ML20029D038

Amend 98 to License NPF-43,revising TS 3.4.3.2.d Which Identifies Allowable Leakage for RCS Pressure Isolation Valves Listed in Table 3.4.3.2-1

ML20029D038
Person / Time
Site:	Fermi
Issue date:	04/22/1994
From:	Marsh L Office of Nuclear Reactor Regulation
To:
Shared Package
ML20029D039	List: ML20029D038 ML20029D042 ML20029D046
References
NUDOCS 9405030319
Download: ML20029D038 (14)

Text

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• r UNITED STATES.

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NUCLEAR REGULATORY-COMMISSION

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WASHINoTON D.C. 20686-0001-5

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DETROIT EDISON COMPANY DOCKET NO. 50-341 FERMI-2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 98 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 May 24, 1993, 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 1:

B.

The f acility will operate in conformity with the application, the l

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

There is reasonable assurance (1) that the activities authorized by' this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commiss' ion'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.

j i

9405030319 940422 PDR ADOCK 05000341 P

PDR-

-l 4

, 2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment amended to read as follows:and paragraph 2.C.(2) of Facility Operating Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 98, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license.

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

3.

This license amendment is effective as of the date of its issua full implementation within 45 days.

FOR THE NUCLEAR REGULATORY COMMISSIO Ledyard B. Marsh, Director Project Directorate 111-1 Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance:

April 22,1994 o

r s

4 9

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ATTACHMENT 'TO LICENSE AMENDMENT' NO. 98 FACILITY OPERATING LICENSE NO. NPF-43

. DOCKET N0. 50-341 LReplace thelfollowing pages ~of the Appendix "A" Technical Specifications with; the attached pages.

contain vertical lines indicating the area of change.The revised page REMOVE

' INSERT 3/4 4-9*

3/4 4-9*

3/4 4-10 3/4 4-10 3/4 4-12 3/4 4-12 3/4 6-28 3/4 6-28 3/4 6-35 3/4 6-35 3/4 6-36*

-3/4 6-36*

3/4 6-47 3/4 6-47 3/4 6-48*

3/4 6-48*

B 3/4 4-2 B 3/4 4-2 B 3/4 4-2a B 3/4 4-2a B 3/4-6-1 B 3/4 6-1 r

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1

• 0verleaf page provided to maintain document completeness.

contained on these pages.

No changes 1

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.4

REACTOR COOLANT SYSTEM 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS (JMITING CONDITION FOR OPERATION 3.4.3.1 The following reactor coolant system leakage detection systems shall be OPERABLE:

The primary containment atmosphere gaseous radioactivity a.

monitoring system channel.

b.

The primary containment sump flow monitoring system consisting of:

1.

The drywell floor drain sump level, flow and pump-run-time system, and 2.

The drywell equipment drain sump level, flow and pump-run-time system.

The drywell floor drain sump level monitoring system.

c.

APPLICABillTY:

OPERATIONAL CONDITIONS 1, 2, and 3.

ACTION:

With only two of the above required leakage detection systems OPERABLE, restore the inoperable detection system to OPERABLE status within 30 days; when the required gaseous radioactive monitoring system is inoperable, operation may continue for up to 30 days provided grab samples of the containment atmosphere are obtained and analyzed at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

SURVElllANCE REOUTREMENTS 4.4.3.1 The reactor coolant system leakage detection systems shall be demonstrated OPERABLE by:

Primary containment atmosphere gaseous monitoring systems-a.

performance of a CHANNEL CHECK at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, a CHANNEL FUNCTIONAL TEST at least once per 31 days and a CHANNEL CAllBRA!!0N at least once per 18 months.

i b.

Primary containment sump flow and drywell floor drain sump level monitoring systems-performance of a CHANNEL FUNCTIONAL TEST at.

least once per 31 days and a CHANNEL CALIBRATION TEST at least once per 18 months.

FERMI - UNIT 2 3/4 4-9 Amendment No. 89

~

REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.3.2 Reactor coolant system leakage shall be limited to:

a.

No PRESSURE BOUNDARY LEAKAGE.

b.

