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Amend 41 to License NPF-43,revising Tech Spec 3/4.3.2 & Associated Tables Re Containment Isolation Function & Secondary Containment Isolation Function to More Clearly Reflect Plant Configuration & Remove Duplication
	ML20247E634
Person / Time
Site:	Fermi
Issue date:	09/07/1989
From:	Thoma J; Office of Nuclear Reactor Regulation
To:	;
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ML20247E637	List: ML20247E634; ML20247E644; ML20247E649;
References
	NUDOCS 8909150338
	Download: ML20247E634 (23)
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r DETROIT EDISON COMPANY WOLVERINE POWER SUPPLY COOPERATIVE, INCORPORATED DOCKET NO. 50-341 FERMI-2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 41~

License No. NPF-43 1.

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

A.

The application for amendment by the Detroit Edison Company (the licensee) dated December 22, 1988, 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; q

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 Specifica-tions as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. NPF-43 is hereby amended to read as follows:

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 41, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license.

DECO shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

8909150338 890907 PDR ADOCK 03000341 P

PDC l

3.

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

FOR THE NUCLEAR REGULATORY COMMISSION

'h-John 0. Thoma, Acting Director Project Directorate III-1 Division of Reactor Projects - III, IV, V & Special Projects Office of Nuclear Reactor Regulation a

Attachment:

Changes to the Technical Specifications Date of Issuance:

September 7, 1989 4

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.1 ATTACHMENT TO LICENSE AMEN 0 MENT NO. 41-FACILITY OPERATING LICENSE NO. NPF-43 DOCKET NO. 50-341 o

/ Replace.the:following pages of the Appendix "A" Technical Specifications with the attached pages.

The revised pages are identified by Amendment number and.

Econtain'a vertical line indicating the area of change..The corresponding overleaf pages are also'provided to maintain document completeness.

REMOVE INSERT 1-1 1-1 3/4 3-9 3/4 3-9 3/4 3-11 3/4 3-11 3/4 3-12 3/4 3-12 3/4 3-13 3/4 3-13 3/4 3-14.

3/4 3-14 3/4 3-14a 3/4 3-15 3/4 3-15 3/4 3-16 3/4 3 3/4 3-17a ~

3/4 3-18 3/4 3-18 3/4 3-19 3/4 3-19 3/4 3-20 3/4 3-20 3/4 3-21 3/4 3-21 3/4 3-22 3/4 3-22 3/4 6-45 3/4 6-45 3/4 6-46 3/4 6-46 i

i 4

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1. 0 DEFINITIONS 3

i l

The following terms are defined so that uniform interpretation of these specifications may be achieved.

The defined terms appear in capitalized type and shall be applicable throughout these Technical Specifications.

l ACTION 1.1 ACTION shall be that part of a Specification which prescribes remedial j

measures required under designated conditions.

i I

AVERAGE PLANAR EXPOSURE I

1. 2 The AVERAGE PLANAR EXPOSURE shall be applicable to a specific planar height 1

and is equal to the sum of the exposure of all the fuel rods in the l

specified bundle at the specified height divided by the number of fuel rods in the fuel bundle.

AVERAGE PLANAR LINEAR HEAT GENERATIN RATE

1. 3 The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) shall be applicable to a specific planar height and is equal to the sum of the LINEAR HEAT GENERATION RATES for all the fuel rods in the specified bundle at the specified height divided by the number of fuel rods in the fuel bundle.

CHANNEL CALIBRATION

1. 4 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds with the necessary range and accuracy to known values of the parameter which the channel monitors.

The CHANNEL CALIBRATION shall encompass the entire channel including the sensor and alarm and/or trip functions, and shall include the CHANNEL FUNCTIONAL TEST.

The CHANNEL CALIBRATION may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is calibrated.

Calibration of instrument channels with resistance temperature detectors (RTD) or thermocouple sensors shall consist of verification of operability of the sensing element and adjustment, as necessary, of the remaining adjustable devices in the channel.

CHANNEL CHECK

1. 5 A CHANEL CHECK shall be the qualitative assessment of channel behavior during operation by observation.

This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter.

CHANNEL FUNCTIONAL TEST

1. 6 A CHANNEL FUNCTIONAL TEST shall be:

a.

Asialog channels - the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY l

including alarm and/or trip functions and channel failure trips.

b.

Bistable channels - the injection of a simulated signal into the sensor to verify OPERABILITY including alarm and/or trip functions.

1 The CHANNEL FUNCTIONAL TEST may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is tested.

FERMI - UNIT 2 1-1 Amendment No. 41

DEFINITIONS CORE ALTERATION 1.7 CORE ALTERATION shall be the addition, removal, relocation or movement of fuel, sources, incore instruments or reactivity controls within the reactor pressure vessel with the vessel head removed and fuel in the vessel.

