ML20205B322

Amends 143 & 134 to Licenses DPR-42 & DPR-60,respectively, Revise TS 3.2 & Table 3.5-2B to Allow Limited Inoperability of Bast Level Channels & Transfer Logic Channels to Provide for Required Testing & Maintenance of Associated Components

ML20205B322
Person / Time
Site:	Prairie Island
Issue date:	03/17/1999
From:	Kim T NRC (Affiliation Not Assigned)
To:
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ML20205B328	List: ML20205B322 ML20205B335
References
NUDOCS 9903310217
Download: ML20205B322 (20)

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t UNITED STATES g

j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20566 4001

.....l NORTHERN STATES POWER COMPANY DOCKET NO. 50-282 PRAIRIE ISLAND NUCLEAR GENERATING PLANT. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 143 License No. DPR-42 1.

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

A.

The app;ication for amendment by Northern States Power Company (the licensee) dated November 25,1998, 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 of the pubile, 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-42 is hereby amended to read as.follows:

9903310217 990317 PDR ADOCK 05000282 P

PDR

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

143

, 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 issuance, with full implementation within 30 days.

FOR THE NUCLEAR REGULATORY COMMISSION

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Tae Kim, Senior Project Manager Project Directorate lil-1 Division of Licensing Project Management Onice of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of issuance: March 17, 1999 4

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f ATTACHMENT TO LICENSE AMENDMENT NO.

143 4

FACILITY OPERATING LICENSE NO. DPR-42 i

DOCKET NO. 50-282 4

Revise Appendix A Technical Specifications by removing the pages identified below and J

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

REMOVE INSERT i

TS.3.2-1 TS.3.2-1 i

Table TS.3.5-2B (Page 6 of 9)

Table TS.3.5-28 (Page 6 of 9)

Table TS.3.5-2B (Page 9 of 9)

Table TS.3.5-2B (Page 9 of 9) 4 B.3.5-1 B.3.5-1 B.3.5-4 B.3.5-4 B.3.5-5 B.3.5-5 B.3.5-6 4

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1-TS.3.2-1 d

3.2 CHEMICAL AND VOLUME CONTROL SYSTEM Annliembility Applies to the operational status of the chemical and volume control system.

Old.ECliER To define those conditions of the chemical and volume control system necessary to assure safe reactor operation and safe COLD SHUTDOWN.

Specification 1

A.

When fuel'is in a reactor and reactor coolant system average temperature is at or below 2009F there shall be at least one flow path to the core for. boric acid injection. If no OPERABLE flow path exists, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

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B.

A reactor shall not be made or unintained critical nor shall the reactor coolant system average temperature exceed 2009F unless the i

following conditions are satisfied (except as specified in 3.2.C or 3.2.D below, or Table TS.3.5-25):

l 1.

Two of the three charging pumps shall be OPERABLE, a

2.

At least one boric acid tank shall be aligned to the unit and shall contain a minimum of 2000 gallons of 11.5% to 13% by weight j

boric acid solution at a temperature of at least 1459F.

3.

System piping valves and pumps shall be OPERABLE to the extent of catablishing two independent flow paths for boric acid injection -- one flow path from the boric acid tanks to the core

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and one flow path from the refueling water storage tank to the core.

4.

Two channels of heat tracing shall be OPERABLE for,the flow paths from the boric acid tanks required to meet the requirements of Specification 3.2.B.3.

5.

Automatic valves, piping, and interlocks associated with the above components which are required to operate for the steam line break i

accident are OPERABLE.

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' Prairie Island Unit 1 Amendment No. 91, 143 Prairie Island Unit 2 Amendment No. Bd.134

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RR 33 00 TAafE TS.3.5-2B (Page 6 of 9) ww s

OD ENGINEERED. SAFETY FEATURE ACTUATION sisinri INSTRUNENTATION "98 o.a.

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ES TOTAL NO.

CHANNELS LHAMELE APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE NODES ACTION w

8.

LOSS OF POWER

a. Degraded Voltage 4/ Bus 2/ Bus 3/ Bus 1.2.3.4

31. 32. 33 4kV Safeguards Bus (2/ phase on (1/ phase 2 phases) on 2 phases)

b. Undervoltage 4/ Bus 2/ Bus 3/ Bus 1.2.3.4

31. 32. 33 4kV Safeguards Bus (2/ phase on (1/ phase t

2 phases) on 2 phases) 9.

