ML20059M844

Amends 113 & 109 to Licenses DPR-19 & DPR-25,respectively, Incorporating ATWS Requirements Into Tech Specs

ML20059M844
Person / Time
Site:	Dresden
Issue date:	10/02/1990
From:	Barrett R Office of Nuclear Reactor Regulation
To:
Shared Package
ML20059M847	List: ML20059M844
References
NUDOCS 9010050325
Download: ML20059M844 (35)

Text

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NUCLEAR REGULATORY COMMISSION j

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COPMONWEALTH EDISON COMPANY i

-DOCKET NO. 50-237 DRESDEN NUCLEAR POWER STATION, UNIT 2 AMENDMENT TO PROVISIONAL OPERATING LICENSE Amendment No. 113 License No. DPR-19 1.

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

A.

The application.for amendment by the Commonwealth Edison Company (the licensee) dated September 29, 1989, and supplemented by a February.1,1990 submittal, complies with the standards and require-ments 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.

Thereisreasonableassurance(1)thatthe-activitiesauthorized by this amendment can be conducted without endangering-the health i

and safety of.the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; t

D.

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

i 2..

Accordingly the license is amended by changes to the Technical Specifi-cationsasIndicatedintheattachmenttothislicenseamendmentand paragraph 3.B. of Provisional Operating License No. DPR-19 is hereby amended to read as follows:

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Technical Specifications The Technical Specifications contained in A)pendix A, as revised through Amendment No.113, are here)y incorporated-in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective no later than 60 days from the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Rick (r

. Barret Director Proje Directorate III-2 Division of Reactor Projects - III, IV, V and Special Projects Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: October 2, 1990 J

l ATTACHMENT T0 LICENSE AMENDMENT NO. 113 PROVISIONAL OPERATING LICENSE NO. DPR-19 DOCKET NO. 50-237 Revise the Appendix A Technini Specif wations by removing the pages identified below and inserting the attached pages. The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change.

REMOVE INSERT 11 11 vii vii viii viii 3/4.2-7a 3/4.2-18a 3/4.2-27a B 3/4.2-33a B 3/4.2-33a B 3/4.2-37 B 3/4.2-37 j

3/4.4-1 3/4.4-1

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3/4.4-2 3/4.4-2 3/4.4-3 3/4.4-3 3/4.4-4 3/4.4-4 B 3/4.4-6 B 3/4.4-6 B 3/4.4-7 B 3/4.4-7 1

.r' DRESDEN II DPR-19 Amendment No. 113 Table of Contents Pale 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 1/2.1-101

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 3.2.8 Core and Containment Cooling Systems - Initiation and Control 3/4.2-1 3.2.C Control Rod Block Actuation 3/4.2-2 3.2.0 Refueling Floor Radiation Monitors 3/4.2-2 3.2.E Post Accident Instrumentation 3/4.2-3 i

3.2.F Radioactive Liquid Effluent Instrumentation 3/4.2-4 i

3.2.G Radioactive Gaseous ~ Effluent Instrumentation 3/4.2-5 3.2.H Recirculate Pump Trip-3/4.2-7a 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

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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 3.3.0 Control Rod Accumulators 3/4.3 3.3.E

. Reactivity Anomalies 3/4.3-12'

~3.3.G Economic Generation Control System

. B 3/4.3-14 3/4.3-13 l

Limiting Conditions for Operation Bases (3.3)

Surveillance Requirement Bases (4.3)

B 3/4.3-21 3.' 4

• Standby Liquid Control System 3/4.4-1 3.4.A Normal Operation 3/4.4 3.4.B Operation with Inoperable Components 3/4.4-2 3.4.C Liquid Poison Tank 3/4.4-3 3.4.0 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 11 t

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l DRESDEN II DPR-19 Amendment No. 113 List of Tables

.P,a2e 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 r

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 3.2.6 Post Accident Monitoring Instrumentation Requirements 3/4.2-17 Table 3.2.7 Instrumentation That Initiates Recirculation Pump Trip 3/4.2-18a Table 4.2.1 Minimum Test and Calibration Frequency for Coro and Containment Cooling Systems i

Instrumentation, Rod Blocks, and Isolations 3/4.2-19 Table 4.2.2 Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements 3/4.2-22 Table 4.2.3 Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements 3/4.2-24

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Table 4.2.4 Post Accident Monitoring Instrumentation Surveillance Requirements 3/4.2 Table 4.2.5 Minimum Test and Calibration Frequency for the Recirculation Pump Trip 3/4.2 '7a Table 4.5.1 Surveillance of the HPCI Subsystem 3/4.5-la 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 t

l 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

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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 Surveillance Requirements for High Energy Piping Outside Containment 3/4.11-3 vii

