTA) 105 Valve Closure X(3)

X(3)

X A or C Isolation Valve Closure 2

Generator Load X(4)

X(4)

X(4)

A or C Rejection 2

Turbine Stop' Valve (LTA) 105 Valve Closure X(4)

X(4)

X(4)

A or C

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Closure i

2 Turbine control -

(GTA) 900 psig X

X X

A or C Loss of control 011 Pressure Notes: (LTA) = Less than or equal to.

(GTA) = Greater than or equal to.

(Notes continue on next two pages) 3/4.1-5 919e 401D

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DRESDEN II DPR-19 Amendment No, p 84 TABLE 4.1.1 SCRAM INSTRUMENTATION FUNCTIONAL TESTS MINIMUM FUNCTIONAL TEST FREQUENCIES FOR SAFETY INSTR. AND CONTROL CIRCUITS Instrunent Cha'nnel Group (3)

Functional Test Mininum Frecuency (4)

Mode Switch in Shutdown A

Place Mode Switch in Shutdown Each Refueling Outage Manual Scras:

A Trip Channel and Alam Every 3 Months Ilut -

• High Flux C

Trip Channel and Alarm'(5)

Before Each Startup (6)

• Inoperative C

Trip Channel and Alarm Before Each Startup (6) i Apfut High Flux 8

Trip output Relays (5)

Once Each Week Inoperative 8

Trip Output Relays Once Each Week Downscale 8

Trip Output Relays (5)

Once Each Week High Flux (15% scram) 8 Trip Output Relays Before Each Startup High Reactor Pressure d-Trip Channel and Alam (1) i High Dr 11 Pressure A

Trip Channel and Alam (1)

Reactor Low Water Level (2) 8 (8)

(1) l High Water Level in Scram Discharge A

Trip Channel and Alarm (7)

Every 3-Months Volumes (Thermal and dp Switch) l Turbine Condenser Low Vacuum A

Trip Channel and Alarm (1)

Main Steam Line High Radiation (2) 8 Trip Channel and Alarm (5) 6nceEachWeek Main Steam Line A

Trip Channel and Alarm (1)

Isolation valve Closure i

GeneratorLoadRejection A

Trip Channel and Alarm (1) l Turbine Stop Valve Closure A

Trip Channel and Alarm (l)

Tuttine Control - Loss of Control A

Trip Channel and Alarm (1)

Oil Pressure Notes: (See next page.)

3/4.1-8 3919a 3401D

l DRESDEN II DPR-19 Am2ndment N3. pf, 84 NOTES:

(For Table 4.1.1) l'.

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

is 2.0x10 ; thereafter, according to Figure 4.1.1 with an 4.1.1) interval not less than one month nor more than three months. The compilation of instrument failure rate data may include data obtained feca other Boiling Water Reactors for which the same design instrument operates in an environment similar to that of Dresden; Unit 2.

l 2.

An instrument check s, hall be performed on low reactor water level once per day and on high steam line radiation once per. shift.

3.

A description of the three groups is included in the Bases of this Specification.

4.

Functional tests are'not required when the systems are not required to be operable or are tripped.

If tests are missed, they shall be performed prior to returning the systems to an operable status.

5.

This instrumentation is exempted from the Instrument Functional Test Definition (1.0.G).

This Instrument Function Test will consist of injecting a simulated electrical signal into the measurement channels.

6.' If reactor start-ups occur more frequently than once per week, the functional test need not be performed; i.e., the maximum functional test frequency shall.be once per week.

7.

Only the electronics portion of the thermal switches will be tested i

using an electronic calibrator during the three month test. A water column or equivalent will be used to test the dp switches.

8.

A functional test of the master and slave trip unit is required monthly (staggered one channel out of 4 every week). A calibration of the trip unit is to be performed concurrent with the functional I

testing.

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DRESDEN II DPR-19 Am:ndm:nt No. % 84 TABLE 4.1.2 SCRAM INSTRUMENTATION CALIBRATIONS MINIMUM CALIBRATION FREQUENCIES FOR REACTOR.,.. PROTECTION INSTRUMENT CHANN y<

Instrurrent Channel Group (1)

Calibration Test Minimus Frecuency (2)

• High Flux IRM C

Comparison to APRM after Heat Every Shutdown (4)'

Balance s

High Flux APRM Output Signal 8

Heat Balance Once Every 7 Days Flow Bias B

Standard Pressure and Voltage Refueling Outage Source High Reactor Pressure A

Standard Pressure Source Every 3 Months

.High Drywell Pressure A

Standard Pressure Source Every 3 Months

. Reactor low Water Level 8

Water Level (5)

Turbine Condenser low Vacuum A

StaridardVacuumSource Every 3 Months Main Steam Line High Radiation

'8 Standard Current Source (3)

Every 3 Months Turbine Conirol - Loss of Control A

Pressure Source

{

Oil Pressure Every 3 Months High Water level in Scram Discharge A

Water Level Volume (4 oniy)

Once per Refueling

. Outage..

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M:

(For Tabie 4.1.2) s 1.

A description of the three groups is included in the bases of this Specification.

2.

Calibration tests are not required when the systems are not required to be operable or are tripped.

If tests are missed, they shall be performed prior to returning the systems to an operable status.

l 3.

The current source provides an instrument channe'l alignment.

Calibration using a radiation source'shall be made during each refueling outage.

• 4.

If reactor startups occur more frequently.than once per week, the functional test need not be performed; i.e., the maximum functional test frequency shall be once per week.

i 5.

Trip units are calibrated monthly concurrently with f.unctional testing (staggered one channel out of 4 every week).

Transmitters are calibrated once per operating cycle.

3/4.1-10

'3919a

,8,401D

DRESDEN II DPR-19 Amendment No. $l, 84 3.1 LIMITING CONDITION FOR OPERATION BASES (Cont'd.)

system can tolerate a single failure and still perform its intended function of scrasuning the reactor. Three APRM instrument channels are provided for each protection trip system.

APRM's #1 and #3 operate contacts in a one subchannel and APRM's

1. 2 and #3 operate contacts in the other subchannel. APRM's #4, #5 and #6 are arranged similarly in the other protection trip system. Each protection trip system has one more/APRM than is necessary to meet the minimum number required per channel. This

(

allows the bypassing of one APRM per protection trip system for i

maintenance, testing or calibration. Additional IRM channels have also been provided to allow for bypassing of one such channel.

The bases for the scram settings for the IBM, APRN, high reactor pressure, reactor low water level, generator load rejection, and turbine stop valve closure are discussed in specification 2.3.

Instrumentation (pressure switches) in the drywell are provided to detect a loss of coolant accident'and initiate the emergency core cooling equipment. This instrumentation is a backup to the water level instrumentation which is discussed in Specification 2.2.

A scram is provided at the same setting as the emergency core cooling system (ECCS) initiation to minimize the energy which must be accommodated during a loss of coolant accident and to prevent the reactor from goin's, critical following the accident.

The control rod drive scram system is designed so that all of the

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water which is discharged from the Reactor by a scram can be l

accommodated in the discharge piping. A part of this system is an l

individual instrument volume for each of the south and north CRD l

accumulators. These two volumes and their piping can hold in j

excess of 90 gallons of water and is the low point'in 'the piping.

No credit was taken for these volumes in the design of the discharge piping relative to the amount of water which must be

,s accomunodated during a scram. During normal operations, the

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discharge volumes are empty; however, should either volume fill with water, the water discharged to the piping from the Reactor may not be accommodated which could result in slow scram +ines or

-partial or no control rod insertion. To preclude this occurrence, level switches have been installed in both volumes which will alarm and scram the Reac' tor when the volume remaining in either instrument volume is approximately 40 gallons.

