ML17191B235

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Amends 170,165,183 & 180 to Licenses DPR-19,DPR-25.DPR-29 & DPR-30,respectively,relocating Requirement to Remove Reactor Protection Sys Shorting Links to Updated Final Safety Analysis Rept
ML17191B235
Person / Time
Site: Dresden, Quad Cities  Constellation icon.png
Issue date: 02/08/1999
From: Rossbach L
NRC (Affiliation Not Assigned)
To:
Shared Package
ML17191B236 List:
References
NUDOCS 9902160330
Download: ML17191B235 (26)
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Text

(

. I UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-<>001 COMMONWEAL TH EDISON COMPANY DOCKET NO. 50-237 DRESDEN NUCLEAR POWER STATION. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 1 70 License No. DPR-19

1.

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

A.

The application for amendment by the Commonwealth Edison Company (the licensee) dated November 30, 1998, as supplemented by letter dated January 8, 1999, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), an.d the Commission's rules and regulations set.

forth in 1 O 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 Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission'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 Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. DPR-19 is hereby amended to read as follows:

9902160330 990208 PDR ADOCK 05000237 P

PDR

. l.

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3.
  • (2)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 1 "!Q, are hereby incorporated in the license. The licensee*

shall operate the facility in accordance with the Technical Specifications.

This license amendment is effective as of the date of its issuance and shall be implemented within 60 days including the relocation of information from the Technical Specifications to the licensee's Updated Final Safety Analysis Report (UFSAR) as described in the licensee's application dated November 30, 1998, as supplemented by letter dated January 8, 1999, and evaluated in the staff's safety evaluation dated February a. 1999.

FOR THE NUCLEAR REGULA TORY COMMISSION Lawrence W. Rossbach, Project Manager Project Directorate 111-2

  • Division of Reactor Projects - Ill/IV Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: February 8, 1999

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555--0001 COMMONWEAL TH EDISON COMPANY DOCKETNO. 50-249 DRESDEN NUCLEAR POWER STATION. UNIT 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 165 License No. DPR-25

1.

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

A.

The application for amendment by the Commonwealth Edison Company (the licensee) dated November 30, 1998, as supplemented by letter dated January 8, 1999, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission'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 Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety ofthe public, and (ii) that such activities will be conducted in compliance with the Commission'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 pf this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 3.8. of Facility Operating License No. DPR-25 is hereby amended to read as follows:

~*-**

B.

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.

165, 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 and shall be implemented within 60 days including the relocation of information from the Technical Specifications to the licensee's Updated Final Safety Analysis Report (UFSAR) as described in the licensee's application dated November 30, 1998, as supplemented by letter dated January 8, 1999, and evaluated in the staff's safety evaluation dated February

~. 1999.

Attachment:

Changes to the Technical Specifications FOR THE NUCLEAR REGULA TORY COMMISSION

/a---* Lv./(~

Lawrence W. Rossbach, Project Manager Project Directorate 111~2 Division of Reactor Projects - Ill/IV Office of Nuclear Reactor Regulation Date of Issuance: February 8, 1999

A TIACHMENT TO LICENSE AMENDMENT NOS. 170 AND 165 FACILITY OPERATING LICENSE NOS. DPR-19 AND DPR-25 DOCKET NOS. 50-237 AND 50-249 Revise the Appendix A Technical Specifications 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 3/4.1-2 3/4.1-6 B 3(4.1-1 B 3/4.1-2 3/4.10-3

. 3/4.10-4 B 3/4.10-1 3/4.12-2 INSERT 3/4.1-2 3/4.1-6 B 3/4.1-1 B 3/4.1-2 3/4.10-3 3/4.10-4 B 3/4.10-1 3/4.12-2

. ~*

0 TABLE 3.1.A-1

a
a

~

m

(/)

REACTOR PROTECTION SYSTEM INSTRUMENTATION

()

0

-f m

0 z

a c

Applicable Minimum "U

a z

OPERATIONAL OPERABLE CHANNEL(s) 0

-f

-f

(/)

Functional Unit MODEis)

~er TRIP SYSTEMca>

ACTION m

()

I\\)

-f Qo 0

(..)

1.

Intermediate Range Monitor:

z

(/) e

a.

Neutron Flux - High

  • 2 3

11

(/)

-f 3,4 2

12 m

~

5 3

13

. b.

Inoperative 2

3 11.