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DRESDEN II DPR-19 Amendment No. 113 List of Tables (continued)

P_agg 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 CO Systems B 3/4.12-19 t

2 Table 3.12-4 Fire Hose Stations B 3/4.12-20 & 21

. Table 6.1.1 Minimum Shift Manning Chart 6-4 1

Table 6.6.1 Special Reports 6-23 i

List of Figures L

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 an Adequate Interval Between Tests B 3/4.1-18 Figure 4.2.2 Test Interval vs. S.vstem' Unavailability B 3/4.2-38 Figure 3.4.1 Deleted 3/4.4-4 Figure 3.4.2 Sodium Pentaborate Solution Temperature Requirements 3/4.4-5

' Figure 3.6.1

, Minimum Temperature Requirements per Appendix G of 10 CFR S0 3/4.6-23 Figure 3.6.2 Thermal Power vs. Core Flow Limits for Thermal Hydraulic Stability Surveillance in-Single Loop Operating 3/4.6-24 t

l Figure 4.6.1 Minimum Reactor Pressurization Tempera'.ure B 3/4.6-29 Figure.4.6.2 Chloride Stress Corrosion Test Results at 500 F B 3/4.6-31' Figure'4.8.1 Owner Controlled / Unrestricted Area Boundtry B 3/4.8-38 l

Figure 4.8.2 Detail of Central Complex B 3/4.8-39' Figure 6.1-1 Offsite'0rganization - Deleted Figure 6.1-2 Station Organization - Deleted viii

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DRESDEN II.

DPR-19.

Amendment No. 113 3.2 LIMITING CONDITION FOR OPERATION 4.2 SURVEILLANCE REQUIREMENTS.

(CONT'D)

(CONT'D)

H.--Recirculation Pump Trip H.

Recirculation Pump Trip Initiation Initiation l

The recirculation pump trip Instrumentation and logic system, initiated by low low systems shall be function-reactor water level or high ally tested and calibrated reactor pressure, limiting as indicated in Table 4.2.5.

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. conditions for operation are specified in Table 3.2.7.

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wr DRESDEN II DPR-19 Amendment No. 113 TABLE 3.2.7 INSTRUMENTATION THAT INITIATES RECIRCULATION PUMP TRIP Minimum Number of Trip Function Operable or Tripped Trip Level Setting Applicable 4.ction Instrument Channels Operational Per Trip System (a)

Mode High Reactor 2.

(GT/E) 1230 psig &

1 (d) 70 Pressure (LT/E) 1250 psig Low Low (GT/E) 84 inches 1 (d) 70 Reactor Water 2

above top of active Level fuel (b)(c)

Action 70 The minimum number of operable trip systems shall be four, two high reactor pressure and two low low reactor water level, except i

that one trip system may be inoperable for up to fourteen days.

I If one trip _ system is inoperable for greater.than fourteen days, or if any two trip systems are made or found to be inoperable, the

-i reactor must be placed in at least the Startup/ Hot Standby Mode in

.the.next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

NOTE:

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

required surveillance without placing the trip system-in the tripped

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-condition provided at least one operable channel in the same trip system is monitoring that parameter.

- b.

Top of active fuel is defined to be 360 inches above vessel zero.

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The trip will occur following a (GT/E) 8 & (LT/E) 10 second delay.

c.

d.

MODE 1 is the RUN MODE-3/4.2-18a

DRESDEN II.

DPR-19 Amendment No. 113 Table 4.2.5 MINIMUM TEST AND CALIBRATION FREQUENCY FOR THE:

RECIRCULATION PUMP TRIP i

Instrument Channel Instrument Instrument ApplicableL

. Functional Calibration Check Operational Test Mode

~ 1.

Reactor High Pressure Q

R D

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2.

Reactor Low Low Water Q

R D

1*

Level L

• MODE 1 is the RUN MODE 1

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.DRESDEN II DPR-19 Amendment No. 113 3.2 LIMITING CONDITION'FOR OPERATION BASES (Cont'd.)

without being affected by normal voltage fluctuations due to pumps starting.. Reset of the relay, approximately 11% above the trip point, indicates that the diesel generator has restored bus voltage and will accept ECCS loads.

The reset signal provides-a permissive for starting ECCS pumps.

The setting for 4KV emergency bus degraded voltage is chosen to detect sustained degraded voltage which may cause equipment ~ damage, while preventing trips caused by voltage fluctuations.

The-reset point for degraded voltage indicates restoration of normal bus voltage.

The recirculation pump trip (RPT) system is required by 10 CFR 50.62 to mitigate the consequences of an Anticipated Transient Without Scram (ATWS).

