Text

Proposed Its,Rev 13
	ML20211B999
Person / Time
Site:	Fermi
Issue date:	08/06/1999
From:	; UNION ELECTRIC CO.
To:	;
Shared Package
ML20211B983	List: ML20211B999; NRC-99-0089, Forwards Rev 13 to Fermi 2 ITS Submittal Dtd 980403.Rev Provide Update for Remaining Open Issues Associated with NRC Review of ITS Submittal.Attachment 1 Contains Brief Abstract of Changes Included in Rev;
References
	NUDOCS 9908250139
	Download: ML20211B999 (200)
	v • d • e

E INSERT THIS PAGE IN FRONT OF VOLUME 2 Volume 2 SECTION 1.0,2.0 & 3.0 -

Remove Replace 1.1 ITS pg 1.1-3 Rev 0 1.1 ITS pg 1.1-3 Rev 13 3.0 DOCS pg 4 Rev 0 3.0 DOCS pg 4 Rev 13 l

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9908250139 990817 PDR ADOCK 05000341 P pg Her13 08/06/99

r Definitions 1.1 1.1 Definitions (continued)

CORE OPERATING LIMITS The COLR is the unit specific document that REPORT (COLR) provides cycle specific parameter limits for the current reload cycle. These cycle specific limits shall be determined for each reload cycle in accordance with Specification 5.6.5. Plant operation within these limits is addressed in individual Specifications.

DOSE EQUIVALENT I-131 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries/ gram) that alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131. 1-132. I 133. I-134 and I-135 actually present. The thyroid dose conversion factors used for this calculation shall

@ be those listed in Table III of TID-14844 AEC.1962. " Calculation of Distance Factors for Power and Test Reactor Sites."

EMERGENCY CORE COOLING The ECCS RESPONSE TIME shall be that time interval SYSTEM (ECCS) RESPONSE from when the monitored )arameter exceeds its ECCS TIME initiation setpoint at t1e 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 means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

ISOLATION SYSTEM The ISOLATION SYSTEM RESPONSE TIME shall be that RESPONSE TIME time interval from when the monitored parameter exceeds its isolation initiation setpoint at the channel sensor until the isolation valves travel to their required positions. Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

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s (continued) l FERMI UNIT 2 1.1 3 Revision 13 08/06/99 I

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DISCUSSION OF CHANGES ITS: SECTION 3.0 LC0 AND SR APPLICABILITY Certain current Specifications contain Actions such as

" Declare (the supported system) inoperable and take the Actions of (its Specification)." In many cases the supported system would likely already be considered j inoperable. The implication of this presentation is that l the Actions of the inoperable supported system would not have been taken without the specific direction to do so. 4 The BWR STS, NUREG 1433 Rev. I was developed, with the Industry input and approval of the NRC, to include LCO 3.0.6, and a new program, Specification 5.5.11, Safety Function Determination l Program to resolve the application of LCOs to support and supported systems. Since the function of ITS LC0 3.0.6 is to clarify existing ambiguities and to maintain actions consistent with change of previous interpretations, this new provision is an administrative change with no impact on safety.

A.7 ITS LC0 3.0.7 provides guidance regarding the meeting of Special Operations LCOs in Section 3.10. These Special Operations LCOs allow specified Technical Specification requirements to be changed (made applicable in part or whole, or suspended) to permit the performance of special tests or operations which otherwise could not be performed. If the Special Operations LCOs did not exist, many of the special tests and operations that. demonstrate plant performance characteristics, special maintenance activities and special evolutions could not be performed. LC0 3.0.7 eliminates the confusion which would otherwise exist as to which LCOs apply during the performance of a special test or operation. ITS LC0 3.0.7 is consistent with the intent of the CTS Special Test Exceptions: however, without ITS LCO 3.0.7 to allow changing the requirements of another LCO, a conflict of requirements could be interpreted to exist. This change provides clarity only and, therefore is an administrative change, with no impact on safety.

A.8 ITS SR 3.0.1 presents the relationship between Surveillance Requirements and meeting the requirements of the LCO. The concepts within CTS 4.0.3 are combined with CTS 4.0.1 into ITS SR 3.0.1.

The second sentence of ITS SR 3.0.1, " Failure to meet a Surveillance, whether such failure is experienced during the !

performance of the Surveillance or between performances of the Surveillance, shall be failure to meet the LCO," is a requirement that is consistent with the intent but not explicitly stated in the CTS.

FERMI UNIT 2 4 REVISION 13, 08/06/99l

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1 INSERT TIIIS PAGE IN FRONT OF VOLUME 3 1

l Volume 3 SECTION 3.1 & 3.2 Remove Replace B 3.1.2 ITS pg B 3.1.2-2 Rev 0 B 3.1.2 ITS pg B 3.1.2 2 Rev 13 B 3.1.2 NUREG M/U pg B 3.1-9 B 3.1.2 NUREG M/U pg B 3.1-9 Rev 13 D 3.1.7 ITS pg B 3.1.7-6 Rev 0 B 3.1.7 ITS pg B 3.1.7-6 Rev 13 B 3.1.7 ITS pg B 3.1.7-7 Rev 13 B 3.1.7 NUREG M/U pg B 3.1-45 B 3.1.7 NUREG M/U pg B 3.1-45 Rev 13 l

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5 Rev 13 08/06/99

Reactivity Anomalies B 3.1.2 BASES BACKGROUND (continued)

The predicted core reactivity is calculated by a 3D core Ofl simulator code as a function of cycle exposure. This calculation is performed for projected operating states and conditions throughout the cycle. The core reactivity is determined from actual plant conditions and is then compared to the predicted value for the cycle exposure.

APPLICABLE Accurate prediction of core reactivity is either an explicit SAFETY ANALYSES or implicit assumption in the accident analysis evaluations (Ref. 2). In particular, SDM and reactivity transients, such as control rod withdrawal accidents or rod drop accidents, are very sensitive to accurate prediction of core reactivity. These accident analysis evaluations rely on compute.r codes that have been qualified against available test oata, operating plant data, and analytical benchmarks.

Monitoring reactivity anomaly provides additional assurance that the nuclear methods provide an accurate representation of the core reactivity.

The comparison between measured and predicted initial core reactivity provides a normalization for the calculational models used to predict core reactivity. If the measured and predicted reactivity for identical core conditions at B0C do not reasonably agree, then the assumptions used in the reload cycle design analysis or the calculation models used to predict reactivity may not be accurate. If reasonable agreement between measured and predicted core reactivity exists at BOC, then the prediction may be normalized to the measured value. Thereafter, any significant deviations in the measured reactivity from the predicted reactivity that develop during fuel depletion may be an indication that the assumptions of the DBA and transient analyses are no longer valid, or that an unexpected change in core conditions has occurred. 1 Reactivity anomalies satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii). j LC0 The reactivity anomaly limit is established to ensure plant !

operation is maintained within the assumptions of the safety j analyses. Large differences between monitored and predicted I

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! FERMI - UNIT 2 B 3.1.2 - 2 Revision 13. 08/06/99 I

1 Reactivity Anomalies B 3.1.2 i BASES I BACKGROUND poisons (mainly xenon and samarium) are present in the fuel. l (continued) The predicted core reactivit i s calculated by Da [ym er::==: codeeas ea' eI

=="4 core simulator function of cycle exposure. This calculation is performed a

f,\ for pro.jected operating states and conditions throughout the cycle. The core reactivity is determined from =nt=! ra

nstti rfirr t actual plant conditions and is then compared l to the predicted value for the cycle exposure.

APPLICABLE Accurate prediction of core reactivity is either an explicit SAFETY ANALYSES or implicit assumption in the accident analysis evaluations (Ref. 2). In particular, SDN and reactivity transients, i such as control rod withdrawal accidents or rod drop i accidents, are very sensitive to accurate prediction of core !

reactivity. These accident analysis evaluations rely on computer codes that have been qualified against available l test data, operating plant data, and analytical benchmarks. I Monitoring reactivity anomaly provides additional assurance that the nuclear methods provide an accurate representation of the core reactivity.

The comparison between measured and predicted initial core reactivity provides a normalization for the calculational I models used to predict core reactivity. If the measured and

_ predicted ed t .:ity for identical core conditions at BOC ^

co not reasonably agree, then the assumptions used in the reload cycle design analysis or the calculation models used to predict; ed 6 2ity may not be accurate. If reasonable greement between measured and predicted core reactivity exists at BDC, then the prediction may be normalized to the measured value. Thereafter. any significant deviations in

.I gg}yf J the sensureds;ii;;.ity from the predicteht f=:ity that develop during fuel depletion may be an indication that the assumptions of the DBA and transient analyses are no longer valid, or that an unexpected change in core conditions has occurred.

Pn 3t tt ==t .

Reactivity anomalies satisfy Criterion 2 of4hi EC Fuii tou wo.succ .

(continued)

< B aj4 sis 8 3.1-9 "iv ;, 04/67/W-Rei(3 l

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SLC System B 3.1.7

). BASES SURVEILLANCE REQUIREMENTS (continued) tested every 36 months at alternating 18 month intervals.

The Surveillance may be performed in separate ste)s to prevent injecting boron into the RPV. An accepta)le method for verifying flow from the pump to the RPV is to pump demineralized water from a test tank through one SLC subsystem and into the RPV. The 18 month Frequency is based l on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating ,

experience has shown these components usually pass the l Surveillance when performed at the 18 month Frequency: ;

therefore, the Frecuency was concluded to be acceptable from !

a reliability stancpoint.

Demonstrating that all piping between the boron solution storage tank and the explosive valve is unblocked ensures l that there is a functioning flow path for injecting the !

sodium pentaborate solution. An acceptable method for !

verifying that the suction piping is unblocked is to pump l from the storage tank to the test tank (this is followed by I draining and flushing the piping with demineralized water).

The 18 month Frequency is acceptable since there is a low l probability that the subject piping will be blocked due to !

precipitation of the boron from solution in the piping.

This is especially true in light of the temperature l verification of this piping required by SR 3.1.7.3. i However. if, in performing SR 3.1.7.3 it is determined that j the temperature of this piping has fallen below the i specified minimum. SR 3.1.7.9 must be performed once within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the piping temperature is restored to a 48'F.

SR 3.1.7.10 Enriched sodium pentaborate solution is made by mixing granular. enriched sodium pentaborate with water. Isotopic i tests on the granular sodium pentaborate to verify the i actual B-10 enrichment must be performed )rior to addition !

to the SLC tank in order to ensure that tie proper B-10 atom percentage is being used.

1 l FERMI - UNIT 2 B 3.1.7 - 6 Revision 13 08/06/99 1

f, SLC System l B 3.1.7 i

t ,

BASES REFERENCES 1. 10 CFR 50.62.

2. UFSAR. Section 4.5.2.4.

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r hl FERMI-UNIT 2 B 3.1.7 - 7 Revision 13. 08/06/99

SLC System B 3.1.7 BASES

\

SURVEILLANCE SR 3.1.7.7 (continued)

REQUIRENENTS requirements, the rate of negative reactivity insertion from the SLC System will adequately compensate for the positive I

reactivity effects encountered during power reduction, cooldown of the moderator, and xenon decay. This test confims one point on the pump design curve and is indicative of overall performance. Such inservice inspections confim component OPERASILITY, trend perfomance, and detect incipient failures by indicating abnormal performance. The Frequency of this Surveillance is 41n accordance with the Inservice Testing Program-or-

....,.c i SR 3.1.7.8 and SR 3.1.7.9 These Surveillances ensure that there is a functioning flow path from the boron solution storage tank to the RPV, including the firing of an explosive valve. The replacement charge for the explosive valve shall be from the same manufactured batch as the one fired or from another batch that has been certified by having one of that batch successfully fired. The pump and explosive valve tested should be alternated such that both complete flow paths are tested every 36 months at alternating 18 month intervals.