3,4 2

12

(..)

5 3

13

ii:

-lo.

I

2.

Average Power Range Monitor<e>:

I\\)

a.

Setdown Neutron Flux - High 2

2 11 3

2 12 5<sl 2

13

b.

Flow Bia$ed Neutron Flux - High 1

2 14 e

c.

Fixed Neutron Flux - High 1

2 14

)>

d.

Inoperative 1, 2 2

11 3

3 2

12 CD 519) 2 13

J
0.

3 CD

J 3..

Reactor Vessel Steam Dome Pressure - High 1, 211) 2 11

a z

"U 0

(/)

!/)

~

4.

Reactor Vessel Water level - low 1, 2 2

11

(..)

ii:

0

-lo..

~

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REACTOR PROTECTION SYSTEM RPS 3/4.1.A TABLE 3.1.A-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION TABLE NOTATION (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 one OPERABLE CHANNEL in the same TRIP SYSTEM is monitoring that parameter.

(b) This function may be bypassed, provided a control rod block is actuated, for reactor protection sys~em logic reset in Refuel and Shutdown positions of the reactor mode switch.

(c) Deleted (d) With THERMAL POWER greater than or equal to 45% of RATED THERMAL POWER.

(e) An APRM CHANNEL is inoperable if there are fewer than 2 LPRM inputs per level or there are less than 50% of the normal complement of LPRM inputs to an APRM CHANNEL.

(f)

This function is not required to be OPERABLE when the reactor pressure vessel head is.

unbolted or removed per Specification 3.12.A.

(g). Required to be OPERABLE only prior to and during required SHUTDOWN MARGIN demonstrations performed per Specification 3.12.B.

(h) This function is not required to be OPERABLE when PRIMARY CONTAINMENT INTEGRITY is not required.

(i)

With any control rod withdrawn. Not applicable to control rods removed per Specification 3.10.1 or 3.1 O.J.

(j)

This function is not required to be OPERABLE when reactor pressure is less than 600 psig.

DRESDEN

  • UNITS 2 & 3 3/4.1-6.

Amendment Nos. 1 70; 165

RPS B 3/4.1 BASES 3/4.1.A REACTOR PROTECTION SYSTEM INSTRUMENTATION The reactor protection system (RPS) automatically initiates a reactor scram to:

a.

preserve the integrity of the fuel cladding,

b.

preserve the integrity of the primary system, and

c.

minimize the energy which must be absorbed and prevent criticality following a loss-of-coolant accident.

This specification provides the Limiting Conditions for Operation necessary to preserve the ability of the system to perform its intended function, even during periods when instrument CHANNEL(s).

may be out-of-service because of maintenance. When necessary, one CHANNEL may be made inoperable for brief intervals to conduct required surveillance.*

The reactor protection system is made up of two independent TRIP SYSTEM(s), each having a minimum of two CHANNEL(s) of tripping devices. Each CHANNEL has an input from at least one instrument CHANNEL which monitors a critical parameter. The outputs of the CHANNEL(s) are combined in a one-out-of-t~o-logic, i.e., an input signal on either one or both of the CHANNEL(s)

  • will cause a TRIP SYSTEM trip. The outputs of the TRIP SYSTEM(s) are arranged so that a trip on both systems is required to produce a reactor scram.. This system meets the intent of the proposed IEEE 279, "Standard for Nuclear Power Plant Protection Systems" issued September 13, 1966.

The system has a reliability greater than that of a two-out-of-three system and somewhat less than that of a one-out-of-two system (reference APED 5179). The bases for the trip settings of the RPS are discussed In the Bases for Specification 2.2.A.

DRESDEN - UNITS 2 & 3 B 3/4.1-1 Amendment Nos. 170; 165

RPS B 3/4.1 BASES The primary reactivity control functions during refueling are the refueling interlocks and the SHUTDOWN MARGIN calculations, which together provide assurance that adequate SHUTDOWN MARGIN is available. The IRMs also provide backup protection for any significant reactivity excursions.

The IRM system provides protection against excessive power levels and short reactor periods in the startup and intermediate power ranges (reference SAR Sections 7.4.4.2 and 7.4.4.3).