BORIC ACID STORAGE TANK f

a. Lo-Lo Level 2 channels 1 sensor 2 sensors 1.2.3.4 34 with 2 per in one sensors per channel channel gg channel in both wa channels i

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b. Automatic Actuation Logic 2

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1.2.3.4 35, 36 gg and Actuation Relays i

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ACTION 30: With the number of OPERABLE channels ACTION 33: If the requirements of ACTIONS 30 or 31

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ELEl one less than the Total Number of cannot be met within the time Channels, declare the associated specified or with the number of t

ks >a auxiliary feedwater pump inoperable and OPERABLE channels three less than the take the action required by Total Number of Channels, declare the Specification 3.4.2. Houaver, one associated diesel generator (s)

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channel may be bypassed for up to 8 inoperable and take the ACTION required i

hours for surveillance t-sting per by Specification 3.7.B.

i Specification 4.1, provided the other i

channel is OPERABLE.

ACTION 34: With the number of OPERABLE channels less than the Total ACTION a: With the number of OPERABLE channels Number of Channels, operation may one less than the Total P-aber of proceed provided the inoperable Channels. operation in the applicable channel is placed in the tripped NODE may proceed provided the condition within 6 hourc and the inoperable channel is placed in the Minimum Channels OPERABLE f

bypassed condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

requirement is met. Restore the inoperable channel to OPERABLE ACTION 32: With the number of OPERABLE channels status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at i

two less than the Total Number of

't NOT SHUTDOWN within the next i

Channels operation in the applicable e

atre and in COLD SHUTDOWN l

NODE may proceed provided the folicwing wis.in the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

conditions are satisfied:

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ACTION 35: With one channel inoperable.

a. One inoperable channel is placed in restore the inoperable channel to the bypassed condition within 6 OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or i

hours. and, be in at least NOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD t

b. The other inoperable channel is SKUTDOWN within the following 30 placed in the tripped condition hours.

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gg within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and, ge i

ACTION 36: Two channels may be inoperable for up M

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c. All of the channels associated with to I hour for surveillance testing per s

j gg the redundant 4kV Safeguards Bus Specification 4.1.

Restore at least

&P eu are OPERABLE.

one channel to OPERABLE status within W

I this I hour or initiate the action w

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necessary to place the affected unit in 4

i HOT SHUTDOWN. and be in at least HOT ts j

ww SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in M

COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

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B.3.5-1 3.5 INSTRUMENTATION SYSTBt DEnnE Instrumentation has been provided to sense accident conditions and to initiate reactor trip and operation of the Engineered Safety Features (Reference 1).

The OPERABILITY of the Reactor Trip System and the Engineered Safety System instrumentation and interlocks ensures that: (1) the associated ACTION and/or reactor trip will be initiated when the parameter monitored by each channel or combination thereof reaches its setpoint. (2) the specified coincidence logic and sufficient redundancy is maintained to permit a channel to be out of service for testing or maintenance consistent with maintaining an appropriate level of reliability of the Reactor Protection and Engineered Safety Features instrumentation and. (3) sufficient system functions capability is available from diverse parameters.

The OPERABILITY of these systems is required to provide the overall reliability, redundancy and diversity assumed available in the facility design for the protection and mitigation of accident and transient conditions. The integrated operation of each of these systems is consistent with the assumptions used in the safety analysis.

Specified surveillance and maintenance outage times have been determined in accordance with WCAP-20271. " Evaluation of Surveillance Frequencies and out of Service Times for the Reactor Protection Instrumentation System".

and supplements to that report. Out of service times were determined based on maintaining an appropriate level of reliability of the Reactor Protection System and. Engineered Safety Features instrumentation.

(Boric Acid Storage Tank instrumentation which provides automatic *.ransfer of safety injee+. ion suction was not modeled in this WCAP.)

The evalua? on of surveillance frequencies and out of service times for the reacto; protection and engineered safety feature instrumentation described in 'JCAP-10271 included the allowance foi testing in bypass. The evaluation assumed that the average amount of time the channels within a given trip function would be in bypass for testing was 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Safety Injection The Safety Injection System is actuated automatically to p'rovide emergency cooling and reduction of reactivity in the event of a loss-of-coolant accident or a steam line break accident.