The design of this system meets the intent of the NRC Safety Eval-uation for NEDE-31096-A (Reference 1).

RPT is provided to minimize reactor pressure in the highly unlikely event of a plant transient coincident with the failure of all control rods to scram.

The rapid flow reduction in-creases core voiding providing negative reactivity.

RPT is only required to operate in the RUN mode since at lower power levels the safety relief valves have sufficient capacity to relieve the steam which continues to be produced during this postulated event.

The low low reactor water level trip includes-a nine second delay to avoid increasing the consequences of a postulated design basis loss of coolant accident.

References 1.

NEDE-31096-A, " Anticipated Transients Without Scram; Response to NRC ATVS Rule,'10 CFR 50.62," dated February 1987.-

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DRESDEN II DPR-19 Amendment No. 113

'4.2.

SURVEILLANCE REQUIREMENT BASES (Cont'o.)

i For instruments 2-263-73A, 73B and 2-2352, 2353,- the logic downstream of the output relay contacts exhibits a one-out-of-two logic and, by utilizing the Availability Criteria identified in NE00-21617-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 1

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 i

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

The analog trip units which provide the initiation signal for RPT are calibrated quarterly.

The reactor pressure and'1evel transmitters which provide the input to the analog trip units are calibrated every refueling outage, concurrent with logic tests that ensure overall system operability. An instrument check is performed on a daily basis. -These frequencies are considered appropriate, commensurate with the' design application and the fact that the RPT system'is a backup to existing.

protective instrumentation.

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DRESDEN II DPR-19 Amendment No. 113 3.4 LIMITING CONDITION FOR OPERATION 4.4 SURVEILLANCE REQUIREMENT r

STANDBY LIQUID CONTn L SYSTEM STANDB1 LIQUID CONTROL SYSTEM Applicability:

Applicabi14ty:

Applies to the operating Applies to the periodic testing status of the standby liquid requirements for the standby control system.

liquid control system.

Objective Objective:

To assure the availabilAty of To verify the operability of an independent reactivity the standby liquid control control mechanism, system.

Specification:

Speeffication:

A.

Normal Operation A.

Normal Operation During periods when fuel The operability of the is in the reactor the standby liquid control standby liquid control system shall be verified system shall be operable by performance of the i

except when the reactor following tests is in the Cold Shutdota Condition and all control 1.

At least once per rods art fully inserted month -

and Specification 3.3.A is met or as specified in Demineralized water 3.4.B below, shall be recycled to the i

test tank. Pump minimum i

flow rate of 40 gpm shall be erified against a rivvem head of 1275 psig.

2.

At least otwq juring each operatink cycle 8

a. Manually initiate the system, except the explosion valves and pump solution in the t

recirculation path, to L

demonstrate that the i

pump suction line from the storage tank is not plugged.

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DRESDEN II DPR 29 l

Amendment No. 113 3:4 LIMITING CONDITION FOR OPERATION 4.4 SURVETLLANCE REQUIREMENT (Cont'd.)

TCont'd.)

b.

Actuate one of the two standby liquid l

control systems using the normal

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actuation switch and pump i

desineralized water i

into the reactor vessel. Pump 4

minimum flow rate shall be verified against a previous test at the same reactor vessel pressure.

The replacement charges will be selected from a batch from which at least one charge has been successfully test fired and which will not exceed five years life when their use is terminated. Both I

systems shall be tested and inspected, including each explosive actuated valve, in the course of two operating cycles.

c.

Test that the setting l

4 of the system pressure relief valves is be-tween 1455 and 1545 psig.

B.

Operation with Inoperable B.

Surveillance with Inoperable Components Components From and after the date When a component becomes that a redundant component inoperable its redundant 3/4.4-2

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DRESDEN II DPR-19 Amendment No. 113 3.4 LIMITING CONDITION FOR OPERATION 4.4 SURVEILLANCE REQUIREMENT (Cont'd.)

(Cont'd.)

i is made or found to be component shall be inoperable, Specification demonst sted to be 3.4.A shall be considered operable immediately and fulfilled, and continued daily thereafter.

operation permitt. d provided that the component is returned to an operable condition within 7 days.

C.

The liquid poison tank shall C.

The availability of the contain a boron bearing solu-proper boron bearing tion of at least 3605 gallons selution shall be verified (3329 gallons to meet shutdown by performance of the j

requirements plus 376 gallons following tests:

i that are contained below the pump suction) of at least 14 wt.

1.

At least once per percent sodium pentaborate month - Boron concentra-decahydrate (Na2 200 c-10H O).

tion shall be determined.

B 2

At all time when the standby In additien, the boron liquid control system is concentration shall be required to be operable, the determined any time water solution temperature including or boron are added or that in the pump suction pip-if the solution tempera-ing shall not be less than ture drops below the t

the temperature presented limits specified by in F2gure 3.4.2 Figure 3.4.2.