In the power range, the APRM system provides required protection (reference SAR Section 7.4.5.2). Thus, the IRM system is not required (and is automatically bypassed) in OPERATIONAL MODE 1, the APRMs cover only the intermediate and power range; and the IRMs provide adequate coverage in the startup and intermediate range. The IRM inoperative function ensures that the instrument CHANNEL fails in the tripped condition upon loss of detector voltage.

Three APRM instrument CHANNEL(s) are provided for each TRIP SYSTEM. APRM CHANNEL(s) #1 and #3 operate contacts in one logic path and APRM CHANNEL(s) #2 and #3 operate contacts in the other logic path of the TRIP SYSTEM. APRM CHANNEL(s) #4, #5 and #6 are arranged similarly in the other TRIP SYSTEM's dual logic paths. Each 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 TRIP SYSTEM for maintenance, testing, or calibration. Additional IRM CHANNEL(s) have also been provided to allow for bypassing of one such CHANNEL.

A reactor mode switch is provided which actuates or bypasses the various scram functions appropriate to the particular plant operating status (reference SAR Section 7. 7.1.2). A bypass in the? Refuel or Startup/Hot Standby operational modes is provided for the turbine condenser low vacuum scram and main steam line isolation valve closure scrams for flexibility during startup and to allow repairs to be *made to the turbine condenser. While this bypass is in effect, protection is provided against pressure or flux increases by the high-pressure scram and APRM 1 5 % scram, respectively, which are effective in Startup/Hot Standby.

The manual scram function is available in OPERATIONAL MODE(s) 1 through 5, thus providing for a manual means of rapidly inserting control rods whenever fuel is in the reactor.

The turbine stop valve closure scram, the turbine EHC control oil low pressure scram, and the turbine control valve fast closure scram occur by design on turbine first stage pressure which is normally equivalent to -45% RATED THERMAL POWER. However, since this is dependent on bypass valve position, the conservative reactor power is used to determine applicability.

Surveillance requirements for the reactor protection system are selected in order to* demonstrate proper function and operability. The surveillance intervals are determined in many different ways, such as, 1) operating experience, 2) good engineering judgement, 3) reliability analyses, or 4) other analyses that are found acceptable to the NRC. The performance of the specified surveillances at the specified frequencies provides assurance that the protective functions associated with each CHANNEL can be completed as assumed in the safety analyses. A surveillance interval of "prior to startup" assures that these functions are available to perform their safety functions during control DRESDEN - UNITS 2 & 3 B 3/4.1-2 Amendment Nos. 1 70; 165

e REFUELING OPERATIONS 3.10 - LIMITING. CONDITIONS FOR OPERATION B.

Instrumentation At least 2 source range monitor(a) (SRM)

CHANNEL(s) shall be OPERABLE and inserted to the normal operating level with:

1.

Continuous visual indication in the control room, and

2.
  • One of the required SRM detectors located in the quadrant where CORE AL TERATION(s) are being performed
  • and the other required SRM detector located in an adjacent quadrant.

APPLICABILITY:

OPERATIONAL MODE 5, unless the following conditions are met:

1. No more than two fuel assemblies are present in each core quadrant associated with an SRM; Instrumentation 3/4.10.8 4.10 - SURVEILLANCE REQUIREMENTS B.

Instrumentation Each of the required SRM channels shall be demonstrated OPERABLE by:

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

a.

Performance of a CHANNEL CHECK.

b.

Verifying the detectors are inserted to the normal operating level, and

c.

During CORE AL TERATION(s),

verifying that the detector of an OPERABLE SRM CHANNEL is located in the core quadrant where CORE AL TERATION(s) are being performed and another is located in an adjacent quadrant.

2.

Performance of a CHANNEL FUNCTIONAL TEST:

a.

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

b.

At least once per 7 days.

3.

Verifying* that the channel count rate is at least 3 cps:

a.

Prior to control rod withdrawal,

b.

Prior to and at least once per 1 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> during CORE ALTERATION(s),

c.

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

a

  • The use of special movable detectors during CORE ALTERATION(s) in place of the normal SAM neutron detectors is permissible as long as these special detectors are connected to the normal SAM circuits.

DRESDEN - UNITS 2 & 3 3/4.10-3 Amendment Nos. 170; 165

~.

REFUELING OPERATIONS 3.10 - LIMITING CONDITIONS FOR OPERATION

2.

While in the core, these two fuel assemblies are in locations adjacent to the SRM; and

3.

In the case of movable detectors, each group of fuel assemblies shall be separated by at least two fuel cell locations from any other fuel assemblies.