! es The Surveillance may be performed in separate steps to prevent injecting boron into the RPV. An acceptable method for verifying flow from the pump to the RPV is to pump domineralized water from a test tank through one SLC subsystem and into the RPV. The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown these components usually pass the Surveillance when performed at the 18 month Frequency; therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

Demonstrating that allMt trr:d piping between the boron solution storage tank and the = tic. f o t b 1 in W ivu yos;,e pr;n is un'.> locked ensures that there is a functioning flow w path for injecting the sodium pentaborate solution. An acceptable method for verifying that the suction piping is unhlncked h to Dump from the stcrage tank to the test ta (415kMad @nz)

/kg /If7T WI1Udarnir1 and f(ush(e9),

dli3Cd (Ndkt (continued) m __ m _ -

l :". 7./4 ';TS B 3.1-45 l W n'/07/95 l

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Rev I.3 o

i INSERT TIIIS PAGE IN FRONT OF VOLUME 4 I t

l Volume 4 SECTION 3.3.1.1-3.3.4.1 Remove Replace 1 B 3.3.1.1 ITS pg B 3.3.1.1-24 Rev 0 B 3.3.1.1 ITS pg B 3.3.1.1-24 Rev 13 B 3.3.1.1 NUREG M/U pg B 3.3.1.1-24 (Insert) Rev 13 3.3.3.2 ITS pg 3.3-30 Rev 0 3.3.3.2 ITS pg 3.3-30 Rev 13 3.3.3.2 NUREG M/U pg 3.3-29 (Insert) 3.3.3.2 NUREG M/U pg 3.3-29 (Insert) Rev 13 l

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l RPS Instrumentation B 3.3.1.1 BASES ACTIONS (continued)

[L1 Required Action D.1 directs entry into the appropriate Condition referenced in Table 3.3.1.11. The applicable Condition specified in the Table is function and MODE or other specified condition dependent and may change as the Required Action of a previous Condition is completed. Each time an inoperable channel has not met any Required Action of Condition A. B or C and the associated Com)letion Time has expired, Condition D will be entered for t1at channel !

and provides for transfer to the appropriate subsequent Condition.

E.1. F.1. G.I. H.1. and H.2 If the channel (s) is not restored to OPERABLE status or placed in trip (or the associated trip system placed in !

trip) within the allowed Completion Time, the plant must be !

alaced in a MODE or other specified condition in which the l _C0 does not apply. Alternately, for Condition H, the MSLs i may be isolated (Required Action H.1), and. if allowed I (i.e., plant safety analysis and minimal steam flow in MODE 2 allows operation with the MSLs isolated), operation ,

with the MSLs isolated may continue. Isolating the MSLs i conservatively accomplishes the safety function of the inoperable channel. The allowed Completion Times are reasonable, based on operating experience, to reach the specified condition from full power conditions in an orderly <

manner and without challenging plant systems. In addition. l the Completion Time of Required Action E.1 is consistent I with the Completion Time provided in LC0 3.2.2. " MINIMUM I CRITICAL POWER RATIO (MCPR)." )

l l FERMI UNIT 2 B 3.3.1.1 - 24 Revision 13. 08/06/99

i RPS Instrumentation-B 3.3.1.1

..s Insert B 3.3.1.1 3

! Alternately, for Condition H, the MSLs may be isolated j (Required Action H.1), and, if allowed (i.e., plant safety l l analysis and minimal steam flow in MODE 2 allows operation with the MSLs isolated), operation with the MSLs isolated may

]

continue. Isolating the MSLs conservatively accomplishes the safety function of the inoperable channel.

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l FERMI - UNIT 2 Page B 3.3 24 (Insert) REVISION 13. 08/06/99l I

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Remote Shutdown System 3.3.3.2

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Table 3.3.3.21 (page 1 of 1)

Remote Shutdown System Instrumentation

)

l INSTRUENT FUNCTION

1. Reactor Vessel Pressure

2. Reactor Vessel Water Level

3. Suppression Chamber Water Temperature

4. Drywell Pressure

5. RHR Heat Exchanger Discharge Flow

6. RCIC Flow l CONTROL FUNCTION

1. Control Rod Drive Pump A

2. Control Rod Drive Pump B

3. RHR Valve E1150 F009

4. RHR Valve E1150 F008

5. RHR Valve E1150 F006A

6. Recirc Pump A Valve B3105 F023A

7. Main Steam Line (D) Relief Valve B2104 F013A

8. Main Steam Line (C) Relief Valve B2104 F013B

9. RHR Valve E1150 F015A

10. RHR Valve E1150 F017A

11. RHR Valve E1150 F004A l

12. RHR Pump A l

13. RHR Valve E1150 F024A i

14. RHR Valve E1150 F028A

15. RHR Valve E1150 F048A

16. RHR Valve E1150 F068A

17. RHR Service Water Pump A l

18. RHR Service Water Pump C 4

19. Cooling Tower Fan A

20. Cooling Tower Fan C

21. RCIC Valve E5150 F059

22. RCIC Valve E5150 F045

23. RCIC Initiate

24. Division II DC Transfer l 25. B0P Transfer 1

26. Division I DC Transfer

27. Division I AC Transfer

28. Swing Bus Transfer l FERMI UNIT 2 3.3 30 Revision 13. 08/06/99

Remote Shutdown Instrumentation 3.3.3.2 Insert 3.3.3.2-1 INSTRUMENT FUNCTION lh

1. Reactor Vessel Pressure

2. Reactor Vessel Water Level

3. Suppression Chamber Water Temperature

4. Drywell Pressure

5. RHR Heat Exchanger Discharge Flow

6. RCIC Flow CONTROL FUNCTION lh

1. Control Rod Drive Pump A

2. Control Rod Drive Pump B

3. RHR Valve E1150 F009

4. RHR Valve E1150 F008

5. RHR Valve E1150 F006A

6. Recirc Pump A Valve B3105 F023A

7. Main Steam Line (D) Relief Valve B2104 F013A

8. Main Steam Line (C) Relief Valve B2104 F013B

9. RHR Valve E1150 F015A

10. RHR Valve E1150 F017A

11. RHR Valve E1150 F004A

12. RHR Pump A

13. RHR Valve E1150 F024A

14. RHR Valve E1150 F028A

15. RHR Valve E1150 F048A

16. RHR Valve E1150 F068A

17. RHR Service Water Pump A

18. RHR Service Water Pump C

19. Cooling Tower Fan A

20. Cooling Tower Fan C

21. RCIC Valve E5150 F059

22. RCIC Valve E5150 F045

23. RCIC Initiate

24. Division II DC Transfer

25. BOP Transfer

26. Division I DC Transfer

27. Division I AC Transfer

28. Swing Bus Transfer FERMI UNIT 2 Page 3.3 29 (Insert) REVISION 13. 08/06/99l

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! INSERT THIS PAGE IN FRONT OF VOLUME 5

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l Volume 5 SECTION 3.3.5.13.3.8.2 Remove Replace l 3.3.6.3 ITS pg 3.3-65 Rev 6 3.3.6.3 ITS pg 3.3-65 Rev 13 3.3.6.3 ITS pg 3.3-67 Rev 6 3.3.6.3 ITS pg 3.3-67 Rev 13 l

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LLS Instrumentation l 3.3.6.3 1

3.3 INSTRUMENTATION g

3.3.6.3 Low Low Set (LLS) Instrumentation LC0 3.3.6.3 The LLS valve instrumentation for each Function in Table 3.3.6.31 shall be OPERABLE.

APPLICABILITY: MODES 1. 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One LLS valve A.1 Restore channel (s) to 14 days inoperable due to OPERABLE status, inoperable channel (s).

B. - -- - NOTE - -

B.1 Restore one tailpipe 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

, Separate Condition pressure switch for

,- entry is allowed for 11 OPERABLE SRVs.

each SRV. including one in each

@', / ......................

Division for an OPERABLE SRV in the One or more safety / lowest setpoint e relief valves (SRVs) group to OPERABLE 4

with one or more status.

d Function 3 channel (s)~

inoperable. ANQ

............N0TE- --- .. -

LC0 3.0.4 is not applicable.

B.2 Restore both tailpipe Prior to pressure switches for entering MODE 2 ,

G 11 OPERABLE SRVs.

including 4 of 5 or 3 from MODE 4 l OPERABLE SRVs with i y the lowest relief '

setaoints, to OPEMBLE status.

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(continued) !

l FERMI - UNIT 2 3.3 65 Revision 13, 08/06/99 l

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l LLS Instrumentation l

3.3.6.3 l i 3.3 INSTRUMENTATION

. 3.3.6.3 Low Low Set (LLS) Instrumentation LC0 3.3.6.3 The LLS valve instrumentation for each Function in Table 3.3.6.3 1 shall be OPE ggQ a+ gg+ m -

A.i +10(gipc M a x ssi b , 14 hour1.62037e-4 days 0.00389 hours 2.314815e-5 weeks 5.327e-6 months $

' (i 6e n weass $49,,-

4 APPLICABILITY: H0 DES 1, 2. and 3. I

'in d a dfo $sw P/*swee a.ch

% eatpree 6 ca& A 03vi sim e oesm6tr to twe s1 L ACTIONS _ sRv)in[o MifYrn&5 e k

k CONDITION fREQUIREDACTION M TIME I4 day 5 A. One LLS valve inoperable due to A.1 Restore channel (s) to OPERABLE status.

44:MEC- h# h inoperable channel (s).

(%c N.I) ,

B. One or more safety / B.1 F - --- NOTE- - --

& // f FEQ6tE SW5e relief valves ( s) LCO 3.0.4 is not includinq 'l *f f.

with on Function M _

applicable. gp g gg g, wg I

channel noperable.