Minimum concentration is 14 wt. percent sodium i

pentaborate.

2.

At least once per day -

Solution volume shall be checked.

3.

At least once per day -

The solution temperature shall be checked.

D.

If specification 3.4.A through C are not met, an orderly shutdown shall be initiated and the reactor r

shall be in the cold shutdown condition within l

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

3/4.4-3 L

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DRESDEN II DPR-19 Amendment No. 113 THIS PAGE WAS LETT INTENTIONALLY BLANK e

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ORESDEd II DPR 19 Amendment No. 113 3.4 LIMITING CONDITION FOR OPERATION BASES i

A.

The design objective of the standby liquid control system is to provide the capability of bringing the reactor from full power to a cold, xenon-free shutdown assuming that none of the withdrawn control rods can be inserted.

To meet this objective, the liquid control system is designed to inject a quantity of boron which produces a concentration of no less than 600 ppm of boron in the reactor core in less than 100 minutes.

600 ppm boron concentration in the reactor core is required to bring the reactor from full power to a 3% delta k or more subcritical condition considering the hot to cold reactivity swing and xenon poisoning.

An additional margin of 25% boron is added to compensate for possible imperfect mixing of the chemical solution in the reactor water, making the total concentration 750 ppm of boron in the reactor core. A minimum quantity of 3329 gallons of solution having a 14 wt. percent sodium pentaborate concentration is required to meet this shutdown requirement.

For a required pumping rate of 40 gpm, 3329 gallons per pump of at least 14 wt. percent solution will be inserted in less than 100 min-utes.

This insertion rate of boron solution will override the rate of reactivity insertion due to cool down of the reactor following the xenon peak.

Two pump operation will enable faster reactor shutdown for ATWS events.

The minimum volume required in the storage tank shall be 3605 gallons (276 are contained below the pump suction and, therefore, cannot be inserted).

The monthly pump minimum flowrate test shall require a minimum flowrate of 40 gpm per pump.

This i

requirement, combined with the solution concentration requirement of at least 14 wt. percent, will demonstrate that the Standby Liquid s

Control System meets the requirements of 10 CFR 50.62.

Boron concentration, solution temperature, and volume are checked on a frequency to assure a high reliability of operation of the system should it ever be required.

Experience with pump operability indi-cates that monthly testing is adequate to detect if failures have occurred.

Components of the system are checked periodically as described above and make a functional test of the entire system on a frequency of less than once during each operating cycle unnecessary.

A test of one installed explosive charge is made at least once during each operating cycle to assure that the charges have not deteriorated, the actuation circuit is functioning properly, the valve functions properly, and no flow blockages exist.

The replacement charge will be selected from a batch for which there has been a successful test firing.

Reccmmendations of the vendor shall be followed in main-taining a five year life of the explosive charges.

A continual check of the firing circuit continuity is provided by pilot lights in the control room.

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Di', DEN 11 DPR-19 Amsadment No. 113

3. 4 LIMITING CONDITION FOR OPERATION BASES (Cont'd.)

I The relief valves in the standby liquid control system prc+ect the system piping and positive displacement pumps which are no.inally designed for 1500 psig protection from over-pressure.

The pressure relief valves discharge back to the standby liquid control solution I

tank.

B.

Only one of the two standby liquid control pumping circuits is needed for proper operation of the system.

If one pumping circuit is found to be inoperable, there is no immediate threat to shutdown capability, and reactor operation may continue while repairs are being made.

Assurance that the remair.'ng system will perform its intended func-tion and that the reliability of the system is good is obtained by demonstrating operation of the pump in the operable circuit at least I

once daily.

C.

The solution saturation temperature of 14 wt. percent sodium penta-borate, is 62'F.

To guard against boron precipitation, the solution including that in the pump suction piping is kept at least 10'F above the saturation temperature by a tank heater and by heat tracing in the pump suction piping.

The 10'F margin is included in Figure 3.4.2.

l Temperature and liquid level alarms for the system are annunciated in the control room.

Pump operability is checked on a frequency to assure a high reliabil-ity of operation of the system should it ever be required.

Once the solution has been made up, boron concentration will not vary unless more boron or more water is added.

Level indication and alarm indicate whether the solution volume has changed which might indicate a possible solution concentration change.

Considering these factors, the test interval has been established.

4.4 SURVEILLANCE REQUIREMENT BASES Periodic tests to demonstrate two pump flow capability are not feasible in the present system configuration and are unnecessary because the flow

~

path integrity can be determined from the test of a single pump.