ACTION:

With the requirements of the above specification not satisfied, immediately.

suspend all operations involving CORE AL TERATION(s) and fully insert all insertable control rods.

Instrumentation 3/4.1 O.B 4.10 - SURVEILLANCE REQUIREMENTS I

DRESDEN - UNITS 2 & 3 3/4.10-4 Amendment Nos. 170; 165

e REFUELING OPERATIONS B 3/4.10 BASES 3/4.1 O.A Reactor Mode Switch Locking the OPERABLE reactor mode switch in the Shutdown or Refuel position, as specified, ensures that the restrictions on control rod withdrawal arid refueling platform movement during the refueling operations are properly activated. These conditions reinforce the refueling procedures and reduce the probability of inadvertent criticality, damage to reactor internals or fuel assemblies, and exposure of personnel to excessive radioactivity.

The addition of large amounts of reactivity to the core is prevented by operating procedures, which are in turn backed up by refueling interlocks on rod withdrawal and movement of the refueling platform. When the mode switch is in the Refuel position,. interlocks prevent the refueling platform from being moved over the core if a control rod is withdrawn and fuel is on a hoist. If the refueling platform is over the core with fuel on a hoist, control rod motion is blocked by the interlocks. With the mode switch in the refuel position only one control rod can be withdrawn.

3/4.10.B Instrumentation The OPERABILITY of at least two source range monitors ensures that redundant monitoring capability is available to detect changes in the reactivity condition of the core, whenever reactor criticality is possible.

The source range monitors (SRM) are provided to monitor the core during periods of station shutdown and to guide the operator during refueling operations and reactor star:tup. Requiring ~wo OPERABLE source range monitors in and adjacent to any core quadrant where fuel or control rods are being moved assures adequate monitoring of that quadrant during such alterations. Requiring a minimum of 3 counts per second whenever criticality is possible provides assurance that neutron flux is being monitored.. The SRM system is designed to provide a signal-to-noise ratio of at least 3: 1 and a count ra~e of at least 3 counts per second.

  • Criticality is considered to be impossible if there are no more than two assemblies in a quadrant and if these are in locations adjacent to the source range monitors (i.e., spatially separated).

Special movable detectors may be used during CORE ALTERATION(s) in place of the normal SRM neutron detectors. These special detectors must be connected to the normal SRM circuits such that the applicable neutron flux indication, control rod blocks and scram signals can be generated~

The special detectors provide more flexibility in monitoring reactivity changes during fuel loading since they can be positioned anywhere within the core during refueling provided they meet th~

location requirements of the specification.

DRESDEN - UNITS 2 & 3 B 3/4.10-1 Amendment Nos. 170;

  • 165 I

SPECIAL TEST EXCEPTIONS 3.12 - LIMITING CONDITIONS FOR OPERATION B.

SHUTDOWN MARGIN D.emonstrations The provisions of Specifications 3.1 O.A and 3.1 O.C and Table 1-2 may be suspended to permit the reactor "!'ode switch to be in the Startup position and to allow more than one control rod to be withdrawn for SHUTDOWN MARGIN demonstration, provided that at least the following requirements are satisfied.

1. The source range monitors are OPERABLE per Specification 3.1 O.B.

2.

The rod worth minimizer is OPERABLE per Specification 3.3.L and is programmed for the SHUTDOWN MARGIN demonstration, or conformance with the SHUTDOWN MARGIN demonstration procedure is verified by a second licensed operator or other technically qualified individual.

3.. The "rod-out-notch-override" control shall not be used during out-of-sequence movement of the control rods.

4.

No other CORE AL TERA TION(s) are in progress.

APPLICABILITY:

OPERATIONAL MODE 5, during SHUTDOWN MARGIN demonstrations.

ACTION:

With the requirements of the above specification not satisfied, immediately place the reactor mode switch in the Shutdown or Refu,el position.

SOM 3/4.12.B 4.12 - SURVEILLANCE REQUIREMENTS B.

SHUTDOWN MARGIN Demonstrations Within 30 minutes prior to and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during the performance of a SHUTDOWN MARGIN demonstration, verify

.that; 1. The source range monitors are OPERABLE per Specification 3.1 O.B,

2.

The rod worth minimizer is OPERABLE with the required program per Specification 3.3.L or a second licensed operator or other technically qualified individual is present and verifies compliance with the SHUTDOWN MARGIN demonstration procedures, and

3.