Normal movement of SRMs, IRMs, TIPS, or special movable detectors is not considered a CORE ALTERATION.

Suspension of CORE ALTERATIONS shall not preclude completion of the movement of a component to a safe conservative position.

CRITICAL POWER RATIO 1.8 The CRITICAL POWER RATIO (CPR) shall be the ratio of that power in the assembly which is calculated by application of the GEXL correlations to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.

DOSE EQUIVALENT I-131 1.9 DOSE EQUIVALENT I-131 shall be that concentration of I-131, microcuries per gram, which alone would produce the same thyroid dose as the quLntity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present.

The thyroid dose conversion factors used for this calcula':on shall be those listed.in Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites."

E-AVERAGE DISINTEGRATION ENERGY 1.10 E shall be the average, weighted in proportion to the concentration of each radionuclides in the reactor coolant at the time of sampling, of the sum of the average beta and gamma energies per disintegration, in MeV, for isotopes, with half lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant.

EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME 1.11 The EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ECCS actuation set-point at the channel sensor until the ECCS equipment is capable of performing its safety function, i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.

Times shall include diesel generator starting and sequence loading delays where applicable.

The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.

FRACTION OF LIMITING POWER DENSITY 1.12 The FRACTION OF LIMITING POWER DENSITY (FLPD) shall be the LHGR ex. sting at a given location divided by the specified LHGR limit for that bundle type.

FRACTION OF RATED THERMAL POWER 1.13 The FRACTION OF RATED THERMAL POWER (FRTP) shall be the measured THERMAL POWER divided by the RATED THERMAL POWER.

FERMI - UNIT 2 1-2

. c,

INSTRUMENTATION 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.2 The isolation actuation instrumentation channels shown in Table 3.3.2-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.2-2 and with ISOLATION SYSTEM RESPONSE TIME as shown in Table 3.3.2-3.

APPLICABILITY:

As shown in Table 3.3.2-1.

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ACTION:

a.

With an isolation actuation instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.2-2, declare the channel inoperable until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.

b.

With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System requirement for one trip system, place the inoperable channel (s) and/or that trip system in the tripped condition

within one hour.

The provisions of Specification 3.0.4 are not applicable.

c.

With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System requirement for both trip systems, place at least one trip system ** in the tripped condition within one hour and take the ACTION required by Table 3.3.2-1.

An inoperable channel need not be placed in the tripped condition where this would cause an isolation to occur.

In these cases, the inoperable channel l

shall be restored to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or the ACTION required by Table 3.3.2-1 for that Trip Function shall be taken.

- The trip system need not be placed in the tripped condition if this would cause an isolation to occur.

When a trip system can be placed in the tripped condition without causing an isolation to occur, place the trip system with the most inoperable channels in the tripped condition; if both systems have the same number of inoperable channels, place either trip system in the tripped condition.

FERMI - UNIT 2 3/4 3-9 Amendment No. 41

L INSTRUMENTATION SURVEILLANCE REQUIREMENTS 4.3.2.1 Each isolation actuation instrumentation channel shall-be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and.at the l

. frequencies shown in Table 4.3.2.1-1.

i; l-4.3.2.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of

-all channels shall be performed at least once per 18 months.

4.3.2.3 The ISOLATION SYSTEM RESPONSE. TIME of each isolation trip function shown in Table 3.3.2-3 shall be demonstrated to be within its limit at least.

once per 18 months.

Each test shall include at least one channel per trip system such that all channels are tested at least once every N times'18 months,.

where N is the total number of redundant channels in a specific isolation trip system.

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ISOLATION ACTUATION INSTRUMENTATION ACTION STATEMENTS ACTION 20 Be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 21 Be in at least STARTUP with the associated isolation valves closed within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 22 Be in at least STARTUP within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 23 Close the affected system isolation valves within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and declare the affected system inoperable.

ACTION 24 Establish SECONDARY CONTAINMENT INTEGRITY with the standby gas treatment system operating within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

ACTION 25 Disable in the closed position the affected system isolation valves within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and declare the shutdown cooling mode of RHR inoperable.

ACTION 26 Restore the manual initiation function to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or close the affected system isolation valves within the next hour and declare the affected system inoperable.

ACTION 27 Restore the manual initiation function to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or establish SECONDARY CONTAINMENT INTEGRITY with the Standby Gas Treatment System operating.

TABLE NOTATIONS When handling irradiated fuel in the secondary containment, during CORE ALTERATIONS, or during operations with a potential for draining the reactor vessel.