Compari-son of single pump test pressures with previous results and correlation of these data with initial two pump test will be used to verify the capa-bility of the piping to support two pump flow.

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UNITE D STATES

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- /g NUCLE AR REGULATORY COMMISSION

> s-WASHINGTON. D C. 20%6

\\.'...f COMMONWEALTH EDISON COMPANY l

DOCKET NO. 50-249 DRESDEN NUCLEAR POWER STATION. UNIT 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.109 License No. DPR-25 1.

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

A.

The application for amendment by the Comonwealth Edison Company (thelicensee)datedSeptenber29 1989, and supplemented by a February 1,1990 submittal, comp 1Ies with the standards and require-1 ments 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 o>erate in corformity with the application the provisions of the Act and the rules and regulations of the l

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 (ii) that such activities will be conducted in compliance with the Comissioh's regulations; 1

D.

The issuance of this amendment will not be inimical to the comon i

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 Commission's regulations and all applicable requirements have been satisfied.

2.

AccordinglyIndicatedintheattachmenttothislicenseamendmentandthe license is a cations as paragraph 3.B. of Facility Operating License No. DPR-25 is hereby amended to read as follows:

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. 1 B.

Technical Specifications I

The Technical Specifications contained in Appendix A, as revised through Amendment No.109, are hereby incorporated in the license.

The Itcensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective no later than 60 days from the date of its iss::4nce.

FOR THE NUCLEAR REGULATORY C0m!S$10N

/

Richaf4<hett, Director c

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i Project irectorate III-2 Division of Reactor Projects - III, IV, Y and Special Projects t

Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: October 2, 1990 1

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ATTACHMENT TO LICENSE AMENDMENT NO. 109 FACILITY OPERATING LICENSE N0. DPR-25 DOCKET NO. 50-249 Revise the Appendix A Technical Specifications by removing the pages identified i

below and inserting the attached pages. The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change.

REMOVE INSERT 11 11 vii vii viii viii 3/4.2-74 3/4.2-18a 3/4.2-27a B 3/4.2-33a B 3/4.2-33a B 3/4.2-37 B 3/4.2-37 3/4.4-1 3/4.4-1 l

3/4.4-2 3/4.4-2 3/4.4-3 3/4.4-3 3/4.4-4 3/4.4-4 B 3/4.4-6 B 3/4.4-6

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ORESDEN III DPR-25 Amendment No. 109 Table of Contents Pm 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 8 1/2.2-2 2.1 Limiting Safety System Settings - Fuel Cladding Integrity 1/2.1-1 Limiting Safety System Settings Bases 8 1/2.1-10 1

2. 2 Limiting Safety Syster,Scttings - Reactor Coolant System 1/2.2-1 Limiting Safety Systen: Settings Bases B 1/2.2-4 3.0 LIMITING CONDITION FOR bPERATION 3.0-1 Limiting Condition fo Operation Bases B

3.0-3 I

3.1 Reactor Protection System 3/4.1-1 Limiting Conditions for Operr. tion 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. 8 Core and Containment Cooling Systems - Initiation and Control 3/4.2-1 3.2.0 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 Gaseous Effluent Instrumentation 3/4.2-5 3.2.H Recirculation Pump Trip 3/4.2-7a l

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 3.3.0 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 3/4.3-14 Surveillances Requirement Bases (4.3)

B 3/4.3-22 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 3.4.0 Liquid Poison Tank 3/4.4-3 3.4.0 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, 6 Core Spray and LPCI Subsystems 3/4.5-1 3.5.B Containment Cooling Subsystsm 3/4.5-5 ii

3 DRESDEN III DPR-25 Amendment No.

109 l

List of Tables i

.P_ag Table 3.1.1 Reactor Protection System (Scram) 3/4.1-5 I

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 i

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 3/4.2-15 Monitoring Instrumentation Table 3.2.6 Post Accident Monitoring Instrumentation Requirements 3/4.2-17 Table 3.2.7 Instrumentation That Initiates Recircula-tion Pump Trip 3/4.1-18a 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 3/4.2-22 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 Table 4.2.5 Minimum Test and Calibration Frequency for Recirculation Pump Trip 3/4.2-27a Table 4.5.1 Surveillance of HPCI Subsystem 3/4.5-7a 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 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 Radioactive 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 Surveillance Requirements for High Energy Piping Outside Containment 3/4.11-3 Table 3.12-1 Fire Detection Instruments B 3/4.12-17 vii

i DRESDEN III DPR-25 Amendment No. 109 List of Tables (continued}

P, age i

Table 3.12-2 Sprinkler Systems 8 3/4.12-18 l

Table 3.12-3 CO Systems B 3/4.12-19 j

2 Table 3.12-4 Fire Hose Stations B 3/4.12-20 & 21 Table 6.1.1 tinimum Shift Manning Chart 6-4 Table 6.6.1 special Reports 6-22 1