For diversity of level sensing methods that will ensure and provide a scram, both differential pressure switches and thermal switches have been incorporated into the design and logic of the system. The

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setpoint for the scram signal has been chosen on the basis of providing sufficient volume remaining to accommodate a scram even 4

with 5 spa leakage per drive into the SDV.

As indicated above, there is sufficient volume in the piping to accomunodate the scram j

without impairment of the scram times or the amount of insertion of the control rods. This function shuts the Reactor down while sufficient volume remains to accomunodate the discharged water and p: ecludes the situation in which a scram would be required but not be able to perform its function properly.

B 3/4.1-12 3919a

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8401D

DRESDEN II DPR-19 Amendment No. pd, 84 4.1 SURVEILLANCE REOUIREMENT BASES (Cont'd.)

Reactor low water level instruments 2-263-57A, 2-263-57B, 2-263-58A, and 2-263-588 have been modified to be an analog trip system. 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 trip system, including reactor low water level, has been established in Licensing Topcial Report NEDO-21617-A (December 1978). With the one-out-of-two-taken-twice logic, NEDO-21617-A states that each trip unit be subjected to a calibration / functional test of one month (staggered one channel out of four every week). An adequate calibration /

surveillance test interval for the transmitter is~once per operating cycle.

Group (C) devices are active only during a given portion of the operational cycle.

For example, the IRM is active during startup and inactive during full-power operation. Thus, the only test that is meaningful is the.one performed just prior to shutdown or startup; i.e., the tests that are performed just prior to use of the instrument.

I calibration frequency of the instrument channel is divided into two groups. These are as follows:

1.

Passive type indicating devices that can be compared with like units on a continuous basis.

2.

Vacuum tube or semiconductor devices and detectors that drift or lose sensitivity.

Experience with passive type instruments in Commonwealth s

Edison generating stations and substations indicates that the specified calibrations are adequate. For those devices which employ emplifiers, etc., drift specifications call for drift to be less than 0.4%/ month; i.e.,

in the period of a month, a drift of 0.4% would occur and thus provide for adequate margin.

l For the APRM system drift of electronic apparatus is not the only consideration in determining a calibration frequency.

Change in power distribution and loss of' chamber'sensitiv'ity dictate a calibration every seven days.

Calibration on this frequency assures plant operation at or below thermal limits, r

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DRESDEN II DPR-19 d

Amendment No. J, 84 4.1 SURVEILLANCE REQUIREMENT BASES (Cont'd.)

A comparison of Tables 4.1.1 and 4.1.2 indicates that six instrument channels have not been included in the latter Table. These are:

Mode Switch in Shutdown, Manual Scram, High Water Level in Scram Discharge Volume dp and Thermal Switches, Main Steam Line Isolation Valve closure, Generator Load Rejection, and Turbine Stop Valve Closure. All of the devices or sensors associated with these scram. functions are simple on-off switches and, hence, calibration is not applicable; l'.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 switches in the scram discharge volume tank. Based on the above, no calibration is required for these six instrument channels.

B.' The NFLPD for fuel fabricated by CE shall be checked once per 1

day to determine if the APRM gains or scram requires adjustment. This may normally be done by checking the LPRM readings, TIP traces, or process computer calculations.

Only a small number of control rods are moved daily and thus the peaking factors are not expected to change significantly and thus a daily c, heck of the MFLPD is adequate.

For fuel fabricated by ENC, the power distribution will be checked once per day to ensure consistency with'the power distribution assumptions of the fuel design analysis for overpower conditions.

During periods of operation beyond these power distribution assumptions, the APRM gains or scram i

settings may be adjusted to ensure consistency.with the fuel design criteria for overpower conditions.

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3919a

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DRESDEN II DPR-19 Amsndm2nt No. %, %, 84 O

TABLE 3.'2.3 INSTRUMENTATION THAT INITIATES ROD BLOCK t

Minimum No. of Operable Inst'.

Channels Per Trio System (1)

Instrument Trip' Level Setting 1

APRM upscale (flow bias) (7)

Less than or equal to D P us 50) (FRP/MFLPD)

(0.58 W l

(See Note 2) 1 APRM upscale (refuel and Less than or equal to Startup/ Hot Standby mode) 12/125 full scale 2

APRM downscale (7)

Greater than or equal to 3/125 full scale 1

Rod block monitor Less than or equal to upscale (flow bias) (7)

(0.65 WD P us 45) l (see Note 2) 1 Rod block monitor Greater than or equal to 5/125 full scale downscale (7) 3 IRM downscale (3)

Grecter than or equal to 5/125 full scale 3

IRM upscale Less than or equal to 108/125 full scale l

'3 IRM detector not fully N/A l

inserted in the core 2 (5)

SRM detector not in startup position (4) 2 (5). (6)

SRM upscale Less than or equal to 105 counts /sec.

1 (per bank) Scram discharge volume (LT/E) 26 inches above water level - high the bottom of the instrum,ent volume l

flotes:

(See Next Pate) 3/4.2-12 39f0a 3843A

DRESDEN II DPR-19 Amendment No.JPI, pd, 84

. TABLE 3.2.3 (Notes)

~

1.

For the Startup/ Hot Standby and Run positions of the Reactor Mode Selector Switch, there shall be two operable or tripped trip systems for each function, except the SRM rod. blocks. IRM upscale, IRM downscale and IRM detector not fully inserted in the core need not be

~

operable.in the "Run" position and APRM downscale APRM upscale (flow

' bias), and RBM downscale need not be operable in the Startup/ Hot Standby mode. A RBM upscale need not be operable at less than 30%

rated thermal power.,0ne channel may be bypassed above 30% rated thermal power provided that a limiting control rod. pattern does not exist. For systems with more than one channel per trip system, if the first column cannot be met for both trip systems, the systems shall be tripped.

For the scram discharge volume water level high rod block, there is one instrument channel per bank.

t 2.

WD Percent of drive flow required to produce a rated core flow of 98 M1b/hr. NFLPD = highest value of FLPD for G.R. fuel.

3.

IRM downscale may be bypassed when it is on its lowest range.

4.

This function may be b'ypassed when the count r' ate is greater than or equal to 100 cys.

5.

One.of the four SEN inputs may be bypassed.

6.

• This SEM function may be bypassed -in the higher IBM ranges when the IRM upscale Rod Block is operable.

7.

Not required while performing low power physics test at atmospheric pressure during or after refueling at power levels not to exceed 5 MWt.

I l

l i

I i

l i

3/4.2-13 I

i 3960s l

3843A I

T ele 4.2.1 ORE 5 DEN !!

DPR-Ig M1hleum TEST AND CAL 1pmA110m FmEQUENCY FOR CDIE AND Amenoment W. %,

g4 CONTAlptNT C00LINC SYSTEMS INSTRLPENTA1108, ROC SLOCK5, AMD ISOLA 110NS Instrument Channel Instrument Instrument Functional fest Calibration Check ECC5 Instrwentation 1.

neactor Len.-Lou beter Level (1)

Once/3 nonths

>ce/ Day 2.

Dryme11 Mi p Pressure (1)

Once/3 Months hone 3.

Reactor tw Pressure (1) once/3 mnths hone

• 4 Contaiment Soray Interlock l

a.

2n Core Heignt (1) (13)

(13) wone b.

Contalnuent High Pressure (1)

Once/3 Montas hone 5.

Lou Pressure core Cooling Pump (1)

>ce/3 months hone Discharge 6.

undervoltage Esergency Bus nefueling matage hefuel Outage Once/3 months 7.

Sustained High menctor Pressure (1)

Once/3 months hone 8.

De3 raced voltage Energency tus Refueling Outage (10)

Refuel Outage Monthly Rod Stocks 1.

APun Counscale (1) (3) mce/3 Months hone 2.

Apun F1m Variele (1) (3)

Refuel Outage hone 3.

Arm tescale (Startup/ Hot Staney)

(2) (3)

(2) (3)

(2) 4.