5 gpm UNIDENTIFIED LEAKAGE.

c.

25 gpm total leakage averaged over any 24-hour period.

d.

Leakag'e specified in Table 3.4.3.2-1 at a reactor coolant system pressure of 1045 i 10 psig from any reactor coolant system pressure isolation valve specified in Table 3.4.3.2-1.

e.

2 gpm increase in UNIDENTIFIED LEAKAGE within any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period during OPERATIONAL CONDITION 1.

f.

2 gpm increase in UNIDENTIFIED LEAKAGE within any 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> period during OPERATIONAL CONDITIONS 2 and 3.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, and 3.

ACTION:

a.

With any PRESSURE B0UNDARY LEAKAGE, be in at least H0T SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b.

With any reactor coolant system leakage greater than the limits in b and/or c, above, reduce the leakage rate to'within the' limits withinl4 hours or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

c.

With any reactor coolant system pressure isolation valve leakage greater than the above limit, isolate the high pressure portion of the affected system from the low pressure. portion within'4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least one other closed manual, deactivated automatic, or check

• valve, or be in at least HOT SHUTDOWN with the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

d.

With one or more of the high/ low pressure interface valve leakage pressure monitors'shown in Table 3.4.3.2-2 inoperable, restore the inoperable monitor (s) to OPERABLE status within 7 days or verify the pressure to be less than the alarm setpoint at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />; restore the inoperable monitor (s) to OPERABLE status within 30. days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

• Which has been verified not to exceed the allowable leakage limit 'at the last refueling outage or after the last time the valve was disturbed, whichever is more recent.

FERMI - UNIT 2 3/4 4-10 Amendment No. S7, p,98

TABLE 3.4.3.2-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALV VALVE NUMBER VALVE DESCRIPTION MAXIMUM LEAKAGE faom) 1.

RHR System E11-F015A E11-F015B LPCI Loop A Injection Isolation Valve Ell-F050A LPCI Loop B Injection Isolation Valve 0.4 LPCI Loop A Injection Line Testable 0.4 Check Valve 10 E11-F050B LPCI Loop B Injection Line Testable Check Valve 10 E11-F008 Shutdown Cooling RPV Suction Outboard Isolation Valve 1

Ell-F009 Shutdown Cooling RPV Suction Inboard Isolation Valve 1

Ell-F608 Shutdown Cooling Suction Isolation Valve 1

2. Core Spray System E21-F005A Loop A Inboard Isolation Valve E21-F005B Loop B Inboard Isolation Valve 1

E21-F006A Loop A Containment Check Valve 1

E21-F006B Loop B Containment Check Valve 1

1

3. High Pressure Coolant Injection System E41-F007 Pump Discharge Outboard 1 solation Valve 1

l E41-F006 Pump Discharge Inboard Isolation Valve 1

j'

4. Reactor Core Isolation Cooling System E51 F012 Pump Discharge Isolation Valve E51 F013 Pump Discharge to Feedwater Header 1

l Isolation Valve 1

l (a) External Leakage from this valve shall be limited to 5 ml/ min.

TABLE 3.4.3.2-2 RfACTOR COOLANT SYSTEM INTERFACE VALVES LEAKAGE PRESSURE MONITORS ALARM VALVE NUMBER SETPOINT SYSTEM Ell-F015A & B, E11-F050A & B fosio) 1 i

Ell-F00B, F009, F608 RHR LPCI s 449 E21-F005A & B, E21-F006A & B RHR Shutdown Cooling s 135

'1 E41-F006, F007 Core Spray s 452 E51-F012, F013 HPCI s 71 RCIC s 71 TERMI - UNIT 2 3/4 4-12 Amendment No. Jf, E5, M,98

~4

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TABLE 3.6.3-1 (Continuent)

A PRIMARY CONTAINMENT ISOLATION VALVES E

MAXIMUM

~

VALVE FUNCTION AT; NUMBER ISOLATION TIME B.

Remote-Manual isolation Valves (c)

(Seconds)

~

E 1.

Main Steam Isolation _ Valves (MSIV1 leakaqe Control Valves

-4 NA ro 821-F434 2.