List of Figures F;gure 2.1-3 APRM Bias Scram Relationship to Normal Operating Conditions B 1/2.1-17 Fig,tre 4.1.1 Graphical Aid in the Selection of an Adequate Interval Between Tests B 3/4.1-18 figure 4.2.2 Test Interval vs. System Unavailability B 3/4.2-38 Figure 3.4.1 Deleted 3/4.4-4 Figure 3.4.2 Sodium Pentaborate Solution Temperature Requirements 3/4.4-5 Figure 3.6.1 Minimum Temperature Requirements per Appendix G of 10 CFR 50 3/4.6-23 Figure 3.6.2 Thermal Power vs. Core Flow limits for Thermal Hydraulic Stability Surveillance In Single Loop Operation 3/4.6-24 Figure 4.6.1 Minimum Reactor Pressurization Temperature B 3/4.6-29 Figure 4.6.2 Chloride Stress Corrosion Test Results at 500'F B 3/4.6-31 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 Offsite Organization - Deleted Figure 6.1-2 Station Organization - Deleted viii

9 DRESDEN III DPR-25 Amendment No. 109 l

3. 2 LIMITING CONDITION FOR OPERATION 4.2 SURVEILLANCE REQUIREMENTS (CONT'D)

(CONT'0)

H.

Recirculation Pump Trip H.

Recirculation Pump Trip Initiation Initiation The recirculation pump trip In:trumentation and logic system, initiated by low low systeos shall be functionally reactor water level or high tested and calibrated as reactor pressure, limiting indicated in Table 4.2.5 conditions for operation are specified in Table 3.2.7 l

i 3/4.2-7a

6 DRESDEN III DPR-25 Amendment No. 109 TABLE 3.2.7 INSTRUMENTATION THAT INITIATES RECIRCULATION PUMP TRIP Minimum Number of Trip Function Operable or Tripped Trip Level Setting Applicable Action Instrument Channels Operational Per Trip System (a)

Mode High Reactor 2

(GT/E) 1230 psig &

1 (d) 70 Pressure (LT/E) 1250 psig low Low (GT/E) 84 inches 1 (d) 70 Reactor Water 2

above top of active Level fuel (b)(c)

Action 70 The minimum number of operable trip systems shall be four, two high reactor pressure and two low low reactor water level, except thct one trip system may be inoperable for up to fourteen days.

If one trip system is inoperable for greater than fourteen days, or if any two trip systems are made or found to be inoperable, the reactor must be placed in at least the Startup/ Hot Standby Mode in the'next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

NOTE:

r a.

A channel may be placed in an inoperable status for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for required surveillance without placing the trip system in the tripped condition provided at least cne operable channel in the same trip system is monitoring that parameter, b.

Top of active fuel is defined to be 360 inches above vessel zero, The trip will. occur following a (GT/E) 8 & (LT/E) 10 second delay.

c.

d.

MODE 1 is the RUN MODE 3/4.2-18a

o DRESDEN III DPR-25 Amendment No. 109 Table 4.2.5 MINIMUM TEST AND CALIBRATION FREQUENCY FOR THE RECIRCULATION PUMP TRIP Instrument Channel Instrument Instrument Applicable Functional Calibration Check Operational Test Mode 1.

Reactor High Pressure Q

R D

1*

I 2.

Reactor Low Low Water Q

R D

1*

Level

• MODE 1 is the RUN MODE 1

?

3/4.2-27a l'

a DRESDEN III DPR-25 Amendment No. 109 1

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

Reset of the relay, approximately 11% above the trip point, indicates that the diesel generator has restored bus voltage and will accept ECCS Loads.

The reset signal provides a permissive for starting ECCS pumps.

i Th0 setting for 4KV emergency bus degraded voltage is chosen to detect sustained degraded voltage which may cause equipment damage, while pre-venting trips caused by voltage fluctuations.

The reset point for degraded voltage indicates restoration of normal bus voltage.

l The recirculation pump trip (RPT) system is required by 10 CFR 50.62 to mitigate the consequences of an Anticipated Transient Without Scram (ATWS).

The design of this system meets the intent of the NRC Safety Evaluation for NEDE-31096-A (Reference 1).

RPT is provided to minimize reactor pressure in the highly unlikely_ event of a plant transient coin-cident with the failure of all control rods to scram.

The rapid flow reduction increases core voiding providing negative reactivity.

RPT is only required to operate in the RUN mode since at lower power levels the safety relief valves have sufficient capacity to relieve the steam which continues to be produced during this postulated event.