& -- i ord

~

Restore p1 >e pressure switcles o Prior to entering MODE 2 M'W"*5 -

OPERABLE status, or 3 from H0DE 4 fDoc y.I)

% S NOTE-eparate Condition C Res re one ta' pipe pr ssure swit to

[14 days entry is allowed for j 0 ERABLE sta us, ieachSpV.,2.......

g..... _ ,

p 0 or mor( S/RVs th o Func on 3 hannel inopera e.

l (continued) l BWR/4 STS 3.3 67 P,ev 1. 04/07/95 6/

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INSERT THIS PAGE IN FRONT OF VOLUME 6 Volume 6 SECTION 3.4 &3.5 Remove Replace l

3.4.1 ITS pg 3.4-1 Rev 10 3.4.1 ITS pg 3.4-1 Rev 13 1

3.4.1 ITS pg 3.4-2 Rev 2 3.4.1 ITS pg 3.4-2 Rev 13 3.4.1 ITS pg 3.4-3 Rev 2 3.4.1 ITS pg 3.4-3 Rev 13 3.4.1 ITS pg 3.4-4 Rev 13 B 3.4.1 ITS pg B 3.4.13 Rev 2 B 3.4.1 ITS pg B 3.4.1-3 Rev 13 B 3.4.1 ITS pg B 3.4.1-4 Rev 10 B 3.4.1 ITS pg B 3.4.1-4 Rev 13 B 3.4.1 ITS pg B 3.4.1-5 Rev 0 B 3.4.1 ITS pg B 3.4.1-5 Rev 13 l

B 3.4.1 ITS pg B 3.4.1-6 Rev 2 B 3.4.1 ITS pg B 3.4.1-6 Rev 13 B 3.4.1 ITS pg B 3.4.1-7 Rev 2 B 3.4.1 ITS pg B 3.4.17 Rev 13 B 3.4.1 ITS pg B 3.4.1-8 Rev 2 B 3.4.1 ITS pg B 3.4.1-8 Rev i3 B 3.4.1 ITS pg B 3.4.19 Rev 2 B 3.4.1 ITS pg B 3.4.1-9 R.ev 13 B 3.4.1 ITS pg B 3.4.1 10 Rev 2 B 3.4.1 ITS pg B 3.4.1 10 Rev 13 3.4.1 CTS M/U (3/4 4-1) pg 2 of 6 Rev 10 3.4.1 CTS M/U (3/4 4-1) pg 2 of 6 Rev 13 3.4.1 DOCS pg 6 Rev 0 3.4.1 DOCS pg 6 Rev 13 3.4.1 NSHC pg 7 Rev 13

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3.4.1 NSHC pg 8 Rev 13 3.4.1 NUREG M/U pg 3.4-1 Rev 10 3.4.1 NUREG M/U pg 3.4-1 Rev 13 3.4.1 NUREG M/U pg 3.4-1 (1)(Insert) Rev 10 3.4.1 NUREG M/U pg 3.4-1 (1)(Insert) Rev 13 B 3.4.1 NUREG M/U pg B 3.4-3 Rev 10 B 3.4.1 NUREG M/U pg B 3.4-3 Rev 13 B 3.4.1 NUREG M/U pg B 3.4-3 (Insert) Rev 10 B 3.4.1 NUREG M/U pg B 3.4-3 (Insert) Rev 13 l-1 l

Rev13 08/06/99

Recirculation Loops Operating 3.4.1 3.4 REACTUR COOLANT SYSTEM (RCS) 3.4.1 Recirculation Loops Operating LC0 3.4.1 The reactor core shall not exhibit core thermal hydraulic instability or operate in the " Scram" or " Exit" Regions, d M

~7-

' a.

Tworecirculationloobeinoperation:s jet pump flows shall with matched recirculation loop v 2

b. One recirculation loop may be in operation provided the following limits are applied when the associated LC0 is g applicable:

1. LC0 3.2.1. " AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)." single loop operation limits

@ 2.

specified in the COLR:

LCO 3.2.2. " MINIMUM CRITICAL POWER RATIO (MCPR)."

single loop operation limits specified in the COLR:

pl 3. LC0 3.3.1.1 " Reactor Protection System (RPS)