RHR Shutdown Coolino Suttion Inboard Isolation Valve Bypass Valve (4)

NA Ell-F608 3.

LPCI Inboard Isolation _ Valves (f)(5)

NA l

loop A:

Ell-F015A Loop B:

Ell-F0158 4.

RHR Pumps Recirculation Motor coerated Valves (b)(g)

,s gg Pumps A/C: Ell-F007A

~*4 Pumps B/D: Ell-F0078

.m 5.

Warmun and Flush Line Isolation Valve (b)

NA Ell-F026B 6.

Reactor Protection System Instrumentation Isolation Valves MA Division I: Ell-F412 Ell-F413 Division II: Ell-F414

[

Ell ~F415 7.

-RHR Pump Torus Suction Isolation Valves (b)

[

Pump A:

Ell-F004A NA

,o Pump B:

Ell-F0048 y

Pump C:.

Ell-F004C y

Pump D:

Ell-F004D

• m t

+a w

-, - +,

s

TABLE 3.6.3-1 (Continued)

PRIMARY CONTAINMENT ISOLATION VALVES 5

MAXIMUM V_ALVE FUNCTION AND NUMBER ISOLATION TIME ro (Seconds)-

D.

OtherIsElationValves l.

' Main Feedwater Reverse Flow Check Valves 821-F010A NA.

B21-F010B B21-F076A B21-F076B 2.

Deleted

,s.

40 3.

RHR Heat Exchancer Relief Valves (b)

Ell-F001A~

NA jF Ell-F001B u,

4.

RHR Heat Exchancer Outlet line-Relief Valves (b)(p)

Ell-F025A NA Ell-F025B-5.

RHR Pump Suction From Recirc Pipino Reverse Flow Check Valve Ell-F408-NA 6.-

RHR Shutdown Cooline Suction Relief Valve (b)(p)

. Ell-F029 gg.

7.

RHR Pump Torus Suction Relief Valves (D)(P) l[

Ell-F030A-

-NA Ell-F030B

=

l[

Ell-F030CS gp Ell-F0300 W

. =.

g; TABLE 3.6.3-1 (Continued)

PRIMARY CONTAINMENT ISOLATION VALVES e

U3 MAXIMUM

[

VALVE FUNCTION AND NUMBER ISOLATION TIME (Seconds)

D.

Other Isolation Valves (Continued) 3.

Core Spray Loop Containment Reverse Flow Check Valves NA E21-F006A E21-F0068 l

9.

Core Spray Loop Pump Suction Relief Valves (b)(p)

NA

- E21-F032A

}{

E21-F0328 10.

Core Spray Loop Pump Discharge' Pressure Relief Valves (b)

NA E21-F011A E21-F012A

- E21-F011B E21-F0128 11.

Excess Flow Check-Valves (#)

NA a.

Jet Pump Instrumentation l

821-F513A B21-F513B B21-F513C B21-F513D B21-F514A B21-F5148 B21-F514C B21-F514D B21-F515A l

-821-F515B L

B21"F515C y

+

c9

-,-wr r

> + -

y w

r e

r

.=r

-4 ! s :.

TABLE 3.6.3-1 (Continued)

PRIMARY CONTAINMENT ISOLATION VALVES TABLE NOTATIONS (Continued)

(k)

Will automatically close when a) RCIC Turbine' Steam Stop Valve.E51-F045 closes or b).RCIC Turbine Governor Trip and Throttle Valve E51-f059 closes.

(I)

Will automatically close as a result of the conditions listed in Note (k) above, as well as when RCIC flow is greater than 130 gpm.

(m)

These valves are actuated by remote manual key-locked switches and.will cut the TIP cable and seal.off the TIP guide tube when actuated. These valves are squib-fired.

-(n)

May be closed remotely as a secondary actuation. mode to reverse ' flow.

(0)

Valves realign automatically on a reactor scram signal.

(P)

Thermal relief valves.

(4)

Locked closed.

(r)

Not subject to Type C leakage tests.