No other CORE ALTERATION(s) are in progress.

.DRESDEN - UNITS 2 & 3 3/4.12-2 Amendment Nos. 170; 165

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. ~1 COMMONWEAL TH EDISON COMPANY AND MIDAMERICAN ENERGY COMPANY DOCKET NO. 50-254 QUAD CITIES NUCLEAR POWER STATION. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 183 License No. DPR-29

1.

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

A.

The application for amendment by Commonwealth Edison Company (the licensee) dated November 30, 1998, as supplemented by letter dated January 8, 1999, complies with the standards and requirements of the Atomic Energy Act of 1954, as amenqed (the Act) and the Commission's rules and regulations set forth

. in 10 CFR Chapter I;.

8.

The facility will operate in conformity with the application, the provisions of t~e Act, and the rules and regulations of the Commission;

. C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission'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 ame.ndment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have beeh satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.8. of Facility Operating License No. DPR-29 is hereby amended to read as. follows:

  • B.

Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 183 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 and shall be implemented within 60 days including the relocation of information from the Technical Specifications to the licensee's Updated Final. Safety Analysis Report (UFSAR) as described in the licensee's application dated November 30, -1998, as supplemented by letter dated January 8, 1999, and evaluated in the staff's safety evaluation dated February 8, 1999.

Attachment:

Changes to the Technical Specifications FOR THE NUCLEAR REGULA TORY COMMISSION Robert M. Pulsifer, Project Manager Project Directorate 111-2 Division of Reactor Projects - Ill/IV Office of Nuclear Reactor Regulation Date of Issuance:

rebruary 8, 1999

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555--0001 COMMONWEAL TH EDISON COMPANY AND MIDAMERICAN ENERGY COMPANY DOCKET NO. 50-265 QUAD CITIES NUCLEAR POWER STATION. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 180 License No. DPR-30

1.

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

A.

The application for amendment by Commonwealth Edison Company (the licensee) dated November 30, 1998, as supplemented by letter dated January 8, 1999, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth *.

in 10 CFR Chapter I;

  • B.

The facility will operate in conformity with the application, the provisions of th'e Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted. without endangerin9 the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission'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 Commission's regulations and all applicable requirements have been satisfied.

  • 2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B. of Facility Operating License No; DPR-30 is hereby amended to read as follows:

B.

Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.

180. are hereby incorporated in the license. The licensee shall operate the facility in accordance with theTechnical Specifieations.

3.

This license amendment is effective as of the date of its issuance and shall be implemented within 60 days including the relocation of information from the Technical Specifications to the licensee's Updated Fin.al Safety Analysis Report (UFSAR) as described in the licensee's application dated November 30, 1998, as supplemented by letter dated January 8, 1999, and evaluated in the staff's safety evaluation dated February 8, 1999.

Attachment:

Changes to the Technical Specifications FOR THE NUCLEAR REGULATORY COMMISSION

  • flA1J2?

Robert M. Pulsifer, Project Manager Project Directorate 111-2 Division of Reactor Projects - Ill/IV Office of Nuclear Reactor Regulation Date of Issuance: February 8, 1999

ATTACHMENT TO LICENSE AMENDMENT NOS. 183 AND 180 FACILITY OPERATING LICENSE NOS. DPR-29 AND DPR-30 DOCKET NOS. 50-254 AND 50-265 Revise the Appendix A Technical Specifications 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 3/4.1-2 3/4.1-6 B 3/4.1-1 B 3/4.1-2 3/4.10-3 3/4.10-4 B 3/4.10-1 3/4.12-2 INSERT 3/4.1-2 3/4.1-6 B 3/4.1-1 B 3/4.1-2 3/4.10-3 3/4.10-4 B 3/4.10-1 3/4.12-2

i*

0 TABLE 3.1.A-1

0 c

-~

)>

a REACTOR PROTECTION SYSTEM INSTRUMENTATION

-~

0 0

0 m

. -0 en Applicable Minimum

0 OPERATIONAL OPERABLE CHANNEL(s) 0 c

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ACTION m

0 en

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0

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1. Intermediate Range Monitor:

z I\\)

~-

a.

Neutron Flux - High 2

3 11 3,4 2

12 m

3:: I 5

3 13

b.