IRM upscale (2) (3)

(2) (3)

(2) 5.

tan o m nscale (2) (3)

(2) (3)

(2) 6.

Im Detector hot Fully Inserted (2) h/A home in the Core 7.

aan upscale (1) (3) mefuel Outage mane 8.

Ren Desnscale (1) (3) mce/3 months hone g.

Sam upscale (2) (3) *

(2) (3)

(2)

10. San Detector mot in Start e Position (2) (3)

~

(2) (3)

(2)

11. Scram Instrument volwe tevel Nigh once/3 Months (g) kone mene o

contairweat monito*4ae 1.

Pressure N

a.. Minus 5 in. He to plus 5 psig hone once/3 Months once/ Day Indicator b.

O to 75 psig Indicator hone mice /3 months hone 2.

Tamperature hone Asfuel Ortage Qice/ Day 3.

Drymell-forms Differential home Once/6 Months (Teso home Pressure (5) (6)

Channels Opere le)

(0-3 psid)

Once/ month (Die Channel Operele) 4 Torus heter tevel (5) (6) none once/6 monthe a.

Plus or eines 25 in. bride mange Indicator b.

18 in. Sight Glass Safetv/telief valve u nitorine 1.

Safety /nelief valve (7) hone Once Per l

Position Indicator 31 Days (Acoustic nenitor) (0) l 2.

Safety /nellef valve Position home Once overy bce Per Indicator (Temperature 18 months 31 Days Manitor (8) 3.

Safety Walve Positlen Indicator (7)

None Once per (Aemustic monitor) (e) 31 Days 4.

Safety valve Positter Indicator home Once overy once por (Temperature monitor) (8) 18 months 31 Days Hein Ste m Line Isolatten 1.

Steam Tunnel Nigh Temperature Refueling h tage Refuel Cutage NIsne 2.

Steam Line High Fler (1)

Dion/3 Ronths

>ce/ Day 3.

Steen Line tene Pressure (1)

Once/3 Runths hone l

4.

Steam Line Nigh Radiation (1) (3)

Once/3 months (8)

Dece/Doy lentation Condenser 1salation 1.

Ste m Line Ni p Pleu (1)

Once/3 Months hone 2.

Consensate Line High Flem (1)

Once/3 Months hone M 1 solation l

1.

Stem Line High Flou (1) (11) (12)

(11) (12) kone 2.

Steam Line Area 6tgh Temperature hefueling Outage Refuel hatage hone l

3.

Lou heettor Pressure (1) (13)

-(13) hone anactor buildine vent fsctat on and SACTS Initiation i

1.

Refueling Floor mediation monitors (1)

Dece/3 months Once/ Day Isotest (See West Two Pages) 3/a.r-1J l

l l

i

Lal&RLLT1C1 LOM-)Ei)

Amendscnt No. pf, Sg, 84 NOTES:

(For Table 4.2.1) (Cont'd.)

9,.

The functional test of the Scram Dischirge 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 calibra. tion is required for these instrument channels.

Functional test sh'all l'clude verification of the sec'o'nd level 10.

n undervoltage (degrade,d. voltage) timer bypass and shall verify 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 durin5 each refueling outage.

12.

Telp units are functionally tested monthly (staggered one channel out of four every week).

A celibration of the trip units is to be pet.?ormed concurrent with the functional testing.

13.

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

A calibration of the trip units is to be performed concurrent with the functional testing.

4 N

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3/4.2-19 3920a 8401D l

i

DRESDEN II DPR-19 Amindm:nt Ns. $d, 84

{

4.2 SURVEILLANCE REOUIREMENT 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 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.

Also, the minimum 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 i

be avoided.

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

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 l

2.

More than one channel should not be bypassed for testing 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 level, has been established in Licensing Topical Report NEDO-21617-A (December,'1978).

For instruments 2(3)-2389A, B, C, D, the one-of-two-taken-twice l

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 operating cycle.

B 3/4.2-32 l

3920s 8401D

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DRESDEN II DPR-19

~

AmsndmentNo.pd,Q3,84 4.2 SURVEILLANCE REOUIREMENT BASES (Cont'd.)

For instruments 2(3)-263-73A, 73B and 2(3)-2352, 2353, the logic downstream 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 1 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.

I i

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l B 3/4.2 '

t 3920a 8401D

i DRESDEN II DPR-19 Amendment N3. k, $$, 84 10"I

~

A gg-2 Cuftvt !

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TEST WITERVAl. (i) haws t

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I TEST INTERVAL VS. SYSTEM UNAVAILABILITY B 3/4.2-34 3920e 8401D I

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DRESDEN II DPR-19 Amendment No. P, 84 f

3.5' LIMITING CONDITION"F,0,,L OPERATION 4.5 SURVEILLANCE REOUIREMENT

.(Cont'd.)

(Cont'd.)

I.

Averste Planar LHGR I.

Averste Planar Linear Heat Generation Rate (APLHCR)

During steady state poder operation, the Average The APLHGR for each type of Planar Linear Heat fuel as a function of Generation Rate.(APLHGR) aver'a'ge planar exposure of all the rods,in any i

for G.E. fuel and average fuel assembly, as a bundle exposure for Exxon function of average fuel shall be determined planar exposure for G,E.

daily during reactor fuel and average bundle operation at greater than exposure for Exxon fuel

. or equal to 25% rated at any axial location, thermal power.

shall not exceed the maximum average planar LHGR shown.la Figure 3.5-1 (consisting of eight curves).

If at any time

.' ~

~

during operation'it is determined by normal surveillance that the limiting value for APLHGR is being exceeded, action shall be initiated within 15.minutos to restore ope' ration to within the prescribed limits.

If the APLHGR is not returned to within the prescribed limits within two (2) 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, the reactor shall be brought to the Cold Shutdown condit' ion within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

Surveillance and

-corresponding action shall continue until reactor operation is within the prescribed limits.

J.

1& GAL LHGR

.J.

Linear Heat Generation Rate

-(LHGR)

During steady state power The'LHGR shall be checked operation, the linear daily during reactor heat generation ~ rate operation at greater,than (LHGR) of any rod in any or equal to 25% rated fuel assembly. fabricated thermal power.

3/4.5-15 3647a 3123A

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DRESDEN II DPR-19 Amendme:t Ns. p, 84 i

Maximum s/erage Planar Linear Heat Generation P:-te WW.PLHGP) versus Exposure 14 0 l

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10,000 20,000 30,000 40,000 Planar Average Exposure WND/STL The above graph is based on the following MAPLHCR summary for fuel type 8D250.

Planar Average Exposure (MWD /ST)

MAPLHCR-(KW/Ft) 200 11.2 l

1,000 11.3 5,000 11.9 10,000 12.1

.15,000 12.2' I

20,000 12.0 l

25,000 11.5 30,000-10.6 35,000 9.6 40,000 9.0 i

~

Figure 3.5-1 (Sheet 1 of 8) 3/4.5-17 l

3687a

'3123A i

[. :..-

DRESDEN II Amendment.ua.pt,SSg_19 Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) versus Exposure 11.6 i

13.O t=

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10,0"0 20,300 30,200 40,000 Planar nver age Expo sur e 8.1+.iD'STL The above graph is based on the following MAPLHGR summary for fuel type SD262.

Planar Average Exposure (MWD /ST)

MAPLHGR (KW/Ft) 200 11.1 1,000 11.3 5,000 11.9 4

10,000 12.1 15,000' 12.2 20,000 12.0 25,000 11.6 30,000 10.7 35,000 9.8 40,000 9.2 Figure 3.5-1 (Sheet 2 of 8)

I 3/4.5-18 3687a

-.31234..

DRESDEN II DPR-19.