(s)

Hydrostatically tested in accordance with Specification 4.4.3.2.2 in' lieu l-of the requirements of Specification 4.6.1.2.

l I

4 1

FERMI-- UNIT 2 3/4 6-47 Amendment.No.98 4

CONTAINMENT' SYSTEM 1 3/4.6.4 VACUUM RELIEF SUPPRESSION CHAMBER - DRYWELL VACUUM BREAKERS LIMITING CONDITION FOR OPERATION 3.6.4.1 All suppression chamber - drywell vacuum breakers shall be closed and OPERABLE.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.

ACTION:

With one of the above required vacuum breakers inoperable for a.

opening but known to be closed, restore the inoperable vacuum breaker to OPERABLE 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 least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b.

With one or more suppression chamber - drywell vacuum breakers open, close the open vacuum breaker (s) within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

With one of the position indicators of any suppression chamber -

c.

drywell vacuum breakers inoperable, verify that all other vacuum breakers are closed within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and:

1.

Verify the vacuum breaker (s) with tne inoperable position indicator to be closed by demonstrating the other indicator to be OPERABLE within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> aw at least once per 14 days thereafter, or 2.

Verify the vacuum breaker (s) with the inoperable position indicator to be closed by conducting a test which osmonstrates that the drywell-to suppression chamber AP is maintained at greater than or equal to 0.5 psi for one hour without makeup within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and at least once per 14 days thereafter.

Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUIDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

d.

With one of the closed position indicators of one or more suppression chamber - drywell vacuum breaker (s) indicating open and the redundant closed position indicator indicating closed after a suppression chamber - drywell vacuum breaker opening as a result of a steam release, within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, cycle the applicable valve (s) to determine which of the redundant indicators is OPERABLE.

s FERMI - UNIT 2 3/4 6-48 Amendment No. JJ, JJ, 96

REACTOR COOLANT SYSTEM i

BASES 3/4.4.4 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.3.1 LEAKAGE DETECTION SYSTEMS The RCS leakage detection systems required by this specification are provided to monitor and detect leakage from the reactor coolant pressure boundary.

These detection systems are consistent with the recommendations of-Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems", May 1973.

i 3/4.4.3.2 OPERATIONAL LEAKAGE The allowable leakage rates from the reactor coolant system have been based on the predicted and experimentally observed behavior of cracks in pipes.

The normally expected background leakage due to equipment design and the detection i

capability of the instrumentation for determining system leakage was also considered.

The evidence obtained from experiments suggests that for leakage somewhat greater than that specified for UNIDENTIFIED LEAKAGE the probability is small that the imperfection or crack associated with such leakage would grow rapidly.

However, in all cases, if the leakage rates exceed the values specified.

or the leakage is located and known to be PRESSURE B0UNDARY LEAKAGE, the reactor will be shut down to allow further investigation and corrective action.

Service sensitive reactor coolant system Type 304 and 316 austenitic stainless steel piping; i.e., those that are subject to high stress or that contain relatively stagnant, intermittent, or low flow fluids, requires additional surveillance and leakage limits.

The additional limit placed upon the rate of increase in UNIDENTIFIED LEAKAGE in OPERATIONAL CONDITION 1 meets the NRC Staff guidance i Generic Letter 88-01, "NRC Position on IGSCC in BWR Austenitic Stainless Steel Piping." The applicability of the Generic Letter 88-01 limit to OPERATIONAL CONDITION 1 only ensures that the expected increases in UNIDENTIFIED LEAKAGE experienced during reactor vessel heatup and pressurization'during startup do not cause unwarranted entries into the applicable ACTION statement.

The rate of increase in UNIDENTIFIED LEAKAGE limit in OPERATIONAL CONDITIO that the above service sensitive reactor coolant system Type 304 and 316 austenitic stainless steel piping is monitored during reactor startup prior to reactor vessel heatup and pressurization.

The surveillance interval for determination of UNIDENTIFIED LEAKAGE in OPERATIONAL CONDITION 1 m guidance in Supplement I to Generic Letter 88-01.

The purpose of the RCS interface valves leakage pressure monitors (LPMs) is ide assurance of the integrity of the Reactor Coolant System pressure r

on valves which form a high/ low pressure boundary.