Inoperative 2

3 11

. 3, 4 2

12 c..>

5 3

13

~

...a.

I I\\)

2. Average Power Range Monitor<e>:

I

a.

Setdown Neutron Flux - High 2

2 11 3

2 12 I.

5<g) 2 13

b.

Flow Biased Neutron Flux - High 1

2 14 e

c.. Fixed Neutron Flux - High 1

2 14 d.* Inoperative 1, 2 2

11

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3 2

12 3

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0

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11

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co 0

REACTOR PROTECTION SYSTEM TABLE 3. 1.A-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION TABLE NOTATION RPS 3/4.1.A (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 one OPERABLE CHANNEL in the same TRIP SYSTEM is monitoring that parameter.

. (b) This function may be bypassed, provided a control rod block is actuated, for reactor protection system logic reset in Refuel and Shutdown positions of the reactor mode switch.

(c) Deleted.

  • (d) With THERMAL POWER greater than or equal to 45% of RATED THERMAL POWER.

(e) An APRM CHANNEL is inoperable if there are fewer than 2 LPRM inputs per level or there are less than 50% of the normal complement of LPRM inputs to an APRM CHANNEL.

(.f)

This function is not required to be OPERABLE when the reactor pressure vessel head is unbolted or removed per Specification 3.12.A.

(g) Required to be OPERABLE only prior to and during required SHUTDOWN MARGIN demonstrations performed per Specification 3.12.B.

(h) This function is not required to be OPERABLE when PRIMARY CONTAINMENT INTEGRITY is not required.

(i)

With any control rod withdrawn. Not applicable to control rods removed per Specification 3.10.1 or 3.1 O.J.

QUAD CITIES - UNITS 1 & 2 3/4.1-6 Amendment Nos. 183; 180

. RPS B 3/4.1 BASES 3/4.1.A REACTOR PROTECTION SYSTEM INSTRUMENTATION The reactor protection system (RPS) automatically initiates a reactor scram to:

a.

preserve the integrity of the fuel cladding,

b.

preserve the integrity of the primary system, and

c.

minimize the energy which must be absorbed and prevent criticality following a loss-of-coolant accident.

This specification provides the Limiting Conditions for Operation necessary to preserve the ability of the system to perform its intended function, even during periods when instrument CHANNEL(s) may be out-of-service because of maintenance. When necessary, one CHANNEL may be made inoperable for brief intervals to conduct required surveillance.

The reactor protection system is made up of two independent TRIP SYSTEM(s), each having a minimum of two CHANNEL(s) of tripping devices. Each CHANNEL has an input from at least one

  • instrument CHANNEL which monitors a critical parameter. The outputs of the CHANNEL(s) are combined in a one-out-of-two-logic, i.e., an input signal on either one or both of the CHANNEL(s)

. will cause a TRIP SYSTEM trip. The outputs of the TRIP SYSTEM(s) are arranged so that a trip on both systems is required to produce a reactor scram. This system meet~ the intent of the proposed IEEE 279, "Standard for Nuclear Power Plant Protection Systems" issued September 13, 1966.

The system has a reliability greater than that of a two-out-of-three system and somewhat less than that of a one-out-of-two system (reference APED 5179). The bases for the trip settings of the RPS are discussed in the Bases for Specification 2.2.A.

The primary reactivity control functions during refueling are the refueling interlocks and the SHUTDOWN MARGIN calculations, which together provide assurance that adequate SHUTDOWN MARGIN is available. The IRMs also provide backup protection for any significant reactivity excursions.

QUAD CITIES - UNITS 1 & 2 B 3/4.1-1 Amendment Nos. 183; 180

RPS B 3/4.1 BASES The IRM system provides protection against excessive power levels and short reactor periods in the startup and intermediate power ranges (reference SAR Sections 7.4.4.2 and 7.4.4.3).

In the power range, the APRM system provides required protection (reference SAR Section 7.4.S.2). Thus, the IRM system is not required (and is automatically bypassed) in OPERATIONAL MODE 1, the APRMs cover only the intermediate and power range; and the IRMs provide adequate coverage in the startup and intermediate range. The IRM inoperative function ensures that the instrument CHANNEL fails in the tripped condition upon loss of detector voltage.