Safety injection in response to a loss-of-coolant accident (LOCA) is provided by a high containment pressure signal backed up by the low pressurizer pressure signal. These conditions would accompany the depressurization and coolant loss during a LOCA.

Safety injection in response to a steam line break is provided directly by a low steam line pressttre signal, backed up by the low pressuriser pressure signal.and, in case of a break within the containment. by the high containment pressure signal.

The safety injection of highly borated water will offset the temperature-induced reactivity addition that could otherwise result from cooldown following.a steam line break.

Prairie Island Unit 1 Amendment No. 97, 177. 143 Prairie Island Unit 2 Amendment No. 84. 194. 134

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3.3.5-4 4

2 3.5 INSTRUMENTATION SYSTEM I

j Bases continued i

l Automatic Transfer of Safety Injection Suetiorl V

The plant is equipped with three boric acid storage tanks for the'two 1,

units. One tank is normally aligned to the safety injection system for each unit. Following initiation of the Engineered Safety Features, the safety injection pumps take suction from the aligned boric acid storage s

tank. When the boric acid storage tank level falls to the lo-lo level, an j

interlock automatically transfers the' safety injection pumps suction fron the boric acid ratorage tank to the refueling water storage tan *.

The g

boric acid storage tank that is not aligned to either unit. including its associated piping and interlocks, is not required to be OPERABLE.-

2.initing Instrument Setpoints 4

1.

The high containment pressure limit is set at about 10% of the max 4 mum internal pressure. Initiation of Safety Injection protects against i

loss of coolant (Reference 2) or steam line break accidents as j

discussed in the safety analysis.

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2.

The Hi Hi containment pressure limit is set at about 50% of the maximum internal pressure for initiation of containment spray and at 3

1 about 30% for initiation of steam line isolation. Initiation of

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Containment Spray and Steam Line Isolation protects against large loss of coolant (Reference 2) or steam line break accidents (Reference 3) as discussed in the safety analysis, j

3.

The pressurizer low pressure limit is set substantially below system i

operating pressure limits. However, it is sufficiently high to protect against a loss of coolant accident as shown in the safety analysis (Reference 2).

4.

The steam line low pressure signal is lead /las compensated and its i

set-point is set well above the pressure expected in the event of a large steam line break accident as shown in the. safety analysis (Reference 3).

s 5.

The high steam line flow limit is set at approximately 20% of nominal full-load flow at the r.o-load pressure and the high-high steam line flow' limit is set at approximately 120% of nominal full-load flow at the full load pressure in order to protect against large steam break accidents. The coincident low T.

setting limit for steam line isolation initiation is set belev its hot shutdown value. The safety analysis shows that these settings provide protection in the event of a large steam break (Reference 3).

6.

Steam generator low-low water level and 4.16 kV Bus 11 and 12 (21 and 22 in Unit 2) low bus voltage provide initiation signals for the Auxiliary Feedwater System. Selection of these setpoints is diseassed in the Bases of Section 2.3 of the Technical Specification.

Prairie Island Unit 1 Amendment No. 91, 153, Ill. 143

'Priirie Island Unit 2-Amendment No. 84, 96, 194. 134

B.3.5-5 l

3.5 TNSTRUMENTATION SYSTEM Banas continued Limiting Instrument Setpoints (continued)

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High radiation signals providing input to the Containment Ventilation i

Isolation circuitry are set in accordance with the Radioactive

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Effluent Technical Specifications. The setpoints are established to

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prevent exceeding the limits of 10 CPR Part 20 at the SITE BOUNDARY.

8.

The' degraded voltage protection setpoint is 194.8% and $96.2% of j

nominal 4160 V bus voltage. Testing and analysis have shown that all j

safeguards loads will operate properly at or above the minimum i

degraded voltage setpoint. The maximum degraded voltage setpoint.is chosen to prevent unnecessary actuation of the voltage restoring scheme at the minimum expected grit voltage. The first degraded voltage time delay of 8 i 0.5 seconds has been shown by testing and analysis to be long enough to allow for no: mal transients (i.e.

motor starting and fault' clearing). It is also longer than the time required to start the safety injection pump at minimum voltage. The second degraded voltage time delay is provided to allow the degraded voltage condition to be corrected within a time frame which vill not cause damage to permanently connected Class 1E loads.