Instrumentation." Function 2.b (Average Power Range A; Monitors Simulated Thermal Power-Upscale) Allowable o

~~~ Value of Table 3.3.1.11 is reset for single loop operation, when in MODE 1: and

4. THERMAL POWER is s 67.2% RTP.

........................... NOTE -- - - - -- ---

Q Application of the required limitations for single loop operation may be delayed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after transition from two recirculation loop operations to single recirculation loop operation.

APPLICABILITY: MODES 1 and 2.

l FERMI UNIT 2 3.4 1 Revision 13 08/06/99

e Recirculation Loops Operating 3.4.1 ACTIONS i CONDITION REQUIRED ACTION COMPLETION TIME A, Recirculation jet pump A.1 Declare recirculation 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months loop flow mismatch not loop with lower flow:

within limits. *not in operation."

l

8. Reactor core operating - - -

---NOTE ---------- -

g in the " Exit" Region. Restart of an idle recirculation loop or ,

resetting a recirculation flow limiter is not allowed.

Q .............................

B.1 Initiate action to Immediately insert control rods yi or increase core flow to restore operation

\ outside the " Exit" il Region. :

C. No recirculation loops C.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months operating while in MODE 2.

(continued) l FERMI - UNIT 2 3.4-2 Revision 13. 08/06/99 V

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Recirculation Loops Operating 3.4.1 s

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. No recirculation loops D.1 Place the reactor Immediately operating while in mode switch in the MODE 1. shutdown position.

Reactor core operating Gl in the " Scram" Region.

d m

Core thermal hydraulic

() instability evidenced.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.1.1 -- --- -

-- NOTE - --- - -- --

Only required to be performed when operating in the " Stability Awareness" 3lN Region.

Verify the reactor core is not exhibiting I hour core thermal hydraulic instability.

(continued) l

. g

Recirculation Loops Operating 3.4.1 SURVEILLANCE FREQUENCY SR 3.4.1.2 - --- -- - -

-- NOTE- - ---------- ---

Not required to be erformed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after both recircul tion loops are in operation.

Verify recirculation loop jet pump flow 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months mismatch with both recirculation loops in operation is:

a. s 10% of rated core flow when operating at < 70% of rated core flow; and

b. s 5% of rated core flow when operating at a 70% of rated core flow.

@( l FERMI UNIT 2 3.4 4 Revision 13. 08/06/99 l

I Recirculation Loops Operating B 3.4.1

BASES APPLICABLE The operation of the Reactor Coolant Recirculation System is SAFETY ANALYSES an initial condition assumed in the design basis loss of coolant accident (LOCA) (Ref.1). During a LOCA caused by a recirculation loop pipe break, the intact loop is assumed to provide coolant flow during the first few seconds of the accident. The initial core flow decrease is rapid because the recirculation pump in the broken loop ceases to aump reactor coolant to the vessel almost immediately. T1e pump in the intact loop coasts down relatively slowly. This pump coastdown governs the core flow response for the next several seconds until the jet pump suction is uncovered.

The analyses assume that both loops are operating at the same flow prior to the accident. However, the LOCA analysis was reviewed for the case with a flow mismatch between the two loops, with the pi with the higher flow. peWhile breakthe assumed to be inand flow coastdown thecore loop response are potentially more severe in this assumed case (since the intact loop starts at a lower flow rate and the core response is the same as if both loops were operating at a lower flow rate), a small mismatch has been determined to be acceptable based on engineering judgement. The recirculation system is also assumed to have sufficient flow coastdown characteristics to maintain fuel thermal margins during abnormal operational transients (Ref. 2), which are analyzed in Chapter 15 of the UFSAR.

A plant specific LOCA analysis has been performed assuming only one operating recirculation loop. This analysis has demonstrated that, in the event of a LOCA caused by a pipe break in the operating recirculation loop, the Emergency Core Cooling System response will provide adequate core cooling provided the APLHGR requirements are modified accordingly (Ref. 3).

The transient analyses of Chapter 15 of the UFSAR have also been performed for single recirculation loop operation (Ref. 3) and demonstrate sufficient flow coastdown characteristics to maintain fuel thermal margins during the g abnormal operational transients analyzed provided the MCPR requirements are modified. During single recirculation loop operation, modification to the Reactor Protection System (r.PS) average power range monitor (APRM) instrument setpoints is also required to account for the different relationships between recirculation drive flow and reactor l

l FERMI - UNIT 2 B 3.4.1 - 3 Revision 13. 08/06/99

Recirculation Loops Operating B 3.4.1 BASES APPLICABLE SAFETY ANALYSIS (continued) l core flow. The APLHGR and MCPR setpoints for single loop o>eration are specified in the COLR. The APRM Simulated T1ermal Power Upscale setpoint is in LC0 3.3.1.1. " Reactor Protection System (RPS) Instrumentation."

Thermal-hydraulic stability analysis (Ref. 5) has concluded that procedures for detecting and suppressing power l oscillations that might be induced by a thermal hydraulic instability are necessary to provide reasonable assurance l that the requirements of Reference 4 are satisfied.

i Recirculation loops operating satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

I LC0 Two recirculation loops are required to be in operation with their flows matched within the limits specified in SR 3.4.1.2 to ensure that during a LOCA caused by a break of the piping of one recirculation loop the assumptions of the LOCA analysis are satisfied. With the limits specified in SR 3.4.1.2 not met, the recirculation loop with the lower flow must be considered not in operation. With only one i recirculation loop in operation, modifications to the !

required APLHGR limits (LC0 3.2.1, " AVERAGE PLANAR LINEAR l HEAT GENERATION RATE (APLHGR)"), MCPR limits (LC0 3.2.2, '

s " MINIMUM CRITICAL POWER RATIO (MCPR)"). APRM Simulated R. Thermal Power-Upscale setpoint (LC0 3.3.1.1) and limitation 2 on THERMAL POWER may be applied to allow continued operation 6 consistent with the assumptions of the safety analysis.

p Operations that exhibit core thermal-hydraulic instability m are not permitted. Additionally, in order to avoid N potential power oscillations due to thermal hydraulic 3 instability, operation at certain combinations of power and W flow are not permitted. These restricted power and flow

%c regions are referred to as the " Scram" and " Exit" regions C and are defined by Bases Figure B 3.4.1-1.

A Note is provided to allow 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months following the transition to single loop operation from two loop operation to establish the applicable limitations in accordance with the single loop analysis. The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months period is sufficient to make the adjustment given the relatively small change required. This transition only results in applying the new single loop allowable values to APRM OPERABILITY. Any ARPM l FERMI UNIT 2 B 3.4.1 - 4 Revision 13, 08/06/99

Recirculation Loops Operating B 3.4.1 BASES LCO (continued) non compliance with the required allowable value after this 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months allowance, results in ACTIONS of LCO 3.3.1.1 being entered; no ACTION of LC0 3.4.1 would apply. Similarly, any operation with APLHGR or MCPR out of limits results in the ACTIONS of LCO 3.2.1 or LC0 3.2.2 being entered; no ACTION of LC0 3.4.1 would apply.

APPLIC GILITY In MODES 1 and 2, requirements for operation of the Reactor Coolant Recirculation System are necessary since there is considerable energy in the reactor core and the limiting design basis transients and accidents are assumed to occur.

In H0 DES 3, 4, and 5, the consequences of an accident are reduced and the coastdown characteristics of the recirculation loops are not important. In addition, sufficient power to create power oscillations that threaten fuel design limits does not exist.

ACTIONS A_d With the requirements for matched recirculation loop flow not met, the recirculation loops must be restored to o>eration with matched flows within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or the loo) with t1e lower flow must be considered not in operation. 51ould a LOCA occur with one recirculation loop at a significantly lower flow than the other loop, the core flow coastdown and resultant core response may not be bounded by the LOCA analyses. Therefore, only a limited time is allowed to restore the loop to operating status.

Alternatively, if the single loop requirements of the LC0 are applied to RPS setpoints, operation with only one recirculation loop would satisfy the requirements of the LC0 and the initial conditions of the accident sequence.

l The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time is based on the low probability of an accident occurring during this time period, on a reasonable time to complete the Required Action. and on frequent core monitoring by operators allowing abrupt changes in core flow conditions to be quickly detected.

l FERMI UNIT 2 B 3.4.1 - 5 Revision 13. 08/06/99 l

Recirculation Loops Operating B 3.4.1 BASES ACTIONS (contfnued)

This Required Action does not require tripping the recirculation pump in the lowest flow loop when the mismatch between total jet pump flows of the two loops is greater than the required limits. However, in cases where large flow mismatches occur, low flow or reverse flow can occur in the low flow loop jet pumps, causing vibration of the jet pumps. If zero or reverse flow is detected, the condition should be alleviated by changing pump speeds to re-establish forward flow or by tripping the pump.

IL1 When operating in the " Exit" region (refer to Figure B 3.4.11), the potential for thermal hydraulic n instabilities is increased and sufficient margin may not be 4 available for operator response to suppress potential power oscillations. Therefore, action must be initiated immediately to restore operation outside of the " Exit" region. Control rod insertion and/or core flow increases

( are designated as the means to accomplish this objective.

Required Action B.1 is modified by a Note that precludes core flow increases by restart of an idle recirculation loop, or by resetting a recirculation flow limiter. Core flow increases by these means would not support timely completion of the action to restore operation outside the

" Exit" Region.

C.l With no recirculation loops in operation in MODE 2, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . In this condition, the recirculation loops are not required to be operating because of the reduced severity of DBAs and minimal dependence on the l recirculation loop coastdown characteristics. The allowed Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is reasonable, based on operating experience, to reach MODE 3 from MODE 2 conditions in an l orderly manner and without challenging plant systems.

l FERMI - UNIT 2 B 3.4.1 - 6 Revision 13. 08/06/99 L.

Recirculation Loops Operating B 3.4.1

) BASES ACTIONS (continued)

D_d If operating with no recirculation pumps in operation in MODE 1 or operating in the " Scram" region (refer to Bases N Figure B 3.4.1-1), or if core thermal-hydraulic instability is detected, then unacceptable power oscillations may result. Therefore, the reactor mode switch must be immediately placed in the shutdown position to terminate the 4l potential for unacceptable power oscillations.

i Thermal hydraulic instability is evidenced by a sustained increase in APRM or LPRM peak to peak noise level reaching 2 A or more times its initial level and occurring with a N characteristic period of less than 3 seconds.

h If entry into this condition is an unavoidable and well b% known consequence of an event, early initiation of the l

Required Action is appropriate. Also it is recognized that a

during certain abnormal conditions, it may become Q operationally necessary to enter the " Scram" or " Exit" region for the purpose of: 1) protecting plant equipment.

which if it were to fail could impact plant safety, or

2) protecting a safety or fuel operating limit. In these cases, the appropriate actions for the region entered would be performed as required.

These requirements are consistent with References 5 and 6.

l SURVEILLANCE SR 3.4.1.1 l REQUIREMENTS l This SR 3rovides frequent periodic monitoring for core thermal lydraulic instability by monitoring APRM and LPRM signals for a sustained increase in APRM or LPRM peak to

-m peak noise level reaching 2 or more times its initial level N and occurring with a characteristic period of less than 3 seconds. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Frequency is based on the small potential for core thermal hydraulic oscillations to occur outside the " Scram" or " Exit" regions. Therefore, frequent

( ' monitoring of the APRM and LPRM signals is appropriate when operating in the " Stability Awareness" region.

g l FERMI - UNIT 2 B 3.4.1 - 7 Revision 13, 08/06/99

Recirculation Loops Operating B 3.4.1 BASES SURVEILLANCE REQUIREMENTS (continued)

This SR is modified by a Note that states performance is only required when operating in the " Stability Awareness" region (refer to Bases Figure B 3.4.1-1) (i.e.. in the power-to flow region that is near regions of higher probability for core thermal hydraulic instabilities). This a is acceptable because outside the " Stability Awareness" A region, power and flow conditions are such that sufficient margin exists to the potential for core thermal-hydraulic instability to allow routine core monitoring. Any unanticipated entry into the " Stability Awareness" region would require immediate verification of core stability since V the Surveillance would not be current.

SR 3.4.1.2 This SR ensures the recirculation loops are within the allowable limits for mismatch. At low core flow (i.e. .