The low low reactor water level trip includes a nine second delay to avoid increasing the consequences of a postulated design baris loss of coolant accident.

References 1.

NEDE-31096-A, " Anticipated Transients Without Scram; Response to NRC ATWS Rule, 10 CFR 50.62." dated February 1987.

L t

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B 3/4.2-33a

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DRESDEN III DPR-25 Amendment No. 109 4.2 SURVEILLANCE REQUIREMENT BASES (Cont'd)

For instruments 3-263-73A, 73B and 3-2352, 2353, the logic downstream of the output relay contacts exhibits a one out-of-two logic and, by utiliz-ing the Availability Criteria identified in NED0-21617-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 ca?ibration/ surveillance test interval for the transmitter is once per operating cycle.

j 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 l

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.

The analog trip units which provide the initiation signal for RPT are cali-brated quarterly.

The reactor pressure and level transmitters which provide the input to the analog trip units are calibrated every refueling outage, concurrent with logic tests that ensure overall system operability.

An instrument check is performed on a daily basis.

These frequencies are con-sidared appropriate, commensurate with the design application and the fact tL. the_RPT system is a backup to existing protective instrumentation.

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o o

DRESDEN III DPR-25 t

Amendment No. 109 1

3.4 LIMITING CONDITION FOR OPERATION 4.4 SURVEILLANCE REQUIREMEhT STANDBY LIQUID CONTROL SYSTEM STANDBY LIQUID CONTROL SYSTEM Applicability:

Applicability:

Applies to the operating Applies _to the periodic testing status of the standby liquid requirements for the standby control system.

liquid control system.

Objective Objective:

To assure the availability of To verify the operability of an independent reactivity the standby liquid control control mechanism.

system.

Specification:

Specification:

A.

Normal Operation A.

Normal Operation During periods when fuel The operability of the is in the reactor the standby liquid control standby liquid control system shall be verified system shall be operable by performance of the except when the reactor following tests:

is in the Cold Shutdown Condition and all control 1.

At least once per rods are fully inserted month -

and Specification 3.3.A ts met or as specified in Demineralized water 3.4.B below.

shall be recycled to the test tank, pump minimum flow rate of 40 l

gpm shall be verified against a system head of 1275 psig, t

2.

At least once during each operating cycle

a. Manually initiate the system, except the i

explosion valves and pump solution in the recirculation path, to demonstrate that P

the pump suction line from the storage l

tank is not plugged.

l 3/4.4-1

e l.

DRES14 '? T1T DPR-25 Amends tat No.109 3.4 LIMITING CONDITION FOR OPERATION 4.4 SURVEILLANCE REQUIREMEhT (Cont'd.)

(Cont'd.)

l b.

Actuate one of the two standby liquid control systems using the normal actuation switch and pump demineralized water into the reactor vessel.

Pump.

minimum flow rate shall be verified against a previous test at the same reactor vessel pressure. The replacement charges will be selected from a batch from which at least one charge has been successfully test fired and which will not exceed five years life when their use is terminated.

Both systems shall be tested and inspected, including each explosive actuated valve, in the course of two ni operating cycles.

c.

Test that the l

setting of the system pressure l

relief valves is between 1455 and 1545 psig.

B.

Operation with Inoperable B.

Surveillance with Inoperable t

Components Components From and after the date When a component becomes that a redundant component inoperable its redundant 3/4.4-2

DRESDEN III DPR-25 Amendment No. 109 3.4 LIMITING CONDITION FOR OPERATION 4.4 SURVEILLANCE REQUIREMENT (Cont'd.)

(Cont'd.)

is made or found to be component shall be inoperable, Specification demonstrated to be 3.4.A shall be considered operable immediately and fulfilled, and continued daily thereafter.

operation permitted provided that the component is returned to an cperable condition within 7 days.

C.

The liquid poison tank C. The availability of the shall contain a boron proper boron bearing bearing solution of at solution shall be verified least 3605 gallons (3329 by performance of the gallons to meet shutdown following tests:

requirements plus 276 gallons that are contained 1.

At least once per below the pump suction) of month - Boron at least 14 wt. percent concentration shall sodium pentaborate decahy-be determined.

In drate (Na2B o038-10H O),

addition, the boron i

2 At all times when the concentration shall be standby liquid control determined any time mystem is required to water or boron are be operable, the solution added or if the temperature including solution temperature that in the pump suction drops below the limits piping shall not be less specified by Figure than the temperature 3.4.2.

Minimum con-presented in Figure 3.4.2.

centration is 14 wt.

percent sodium pentaborate.

2.

At least once per day -

Solution volume shall be checked.

3.

At least once per day -

The solution temperature shall be checked.

D.