Ame dmelt No. $l, 84 Maxim 0m hverage Planar Linear Heat Generation Rate f.lW LHGPI versus Evposure g ga t

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G 10.000 20.000 30,000 40,000 Plane Heere.;e E,.p _ sere fik 4 STr The above graph is based on the following MAPLHGR sunusary for fuel type P8DRB265H.

Planar Average Exposure (MWD /ST)

MAPLHGR (KW/Ft) 200 11.5 1,000 11.6 5,000 11.9 10,000 12.1 l

15,000 12.1' l

20,000 11.9 l

25,000 11.3 30,000 10.7 35,000 10.2 40,000 9.6 45,000 8.9 l

l Figure 3.5-1 (Sheet 3*of 8)

(-

3/4.5-19 i

l 1687a e

DRESDEN II DPR-19 Ame dment No. %, 84 Maxim.'.m Average Planar. Linear Heat Generation Pate P

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'0.000 00,000 30,400 40,000 Planar 6tver ege Expesur e O ilID 5 T o -..

The above graph is based on the following MAPLHGR susunary for fuel type P8DRB265L.

PlanarAverageRaposure(MihD/ST)

MAPLHGR (KW/Ft) 200 11.6 1,000 11.6 5,000 12.0 10,000.

12.1 15,000

'12.1 20,000 11.9 25,000 11.3 30,000 10.7 35,000 10.2 40,000 9.6 Figure 3.5-1 (Sheet 4 of 8) 3/4.5-20 3687a 3123A

DRESDEN II DPR-19 l

Amendmelt No. pd, 84 Mayimum averade Planar Linear Heat Generation Rate IJ. f (MAPLHGR) versus Evposure q.

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10,300 20,090 30,090 40,000 Flaner Mverage Exposure U ND/ST) 5 The above graph is based on the following MAPLHGR susunary for fuel type 8DR8265L.

Planar Average Exposure (MWD /ST)

MAPLHCR (KWFt)

~

~

200 11.6 1,000 11.6 5,000 11.8 10,000 11.9 15,000 11.9 20,000 11.7 25,000 11.3 30,000 10.7 35,000 10.2 40,000 9.6 Figure 3.5-1 (Sheet 5 of 8) 3/4.5-21 3687a mcm -

l DRESDEN II DPR-19 Ameldmeat No. fl, 84 9

Ma>.irnym Average Planar Linear Heat Generation Pate a..g R'#LHGP1 verv.e E.g.059re

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10.309 2 d, 'J G C 5 0, d e'J 40,000 P i e n t.i merMe E..pusv>e 4.1 MD/$i n The above graph is based on the following MAPLHCR sununary for fuel type s-P8DRB282.

Platar Average Exposure (MWD /ST)

MAPLHGR (KW/Ft) 200 11.2 1,000 11.2 5,000 11.8 10,000 12.0 15,000'

• 12.0 20,000 11.8 25,000 11.3 30,000 11.1 35,000 10.4 40,000 9.8 F15ure 3.5-1 (Sheet 6 of 8) 3/4.5-22 3687a 3123A

4 DRESDEN II DPR-19 AmeIdmertNo.f},84 Ha>.imum Average Planar Linear Heat Generation Rate (MAPLHGR) versus E.)00sure 4.i s.

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The above graph ir based on the following NAPLHGR susumary for fuel type ENC 8Z8 (all types).

Bundle Average Exposure (NWD/MTU)

MAPLHCR (KW/Ft) 0 13.0 15,000 13.0 18,000 12.85 20,000 12.60 25,000 11.95 30,000 11.20 35,000 10.45 Figure 3.5-1 (Sheet 7 of 8) 3/4.5-23 3687a 3123A 6

DRESDEN II DPR-19 Amendmelt N3. P2, 84 l'laximum Average Planar. Linear Generati.on Rate

&WPLHGP.1 verses Ex90sure i4 "

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1 6.

1 I~

ri8

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i l

i ll l'

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i lt 17 i

nc

....t, 2.f,.y e..

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4e.oon Ben d !e m e r e.ee E po ser e f.IIIID/IIT i The above graph is based on the following MAPLHGR summary for fuel type ENC 919 LTA.

Bundle Average Exposure (MWD /MTU)

MAPLHGR (KW/Ft) 0 10.24 l

15,000 10.24 l

18,000 10.12 l

20,000, 9.92 25,000

• 9.41 i

30,000 8.82 35,000 8.23 l

Fisure 3.5-1 (Sheet 8 of 8) l l

3/4.5-24 3687a 3123A l

DRESDEN II DPR-19 Amendment Ns. pd, 84 3.5 LIMITING CONDITION FOR OPERATION 4.5 SURVEILLANCE REOUIREMENT (Cont'd.)

(Cont'd.)

K.

Minimum Critical Power K.

Minimum Critical Power Ratio (MCPR)

Ratio (MCPR) i During steady state MCPR shall be determined operation at rated core daily during a reactor flow, MCPR shall be pos,ar operation at greater greater than or equal to; than or equal to 25% rated 1.34 for IN-1 8x8 and G.E.

the. mal power and following 8x8 Fuel types 1.35 for any change in power level G.I. 8x8R 1.38 for IN-1 9x9 LTA or distribution that would cause operation with a For core flows other than limiting control rod rated, the MCPR operating pattern as described in limit shall be as follows:

the bases for specification

{

• 3.3.B.5.

1.

Manual Flow Control -

the MCPR Operating Limit shall be the value from Figure 3.5-2 Sheet 1 or the above rated core i

3 flow value, which ever is greater.

2.

Automatic Flow Control -

the MCPR Operating Limit shall be the value from

~-~

Figure 3.5-2 Sheet 1, sheet 2 or the above rated core flow value, whichever is greatest.

If at any time during steady state power operation, it I '

is determined that the limiting value for MCPR is being exceeded; action shall be initiated within 15 minutes to restore operation to within the prescribed limits.

If+the

~

steady state MCPR is not returned to within the prescribed limits within two (2) hours, the reactor shall be brought to the Cold shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Surveil-lance and corresponding action shall continue until reactor operation is within the prescribed limits.

3/4.5-25 3687a

- 3123A

_n,..,,___

,,,_.u

~

DRESDEN II DPR-19

~

Amendment No. pf, 84,

3.5 LIMITING CONDITION FOR OPERATION 4.5 SURVEILLANCE REOUIREMENT (Cont'd.)

(Cont'd.)

In the eveat the' average 90% scram insertion time determined bJ Spec. S.3.C foe all operable control rods exceeds 2.74. seconds, the MCPR limit shall be

~

increased by the amount

, equal to 10.092T - 0.252]

where T equals the average l

90% ceram insertion tine for the most recent half-core or full core surveillance data from Spec. 4.3.C.

li L.

Condensate Pump Room L.

Condensate Pumo Room.

Flood Protaction Flood Protection 1.

The system is installed 1.

The following to prevent or surveillance t

mitigate the

\\

requirements shall be i

consequence's of observed to assure that flooding'of the the condensate pump condensate pump room room flood protection l

shall be operable is operable.

l prior to startup of the reactor.

a.

The testable penetra-tions through the walls of CC3W pump yaults shall be checked dur-j.

ing each operating cycle by pressurizing to 15 plus or minus 2 psig and checking for leaks using a soap bub-ble solution. The cel-teria for acceptance should be no visible leakage thro.ush the l

soap bubble solution.

The bulkhead door shall be checked during each operating cycle by hydrostatic-ally testing the door at 15 plus or minus 2 psis and checking to verify that leakage around the door is less than one gallon per hou 3/4.5-26 l

3687a

~ 3123A

...Lz..

DRESDEN II DPR-19 Ame:dmelt No. pd, 84 1.7 j

l.6 s

. N 1.5 N

~

~

~

s s

N s

N a.

1.4

--- ---)h-s s A

s 40 s

a

.N e

s u

s 0

1.3~-

N

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~

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E.