< 70% of rated core flow) the MCPR requirements provide larger margins to the fuel cladding integrity Safety Limit ,

such that the potential adverse effect of early boiling l transition during a LOCA is reduced. A larger flow mismatch !

can therefore be allowed when core flow is < 70% of rated 1 core flow. The recirculation loop jet pump flow. as used in I this Surveillance, is the summation of the flows from all of I the jet pumps associated with a single recirculation loop.

l The mismatch is measured in terms of percent of rated core i flow. If the flow mismatch exceeds the specified limits.

l the loop with the lower flow is considered "not in l operation". The SR is not required when both loops are not

in operation since the mismatch limits are meaningless l l during single loop or natural circulation operation. The )

l Surveillance must be performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after both l l loo)s are in operation. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency is consistent !

l wit 1 the Surveillance Frequency for jet pump OPERABILITY l verification and has been shown by operating experience to I be adequate to detect off normal jet pump loop flows in a timely manner.

1 h l FERMI - UNIT 2 B 3.4.1 - 8 Revision 13. 08/06/99

E l

l Recirculation Loops Operating B 3.4.1 l

BASES REFERENCES 1. UFSAR, Section 6.3.3.

2. NEDE 23785 P A, " SAFER /GESTR Models for the Evaluation of the Loss of Coolant Accident," Revision 1, October 1984.

3. ME 56-0386. " Fermi 2 Single Loop Operation Analysis,"

Rev. 1, April 1987, and NEDC 32313-P, "Enrico Fermi Energy Center Unit 2 Single Loop Operatlon." September 1994.

4. 10 CFR 50, Appendix A. GDC 12.

5. MtC Generic Letter 94 02, "Long-Term Solutions and U) grade of Interim Operating Recommendations for Tiermal Hydraulic Instabilities in Boiling Water Reactors," July 1994.

6. BWROG Letter 94078, "BWR Owners' Group Guidelines for Interim Corrective Action," June 1994.

l l

l FERMI UNIT 2 B 3.4.1 - 9 Revision 13. 08/06/99

Recirculation Loops Operating B 3.4.1 BASES l

i l

THERMAL POWER vs CORE FLOW Figure B 3.4.1 1 ll FERMI - UNIT 2 B 3.4.1 - 10 Revision 13, 08/06/99

1 l

spscoFicxparJ B. '{.1

3 /4. 4 REACTOR COOLANT SYS' TEM

! 3/4 4.1 RECIRCULATION SYSTEM I

RECIRCULATION LOOPS l LIMITING- CONDITION FOR OPERATION l.CO 393 2.'.1.P Two reactor coolant system recirculation loops shall be in operation.

APPLICABILfTY: OPERATIONAL CONDITIONS 1 and 2*.

ACTION:

I a. With one reactor coolant system recirculation loop not in operation:

LLo L41 1. Within 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months s: l Ail -

i NtTE~ '

s

}a) P1 the individual re I culation moL cforoller'forthe] $

pump flow

_ op atino recirculatio pump in the Manual p l l

b G 3'4.l.bg b) Reduce THERMAL POWER to less than or equal to 67.2% of RATED THERMAL POWER. )

I

> j 1

i Limit the speed %f the operating recirt l

or'eaual to 75X of rated pump speed, ydation pump to les/than) l j

'd) Inc ase Ine MINIMU LRITICAL POWtx Hall tr%,rx) darety Li tt th value for sing loop operation reg red by Specifica n l

W debO: MLHGt 8 MCFR. 5 L ce $= *l b I l I' Y'bb'I \, fS- - r l i

3 e) Change the Average Power Range Monitor (APRM) Simulated T ermal,.

@ M*l'g*3 Power - Upscale Flow Biased Scram [rmd "od :"sn "tr seirci-tu !'

4ER-Allowable Values to those applicable for single retircul at' ion l loop operation per Specifications 2.2.1 and 3.3.6.

l f) Perform Surveillt.nce Requirement 4.4.1.1.4 if THERMAL POWER is gggag less than or equal to 30% of RATED THERMAL POWER or the g*q ,,o f recirculation loop flow in the operating loop is less than or

{

v equal to 50% of rated loop flow.

2. -othe + k- " at 1:::: = :: = ,m ,, an i n ine um u ,,_ .

3 b.

1 With no reactor coolant system recirculation loop in operation while in i NMD OPERATIONAL CONDITION 1, immediately place the Reactor Mode Switch in the SHUTDOWN position.

c.

With no reactor coolant system recirculation loops in operation, while in OPERATIONAL CONDITION 2, initiate measures to place the unit in at least hc.T10tdC HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . .

1

{See cial Tes/ Exception 3./0.4.]

I e

FERMI - UNIT 2 3/4 4-1

~

Amendment No. EJ,M ,ES,E3, EJ.J0J, 122 PAGE 2- 0F 06 Noj

DISCUSSION OF CHANGES ITS: SECTION 3.4.1 RECIRCULATION LOOPS OPERATING L.3 (continued)

2) If a unit shutdown is required, the shutdown is only required (per ITS 3.3.1.1 Actions) to Mode 2. This is appropriate since the APRM flow biased scram setpoint that is affected by single loop operation is only applicable in Mode 1. Any further reduction in Thermal Power provides no assumed or credited compensatory measures.

L.4 CTS 3.4.1.1 Action a provides a 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months allowance to adjust various limits for single loop operation. In NUREG 1433 the typical limits that are allowed this 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months allowance include applying single loop limits for APLHGR and MCPR, which are not currently included in the CTS allowance. The Fermi Unit 2 ITS includes the new provision for applying single loop APLHrA and MCPR limits after a 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months allowance. This 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months allows sufficient time for a controlled transition while limiting the period where limits may h not be met. During this brief period the probability of an accident is minimal. Furthermore, the CTS already provides this same time period for other parameters to be not met. As such, including two additional parameters that may also require a transition period to establish new single-loop limits will not result in any significant impact on safety.

I RELOCATED SPECIFICATION _S

{

None IECHNICAL SPECIFICATION BASES The CTS Bases for this Specification have been replaced by Bases that reflect the format and applicable content of ITS 3.4.1 consistent with the BWR STS, NUREG-1433. Rev. 1.

i FERMI - UNIT 2 6 REVISION 13. 08/06/99l

l NO SIGNIFICANT HAZARDS EVALUATION i ITS: SECTION 3.4.1 - RECIRCULATION LOOPS OPERATING m l TECHNICAL CHANGES - LESS RESTRICTIVE (SDecif'eation 3.4.1 "L.4" Labeled Comments / Discussions)

Detroit Edison has evaluated the proposed Technical Specification change I identified as "Less Restrictive" in accordance with the criteria specified by 10 CFR 50.92 and has determined that the proposed change does not involve a significant hazards consideration.

I The bases for the determination that the proposed change does not involve a l significant hazards consideration is an evaluation of these changes against )

each of the criteria in 10 CFR 50.92. The criteria and the conclusions of the evaluation are presented below. 1

1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?

This change allows APLHGR and MCPR. which are not in compliance with single-loop required limits, increased restoration times. This change ,

will not result in a significant increase in the probability of an accident previously evaluated because APLHGR and MCPR values are not n considered initiators of any previously analyzed accident. This change j l I will not result in a significant increase in the consequences of an fr accident previously evaluated because the consequences of an event with ,

APLHGR or MCPR out of limits remains the same regardless of the allowed Completion Time.

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

This proposed change will not involve any physical changes to plant systems, structures, or components (SSC), or changes in normal plant l operation. Therefore, this change will not create the possibility of a l new or different kind of accident from any accident previously evaluated.

1 i

FERMI - UNIT 2 7 REVISION 13. 08/06/99l

NO SIGNIFICANT HAZARDS EVALUATION ,

ITS: SECTION 3.4.1 - RECIRCULATION LOOPS OPERATING l TECHNICAL CHANGES - LESS RESTRICTIVE l (Specification 3.4.1 "L.4" Labeled Comments / Discussions)

3. Does this change involve a significant reduction in a margin of safety?

. The proposed change does not involve a significant reduction in a margin l of safety because the 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months restoration time has been previously accepted for operation with other single-loop parameters not within k 3 1

limits. Applying the same 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months allows sufficient time for a controlled transition while limiting the period where limits may not be met.

l l

1 l

l l

i l

i

! I l

FERMI - UNIT 2 8 RAIISION 13 08/06/99l

Recirculation Loops Operating 3.4.1

'The (YAch C $M COie Ni Cr$\

3.4 REACTOR COOLANT SYSTEM (RCS) cou.thennel hydroidic. in5kbildy ,

/

Recirculation Loops Operating [

03.4.1 3.4.3 h=Two recirculation< >loops

(

~o ,. a.,e LCO w satchedfflows shall be in P

pd67wg4eB_ _

h* 0ne recirculation loop may be in operation provided the following limits are applied when the associated LCO is applicable:

y. LC0 3.2.1, " AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)," single loop operation limits (specified in t,p theCOLRg Q.g. LCO3.2.2,"MINIMUMCRITICALPOWERRATIO(MCPR)." single ( IM LM) loop operation limits ppecified in the COLRg; JEML*.

.,p. LC0 3.3.1.1, " Reactor Protection System (RPS) upute, hug Instrumentation," Function 2.b (Average Power Ra ,,g, e)

Monitors f+ar*+=wrf Simulated Thermal Power- ,

Allowable Value of_ Table 3.3 d.1-1_f t for single loop operatio -

in (no6e i o s e APPLICABILITY: MODES I and 2. !

ACTIONS

/ 3 4, l - /

~

g' CONDITION REQUIRED ACTION COMPLETION TIME f

Re trements of e A.1 tisfy the 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months LC not met, equirements of th LCO.

p.6

" : --' - : . . ..i n . , i s.= .

(continued)

WsEcur 5M ]

l l

BWR/4 STS 3.4-1 Rev 1, 04/07/95 l

db

Recirculation Loops Operating 3.4.1 (CT5f INSERT 3.4.1 - 1 l l

LC0 g '

3,y. l.l

b. 4 ,g @

4. THERMAL POWER is s 67.2% RTP. D

..........................N0TE ------------ - --- --

2'2 l#)

Application of the required limitations for single loop l operation may be delayed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after transition from two recirculation loop operation to single ,

recirculation loop operation. 3'4'I I

.........................................................FNa1 FERMI UNIT 2 Page 3.4 1 (1) (INSERT) REVISION 13. 08/06/99l

Recirculation Loops Operating -

B 3.4.1

, BASES APPLICABLE A plant specific LOCA analysis has been performed assuming SAFETY ANALYSES only one operating recirculation loop. This analysis has (continued) demonstrated that, in the event of a LOCA caused by a pipe break in the operating recirculation loop, the Emergency Core Cooling System response will provide adequate core 1 cooling: provided the APLHGR requirements are modified accordingly (Ref. 3). l The transient analyses of Chapter 15 of the SAR have also been performed for single recirculation loop operation (Ref. 3) and demonstrate sufficient flow coastdown characteristics to maintain fuel thermal margins during the abnormal operational transients analyzed provided the MCPR requirements are modified. During single rec *irculation loop operation, modification to the Reactor Protection System (RPS) average power range monitor setpoints is also required to accou(APRM) instrument nt for the different relationships between recirculation drive flow and reactor

2, , , ' core flow. The APLHGR and MCPR setpoints for single _ loop r

__ operation are specified in the COLR. The APRM waew-******'

fimulatedTjffp%/ tsetpoint is in LCO 3.3.1.1, " Reactor '

TetJeei Protection Sfsteli ( 5 Ins +n - M ation."

6 5.4.1~2- F YRecirculation loops ope ~(~UE5'*!O erion 2 of 4he-

= M k,7 E.i_..i.. lo cTR sa34(c)(a$0 LC0 Two recirculation loops are required to be in operation with their flows matched within the limits specified in SR 3.4.1,k to ensure that during a LOCA caused by a break of the piping of one recirculation loop the assumptions of the LOCA analysis are satisfied. With the limits specified in 5R 3.4.1Vnot met, the recirculation loop with the lower flow must be considered not in o f recirculation loop in operation,peration. With only modifications to theone required APLHGR limits (LCO 3.2.1, ' AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)'), MCPR limits (LCO 3.2.2 '

Simulated Thermal Power 4)W911 setpoint ([Lc0 3.3.1.1)hmay - l k

--> 3 applied to allow continued tiperation consistent with the <

assumptio of arence-3 % ^

s fd7 on oddIWalaYi'd 1

g,,

6 53.\3' x y -- >

C G

(continued)

!""/4 3YS- B 3.4-3 h 1, 04/0f/35-Tfev I3

l l

Recirculation Loops Operating B 3.4.1 1

Insert B 3.4.1 - 2 l

l Thermal hydraulic stability analysis (Ref. 5) has concluded that procedures for detecting and suppressing power oscillations that might be induced by a thermal hydraulic instability are necessary to provide reasonable assurance that

the requirements of Reference 4 are satisfied.

Insert B 3.4.1 3 Operations that exhibit core thermal hydraulic instability are not permitted. Additionally in order to avoid potential O power oscillations due to thermal hydraulic instability. N operation at certain combinations of power and flow are not permitted. These restricted power and flow regions are referred to as the " Scram" and " Exit" Regions. and are defined b by Bases Figure B 3.4.1-1.

A Note is provided to a'llow 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months following the transition \

to single loop operation from two loop operation to establish '

the applicable limitations in accordance with the single loop l-h analysis. The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months period is sufficient to make the adjustments given the relatively small changes required. This transition that results in applying the new single-loop allowable values to APRM OPERABILITY. is such that any APRM g

non-compliance with the required allowable value after this 4-hour allowance results in ACTIONS of LC0 3.3.1.1 being entered; no ACTION of LC0 3.4.1 would apply. Similarly, any D operation with APLHGR or MCPR out of limits results in the ACTIONS of 3.2.1 or 3.2.2 being entered: no ACTION of s LC0 3.4.1 would apply.

l l

FERMI UNIT 2 Page B 3.4 3 (INSERT) REVISION 13 08/06/99l l