Three APRM instrument CHANNEL(s) are provided for each TRIP SYSTEM. APRM CHANNEL(s) #1 and #3 operate contacts in one logic path and APRM CHANNEL(s) #2 and #3 operate contacts in the other logic path of the TRIP SYSTEM. APRM CHANNEL(s) #4, #5 and #6 are arranged similarly in the other TRIP SYSTEM's dual logic paths. Each 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 TRIP SYSTEM for maintenance, testing, or calibration. Additional IRM CHANNEL(s) have also been provided to allow for bypassing of one such CHANNEL.

A reactor mode switch is provided which actuates or bypasses the various scram functions appropriate to the particular plant operating status (reference SAR Section 7. 7.1.2). A bypass in

  • the Refuel or Startup/Hot Standby operational modes is provided for the turbine condenser low vacuum scram and main steam line isolation valve closure scrams for flexibility during startup and to allow repairs to be made to the turbine condenser. While this bypass is in effect, protection is
  • provided against pressure or flux increases by the high-pressure scram and APRM 1 5 % scram, respectively, which ar,e effective in Startup/Hot Standby.

The manual scram function is available in OPERATIONAL MODE(s) 1 through 5, thus providing for a.

manual means of rapidly inserting control rods whenever fuel is in the reactor.

The turbine stop valve closure scram, the turbine EHC control oil low pressure scram, and the turbine control valve fast closure scram occur by design on turbine first stage pressure which is normally equivalent to -45% RATED THERMAL POWER. However, since this is dependent on bypass valve position, the conservative reactor power is used to determine applicability.

Surveillance requirements for the reactor protection system are selected in order to demonstrate proper function and operability.. The surveillance intervals are determined in many different ways, such as, 1) operating experience, 2) good engineering judgement, 3) reliability analyses, or 4) other analyses that are found acceptable to the NRC. The performance of the specified surveillances at the specified frequencies provides assurance that the protective functions associated with each CHANNEL can be completed as assumed in the safety analyses. A surveillance interval of "prior to startup" assures that these functions are available to perform their safety functions during control QUAD CITIES -

UNITS 1 & 2 B 3/4.1-2 Amendment Nos.183; 180

I REFUELING OPERATIONS

3. 10 - LIMITING CONDITIONS FOR OPERATION B.

Instrumentation At least 2 source range monitor(a> (SRM)

CHANNEL(s) shall be OPERABLE and inserted to the normal operating level with:

1. Continuous visual indication in_ the control room, and
2.

One of the required SRM detectors located in the quadrant where CORE AL TERATION(s) are being performed and the other required SRM detector located in an adjacent quadrant.

APPLICABILITY:

OPERATIONAL MODE 5, unless the following conditions are met:

1.

No more than two fuel assemblies are present in each core quadrant associated with an SRM;

  • Instrumentation 3/4.1 O.B 4.10 - SURVEILLANCE REQUIREMENTS B.

Instrumentation Each of the required SRM channels shall be demonstrated OPERABLE by:

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

a.

Performance* of a CHANNEL CHECK.

b.

Verifying the detectors are inserted to the normal operating level, and

c.

During CORE ALTERATION(s),

verifying that the detector of an OPERABLE SRM CHANNEL is located in the core quadrant where CORE AL TERATION(s) are being performed and another is located in an adjacent quadrant.

2.

Performance of a CHANNEL FUNCTIONAL TEST:

a.

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

CORE AL TERATIQN(s), and

b.

At least once per 7 days.

3.

Verifying that the channel count rate is at least 3 cps:

a.

Prior to control_ rod withdrawal,

b.

Prior to and at least once per 1 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> during CORE AL TERATION(s),

c.

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

a The use of special movable detectors during CORE AL TERATION(s) in place of the normal SRM neutron detectors is permissible as long as these special detectors are connected to the normal SRM circuits.

QUAD CITIES - UNITS l & 2 3/4.10-3 Amendment Nos. 183; 180 I

REFUELING OPERATION-Instrumentation 3/4.10.8

. 3.10 - LIMITING CONDITIONS FOR OPERATION 4.10 - SURVEILLANCE REQUIREMENTS

2. While in the core, these two fuel.
  • assemblies are in locations adjacent to the SRM; and
3. In the case of movable* detectors, each group of fuel assemblies shall be separated by at least two fuel cell locations from ~my other fuel assemblies.

ACTION:

With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE AL TERA TION(s) and fully insert all insertable control rods.