The LPM is desi dt alarm on increasing pressure on the low pressure side of the high/l in rfa e to provide indication to the operator of abnormal interfac ha The Surveillance Requirements for RCS pressure isolation valves provide added i

assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA.

Leakage from the RCS pressure isolation valves is IDENTIFIED LEAKAGE and will be considered as a portion of the allowed limit.

FERMI - UNIT 2 B 3/4 4-2 Amendment No. H, M,98

)

. REACTOR COOLANT SYSTEM BASES 3/4.4.3.2 OPERATIONAL LEAKAGE (Continued)

A reduced. leakage acceptance criteria and an external leakage acceptance criteria are specified for the LPCI Injection Isolation Valve,.E11-F015 A and B, to assure adequate' water is maintained inboard of these valves such that~ the associated primary containment penetration can be classified as a water tested-penetration under Appendix J to 10CFR50.

3/4.4.4 CHEMISTRY The water chemistry limits of the reactor coolant. system are established to prevent damage to the reactor materials in contact with the coolant.

Chloride-limits are specified to prevent stress corrosion cracking of the stainless' steel.

The effect of chloride is not as great when the oxygen concentration in the coolant is low, thus the 0.2 ppm limit on chlorides is permitted during POWER OPERATION.

for stress corrosion to occur is not present so a 0.5 ppm conce chlorides is not considered harmful during these periods.

this parameter are an indication of abnormal conditions. Conductivity When the-conductivity.is within limits, the pH, chlorides and other impurities affecting conductivityLmust also be within their acceptable limits. With the conductivity meter inoperable, additional samples must be analyzed to ensure that the chlorides are not exceeding the limits.

The surveillance requirements provide adequate assurance that concentrations in excess of the limits will be detected in sufficient > time to take corrective action.

i O

FERMI - UNIT 2 B 3/4 4-2a Amendment No. 17, $9,98

s.

3/4.6 CONTAINMENT SYSTEMS BASES 3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.]

PRIMARY CONTAINMENT INTEGRITY PRIMARY CONTAINMENT INTEGRITY ensures that the rele materials from the containment atmosphere will be restricted to those l paths and associated leak rates assumed in the safety analyses.

restriction, in conjunction with the leakage rate limitation, will limit the This SITE B0UNDARY radiation doses to within the limits of 10 CFR Part accident conditions.

PRIMARY CONTAINMENT INTEGRITY is demonstrated by leak ra by OPERABLE containment automatic isolation val during accident conditions are closed by locked valves, blank flanges or deactivated automatic valves secured in the closed position.

cap with acceptable sealant in addition to the co For test, vent and provides protection equivalent to a blank flange.

3/4.6.1.2 PRIMARY CONTAINMENT LEAKAGE The limitations on primary containment leakage rates ensure.that the tota analyses at the peak accident pressure of 56.5 psig, P. co demonstrates maximum expected pressure is less than 56.5 psig. Updated ana a

As an added less than or equal to 0.75 La during performance of the leakage tests. account for possible degradation of the containment leakage barrie Operating experience with the main steam line isolation valves has indicate therefore the special requirement for testing these valves.

The surveillance testing for measuring leakage rates is consistent with the requirements of Appendix J of 10 CFR Part 50 with the exception of exempt granted for main steam isolation valve leak testing, testing the airlocks after each opening, testing the Low Pressure Coolant Injection Inboard Isolation l

Valves, and analyzing the Type A test data.

.l-l Appendix J to 10 CFR Part 50, Paragraph III.A.3, requires that all Type A tests be conducted in accordance with the provisions of N45.4-1972, " Le Rate Testing of Containment Structures for Nuclear Reactors."

analytical techniques. requires that Type A test data be analyzed using point N45.4-1972 4

Specification 4.6.1.2a. requires use of the mass plot analytical technique.

technique, since it yields a confidence interval which is a sm the calculated leak rate; and the interval decreases as more' data sets are to the calculation.

The total time and point-to-point techniques may give confidence intervals, which are large fractions of the calculated leak rate, a the intervals may increase as more data sets are added.

FERMI - UNIT 2 B 3/4 6-1~

Amendment No. E, #, E7,98