INSERT THIS PAGE IN FRONT OF VOLUME 7 Volume 7 SECTION 3.6 -

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Volume 7 SECTION 3.6 (cont'd)

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PCIVs 3.6.1.3

) ACTIONS (continued)

C0lOITION REQUIRED ACTION -COMPLETION TIME

=

D. One or more D.1 Restore leakage rates 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for penetration flow )aths to within limit. leakage'on with one or more TIVs hydrostatically N inoperable due to tested line secondary containment without a closed O -

bypass leakage rate, system MSIV leakage rate.

T

-b-purge valve leakage M rate, rostatically tested ine leakage 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for rate, or EFCV leakage secondary y rate not within limit. containment bypass leakage M

n 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months for MSIV WI leakage lD M

N t 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for s purge valve leakage 8NQ C' "

72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months for leakage on g hydrostatically tested line on a w

v . closed system and EFCV leakage b!!G (continued) l FERMI. UNIT 2 3.6 11 Revision 13. 08/06/99

PCIVs 3.6.1.3 O ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME I

D. (continued) - - - -

NOTES------------

1. Isolation devices in high radiation areas may be verified by use of administrative means.

2. Isolation devices that are locked, sealed, or h otherwise secured may be verified by use of administrative means.

3. Only applicable to penetration flow paths isolated to restore leakage to within limits.

D.2 Verify the affected Once per 31 days penetration flow path for isolation is isolated. devices outside primary containment 8NQ Prior to entering MODE 2 or 3 from MODE 4, if primary containment was de inerted while in MODE 4 if not performed within the previous 92 days, for isolation devices inside primary containment (continued) j FERMI UNIT 2 3.6 12 Revision 13. 08/06/99

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PCIVs 3.6.1.3 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME i

&l E. Required Action and E.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion

. Time of Condition A. 8NQ e3 B. C. or D not met in MODE 1, 2. or 3. E.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months (n(

Ek:l F. Required Action and F.1 Initiate action to Immediately associated Completion isolate RHR-Shutdown

! Time of Condition A. Cooling System.

l B. C. or D not met for RHR-SDC PCIV(s) QB required to be Il OPERABLE during MODE 4 F.2 Initiate action to Immediately l or 5. restore valve (s) to OPERABLE status.

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hI FERMI UNIT 2 3.6 13 Revision 13 08/06/99 ;

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PCIVs 3.6.1.3

) SURVEILLANCE REQUIREENTS SURVEILLANCE FREQUENCY SR 3.6.1.3.1 ----- -

-- - --NOTE-- ----------------

Qs I Not required to be met when the isolation valves for one purge or containment pressure control supply line and one

, purge or containment pressure control exhaust line are open for inerting. de-V inerting. pressure control. ALARA or air quality considerations for personnel entry, or Surveillances that require the valves to be open.

Verify each drywell and suppression 31 days chamber purge system and containment pressure control isolation valve is closed.

SR 3.6.1.3.2 -----

---.N0TES ---- ---- -------

1. Valves and blind flanges in high radiation areas may be verified by use of administrative means.

2. Not required to be met for PCIVs that are open under administrative controls.

n Verify each primary containment isolation 31 days bs' manual valve and blind flange that is located outside primary containment and g is not locked, sealed, or otherwise t, secured and is requi.ed to be closed V during accident conditions is closed.

(continued)

-l FERMI - UNIT 2 3.6 14 Revision 13. 08/06/99

i PCIVs 3.6.1.3

) SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.1.3.3 ------- - -- --

NOTES----- ---- - --- -

1. Valves and blind flang2s in high radiation areas may be verified by use of administrative means.

2. Not required to be met for PCIVs that are open under administrative controls.

Y' Verify each primary containment isolation Prior to n -

manual valve and blind flange that is entering MODE 2 located inside primary containment and is or 3 from b< not locked, sealed, or otherwise secured MODE 4 if Da and is required to be closed during primary t accident conditions is closed. containment was L de inerted while in MODE 4 if not performed within the previous 92 days l SR 3.6.1.3.4 Verify continuity of the traversing 31 days incore probe (TIP) shear isolation valve explosive charge.

O d

SR 3.6.1.3.5 Verify the isolation time of each power In accordance Q] operated automatic PCIV, except for with the h MSIVs, is within limits. Inservice C( Testing Program (continued) l l

l l FERMI - UNIT 2 3.6 15 Revision 13. 08/06/99 i

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PCIVs 3.6.1.3

/m

) SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.1.3.6 Perform leakage rate testing for each 184 days primary containment purge valve with resilient seals. AND 1

i Once within 92 days after opening the valve l SR 3.6.1.3.7 Verify the isolation time of each MSIV is In accordance

= 3 seconds and s 5 seconds. with the l

Inservice Testing Program l

l SR 3.6.1.3.8 Verify each automatic PCIV actuates to 18 months l the isolation position on an actual or simulated isolation signal.

l SR 3.6.1.3.9 Verify each reactor instrumentation line 18 months l

EFCV actuates on a simulated instrument 1

line break to restrict flow.

1 SR 3.6.1.3.10 Remove and test the explosive squib from 18 months on a each shear isolation valve of the TIP STAGGERED TEST System. BASIS 1

l (continued) l l

l l FERMI - UNIT 2 3.6 16 Revision 13. 08/06/99 V

PCIVs 3.6.1.3

.= %

l- -) SURVEILLANCE REQUIREENTS (continued)

SURVEILLANCE FREQUENCY O

h SR 3.6.1.3.11 Verify the combined leakage rate for all In accordance Y

Ql secondary containment bypass leakage paths that are not provided with a seal with the Primary system is s 0.04 L, when pressurized to Containment

= 56.5 psig. Leakage Rate p Testing Program and Inservice fl Testing Program L:

SR-3;6.1.3.12 Verify combined MSIV leakage rate for all In accordance four main steam lines is s 100 scfh when with the tested at = 25 psig. Primary Containment i Leakage Rate

! Testing Program SR 3.6.1.3.13 ..........--- NOTE -- - --- ----

Only required to be met in MODES 1, 2 and 3.

l Verify combined leakage rate through In accordance hydrostatically tested lines that with the penetrate the primary containment is Primary I within limits. Containment l 1 -. Leakage Rate l Testing Program

(

l 1 FERMI - UNIT 2- 3.6 17 Revision 13. 08/06/99 j l

E PCIVs B 3.6.1.3 f'

BASES ACTIONS (continued) l D.1 and D.2 With one or more penetration flow paths with one or more PCIVs inoperable due to secondary containment bypass leakage

^ rate (SR 3.6.1.3.11). MSIV leakage rate (SR 3.6.1.3.12),

f d purge valve leakage rate (SR 3.6.1.3.6), hydrostatically tested line leakage rate (SR 3.6.1.3.13), or EFCV leakage h rate (SR 3.6.1.3.9) not within limit, the assumptions of the LQ l

safety analysis may not be met. Therefore, the leakage must l be restored to within limit. Restoration can be ;

\ accomplished by repairing the leaking PCIV(s) (and exiting i Condition D).

The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Completion Time for leakage on hydrostatically tested lines and for secondary containment bypass leakage is k reasonable considering the time required to restore the e

leakage by isolating the penetration and the relative importance of leakage to the overall containment function.

l For MSIV leakage, an 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Completion Time is allowed. The i Completion Time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months for MSIV leakage allows a period of time to restore the MSIVs to OPERABLE status given the fact that the MSIV closure will result in isolation of the g main steam line(s) and potential for plant shutdown. The 24 w hour Completion Time for purge valve leakage is acceptable T considering the purge valves remain closed so that a gross breach of the containment does not exist. The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time for leakage on hydrostatically tested lines ,

on a closed system is acceptable based on the available water seal expected to remain as a gaseous fission product i

boundary during the accident and the associated closed

! system. The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time for EFCV leakage is acce) table based on the instrument and small pipe diameter l

\ of t1e penetration (hence, reliability) to act as a penetration isolation boundary and small pipe diameter of the affected penetrations.

Alternately, the leakage can be restored to within limit by l

isolating the penetration that caused the limit to be

exceeded by use of one closed and de activated automatic valve, closed manual valve, or blind flange. When a i penetration is isolated, the leakage rate for the isolated penetration is assumed to be the actual pathway leakage through the isolation device. If.two isolation devices are used to isolate the penetration, the leakage rate is assumed to be the lesser actual pathway leakage of the two devices.

i$l In this case (isolation of the affected penetration), the UNIT 2 B 3.6.1.3 - 9 Revision 13, 08/06/99 l FERMI l

PCIVs B 3.6.1.3 BASES ACTIONS (continued) leaking PCIV(s) remain inoperable due to leakage and Condition D remains applicable. Required Action D.2 must also be performed to verify the penetration is isolated on a periodic basis. This is necessary to ensure that primary containment penetrations required to be isolated following h an accident are isolated. The Completion Time of "once per 31 days for isolation devices outside primary containment" is appropriate because the devices are operated under administrative controls and the probability of their misalignment is low. For the devices inside primary containment, the time period specified " prior to entering MODE 2 or 3 from MODE 4. if primary containment was de inerted while in MODE 4. if not performed within the previous 92 days" is based on engineering judgment and is considered reasonable in view of the inaccessibility of the devices and other administrative controls ensuring that device misalignment is an unlikely possibility.

Required Action D.2 is modified by three Notes. Note 1 applies to valves and blind flanges located in high radiation areas and allows them to be verified by use of administrative means. Allowing verification by administrative means is considered acceptable, since access to these areas is typically restricted. Therefore, the 3robability of misalignment of these valves, once they have 3een verified to be in the pro)er position, is low. Note 2 ap) lies to isolation devices t1at are locked, sealed, or otlerwise secured in position and allows these devices to be verified closed by use of administrative means. Allowing verification by administrative means ic considered acceptable since the function of locking, sealing, or securing components is to ensure that these devices are not inadvertently repositioned. Note 3 states that verification that the penetration is isolated a> plies only to penetration flow paths isolated to restore lea (age within limits.

l FERMI - UNIT 2 B 3.6.1.3 - 10 Revision 13 08/06/99

PCIVs B 3.6.1.3 BASES ACTIONS (continued) 3l E.1 and E.2 If any Required Action and associated Completion Time cannot n be met in MODE 1, 2. or 3, the plant must be brought to a MODE in which the LC0 does not apply. To achieve this bd status, the plant must be brought to at least MODE 3 within 4tu 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and to MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating Q experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

l M If any Required Action and associated Completion Time cannot be met, the unit must be placed in a condition in which the LCO does not apply. Action must be immediately initiated to isolate the RHR Shutdown Cooling System. However, if the shutdown cooling function is needed to provide core cooling, these Required Actions allow the penetration flow path to remain unisolated provided action is immediately initiated to restore the valve to OPERABLE status or to isolate the RHR Shutdown Cooling System (i.e., provide alternate decay heat removal capabilities so the penetration flow path can be isolated). Actions must continue until the valve is restored to OPERABLE status or the RHR Shutdown Cooling System is isolated.

SURVEILLANCE SR 3.6.1.3.1 REQUIREMENTS This SR ensures that the drywell and suppression chamber I '

purge system isolation valves (6 inch,10 inch, 20 inch, and 24 inch) and the containment pressure control valves (1 inch) are closed as required or, if spen, open for an allowable reason. If a purge or containment pressure 7 control valve is o)en in violation of this SR, the valve is D considered inopera)le. If the inoperable valve is not h otherwise known to have excessive leakage when closed, it is not considered to have leakage outside of limits. Primary y

kl containment purge and containment pressure control valves are only required to be closed in MODES 1, 2. and 3 (i.e..

no isolation instrumentation functions of LCO 3.3.6.1 are fl required to be OPERABLE for isolation of these valves outside of MODES 1, 2, and 3). If a LOCA inside primary l FERMI - UNIT 2 B 3.6.1.3 - 11 Revision 13. 08/06/99

PCIVs B 3.6.1.3 BASES SURVEILLANCE REQUIREMENTS (continued) containment occurs in these MODES, the purge valves may not be capable of closing before the pressure pulse affects systems downstream of the purge valves. At other times m (e.g., during handling of irradiated fuel).. pressurization concerns are not present and the purge and containment pressure control valves are allowed to be open. The SR is modified by a Note stating that the SR is not required to be met when the purge or containment pressure control valves

'l are open for the stated reasons. The Note states that these valves may be opened for inerting, de inerting pressure control. ALARA or air quality considerations for personnel entry, or Surveillances that require the valves to be open.