The un'arvoltage setpoint is 75 2.5% of nominal bus voltage..The minimum setpoint ensures equipment operates above the limiting value of 75% (of 4000 V) for one minute operation. The 75% maximum setpoint is chosen to prevent unnecessary actuation of the voltage restoring scheme during voltage dips which occur during motor starting. The undervoltage time delay of 4 1.5 seconds has been shown by testing and analysis to be long enough to allow for normal transients and short enough to operate prior to the degraded voltage logic, providing a rapid transfer to an alternate sourte.

Instrument Operating Conditions During plant operations, the complete instrumentation systems will normally be in service. Reactor safety is provided by the Reactor Protection System, which automatically initiates appropriate action to prevent exceeding established limits. Safety is not compromised, however, by continuing operation with certain instrumentation channels out of service since provisions were made for this in the plant design. This specification outlines limiting conditions for operation necessary to

. preserve the effectiveness of the Reactor Control and Protection System when any one or more of the channels is out of service.

Almost all reactor protection channels are supplied with sufficient redundancy to provide the capability for CHANNEL CALIBRATION and test at power. Exceptions are backup channels such as reactor coolant pump breakers. The removal of one trip channel'on process control equipment is accomplished by placing that channel bistable in a tripped mode: e.g.. a two-out-of-three circuit becomes a one out of two circuit. The source and intermediate range nuclear instrumentation system channels are not Prairie Island Unit 1 Amendment No. 91, 163, Ill 143 Prairie Island Unit 2:

Amendment No. M, N. JN, 134

3.3.5-6 3.5 INSTRLMENTATION SYSTEM Bases continued Instrument Operating Conditions (continued)

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intentionally placed in a tripped mode since these are one-out of-two trips, and the trips are therefore bypassed during testing. Testing does not trip the system unless a trip condition exists in a concurrent channel.

References 1.

USAR. Section 7.4.2 2.

USAR. Section 14.6.1 3.

USAR. Section 14.5.5 4.

FSAR. Appendia I Prairie Island Unit 1 Amendment No. IH, III, 142 i

Prairie Island Unit 2 Amendment No. 96, IN,134 l

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t-UNITED STATES g

j NUCLEAR REGULATORY COMMISSION i

2 WASHINGTON, D.C. 30seH001 i

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l NORTHERN STATES POWER COMPANY DOCKET NO. 50-306 PRAIRIE ISLAND NUCLEAR GENERATING PLANT. UNIT 2 i

j AMENDMENT TO FACILITY OPERATING LICENSE 1

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Amendment No.134 License No. DPR 60 1.

The Nuclear Regulatory Commission (the Commission) has found that; a

A.

The application for amendment by Northern States Power Company (the j

licensee) dated November 25,1998, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the i

Commission's rules and regulations set forth in 10 CFR Chapter I;.

i B.

The facility will operate in conformity with the application, the provisions of the j

Act, and the rules and regulations of the Commission; 1

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 defew 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-60 is hereby amended to read as follows:

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t Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. '134

, are hereby incorporated in the license. The licensee l

shall operate the facility in accordance with the Technical Specifications.

l 3.

This license amendment is effective as of the date of issuance, with full implementation l

within 30 days.

FOR THE NUCLEAR REGULATORY COMMISSION M

Tae Kim, Senior Project Manager l

Project Directorate ill-1 l

Division of Licensing Project Management Office of Nuclear Reactor Regulation Attachm..nu Changes to the Technical j

Specifications l

Date of Issuance: March 17, 1999 L

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ATTACHMENT TO LICENSE AMENDMENT NO.134 FACILITY OPERATING LICENSE NO. DPR-60 DOCKET NO. 50-306 Revise Appendix A Technical Soecifications by removing the pages identified below and inserting the attached pages. i nc mvised pages are identified by amendment number and contain vertical lines indicating the area of change.

REMOVE INSERT i

TS.3.2-1 TS.3.2-1 Table TS.3.5-28 (Page 6 of 9)

Table TS.3.5-2B (Page 6 of 9)

Table TS.3.5-2B (Page 9 of 9)

Table TS.3.5-2B (Page 9 of 9) i B.3.5-1 B.3.5-1 B.3.5-4 B.3.5-4 B.3.5-5 B.&.5-5 B.3.5-6 a

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T TS.3.2-1 3.2 CHEMICAL AND VOLUME CONTROL SYSTEM Anelicab414tv Applies to the operational status of the chemical and volume control system.