N o

s ad

\\

s g, -

g U

1.2

% ss,

% g s

1.1 N

I 1.1 30 40 50 60 70 80 90 100 Total Core Recirculation Flow (% Rated, 98 alb/hr)

MCPR Limit For Reduced Core Flow Fisure 3.5-2 (Sheet 1 of 2) 3/4.5-27

- 3687a 3123A 4

DRE3 DEN II DPR-19 a

j Amendment No. y, 84 t

1.8 bllllllllll s.e not. i 1.7 l,<,,

See Note 2

__', s, 1.6

~

g

,l 1.5 3,, gag, 3.

~

s aa y

s

,5 1.4

- -pg.

1

.et..

g

1) If MCPR Operating Limit = 1.4 at Rated conditions.

y y,3

2) If MCPR operating Limit = 1.3 at Rated conditions.

3) If HCPR Operating Lleit = l.2 at Rated 1.2 conditions.

- n.,

l.8 I ta l lJL ID uJ Li t t u1 LLL1J 11111U111U l t t i t i t tn U n t]. _

.],,_

10 20 30 40 50 60 70 80 90 100 Tota 1 Core Recirculating Flow (1 Rated, 98 alb/ht)

MCPR Limit For Automatic Flow Control Figure 3.5-2 (Sheet 2 of 2) 3/4.5-28 3687a 3123A

DRESDEN II DPR-19 Ame:dment N2. % 84 3.5 LIMITING CONDITION FOR OPERATION, 4.5 SURVIILLANCE REOUIREMENT (Cont'd.)

(Cont'd.)

b.

The CC3W Vault Floor drain shall be checked during each operating cycle by assuring that water can be run through the drain line and actuating the air operated valves by operation of the following sensor:

i.

loss of air

11. high level in the condensate pump room (5'0")

4 c.

The condenser pit five foot. trip shall have a trip setting of less than or equal to five feet zero inches. The five foot trip circuit for each channel s

shall be checked once every three

' months. The 3 and 1 foot alarms shall have a setting of less than or equal to three feet zero inches and less

. than or. equal to 1 foot 0 inches. A logic system functional test, including all alarms, shall be performed during the refueling outage.

l 3/4.5-29 3687a 3123A s

.,,n

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

DRESDEN II DPR-19 AmendmentN3.pt,84 3.5 LIMITING CONDITION FOR OPERATION 4.5 SURVEILLANCE REQUIREMENT (Cont'd.)

(Cont'd.)

2.

The condenser pit water level switches shall trip the condenser

. circulating water pumps and alarm in the control room if water level in the condenser pit exceeds'a level of 5 feet above the pit floor.

If a failure occurs in one of these trip and alarm circu1Es, the failed circuit shall be immediately placed in a trip condition and reactor operation shall be permissible for the following seven days unless the circuit is sooner made opsepble.

3.

If Specification 3.5.L.1 and 2 cannot be met, reactor startup shall not commence or if operating, an orderly shutdown shall be initiated and the reactor shall be in a cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

i 1

t 1

j i

l i

j 3/4.5-30 1

i l

3647a 3123A l

_ _. ~. _ _ _. _

DRESDEN II DPR-19 Amendment No. pd, 84 3.5 LIMITING CONDITION FOR OPERATION BASES A.

Core Sorav and LPCI Mode of the RHR System - This specification assures that adequate emergency cooling capability is available.

Based on the loss of coolant analyses included in References (1) and (2) in accordance with 10CFR50.46 and Appendix K, core cooling systems provide sufficient cooling to'the core to dissipate the energy associated with the loss of coolant I

accident, to limit the calculated peak clad temperature to less than 2200*F, to assure that core geometry remains intact, to limit the core wide clad metal-water reaction to less than 1%, and to limit the calculated local metal-water reaction to i

less than 17%.

The allowable repair times are established so that the average risk rate for repair would be no greater than the basic risk 4

rate. The method and concept are described in Reference (3).-

(

Using the results developed in this reference, the repair Period is found to be less than 1/2 the test interval.

This assumes that the core spray and LPCI ' subsystems constitute a 1 out of 3 system, however, the combined effect of the two systems to limittexcessive clad temperatures must also be considered.

The test interval specified in Specification 4.5 was'3 months.

Therefore, an allowable repair period which maintains the basic risk considering' single failures should be less than 45 days and this specification is within this period.

For multiple failures, a shorter interval is specified and to improve the assurance that the remaining (1) " Loss of Coolant Accident Analyses Report for Dresden Units 2, 3 and Quad-Cities Units 1, 2 Nuclear Power Stations," NED0-24146A, Revisions 1, April 1979.

i (2) NEDo-20566, General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10CFR50 Appendix K.

4 (3) APED " Guidelines for Determining Safe Test Intervals and Repair Times for Engineered Safeguards" - April 1969 I.M.

Jacobs and P.W. Marriott.

8 3/4.5-31 M

3687a

- 3123A

-t

..-,,.-,___._,..,y_.__

-, _ _ - - _ ~..

..m

DRESDEN II DPR-19 Amendment No. pd, 84 3.5 Lin111NG CONDITION FOR OPERATION RASES (Cont'd.)

systems will, function, a daily test is called for. Although it is recognized that the information given in reference 3 provides a quantitative method to estimate allowable repair times, the lack of operating data to support the analytical approach prevents complete acceptance of this' method at this time. Therefore, the times stated in the spec,1fic items were established with due regard to judgement.

Should one core spray subsystem become inoperable, the remaining core spray and the entire LPCI system are available should the reactor core cooling arise. To assure that the remaining core spray and LPCI subsystems and the diesel generators are available they are demonstrated to be operable immediately. This demonstration includes a manual initiation of the pumps and associated valves and diesel generators.

Based on judgements of the reliability of the remaining systems; i.e. the core spray and LPCI, a 7-day repair period was obtained.

~

Should the loss of one LPCI pump occur, a nearly full complement of' core and containment cooling equipment is available. Three.LPCI pumps in conjunction with the core

~

spray subsystem will perform the core cooling function.

Because of the availability of the majority of the core

~

cooling equipment, which will be demonstrated to be operable, a 30-day repair period is justified.