The high condenser pressure input to the isolation actuation instrumentation may be bypassed during reactor shutdown or for reactor startup when condenser pressure is above the trip setpoint.

Actuates dampers shown in Table 3.6.5.2-1.

(a) A channel may be placed in an inoperable status for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for required surveillance without placing the channel or trip system in the tripped condition provided at least one other OPERABLE channel in the same

-trip system is monitoring that parameter.

In addition, for the HPCI system and RCIC system isolation, provided that the redundant isolation valve, inboard or outboard, as applicable, in each line is OPERABLE and all required actuation instrumentation for that valve is OPERABLE, one channel may be placed in an inoperable status for up to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months for reqs. red surveillance without placing the channel or trip system in the tripped condition.

(b) Also starts the standby gas treatment system.

(c) A channel is OPERABLE if 2 of 4 detectors in that channel are OPERABLE.

(d) This level signal actuates Groups 2, 10, 11, 12, 14, 16, 17, 18, and ***.

(e) This level signal actuates Groups 4, 13 and 15.

FERMI - UNIT 2 3/4 3-14

, Amendment No. 41

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I TABLE 3.3.2-1 (Continued)

ISOLATION ACTUATION INSTRUMENTATION ACTION STATEMENTS (f) Isolates'with simultaneous RCIC Steam Supply Pressure-Low (Isolation Instrumentation) and Drywell Pressure-High'(ECCS Actuation Instrumentation).

(g) Isolates with simultaneous HPCI Steam Supply Pressure-Low (Isolation Actuation Instrumentation) and Drywell Pressure-High (ECCS Actuation Instrumentation).

-(h) Reserved.

(i) Secondary. Containment Isolation Pushbottons.

(j) Thjspressuresignalactuates. Groups 2, 12, 13, 14, 15, 16, 17, 18, and With time delay of 45 seconds.

FERMI - UNIT 2 3/4 3-14a Amendment No. 41

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TABLE 3.3.2-3 ISOLATION ACTUATION SYSTEM INSTRUMENTATION RESPONSE TIME.

TRIP FUNCTION RESPONSE TIME (Seconds)#

1.

PRIMARY CONTAINMENT ISOLATION l

a.

Reactor Vessel Low Water Level 1)

Level 3

< 13(a) 3 2)

Level 2 7 13(a)**

3)

Level 1-7 1.0*/< 13(a)**

1 b.

Drywell Pressure - High h13(a)-

I c.

Main Steam Line m

1)

Radiation - High(b) 713(a)]

< 13(a) 2)

Pressure - Low

3) ~ Flow - High 7 13(a)**

d.

Main Steam Line Tunnel Temperature - High NA e.

Condenser Pressure - High NA f.

Turbine Bldg. Area' Temperature - High NA g.

Deleted I

h.

Manual Initiation NA 2.

REACTOR WATER CLEANUP SYSTEM ISOLATION a.

A Flow - High NA l

b.

Heat Exchanger / Pump /High Energy Piping Area Temperature - High NA c.

Heat Exchanger / Pump / Phase Separator

{

Area Ventilation Temperature AT - High NA d.

SLCS Initiation NA e.

Reactor Vessel Low Water Level - Level 2 5 13(a) f.

Deleted g.

Manual Initiation NA 3.

REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION a.

RCIC Steam Line Flow - High 1 13(a) b.

RCIC Steam Supply Pressure - Low 5 13(a) c.

RCIC Turbine Exhaust Diaphragm Pressure - High NA d.

RCIC Equipment Room Temperature - High NA e.

Manual Initiation NA l

FERMI - UNIT 2 3/4 3-18 Amendment No. 21, 41

_____.____._..-n._-___

L TABLE 3.3.2-3 (Continued)

ISOLATION ACTUATION SYSTEM INSTRUMENTATION RESPONSE TIME TRIP FUNCTION RESPONSE TIME (Seconds)#

4.

,HIGH PRESSURE COOLANT INJECTION SYSTEM ISOLATION a.

HPCI Steam Flow - High

<13(a) b.

HPCI Steam Supply Pressure - Low 713(a) c.

HPCI Turbine Exhaust Diaphragm Pressure - High NA d.

HPCI Equipment Room Temperature - High NA e.

Manual Initiation-NA

.5.

RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION a.

Reactor Vessel Low Water Level - Level 3 NA

' b.

Reactor Vessel (Shutdown Cooling Cut-in Permissive Interlock) Pressure - High NA c.

Manual Initiation NA 6.

SECONDARY CONTAINMENT ISOLATION a.