The purge valves (6 inch.10 inch, 20 inch, and 24 inch) and h the containment pressure control valves (1 inch) are capable m of closing in the environment following a LOCA. Therefore, these valves are allowed to be open for limited periods of f time. The 31 day Frecuency is consistent with other PCIV requirements discussee in SR 3.6.1.3.2.

SR 3.6.1.3.2 This SR verifies that each primary containment isolation manual valve and blind flange that is located outside i primary containment and is not locked, sealed. or otherwise secured and is required to be closed during accident conditions is closed. The SR helps to ensure that post accident leakage of radioactive fluids or gases outside the primary containment boundary is within design limits.

This SR does not require any testing or valve manipulation.

Rather it involves verification that those PCIVs outside primary containment, and capable of being mispositioned, are in the correct position. Since verification of valve position for PCIVs outside primary containment is relatively easy, the 31 day Frequency was chosen to provide added assurance that the PCIVs are in the correct positions.

Two Notes have been added to this SR. The first Note allows valves and blind flanges located in high radiation areas to be verified by use of administrative controls. Allowing verification by administrative controls is considered acceptable since access to these areas is typically restricted during MODES 1, 2. and 3 for ALARA reasons. I Therefore, the 3robability of misalignment of these PCIVs. !

once they have 3een verified to be in the proper position. l 1s low. A second Note has been included to clarify that

PCIVs B 3.6.1.3 BASES SURVEILLANCE REQUIREMENTS (continued)

PCIVs that are open under administrative controls are not required to meet the SR during the time that the PCIVs are open. This SR does not apply to valves that are locked, sealed, or otherwise secured in the closed position since y these were verified to be in the correct position upon locking, sealing, or securing.

SR 3.6.1.3.3 1

gq This SR verifies that each primary containment isolation manual valve and blind flange that is located inside primary Ql containment and is not locked, sealed, or otherwise secured and is required to be closed during accident conditions is closed. The SR helps to ensure that post accident leakage of radioactive fluids or gases outside the primary b containment boundary is within design limits. For PCIVs 1 inside primary containment, the Frequency defined as " prior to entering MODE 2 or 3 from MODE 4 if primary containment was de inerted while in MODE 4, if not performed within the previous 92 days" is appropriate since these PCIVs are operated under administrative controls and the probability 4 of their misalignment is low. This SR does not apply to valves that are locked, sealed, or otherwise secured in the closed position since these were verified to be in the correct position upon locking, sealing, or securing.

Two Notes have been added to this SR. The first Note allows valves and blind flanges located in high radiation areas to be verified by use of administrative controls. Allowing verification by administrative controls is considered acceptable since the primary containment is inerted and access to these areas is typically restricted during MODES 1, 2, and 3 for ALARA reasons. Therefore, the probability of misalignment of these PCIVs, once they have been verified to be in their proper position, is low. A second Note has been included to clarify that PCIVs that are o)en under administrative controls are not required to meet t1e SR during the time that the PCIVs are open.

l FERMI UNIT 2 B 3. 6.1.3 - 13 Revision 13 08/06/99

PCIVs B 3.6.1.3 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.6.1.3.4 The traversing incore probe (TIP) shear isolation valves are actuated by explosive charges. Surveillance of explosive charge continuity provides assurance that TIP valves will actuate when required. Other administrative controls, such as those that limit the shelf life of the explosive charges, must be followed. The 31 day Frequency is based on operating experience that has demonstrated the reliability of the explosive charge continuity.

SR 3.6.1.3.5 y Verifying the isolation time of each power operated automatic PCIV is within limits is required to demonstrate OPERABILITY. MSIVs may be excluded from this SR since MSIV full closure isolation time is demonstrated by SR 3.6.1.3.7.

The isolation time test ensures that the valve will isolate in a time period less than or equal to that assumed in the A safety analyses. The isolation time and Frequency of this bd SR are in accordance with the requirements of the Inservice Testing Program. I SR 3.6.1.3.6 For primary containment ) urge valves with resilient seals (6 inch. 10 inch, 20 incl. and 24 inch), additional leakage !

rate testing beyond the test requirements of 10 CFR 50 Appendix J. Option B (Ref. 3), is required to ensure 4 OPERABILITY. This will ensure that leakage is s 0.05 L when tested at P Operating experience has demonstrated that this type of seal has the potential to degrade in a shorter time period than do other. seal types. Based on this observation and the importance of maintaining this penetration leak tight (due to the direct path between primary containment and the environment), a Frequency of 184 days was established.

Additionally, this SR must be performed once within 92 days after opening the valve. The 92 day Frequency was chosen recognizing that cycling the valve could introduce additional seal degradation (beyond that which occurs to a valve that has not been opened). Thus, decreasing the interval (from 184 days) is a prudent measure after a valve has been opened.

l FERMI UNIT 2 B 3.6.1.3 - 14 Revision 13. 08/06/99

PCIVs B 3.6.1.3 BASES SURVEILLANCE REQUIREMENTS (continued)

The primary containment purge valves are only required to meet leakage rate testing requirements in MODES 1, 2, and 3.

(i.e., no isolation instrumentation functions of LC0 3.3.6.1 are required to be OPERABLE for purge system isolation outside of MODES 1, 2, and 3). If a LOCA inside primary containment occurs in these MODES, purge valve leakage must be minimized to ensure offsite radiological release is within limits. At other times (e.g., during handling of irradiated fuel). pressurization concerns are not present and the purge valves are not required to meet any specific leakage criteria.

SR 3.6.1.3.7 Verifying that the isolation time of each MSIV is within the specified limits is required to demonstrate OPERABILITY. ,

The isolation time test ensures that the MSIV will isolate )

in a time period that does not exceed the times assumed in I the DBA analyses. This ensures that the calculated radiological consequences of these events remain within 10 CFR 100 limits. The minimum stroke time ensures that isolation does not result in a pressure spike more rapid than assumed in the transient analyses. The Frequency of this SR is in accordance with the requirements of the Inservice Testing Program.

SR 3.6.1.3.8 Automatic PCIVs close on a primary containment isolation signal to prevent leakage of radioactive material from o primary containment following a DBA. This SR ensures that each automatic PCIV will actuate to its isolation position bd on a primary containment isolation signal. The LOGIC SYSTEM

h. FUNCTIONAL TEST in SR 3.3.6.1.5 overlaps this SR to provide Chl complete testing of the safety function. The 18 month Frequency was developed considering it is prudent that this Surveillance be performed only during a unit outage since isolation of >enetrations would eliminate cooling water flow and disrupt tie normal operation of many critical components. Operating ex>erience has shown that these l components usually pass t1is Surveillance when performed at I the 18 month Frequency. Therefore, the Frequency was !

concluded to be acceptable from a reliability standpoint.

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(( FERMI-UNIT 2 B 3.6.1.3 -15 Revision 13. 08/06/99 l

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PCIVs B 3.6.1.3 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.6.1.3.9 This SR requires a demonstration that each reactor .

instrumentation line excess flow check valve (EFCV) is OPERABLE by verifying that the valve restricts flow on a simulated instrument line break. This SR provides assurance that the instrumentation line EFCVs will perform so that predicted radiological consequences will not be exceeded during the postulated instrument line break event evaluated in Reference 5. The 18 month Frequency is based on the typical performance of this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown that these components usually pass this Surveillance when performed at the 18 month Frequency.

Therefore, the Frecuency was concluded to be acceptable from a reliability stancpoint.

SR 3.6.1.3.10 The TIP shear isolation valves are actuated by explosive charges. An in place functional test is not possible with this design. The explosive squib is removed and tested to provide assurance that the valves will actuate when required. The replacement charge for the explosive squib shall be from the same manufactured batch as the one fired or from another batch that has been certified by having one of the batch successfully fired. No squib will remain in service beyond the expiration of its shelf life or its operating life. The Frequency of 18 months on a STAGGERED TEST BASIS is considered adequate given the administrative controls or. replacement charges and the frequent checks of circuit continuity (SR 3.6.1.3.4).

SR 3.6.1.3.11 I This SR ensures that the leakage rate of secondary containment bypass leakage paths is less than the specified leakage rate. This provides assurance that the assumptions in the radiological evaluations of Reference 1 are met. The leakage rate of each by) ass leakage path is assumed to be 1 the maximum pathway leacage (leakage through the worse of OQl the two isolation valves) unless the penetration is isolated by use of one closed and de activated automatic valve.

l closed manual valve, or blind flange. In this case, the !

i kl FERMI-UNIT 2 B 3.6.1.3 - 16 Revision 13. 08/06/99

PCIVs B 3.6.1.3 BASES SURVEILLANCE REQUIREMENTS (continued) leakage rate of the isolated bypass leakage path is assumed a to be the actual pathway leakage through the isolation device. If both isolation valves in the penetration are btf closed, the actual leakage rate is the lesser leakage rate of the two valves. The frequency is required by the Primary Containment Leakage Rate Testing Program. This SR simply D imposes additional acceptance criteria. Additionally, some secondary containment bypass paths (refer to UFSAR 6.2.1.2.2.3) use non PCIVs and therefore are not addressed by the testing Frequency of 10 CFR 50. Appendix J. testing.

To address the testing for these valves, the Frequency also includes a requirement to be in accordance with the Inservice Testing Program.

Secondary containment bypass leakage is also considered part l of L,.

SR 3.6.1.3.12 The analyses in References 1 and 4 are based on leakage that is less than the specified leakage rate. Leakage through all four main steam lines must be s 100 scfh when tested at api (25 psig). This ensures that MSIV leakage is properly accounted for to assure safety analysis assumptions, regarding the MSIV LCS ability to provide a positive pressure seal between MSIVs. remain valid. This leakage test is performed in lieu of 10 CFR 50. Appendix J. Type C test requirements, based on an exemption to 10 CFR 50 Appendix J. As such, this leakage is not combined with the Type B and C leakage rate totals. The Frequency is required by the Primary Containment Leakage Rate Testing Program.

... p 1

l

)

PCIVs B 3.6.1.3 i

BASES SURVEILLANCE REQUIREMENTS.(continued)

SR 3.6.1.3.13 Surveillance of hydrostatically tested lines provides assurance that the calculation assumptions of Reference 2 p' are met. The acceptance criteria for the' combined leakage V of all hydrostatically tested lines is 1 gpm times the b number of valves per penetration, not to exceed 3 gpm, when s (= 62.2 psig). Additionally, a combined '

G tested leakageatrate 1.1lim P,it of 5 5 gpm when tested at 1.1 P k '

(= 62.2 psig) is applied for all hydrostatically tested PCIVs that penetrate containment. The combined leakage rates must be demonstrated in accordance with the leakage rate test Frequency required by Primary Containment Leakage Rate Testing Program.

This SR has been modified by a Note that states that these valves are only required to meet the combined leakage rate in MODES 1, 2, and 3, since this is when the Reactor Coolant System is pressurized and primary containment is required.

In some instances, the valves are required to be capable of automatically closing during MODES other than MODES 1, 2 and 3. However, specific leakage limits are not applicable in these other MODES or conditions.

REFERENCES 1. UFSAR, Chapter 15.

2. UFSAR, Table 6.2-2.

3. 10 CFR 50 Appendix J. Option B.

4. UFSAR, Section 6.2.

5. UFSAR. Section 15.6.2.

I j

1 l

I l

B 3.6.1.3 - 18 Revision 13, 08/06/99 hFERMI-UNIT 2

specs Pt cATHN 34,l 3 l

/%

) CONTAINMENT SYSTEMS pgYWELL AND SUPPRESSION CHAMBER PURGE SYSTEM LIMITING CONDITION FOR OPERATION sgg,g3,( 3.6.1.8 The drywell and suppression chamber purge system {-0 ...a. N ...J.,

g ^^ ind, E-f 2' '-" ""^t) may be in operation with the supply and exhaust ' - -

isolation valves in one supply line and one exhaust line open for inerting, I deinertingorpressurecontrol. Nitrogen VENTING / makeup and"pressure control E!ME*M""?" '# '3" "!"4 kW ; " "40 APPLICABILITY: OPERATIONAL CONDITIONS 1,g2 and 3. j l (ADe actions Nore t y _['g 'o EILQB:

G(Mcod Acnons 4cTrorlf HOTE plo1E 3L >} Y

a. With a drywell and suppression chamber purge system supply and/or g

3 ricT1oM k sa i exhaust isolation valve open, except as permitted above, close the valve (s) or otherwise isolate the penetration (s) within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or l be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 'and in COLD h l Ac.1Too E SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b. With a drywell and suppression chamber purge system supply and/or ls 0- i kgD exhaust isolation valve (s) with resilient material seals having a 1 l measured leakage rate exceeding the limit of Specification ho r 4.6.1.8.2, restore the inoperable valve (s) to OPERABLE status d p.,