Obdeetive To define those conditions of the chemical and volume control system necessary to assure safe reactor operation and safe COLD SHUTDOWN.

Specification A.

When fuel is in a reactor and reactor coolant system average temperature is at'or below 200*F there shall be at least one flow path to the core for boric acid injection. If no OPERABLE flow path exists, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

B.

A reactor shall not be made or maintained critical nor shall the reactor coolant system average temperature exceed 2009r unless the following conditions are satisfied (except as specified in 3.2.C or 3.2.D below, or Table TS.3.5-23):

l 1.

Two of the three charging pumps shall be OPERABLE.

2.

At least one boric acid tank shall be aligned to the unit and shall contain a minimum of 2000 gallons of 11.5% to 13% by weight boric acid solution at a temperature of at least 145'F.

3.

System piping, valves and pumps shall be OPERABLE to the extent of establishing two independent flow paths for boric acid injection -- one flow path from the boric acid tanks to the core and one flow path from the refueling water storage tank to the core.

4.

Two channels of heat tracing shall be OPERABLE for,the flev paths l

from *.he boric acid tanks required to meet the requirements of Specification 3.2.B.3.

5.

Automatic valves, piping, and interlocks associated with the above l

components which are required to operate for the steam line break accident are OPERABLE.

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' Prairie Island Unit'l Amendment No. 91, 143 Prairie Island Unit 2 Amendment No. 84, 134 4

• u T11:s er Ed 08 TABLE TS.3.5-2B (Page 6 of 9) tN yy ENGINEERED SAFETY FEATURE ACTUATION sisir.n INSTRUMENTATION EE ce MININUN i

SS TOTAL NO.

CHANNELS GBeHNELS APPLICABLE FUNCTIONAL UNIT OF. CHANNELS TO TRIP OPERABLE MODES ACTION new 8.

LOSS OF POWER l

a. Degraded Voltage 4/ Bus 2/ Bus 3/ Bus 1.2.3.4

31. 32, 33 4kV Safeguards Bus (2/phas,e on (1/ phase p

2 phases) on 2 phases)

b. Undervoltage 4/ Bus 2/ Bus 3/ Bus 1.2.3.4

31. 32. 33 4kV Safeguards Bus (2/ phase on (1/ phase 1

2 phases) on 2 i

phases) 9.

BORIC ACID STORAGE TANK

a. Lo-Lo Level 2 channels 1 sensor 2 sensors 1.2.3.4 34 with 2 per in one sensors per channel channel gg channel in both i

.a channels EE gg

b. Automatic Actuation Logic 2

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1.2.3.4 35, 36 gg and Actuation Relays y

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Action Statements 99 c6 ou c: gg ACTION 30: With the number of OPERABLE channels ACTION 33: If the requirements of ACTIONS 30 or 31 Er one less than th.

Ntal Number of cannot be met within the time Channels, declare he associated specified. or with the number of k2 w auxiliary feedwater pump inoperable and OPERABLE channels three less than the take the action required by Total Number of Channels, declare the i

specification 3.4.2. However, one associated diesel generator (s) channel may be bypassed for up to 8 inoperable and take the ACTION required hours for surveillance testing per by Specification 3.7.B.

Specification 4.1. provided the other channel is OPERABLE.

ACTION 34: With the number of OPERABLE channels less than the Total ACTION 31: With the number of OPERABLE channels Number of Channels. operation may one less than the Total Number of proceed provided the inoperable Channels. operation in the applicable channel is placed in the tripped NODE may proceed provided the condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and the inoperable channel is placed in the Minimum Channels OPERABLE bypassed condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

requirement is met. Restore the inoperable channel to OPERABLE e

ACTION 32: With the number of OPERABLE channels status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at two less than the Total Number of least NOT SHUTDOWN within the next Channels, operation in the applicable 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN NODE may proceed provided the following within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

conditions are satisfied:

ACTION 35: With one channel inoperable, j

a. One inoperable channel is placed in restore the inoperable channel to the bypassed condition within 6 OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or hours. and.

be in at least NOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD

b. The other indperable channel is SHUTDOWN within the following 30 placed in the tripped condition hours.

7'4g within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and.