If the LPCI subsystem is not available, at least 2 LPCI pumps must be available to fulfill the containment cooling function. The 7-day repair period is set on this basis.

~~~

3.

Containment coolina Service Water - The containment heat l.,

removal portion of the LPCI/ containment cooling subsystem is provided'to remove heat energy from the containment in the event of a l'oss of coolant accident.

For the flow specified, the containment long-term pressure is limited to less than 8 psig and, therefore, is more than ample to provide the required heat removal capability.

(Ref. Section 5.2.3.2 SAR).

The containment cooling subsystem consists of two sets of 2

' service water pumps, 1 heat exchanger and 2*LPCI pumps.

Either set of equipment is capable of performing the containment cooling function. Loss of one containment cooling service water pump does not seriously jeopardize the containment cooling capability as any 2 of the remaining three pumps can satisfy the cooling requirements. Since there is some redundancy left a 30-day repair period is adequate.

Loss B 3/4.5-32 3687a 3123A

DRESDEN II DPR-19 Amendment No. yl, 84 3.5 LIMITING CONDITION FOR OPERATION BASES (Cont'd.)

of I containment cooling subsystem leaves one remaining system to perform the containment cooling function. The operable system is demonstrated to be operable each day when the above 4

~

condition occurs.

Based on the facts that when one containment cooling subsystem becomes inoperable only one

~

system remains which is tested daily. A 7-day, repair period was specified.

C.

Hiah Pressure Coolant In3ection - The high pressure coolant injection subsystem is provided to adequately cool the core for all pipe breaks smaller than those for which the LPCI or core spray subsystems can protect the core.

The HPCI meets this requirement without the use of off-site electrical power. For the pipe breaks for which the HPCI is i

intended to function the core never uncovers and is continuously cooled and thus no clad damage occurs.

(Ref.

Section 6.2.5.3 SAR).

The repair times for the limiting conditions of operation were set considering the use of the l

HPCI as part'of the isolation cooling' system.

D.

Automatic Pressure Relief - The relief valves of the automatic pressure relief subsystem are a back-up to the HPCI subsystem.

They enable the core spray or LPCI to provide protection against the small pipe break in t.he event of HPCI failure, by depressurizing the reactor vessel rapidly enough to actuate the core sprays or LPCI.

The core spray and/or LPCI provide sufficient flow of coolant to adequately cool the core.

Loss of 1 of the relief valves affects the pressure relieving capability and therefore a 7 day repair period is specified.

Loss of more than 1 relief valve significantly reduces the pressure relief capability and thus a 24-hour repair period is specified.

E.

Isolation Coolina System - The turbine main condenser is normally available.

The isolation condenser is provided for core decay heat removal following reactor isolation and scram.

The isolation condenser has a heat removal capacity sufficient to handle the decay heat production at 300 seconds following's scram. Water will be lost from the reactor vessel through the relief valves in the 300 seconds.following isolation and scram.

This represents a minor loss relative to the vessel inventory.

B 3/4.5-33 3687a

- 3123A l

DRESDEN II DPR-19 AmendmentNo.pd,84 3.5 LIMITING CONDITION FOR OPERATION PASES (Cont'd.)

The system may be manually initiated at any time. The system is automatically initiated on high reactor pressure in excess of 1060 psig sustained for 15 seconds.

The time delay is

. rovided to prevent unnecessary actuation of the system during p

anticipated turbine trips. Automatic initiation is provided to minimize.the coolant loss following isolation from the main condenser.

To be considered operable the shell side of the isolation condenser must contain at least 11,300 gallons of water. Make-up water to the shell side of the isolation i

condenser is provided by the condensate transfer pumps from the condensate storage tank. The condensate transfer pumps are operable from on-site power. The fire protection system is also"available as make-up water.

An* alternate method of i

cooling the core upon isolation from the main condenser is by using the relief valves and HPCI subsystem in a feed and bleed manner. Therefore, the~high pressure relief function and the

[

HPCI must be available together to cope with an anticipated transient so the LCO for HPCI and relief valves is set upon this function rather than their fun'etion as depressurization means for a small pipe break.

>+

s l

Emergency Coolina Availability - The purpose of Specification F.

D is to assure a minimum cf core cooling equipment is l

available at all times.

If, for example, one core spray were out of service and the diesel which powered the opposite core spray were out of service, only 2 LPCI pumps would be available.

Likewise, if 2 LpCI pumps were out of service and 2 containment service water pumps on the opposite-sida. wore also out of service no containment cooling would be available.

It is during refueling outages that major maintenance is performed and during such time that all low s

pressure. core cooling systems may be out of service. This specification provides that should this occur, no work will be performed on the primary system which could lead to draining the vessel. This work would include work on certain control rod drive components and recirculation system. Thus, the specification precludes the events which could require core cooling.

Specification 3.9 must also be consulted to

, determine other requirements for the diesel,senerators.

Dresden Units 2 and 3 share certain process systems such as the makeup domineralizers and the radwaste system and also some safety systems such as the standby gas treatment system, batteries, and diesel generators. All of these systems have been sized to perform their intended function considering the simultaneous operation of both units.

I

~

5 3/4.5-34 3687a 3123A

DRESDEN II DPR-19 AmendmentNo.p,84 3.5 LIMITING CONDITION FOR OPERATION BASES (Cont'd.)

~

For the safety related shared features of each plant, the Technical Specifications for that unit contain the operability and surveillance requirements for the shared feature; thus, i

the level of operability for one unit is maintained l

independently of the status of the other.

For example, the shared diesel (2/3 diesel) would be mentioned in the i

specifications for both Units 2 and 3 and even if Unit 3 were in the Cold Shutdown Condition and needed no diesel power, l

readiness of the 2/3 diesel would be required for continuing j

Unit 2 operation.

l f

C.

Specification 3.5.F.4 provides that should this occur,'no work 2

will be. performed which could preclude adequate emergency cooling capability being available. Work is prohibited unless it is in accordance with specified procedures which limit the period that the control rod drive housing is open and assures that the worst possible loss of coolant resulting from the work will not result in uncovering the reactor core.

Thus, i

this specification assures adequata core cooling.

Specificatica 3.9 must be consulted t'o determine other f.

requirements for the diesel generator.

~

Specification 3.5.'F.5 provides. assurance that an adequate supply of coolant water is immediately available to the low I

~~

{.

pressure core cooling systems and that the core will remain i

covered in the event of a loss of coolant accident while the reactor is depressurized with the head removed.

N.

Maintenance of Filled Discharge Pipe - If the discharge' piping of the core spray, LPCI, and NPCI are not filled, a water hammer can develop in this piping when the pump and/or pumps are started.

l 1.

Averate Plan'ar LNGR This specification assures that the peak cladding temperature following a postulated design basis loss-of-coolant accident will not exceed the'2200*F limit specified in 10CFR50 Appendix I

K considering the postulated affects of fuel pellet l

densification.

The peak cladding temperature following a postulated loss-of-coolant accident is primarily a function of the average LHCR of all the rods in a fuel assembly at any axial location and is only dependent secondarily on the rod to rod power distribution within a fuel assembly.

Since expected 4

B 3/4.5-35 3687a i

3123A

-,.--.---L.__-_

DRESDRN II DPR-19 Amendment N3. $d, 84 3.5.

LIMITING CONDITION FOR OPERATION BASES (Cont'd.)

localvariationsinpowerdistIributionwithinafuelassembly affect the ca.lculated peak clad temperature by less than plus or minus 20*F relative to the peak temperature for a typical fuel design, the limit on the average planar LHGR is sufficient to assure that calculated temperatures are below I

the 10CFR50, Appendix K limit.

The maximum average planar LHGRs shown in Figure 3.5.1 are based on calculations employing the models described in Reference (1) and in reference (2).

Power operation with APLHCRs at or below those shown in Fig. 3.5.1 assures that the peak cladding temperature following a postulated loss-of-coolant accident will not exceed the 2200*F limit.

The maximum average planar LHCRs for G.E. fuel plotted in Fig.

3.5.1 at higher exposures result in a calculated peak clad temperature of less than 2200*F.

However, the maximum average planar LHGRs are shown on Fig. 3.5.1 as limits because conformance calculations have not been performed to justify operation at LHGRs in excess of tho'se.shown.

l J.