QUAD CITIES -

UNITS 1 *& 2 3/4.10-4 I

Amendment Nos. 183; 180

. *---":"----- **~--~--: *:"'""~'

IUELING OPERATIONS B 3/4.10 BASES 3/4.1 O.A Reactor Mode Switch Locking the OPERABLE reactor mode switch in the.Shutdown or Refuel position, as specified, ensures that the restrictions on control rod withdrawal and refueling platform movement during the refueling operations are properly activated. These conditions reinforce the refueling procedures and reduce the probability of inadvertent criticality, damage to reactor internals or fuel assemblies, and exposure of personnel to excessive radioactivity.

The addition of large amounts of reactivity to the core Is prevented by operating procedures, which are in turn backed up by refueling interlocks on rod withdrawal and movement of the refueling platform. When the mode switch is in the Refuel position, interlocks prevent the refueling platform from being moved over the core if a control rod is withdrawn and fuel is on a hoist. If the refueling platform is over the core with fuel on a hoist, control rod motion is blocked by the interlocks. With the mode switch in the refuel position only one control rod can be withdrawn.

3/4.10.B Instrumentation The OPERABILITY of at least two source range monitors ensures that redundant monitoring capability is available to detect changes in the reactivity condition of the core, whenever reactor criticality is possible.

The source range monitors (SRM) are provided to monitor the core during periods of station shutdown and to guide the operator during refueling operations and reactor startup.. Requiring two OPERABLE source range monitors in and adjacent to any core quadrant where fuel or control rods are being moved assures adequate monitoring of that quadrant during such alterations. Requiring a minimum of 3 counts per second whenever criticality is possible provides assurance that neutron flux is being monitored. The SRM system is designed to provide a signal-to-noise ratio of at least 3: 1 and a count rate of at least. 3 counts per second. Criticality is considered to be impossible if there are no more than two assemblies in a quadrant and if these are in locations adjacent to the source range monitors (i.e., spatially separated).

Special movable detectors may be used during CORE ALTERATION(s) in place of the normal SRM neutron detectors. These special detectors must be connected to the normal SRM circuits such that the applicable neutron.flux indication, control rod blocks and scram signals can be generated.

The special detectors provide more flexibility in monitoring reactivity changes during fuel loading since they can be positioned anywhere within the core during refueling provided they meet the location requirements of the specification.

  • QUAD CITIES -

UNITS 1 & 2 B 3/4.10-1 Amendment Nos. 183; 180

. -**---*---*---- -----*-------*- ---~---,---------*-...---------- ----*

I

SPECIAL TEST EXCEPTIONS 3.12 - LIMITING CONDITIONS FOR OPERATION B.

SHUTDOWN MARGIN Demonstrations The provisions of Specifications 3.1 O.A and 3.1 O.C and Table 1-2 may be suspended to permit the reactor f!10de switch to be in the Startup position and to allow more than one control rod to be withdrawn for

  • SHUTDOWN MARGIN demonstration, provided that at least the following requirements are satisfied.
1. The source range monitors are OPERABLE per Specification 3.1 O.B.
2. The rod worth.minimizer is OPERABL,.E.

per Specification 3.3.L and is programmed for the SHUTDOWN MARGIN demonstration, or conformance with the SHUTDOWN MARGIN demonstration procedure is verified by a second licensed operator

. or other technically qualified individual.

3. The "rod~out-notch-override" control shall not be used during out-of-sequence movement of the control rods.
4.

No other CORE AL TERATION(s) are in progress.

APPLICABILITY:

OPERATIONAL MODE 5, during SHUTDOWN MARGIN demonstrations.

ACTION:

With the requirements of the above specification not satisfied, immediately place the reactor mode switch in the Shutdown or Refuel position.

SDM 3/4.12.B 4.12 - SURVEILLANCE REQUIREMENTS B.

SHUTDOWN MARGIN Demonstrations Within 30 minutes prior to and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during the performance of a SHUTDOWN MARGIN demonstration, verify that; 1. The source range monitors are OPERABLE per Specification 3.1 O.B,

2.

The rod worth minimizer is OPERABLE with the required program per

  • Specification 3.3.L or a second licensed operator or other technically qualified individual is present and verifies compliance with the SHUTDOWN MARGIN demonstration procedures, and
3.

No other CORE AL TERATION(s) are in progress.

QUAD CITIES - UNITS 1 & 2 3/4.12-2 Amendment Nos.183; 180

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