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ACTION 36: Two :hannels may be inoperable for up M

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c. All of the channels associated with to I hour for surveillance testing per

• s gg the redundant 4kV Safeguards Bus Specification 4.1.

Restore at least S."

oo are OPERABLE.

one channel to OPERABLE status within W

this 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or initiate the action u

jpg necessary to place the affected unit in a

HOT SHUTDOWN. and be in at least HOT ts SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in l

]R;2 COLD SHUTDOWN.within the following 30 I

hours.

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B.3.5-1 3.5 INSTRUMENTA1 ION SYSTEM anans Instrumentation has been provided to sense accident conditions and to

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initiate reactor trip and operation of the Engineered Safety Features (Reference 1).

The OPERABILITY of the Reactor Trip System and the Engineered Safety System instrumentation and interlocks ensures that: (1) the associated ACTION and/or reactor trip will be initiated when the parameter monitored by each channel or combination thereof reaches its setpoint. (2) the'specified coincidence logie and sufficient redundancy is maintained to permit a channel to be out of service for testing or maintenance consistent with maintaining an app ~ropriate level of reliability of the Reactor Protection and Engineered Safety Features instrumentation and. (3) sufficient system functions capability is 4

available from diverse parameters.

The OPERABILITY of these systems is required to' provide the overall reliability, redundancy and diversity assumed available in the facility design for the protection and mitigation of accident and transient conditions. The integrated operation of each of these systems is consistent with the assumptions used in the safety analysis.

4 Specified surveillance and maintenance outage times have been determined in accordance with WCAP-10271. " Evaluation of Surveillance Frequencies and Out of Service Times for the Reactor Protection Instrumentation System".

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and supplements to that report. Out of' service times were determined based on maintaining an appropriate level of reliability of the Reactor Protection System and Engineered Safety Features instrumentation.

(Boric Acid Storage Tank instrumentation which provides automatic transfer of safety injection suction was not modeled in this WCAP.)

i The evaluation of surveillance frequencies and out of service times for the reactor protection and engineered safety feature instrumentation described in WCAP-10271 included the allowance for testing in bypass. The evaluation assumed that the average amount of time the channels within a given trip function would be in bypass for testing was 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Safety Injection The Safety Injection System is actuated automatically to provide emergency j

cooling and reduction of reactivity in the event of a loss-of-coolant accident or a steam line break accident.

Safety injection in response to a loss-of-coolant accident (LOCA) is provided by a high containment pressure signal backed up by the low pressuriser pressure signal. These conditions would accompany the depressurization and coolant loss during a LOCA.

Safety injection in response to a steam line break is provided directly by 1.

a low steam line pressure signal, backed up by the low pressuriser pressure signal and, in case of a break within the containment. by the p

high containment pressure signal.

{

The safety injection of highly borated water will offset the a

temperature-induced reactivity addition that could otherwise result from cooldown following a steam line break.

j Prairie Island Unit 1 Amendment No. PT. 177, 143 Prairie Island Unit 2 Amendment No. 84. 164, 134 i

B.3.5 4 3.5 INSTRUMENIAIIQtLIIIIIH Aatas continued Automatic Transfer of Safety Injection Suction j

The plant is equipped with three boric acid storage tanks for tho'two units. One tank is normally aligned to the safety injection system for each unit. Following initiation of the Engineered Safety Features, the safety injection pumps take suction from the aligned boric acid storage tank. When the boric acid storage tank level falls to the lo-lo level, an interlock automatically transfers the' safety injection pumps suction from the boric acid storage tank to the refueling water storage tank. The boric acid storage tank that is not aligned to either unit, including its associated piping and interlocks, is not required to be OPERABLE.

Limiting Instrument Satpoints 1.

The high containment pressure limit is set at about 10% of the maximum internal pressure. Initiation of Safety Injection protects against loss of coolant (Reference 2) or steam line break accidents as discussed in the safety analysis.

2.

The Hi-Hi containment pressure limit is set at about 50% of the maximum internal pressure for initiation of containment spray and at about 30% for initiation of steam line isolation. Initiation of Containment Spray and Steam Line Isolation protects against large loss of coolant (Reference 2) or steam line break accidents (Reference 3) as discussed in the safety analysis.

3.

The pressurizer low pressure limit is set substantially below system operating pressure limits. However, it is sufficiently high to protect against a loss of coolant accident as shown in the safety analysis (Reference 2).