If specification 3.4.A through C are not met, an orderly shutdown shall be initiated and the reactor shall be in the cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

l 3/4.4-3 4

- - - - - - - - - - ~ - - - - - - - - - - - - - - - - - - - - - - - - - - - -

- - - - ~ - " - - - - - - - - - - - - -

e

-e DRESDEN III DPR-25 Amendment No. 109 4

TIIIS PAGE WAS LEl'T INTENTIONALLY BLANN 3/4.4-4

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5*

l DRESDEN III OPR-25 Amendment No. 109 3.4, LIMITING CONDITION FOR OPERATION BASES A.

The design objective of the standby liquid control sy 6em is to pro-vide the capability of bringing the reactor from full power to a cold, xenon-free shutdown assuming that none of the withdrawn control rods can be inserted.

To meet this objective, the liquid control system is designed to inject a quantity of boron which produces a concentra-tion of no less than 600 ppm of boron in the reactor core in less than 100 minutes.

600 ppm boron concentration in the reactor core is required to bring the reactor from full power to a 3% delta k or more subcritical condition considering the hot to cold reactivity swing e

and xenon poisoning.

An additional margin of 25% boron is added to compensate for possible imperfect mixing of the chemical solution in the reactor water, making the total concentration 750 ppm of boron in the reactor core.

A minimum quantity of 3329 gallons of solution having a 14 wt percent sodium pentaborate concentration is required to meet this shutdown requirement.

For a required pumping rate of 40 gpm, 3329 gallons per pump of at

's least 14 wt. percent solution will be inserted in less than 100 minutes.

This insertion rate of boron solution will override the i

rate of reactivity insertion due to cool down of the reactor following the xenon peak.

Two pump operation will enable faster reactor 1

shutdown for ATWS events.

The minimum volume required in the storage t

tank shall 4 3605 gallons (276 are contained below the pump suction and, therefore, cannot be inserted).

The monthly pump minimum flow-rate test shall require a minimum flowrate of 40 gpm per pump.

This requirement, combined with the solution concentration requirement of at least 14 wt. percent, will demonstrate that the Standby Liquid Control System meets the requirements of 10 CFR 50.62.

Boron concentration, solution temperature, and volume are checked on a frequency to assure a high reliability of operation of the system should it ever be required.

Experience with pump operability indi-cates that monthly testing is adequate to detect if failures have occurred.

Components of the system are checked periodically as described above and make a functional test of the entire system on a frequency of less e

than once during each operating cycle unnecessary.

A test of one installed explosive charge is made at least once during each operating cycle to assure that the charges have not deteriorated, tne actuation circuit is functioning properly, the valve functions properly, and no flow blockages exist.

The replacement charge will be selected from a' batch for which there has been a successful test firing.

Recommenda-tions of the vendor shall be followed in maintaining a five year life of the explosive charges.

A continual check of the firing circuit continuity is provided by pilot lights in the control room.

B 3/4.4-6

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." ; -o DRESDEN III DPR-25 Amendment No. 109 i

i 3.4 LIMITING CONDITION FOR OPERATION BASES (Cont'd.)

The relief valves in the standby liquid control system protect the System piping and positive displacement pumps which are nominally designed for 1500 psig protection from over pressure.

The pressure relief valves discharge back to the standby liquid control solution tank.

B.

Only one of the two standby liquid control pumping circuits is needed for proper operation of the system.

If one pumping circuit is found to be inoperable, there is no immediate threat to shutdown capability, and reactor operation may continue while repairs are being made.

Assurance that the remaining system will perform its intended function and that the reliability of the system is good is obtained by demon-strating operation of the pump in the operable circuit at least once i

daily.

C.

The solution saturation temperature of 14 wt. percent sodium penta-borate is 62'f.

To guard against boron precipitation, the s0lution including that in the pump suction piping is kept at least 10'F above the saturation temperature by a tank heater and by heat tracing in the pump suction piping.

The 10'F margin is included in Figure 3.4.2.

Temperature and liquid level alarms for the system are annunciated in the control room.

Pump operability is checked on a frequency to assure a high reliabil-i ity of operation of the system should it ever be required.

l Once the solution has been made up, boron concentration will not vary unless more boron or more water is adoed.

Level indication and alarm indicate whether the solution volume has changed which might indicate a possible solution concentration change.

Considering these factors, j

the test interval has been established.

4.4 SURVEILLANCE REQUIREMENT BASES l

Pe'riodic tests to demonstrate two pump flow capability are not feasible in the present system configuration and are unnecessary because the flow path i

integrity can be determined from the test of a single pump.

Comparison of l.

single pump test pressures with previous results and correlation of these l-data with initial two pump test will be used to verify the capability of L

the piping to support two pump flow.

B 3/4.4-7 i

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