Local LHGR This specification assures that the maximum linear h'est generation rate in any fuel rod fabricated,by G.E.' is less than the design linear heat generation rate even if fuel pellet densification is postulated.

For fuel fabricated by ENC, protection of the MCPR and NAPLHOR limits and operation within the power distributio'n'asssaptions of the Fuel Design Analysis provides adequate protection against eladding strain limits, hence the LHGR limitation for,

GE fuel is unnecessary for the protection of ENC fuel.

(1)

" Loss of coolant Accident Analyses Report for Dresden Units 2, 3 and Quad-Cities Units 1, 2 Nuclear Power Stations," NEDO-24146A, Revision 1. April, 1979.

(2)

IN-NF-82-88 "Dresden Unit 2 LOCA Analysis Using the INC EXEN/3WR tvaluation Nodel MAPLNGR Results" 3 3/4.5-36

'3687a 3123A e

__.=. - - -

1 DRRSDRN II DPR-19 AmendmentN3.pd,84 1

l f

3.5 LIMITING CONDITION FOR OPgRATION BASRS (Cont'd.)

i l

K.

Minimum Celtical Power Ratio (MCPR) 4 The steady-state values for NCPR specified in the specifica-4 tion were determined using the THRRMEX thermal limits methodology described in IN-NF-80-19, Volume 3.

The safety limit implicit in the operating limits is established so that during sustained operation at the NCPR safety" limit, at least 99.9% of the fuel rods in the core are expected to avoid boiling transition.

The Limiting Transient delta CPR implicit in the operating limits was calculated such that the occur-rence of the limiting transient from the operating limit will i

i not result in violation of the NCPR safety limit in at least 95% of the random statistical combinations of uncertainties.

l l

I Transient events of each type anticipated during operation of a BWR/3 were evaluated to determine which is most restrictive l

in terms of thermal margin requirements. The generator load rejection / turbine trip without bypass is typically the limit-ing event.

The thermal margin effects of the event are evaluated with the THRRMIX Nothodology and appropriate NCPR limits consistent with the IN-3 critical power correlation are 1

i determined.

Several factors influence which transient results

.in the largest reduction in critical power ratio, such as the cycle-specific fuel loading, exposure and fuel type. The cur-i rent cycle's reload licensing analyses identifies the limiting L

I transient for that cycle.

i As described in Specification 4.3.C.3 and the associated

{

l Bases, observed plant data were used to determine the average scram performance used in the transient analyses for determin-ing the NCPR operating Limit.. If the current cycle scram time

~

performance falls outside of the distribution assumed in the analyses,' en adjustment of the NCPR limit may be required to I

maintain margin to the NCPR Safety Limit during transients.

compliance with the assumed distribution and adjustment of the i

BCPR operating Limit will be performed in accordance with Technical Specifications 4.3.C.3 and 3.5.K.

For core flows less than rated, the NCPR operating Limit l

established in the specificatLon is adjusted to ' provide '

protection of the McPR safety Limit in the event of an f

i uncontrolled recirculation flow increase to the physleal limit i

of pump flow. This protection is provided for manual and L

l automatic flow control by choosing the NCPR operating limit as the value from Figure 3.5-2 sheet 1 or the rated core flow value, whichever is greater.

For Automatic Flow Control, in addition to protecting the NCPR Safety Limit during the' flow a 3/4.5-37

_ 3637a i

3123A

DRESDEN II DFl-19 Amendme:tN3.pd,84 3.5 LIMITING CONDITION FOR OPERI. TION BASES (Cont'd.)

run-up event', protection is provided against violating the rated flow NCPR Operating Limit during an automatic flow increase to rated core flow. This protection is provided by the reduced flow NCPR limits shown in Figure 3.5-2 Sheet 2 where the curve corresponding to the current rated flow NCPR limit is used (linear interpol.. tion between the NCPR limit lines depicted is permissible)

Therefore, for Automatic Flow control, the MCPR Operating Limit is chosen as the value from Figure 3.5-2 Sheet 1, Sheet 2 or the rated flow value, whichever is greatest.

It should be noted that if the rated flow NCPR Limit must be increased due to degradation of control rod scram times during the current cycle, the new value of the rated flow NCPR limit is applied when using Figure 3.5-2 Sheet 2.

L.

Flood Protection Condensate pump room flood protection will assure the availability of the containment cooling service water system (CCSW) during a postulated incident'of flooding in the turbine building. The redundant level switches in the condenser pit will preclude 'any, postulated flooding of the turbine building to an elevation above river water level. The level switches

. provide alarm and circulating water pump trip in the event a water level is detected in the condenser pit.

s B 3/4.5-38 3487a 3123A

.m

1 DRESDEN II Dpt-19 Ame:dmentN3.pd,84 4.5 SURVEILLANCE REQUIREMENT BASES t

(A thru F)

The testing interval for the core and c'ontainment coolins systems is based on quantitative reliability analysis, judgement and practicality.

The core cooling systems have not been designed to be fully testable during oper,ation.

For example the core spray final admission valves do not open untti reactor pressure has fallen to 350 psis thus during operation even if high drywell pressure were stimulated the final valves would not open.

In the case of the HPCI, automatic initiation during power operation would result in pumping cold water into the reactor ves'sel which is not desirable.

i The systems can be automatically actuated during a refueling i

outage and this will be done. To increase the availability of the individual components of the core and containment cooling

• systems the components which make up the system i.e.,

l Instrumentation, pumps, valve operators, etc., are tested more i

frequently. The instrumentation is functionally tested each month.

Likewise the pumps and motor-operated valves are also s

tested each month'.to assure their operability. The combination of a yearly simulated automatic actuation test and monthly' tests of the pumps and valve operators is deemed to be adequate testing of these systems.

With components or subsystems out-of-service overall core and containment cooling relisbility is maintained by demonstrating the operability of the remaining coollas equipment. The degree of operability to be demonstrated depends on the nature of the reason for the out-of-servlee equipmeat.

For routine out-of-service periods caused by preventative maintenance,

~

etc., the pump and valve operability checks will be performed to demonstrate operability of the remaining components.

Nowever, it' a failure, destga defielency, etc., caused the i

out-of-service period, them the demonstration of operability should be thorough enough to assure that a similar problem does not exist on the remalains components. For example, if I

an out-of-service period were caused by failure of a pump to i

deliver rated capacity due to a design deficieriy, the ot'her i

pumps of this type might be subjected to a flow rate test in addition to the operability checks.

1 The requirement of 180 psig at 3500 spo at the containment cooling service water (CCSW) pump discharge provides adequate I

margin to ensure that the LpCI/CCSW system provides the design r

B 3/4~.5-39 l

I 3487a i

3123A en

DRdSDEN II DPR-19 d

Amendment N3. 9, 84 4,. 5 SURVEILLANCE REQUIREMENT BA333 (Coht'd.)

bases cooling water flow and maintains 20 psig differential pressure at the containment cooling heat exchanger.

This l

a differential pressure precludes reactor coolant from entering l

the river water side of the containment cooling heat exchangers.

The verification of Main Steam Relief Valve operability during manual actuat' ion surveillance testing must be made independent of temperatures indicated by thermocouples downstream of the relief valves.

It has been found that a temperature increase i

may result,with the valve still closed.

This is due to steam being vented through the valve actuation mechanism during the surveillance test.

By first opening a turbine bypass valve, and then observing its closure response during relief valve actuation, positive verification can be made for the relief i

T valve opening and passing steam flow.

Closure response of the.

l turbine control valves during relief valve manual actuation would likewise serve as an adequate; verification for relief i

valve opening. This test method may be performed over a wide range of reactor pressure greater than 150 psis. Valve i-operation below 150 psis is limited by the spring tension exhibited by the relief valves.

G.

Deleted j

H.

Maintenance of Filled Discharae Pine i

The surveillance requirements to assure that the discharge i

piping of the core spray, LPCI, and NPCI systems are filled provides for a visual observation that water flows from a high -

lN point vent.

This ensures that the line is in a full con-dition.

Between the monthly intervals at which the lines are vsnted, instrumentation has been provided to monitor the presence of water in the discharge piping. This instrumenta-tion will be calibrated on the same frequency as the safety system instrumentation. This period of periodic testing I

ensures that during the intervals between the monthly checks the status of the discharge piping is monitored on a con-

.tinuous basis.

I.

Averaste Planar LMGR.