Reactor Vessel Low Water Level - Level 2

<13(a)

Fuel Pool Ventilation Exhaust Radiation - Higb(b) 713(a) b.

Drywell Pressure - High 713(a) c.

d.

Manual Initiation HA (a) The isolation system instrumentation response time shall be measured and recorded as a part of the ISOLATION SYSTEM RESPONSE TIME.

Isolation system instrumentation response time specified includes diesel generator starting and sequence loading delays.

(b) Radiation detectors are exempt from response time testing.

Response time shall be measured from detector output or the input of the first electronic component in the channel.

Isolation system instrumentation response time for MSIVs only.

No diesel generator delays assumed for MSIVs.

- Isolation system instrumentation response time for associated valves except MSIVs.

Isolation system instrumentation response time specified for the Trip Function actuating each valve group shall be added to isolation time shown in Table 3.6.3-1 and 3.6.5.2-1 for valves in each valve group to obtain ISOLATION SYSTEM RESPONSE TIME for each valve.

FERMI - UNIT 2 3/4 3-19 Amendment No. 41

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TABLE 3.6.3-1 (Continued)

PRIMARY CONTAINMENT ISOLATION VALVES TABLE NOTATIONS (Continued) 8.

Group 8 - Reactor Core Isolation Cooling-(RCIC) System RCIC Steam Line Flow - High RCIC Steam Supply Pressure - Low RCIC Turbine Exhaust Diaphragm Pressure - High RCIC Equipment Room Temperature - High 9.

Group 9 - Reactor Core Isolation Cooling (RCIC) Vacuum Breakers-a Drywell Pressure - High with simultaneous RCIC i

Steam Supply Pressure - Low 10.

Group 10 - Reactor Water Cleanup (RWCU) System (Inboard)

RWCU Differential Flow - High RWCU Area Temperature - High RWCU Area Ventilation' Differential Temperature - High Reactor Vessel Low Water Level - Level 2 11.

Group 11 - Reactor Water Cleanup (RWCU) System (Outboard)

SLCS Initiation (not a containment isolation signal)

RWCU Differential Flow - High RWCU Area Temperature - High RWCU Area Ventilation Differential Temperature - High Reactor Vessel Low Water Level - Level 2 12.

Group 12 - Torus Water Management System (TWMS)

Reactor Vessel Low Water Level - Level 2 Drywell Pressure - High 13.

Group 13 - Drywell Sumps Reactor Vessel Low Water Level - Level 3 I

Drywell Pressure - High 14.

Group 14 - Drywell and Suppression Pool Ventilation System i

Reactor Vessel Low Water Level - Level 2 i

Drywell Pressure - High i

15.

Group 15 - Traversing In-Core (TIP) System Reactor Vessel Low Water Level - Level 3 Drywell Pressure - High FERMI - UNIT 2 3/4 6-45 Amendment No. 21, 41 i

TABLE 3.6.3-1 (Continued)

PRIMARY CONTAINMENT ISOLATION VALVES TABLE, NOTATIONS (Continued)

-15.

Group 15 - Traversing In-Core (TIP) System (Continued)

NOTE:

Either of these signals initiate TIP withdrawal which results in automatic closure of the TIP Ball Valves when the TIP probe has entered the shield cask.

16. Group 16 - Nitrogen Inerting System Reactor Vessel Low Water Level - Level 2 Drywell Pressure - High

-1 17.

Group 17 - Recirculation Pump System and Primary Containment Radiation Monitoring System Reactor Vessel Low Water Level - Level 2 Drywell Pressure - High 18.

Group 18 - Primary Containment Pneumatic Supply System Reactor Vessel Low Water Level - Level 2

.Drywell Pressure - High (b) These valves are hydrostatically leak tested.

(c) Deleted.

(d) Also closes automatically as a result of Torus Room Floor Drain Sump Level - High - High and Drywell Floor Drain Sump Level - High - High.

(e) These valves may be closed remotely from one of the following locations:

1) control room.

2)'

their respective local panels.

(f) Will automatically reposition as a result of the actuation of the LPCI Loop Selection Logic.

(g) Will automatically close when the corresponding RHR loop flow is greater than 1500 gpm.

(h) Will automatically close when the corresponding core spray loop flow is greater than approximately 775 gpm.

(i) Will automatically close when a) HPCI Turbine Steam Stop Valve E41-F067 closes or b) HPCI Turbine Steam Supply Isolation Valve E41-F001 closes.

(j) Will automatically close as a result of the condition listed in Note (i),

above, as well as when HPCI flow is greater than 1200 gpm.

FERMI - UNIT 2 3/4 6-46 Amendment No. 17, 21, 41

_-

ML20247E634

Text

Attachment:

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