4.

The steam line low pressure signal is lead /las compensated and its set point is set well above the pressure expected in the event of a large steam line break accident as shown in the. safety analysis (Referance 3).

5.

The high steam line flow limit is set at approximate 1y'20% of nominal full-load flow at the no-load pressure and the high-high steam line flow limit is set at approximately 120% of nominal full-load flow at the full load pressure in order to protect against large steam break accidents. The coincident low T., setting limit for steam line isolation initiation is set below its hot shutdown value. The safety analysis shows that these settings provide protection in the event of a large steam break (Reference 3).

6.

Steam generator low-low water level and 4.16 kV Bus 11 and 12 (21 and 22 in Unit 2) low bus voltage provide initiation signals for the Auxiliary Feedwater System. Selection of these setpoints is discussed

~in the Bases of Section 2.3 of the Technical Specification.

Prairie Island Unit 1 Amendment No. 91, 103, III, 143 Pr&irie Island Unit 2 Amendment No. 54, 96, 164, 134

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B.3.5-5 I

3.5 INSTRUMENTATION SYSTEN l

3g333 continued Limiting Instrument Setpoints (continued) 3 7.

High radiation signals providing input to the Containment Ventilation Isolation circuitry are set in accordance with the Radioactive Effluent Technical Specifications. The setpoints are established to prevent exceeding the limits of 10 CFR Part 20 at the SITE BOUNDARY.

8.

The degraded voltage protection setpoint ir 194.8% and $96.2% of nominal 4160 V bus voltage. Testing and analysis have shown that all safeguards loads will operate properly at or above the minimum degraded voltage setpoint. The maximum degraded voltage setpoint is i

chosen to prevent unnecessary actuation of the voltage restoring scheme at the minimum expected grid vol' age. The first degraded t

voltage time delay of 8 i 0.5 seconds has been shown by testin5 and analysis to be long enough to allow for normal transients (i.e.. motor starting and fault clearing). It is also longer than the time required to start the safety injection pump at minimum voltage. The second degraded voltage time delay is provided to allow the degraded voltage condition to be corrected within a time frame which trill not cause damage to permanently connected Class 1E loads.

The undervoltage setpoint is 75 i 2.5% of nominal bus voltage. The minimum setpoint ensures equipment operates above the limiting value of 75% (of 4000 V) for one minute operation. The 75% marinum setpoint is chosen to prevent unnecessary actuation of the voltage restoring scheme during voltage dips which occur during motor starting. The undervoltage time delay of 4 1.5 seconds has been shown by testing and analysis to be long enough to allow for normal transients and short enough to operate prior to the degraded voltage logic providing a rapid transfer to an alternate source.

Instrument Operating Conditions During plant operations, the complete instrumentation systems will normally be in service. Reactor safety is provided by the Reactor Protection System, which automatically initiates appropriate action to prevent exceeding established limits. Safety is not erpromised, however, by continuing operation with certain instrumentation. Snannels out of service since provisions were made for this in the plant design. This specification outlines limiting conditions for operation necessary to preserve the effectiveness of the Reactor Control and Protection System when any one or more of the channels is out of service.

Almost all reactor protection channels are supplied with sufficient redundancy to provide the capability for CHANNEL CALIBRATION and test at power. Exceptions are backup channels such as reactor coolant pump

. breakers. The removal of one trip channel on process control equipment is accomplished by placing that channel bistable in a tripped moder e.g.

a two-out-of-three circuit becomes a one out-of-two circuit. The source and intermediate range nuclear instrumentation system channels are not Prairie Island Unit *1 Amendment No. 91, 103, fil. 143 Prairie Island Unit 2-Amendment No. 84, 96, 164. 134

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B.3.5-6

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3.5 INSTRIDENTATION SYSTEM Bases continued Instrument Operating Conditions (continued) l intentionally placed in a tripped mode since these are one-out of-two trips. and the trips are therefore bypassed during testing. Testing does not trip the system unless a trip condition exists in a concurrent channel.

I Rafarances 1.

USAR. Section 7.4.2 2.

USAR. Section 14.6.1 3.

USAR Section 14.5.5 4.* FSAR Appendix I l

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1 Prairie Island Unit 1 Amendment No. 103, 111, 143 Prairie Island Unit 2 Amendment No. 96, 164, 134 t

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