At core thermal power levels less than or equal to 25 per cent, operating plant experience and thermal hydraulic analyses indicate that the resulting average planar LHGR is below the maximum average planar LHGR by a considerable, l

a 3/4.5-40 3687a 3123A a

DRESDEN II DPR-19 Amendment ND. pd, 84 4.5 SURVEILLANCE REQUIRENENT BASES (Cont'd.)

margin; therefore, evaluation of the average planar LHCR below this power le' vel is not necessary. The daily requirement for calculating average planar LHCR above 25 per cent rated thermal power is sufficient since power distribution shifts are slow when there have not been significant power or control rod changes.

J.

Local IRGE The LHCR for G.E. fuel shall be checked daily during reactor operation at greater than or equal to 25 per cent power to determine if fuel burnup or control rod movement has caused changes in power distribution. A limiting LHGR value is precluded by a considerable margin when employing a permissible control rod pattern below 25% rated thermal power.

K.

Minimum Critical Power Ratio (MCPR)

At core thermal power levels less tha.n or equal to 25 per cent, the reactor will be operating at minimum recirculation pump speed and the moderator void content will be very small.

For all designated control rod patterns which may be employed at this point, operating plant experience and thermal hydraulic analysis indicates that the resulting MCPR value'is

~~"

in excess of requirements'by a considerable margin. With this low void content, any inadvertent core flow increase would only place operation in a more conservative mode relative to McFR.

The daily requirement for calculating McPR above 25 percent rated thermal power is sufficient since power distribution shifts are very slow when there have not been significant power or' con, trol rod changes.

In addition, the kg correction applied to the LCo provides margin for flow increase from low flows.

l L.

Flood Protection l

l The watertight bulkhead door and the penetration seals for pipes and cables penetrating the vault walls have been designed to withstand the maximum flood conditions. To assure that their installation is adequate for maximum flood conditions, a method of testing each seal has been devised.

B 3/4.5-41 3647a 3123A L

DRESDEN II DPR-19 Amendme t N2. pd,84 4.5 SURVEILLANCE REOUIREMENT BASES (Cont'd.)

To test a pipe seal, another test seal is installed in the opposite side.of the penetration creating a space between the two seals that can be pressurized.

Compressed air is then supplied to a fitting on the test seal and the. space inside the sleeve is pressurized to approximately 15 psi. The outer face of the permanent seal is then tested forl leaks using a soap bubble solution.

On completion' of the test. the test seal is removed for use on other pipes and penetrations of the same size.

In order to test the watertight bulkhead doors, a test frame must be installed around each door. At the time of the test, a reinforced steel box with rubber gasketing is clamped to the wall around the door.

The fixture is then pressurized to approximately 15 psis to test.for leak tightness.

Floor drainage of each vault is accomplished through a carbon steel pipe which penetrates the vault. When open, this pipe will drain the vault floor to a floor drain sump in the condensate pump room.

Equipment drainss'e from the vault coolers and the CCSW pump bedplates will also.be routed to the vault floor drains.

The old. equipment drain pipes will be permanently capped to pre-clude the possibility of back-flooding the vault.

As a means of preventing backflow from outside the vaults in the event of a flood, a check valve and an air operated valve are installed in the 2" vault floor drain line 6'0" above the floor of the condensate pump room.

f The check valve is a 2" swing check designed for 125 psig

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service.

The air operated valve is a control valve designed for a 50 psi differential pressure.

The control valve will be in the normally open position in the energized condition and will close upon any one of the following:

a.

Loss of air or power b.

High level (5'0") in the condensate' pump room Closure of the air o'perated valve on high water level in the condensate pump room is effected by use of a level switch set at a water level of 5'0".

Upon actuation, the switch will close the control valve and alarm in the control room.

l B 3/4.5-42 3647a

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- DRESDEN II DPR-19

/

Amendment No. J, 84 4.5 SUEVEILLANCE REQUIREMENT BASES (Cont'd.)

The operator will also be aware of problems in the vaults /

condensate penp room if the high level alarm on the equipment drain sump is not terminated in a reasonable amount of time.

It must be pointed out that these alarms provide information to the operator but that operator action upon the above alarms is not a necessity for reactor safety since the other provisions provide adequate protection.

A system of level switches has been installed in the condenser pit to indicate and control flooding of the condenser area.

The following switches are installed:

Level Function I

a.

l'0" (1 switch)

Alarm, Panel l

Hi-Wat2r-Condenser Pit b.

3'0" (1 switch)

Alarm, Panel High-Cire.

• Water Condenser Pit c.

5'0" (2 redundant Alarm and Circ. Water Pump I

• switch pairs)

Trip 1

l Level (a) indicates water in the condenser pit from either the hotwell or the circulating water system.

Level (b) is above the hotwell capacity and indicates a probable circulating j

water failure.

l Should the switches at level (a) and (b) fail or the operator fall to trip the circulating water pumps on alarm at level (b), the actuation of either level switch pair at level (c) shall trip the circulating water pumps automatically and alarm.

in the control room.

These redundant level switch pairs at 1evel (c) ate designed and installed to IEEE-279, " Criteria for Nuclear Power Plant Protection Systems." As the cir-culating water pumps are tripped, either manually or auto-matica11y, at level (c) of 5'0", the maximum water level reached in the condenser pit due to pumping will be at the 491'0" elevation (10' above condenser pit floor elevation 481'0"; 5' plus an additional 5' attributed to pump coastdown).

l In order to prevent overheating of the CCsw pump motors, a vault cooler is supplied for each pump.

Each vault cooler is l

designed to maintain the vault at a maximum 105'F temperature during operation of its respective pump.

For example, if CCSW pump 28-1501 starts, its cooler will also start and compensate B 3/4.5-43 3687a i

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DRESDEN II DPR-li AmendmentNo.pd,84 t

4.5' SURVEILLANCE REOUIRE?ENT BASES (Copt'd.)

for the heat supplied to the vault by the 2B pump motor keeping the v'ault at less than 105*F.

Each of the coolers is supplied with cooling water from its respective pump's discharge line.

After the water has been passed through the cooler, it returns to its respective pump's suction line.

In this way, the vault coolers are supplied with cooling water totally inside the vault. The cooling water quantity needed for each cooler is approximately 1% to 5% of the design flow of the pumps so that the recirculation

'of this small amount of heated water will not affect pump or cooler cperation.

operation of the fans and coolers is required during pump 4

operability testing and thus additional surveillance is not required.

Verification that acc'ess doors to each vault are closed, following entrance by personnel, is: covered by station operating procedures.

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B 3/4.5-44 3687a

-3123A

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DRESDgN II DPR-19 AmendmentNo.pd,84

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5.0 DESIGN FEATURES 5.1 Site Dresden Unit 2 is' located at the Dresden Nuclear Power Station which consists of a tract of land of approximately 953 acres located in the northeast quarter of the Morris 15-minute quadrangle (as designated by the United States Geological Survey),

Goose Lake Township, Grudy County, Illinois.

The, tract is situated in portions of Sections 25, 26, 27,,34, 35, and 36 of Township 34 North, Range 8 East of the Third Principal Meridian.

5.2 Reactor A.

The core shall consist of not more than 724 fuel assemblies B.

The reactor core shall contain 177 cruciform-shaped control rods. The control material shall be boron carbide powder (B C) compacted to approximately 70% of theoretical density, 4

or Hafnium metal.

5'. 3 Reactor Yessel E

The reactor vessel shall be as described in Table 4.1.1 of the SAR.

The. applicable' design codes shall be as described in Table 4.1.1 of the SAR.

5.4 Containment A.

The principal design parameters and applicable design codes for the primary containment shall be as given in Table _,5.2.1 of the SAR.

B.

The secondary containment shall be as described in Section

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5.3.2 of,the SAR and the applicable codes shall be as described in Section 12.1.1.3 of the SAR.

C.

Penetrations to the primary containment and piping passing through such penetrations shall be designed in accordance with standards set forth,in Section 5.2.2 of the SAR and the applicable codes shall be as described in Section 12.1.1.3 of j

the SAR.

l l

5.5 Fuel Storage A.

The new fuel storage facility shall be such that the K gy dry is less than 0.90 and flooded is less than 0.95.

l B.

The K,gg of the spent fuel storage pool shall be less than or equal to 0.95.

l 5-1 3696a 3124A

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