pg hCR0d g within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT SHUTDOWN within the next 12

. hours and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . lju r4 *g (Aco1 A c m ia B '> L.1 c SURVEILLANCE PE0VIREMENTS x

5 4 % .113,1 4.6.1.8.1 [Before beina opened br purge / vent operation b e SETS) the n tJoTg - drywell ano suppression enamDer purge supply and exhaust butterfly isolation 7 valves shall be verified not to have been open for purge / vent operatic R thre W CT: % . == t M . ^ ^ , ,,,,, , , . , t l , -- to: ?!5 d:,;.- l l! '*-

. 4 34 3.G.l3.4 . .6.1.8.2

>m _ At leas,_t

-< .m _t once per[92

.a ..m days}

ation for each ? ...J., ;;.0

..e drywell and suppression chamber purge supply and exhaust isolation vahn with resilient material seals shall be demonstrated OPERABLE by verifywg that_the measured leakage rate 670" -1

{ . .. . g.2 6 *- : r:: u m " n, f( J.6J,3s / __* Primary containment nitrogen VENTING ano pressure control is permitted 8 I

through the 1-inch valves :nd i: n:t : dje*+ 'e t% 90 her: per 255 d:- I E

W

l,%

FEE.P.! - L' NIT 2 2/4 6-14 Amendment No. EB '

[('8/ h PAGE [0 - 0F 09 jdf

Spec.lFicA'rtoA) 34.1.3 m

]

CONTAINMENT SYSTEMS 3/4.6.3 PRIMARY CONTAINMENT ISOLATION VALVES kel y .

LIMITING CONDITION FOR'0PERATION '

[r'eAchrDwMdh-+o.soppratsAmckspbu erm'm breekW)5 3,.6.3 Each primary containment isolation valve *:d 3 ._.____n..._mo ..3,-

--r-*-- '-r*----t

-- lH shall be OPERABLE.** 1 APPLICABILITY: OPERATI L CONDITIONS 1, 2, and 3.h d M 8 # -

I bo 444: When reguird b loshu M hea .3 f,,

M.T McTtods poTE 2. A.8 y pg g k MO ND Ato m 5 3,4

a. With onene enrelof the primary con ainment isolation valves L' inoperable.pnaintain at isast one isosas10n valve Un KABLE in eacy (affectea nenetration that is open ant l(Elthin 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months eitner:

F --m .. . . . _.__._------m x l

2. Isolate each affected penetration by use of at least one RegvWJAcNM deactivated automatic valve secured in the isolated d' g,l position,* or v, .

Isolateeachaffectedphrationbyuseofatleastone i2

, 3.

ocke closed manual v=1u= ~ h1=a6 M=aaa * * "

n

} ,

i GF c h eck y*Q.%ttuHow 5440/4 l. 10 g Otherwise, be in at least HOT SHUTDOWN within the next 12 nours w.

AC'Tiog g and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . l b With one or more of the reactor' instrumentation line excess flow

$ God C.,0 check valves inoperable, operation may continue,.... m...

. . . . . . . . l (n e-- =--- _ - ,

_c_--7 :::'te:M g rovided that within l rs either:

. ine inoperapie vaive is returned to OPERABLE status, arr j b 0 8 (2. The instrument itne is isolated [and the assoctasea jh*

gg m 3 instrument is aeciarea inoperapie.

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

$0 k77dM O. b) 5 SenorJ NofE I *lsolation valves closed to satisfy these requirements may be reopened on an intt_rmittent basis under administrative control.

Nmr 1 + .1 *4E:20 = :::!9 closed valves may be opened on an intermittent basis under j 1 d

,gg g ,g,3,7 administrative ontrol. , 1

. m. .n.3 .a q FERMI - UNIT 2 3/4 6 20 Amendment No. 4 , 102 hh PAGE 7 0F 09

DISCUSSION OF CHANGES ITS: SECTION 3.6.1.3 - PCIVs TECHNICAL CHANGES HORE RESTRICTIVE M.1 CTS 4.6.1.1.b,2 requires verification of valve position for valves located within high radiation areas "during each Cold Shutdown if not performed within the previous 31 days." ITS SR 3.6.1.3.2 requires a verification every 31 days, but allows the verification to be "by administrative means" for those valves located in high radiation areas (i.e., Note 1 to SR 3.6.1.3.2). Since the high radiation area is a controlled access area, and each closed valve is also administratively controlled, the 31-day verification could consist of an on paper verification that the position of those valves has not changed. This represents an increased surveillance frequency since the ITS will not allow deferral of all verification until the next Cold Shutdown, and as such, is a more restrictive change, with no negative impact on safety.

M.2 CTS 3.6.1.2 Actions for primary containment leakage above established limits only restrict reactor coolant heatup beyond 200*F (which would allow a startup and control rod withdrawal from cold conditions, i .e. , < 200*F). Furthermore, if primary containment leakage above established limits were discovered while operating, the CTS Actions are non specific as to the appropriate required actions. Under the same conditions ITS LC0 3.0.4 will not allow a reactor startup to commence with containment leakage outside limits; just as CTS requires. However, a restriction is added by ITS 3.6.1.3 Action E to require a plant shutdown to Mode l 4, if leakage rates are discovered outside established limits and cannot be corrected within the times provided in ITS Action D. l The Completion Times provided in ITS Action D (i.e., (a) MSIV l leakage is required to be restored within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , consistent with restoration (cr isolation) time for inoperable MSIVs: and (b) hydrostatically tested valve leakage is required to be restored within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months if on a closed system and 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months if not on a closed system, based on the available water seal) present a restriction beyond CTS requirements. Additionally, Required Action D.2 is included in the ITS to require periodic verification of the closed status of any valve isolated to restore leakage to within limits. While ITS 3.6.1.3 is more restrictive, it provides appropriate Actions, commensurate with the type of leakage failure experienced, to ensure the plant is placed in a configuration consistent with the design basis. Therefore, this change will

~, result in an improvement in plant safety and has no negative impact on safety.

FERMI - UNIT 2 4 REVISION 13, 08/06/99l

l l

PCIVs 3.6.1.3 Insert 3.6.1.3-2 l 2. Isolation devices ,

i that are locked, sealed. '

or otherwise secured '

may be verified by use of administrative means.

)

Insert 3.6.1.3 3 /g7sh 7'/

l CONDITION REQUIRED ACTION COMPLETION TIME D. One or more secondary D.1 Restore leakage rate 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for leakage on containment bypass to within limit. hydrostatically tested leakage rate, MSIV line without a closed /g ,.7 leakage rate, purge valve leakage rate, system

%s hydrostatically 6NQ tested line leakage rate, or EFCV leakage 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for secondary rate not within containment bypass (obcM) limit. leakage MQ 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months for MSIV hocE2) leakage MQ

/n i g 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for purge $ Act b valve leakage (

1 MQ 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months for leakage on hydrostatically tested line on a /g4,,3 closed system and EFCV( , g leakage '

MQ

_ (continued) l FERMI UNIT - 2 Page 3.6-10(1) (INSERT) REVISION 13 l 08/06/99ll

Pcn/5

p. i .3

, CONDITION REQUIRED ACTION COMPLETION TIME O

D. (continued) ...........N0TES.-- ..-....

1. Isolation devices in high radiation areas may be verified by use of administrative means.

2. Isolation devices that are locked, sealed, or otherwise secured may be verified by use of administrative i means.

3. Only applicable to penetration flow paths isolated to l restore leakage to within limits.

D.2 Verify the affected Once per 31 days for penetration flow isolation devices path is isolated. outside primary containment AND Prior to entering MODE 2 or 3 from MODE 4. if primary containment was de.inerted while in MODE 4. if not performed within the previous 92 days, for isolation devices inside primary containment l

l FERMI UNIT 2 Page 3.6 10(2) (INSERT) REVISION 13 08/06/99l

[

PCIVs B 3.6.1.3 BASES ACTIONS C.1 and C.2 (continued) 72 - - -

~

MODES 1, 2, and 3.+ The Completion Time of W hours'is b

"$te cloStd S YS } reasonable considering the instrument and the small pipe y fn/ 5P (MAY ik i diameter of penetration (hence, reliability) to act as a -

(LT M p .g o f j penetration isolation boundary and the small pipe diameter-of the affected penetrations.~ In the event the affected l!E g

g p Y. penetration flow-path is isolated in accordance with Required Action C.I, the affected penetration must be -

verified to be isolated on a periodic basis. This is k f,fo necessary to ensure that primary containment penetrations required to be isoMed_fpllowina an accident are iselcted.

1 W Compliti56 Time of once per 31 days for verifying each affected penetration is isolated is appropriate because the valves are operated under administrative controls and the probability of their aisalignment is low.

Condition C is modified by a Note indicating that this Condition is only applicable to penetration flow paths with only one PCIV. For penetration flow paths with two PCIVs, Conditions A and B provide the riate Raa" W d Actions.

Gs. Mek_ Q

. Required Action C.2 is modified Note *ths* applies to valves and blind flanges located in high radiation areas and allows them to be verified by use of administrative means.

Allowing verification by administrative means is considered acceptable, since access to these areap is typically 9 restricted. Therefore, the probabilitysof misalignment of

) these valves, once they have been verified to be in the proper position, is low. ,

pisce l#SE 6 g g' .

M

% a d 0,2. s wt.3%

e Wi th "- r--t --d:k-- " H- -- ' - ' --r7^ i

?-% :te not within limit, the assumptions of the safety' j h f analysis may not be met. Therefore, the leakage must be h restored to within limit,;;i".'- e ' %5t n t'- - -4) x

f. $

l ter-'^M by isolating the penetration that caused the

~

limit to be exceeded by use of one closed and de-activated O automatic valve, closed manual valve, or blind flange. When W a penetration is isolated, the leakage rate for the isolated '

penetration is assumed to be the actual pathway leakage I through the isolation device. If two isolation devices are used to isolate the penetration, the leakage rate is assumed (continued)

- l/4 U S- B 3.6-21 h 1, 04/07/95r 1 '

Rev is

, Bev II A RBI 5~

PCIVs B 3.6.1.3 m.

Insert B 3.6.1.3-13 j Note 2 applies to isolation devices that are locked, sealed, or otherwise secured in position and allows these devices to be verified closed by use of administrative means. Allowing i verification by administrative means is considered acceptable since the function of locking, sealing. or securing components is to ensure that these device:; arc rct inadvertently repositioned, r l

@d I Insert B 3.6.1.3 4a v One or more penetration flow paths with one or more PCIVs inoperable due to secondary containment bypass leakage rate (SR 3.6.1.3.11). MSIV leakage rate (SR 3.6.1.3.12), purge valve leakage rate (SR 3.6.1.3.6). hydrostatically tested g line leakage rate (SR 3.6.1.3.13), or EFCV leakage rate (SR 3.6.1.3.9)

/

Insert B 3.6.1.3 4b Restoration can be accomplished by repairing the leaking g j PCIV(s) (and exiting Condition D).

Alternately, the leakage can be restored to within limit i

FERMI UNIT 2 Page B 3.6 21 (INSERT) REVISION 13 08/06/99l

PCIVs B 3.6.1.3 r

e usetLr 8 34.1.3 - SA. s 4cs4cd I;4e5 witkavt A closed Sy. Hem ACTION

~

~D.J (continued) d hM MMO *#+*4 a %

bvpss leakap 6 p,f k The t

o be4the ShiTay leakage of the two devices. .!

hourlesser actualTimeris Completion reasonable considering the X time required to restore th'e leakage by isolating the penetration and the relative importance of sem.d rf- PC/V M s T:t l

gggg e-twat i;;p;e leakage to the overall containment g3s. :5 --Q -

'*"'**") fh

~

E1 e...

Ir. . ^ event one or more containment purge y- s are not within e purge valve lealiigi~HmTts7p- e valvi Isaksgs must be ored to within limits or affected penetration ' be isolated. Th thod of isolation must j De by the use o' least one ation barrier that cannot be adversely affect a ngle active failure. Isolation barriers that meet thi losed and de-actlyated automa va iterion closedare a >@l valve, and blind manua flange % If a e valve w resilient seals is utilized to sat'sfy R red Action E.I, must have been f

{.3 demonstra to meet the leakage re rements of

- SR 3.6 . 7. The specified Completto ime is reasonable, \ / !

[ cons ering that one containment purge va s hat a gross breach of containment does n exist.

remains closed g / i i ~ accordance with Required Action E.z, Inis penetratioy

' f1 ath must be verified to be isolated on a perioyc' basis. The periodic verification is necessary to fisure that con nment penetrations required to be i ated following a ccident, which are no longer able of being automatically lated, will be in the ation position should an event o r. This Require etion does not require any testing valve mani ation. Rather, it involves verification t th isolation devices outside ;

containment and potentia 11 apable of being mispositioned

{ are in the correct posi n. r the isolation devices '

inside containment, e time per specified as " prior to entering NODE 2 from MODE 4 if t performed within th 1 previous 92 d

is based on engineer judgment and is Iconsidere easonable in view of the inac sibility of 1 isolat devices and other administrative c Rrols that wghalvnnesibilitvnsure that isolation device misalignment fAan

-u --

M (continued)

MD/4 FS B 3.6-22 "r; I, 04/07/05 I

fteV 13ll !

licy G. j

\

PCIVs B 3.6.1.3 INSERT ~B 3.6.1.3-5a In this case (isolation of the affected penetration), the leaking PCIV(s) remain inoperable due to leakage and ,

Condition D remains applicable. Required Action D.2 must I also be performed to verify the penetration is isolated on a periodic basis. This is necessary to ensure that primary containment penetrations required to be isolated following an accident are isolated. The Completion Time of "once per 31 days for isolation devices outside primary containment" is appropriate because the devices are operated under administrative . controls and the probability of their

~^

misalignment is low. For the devices inside pr1 mary containment, the time period specified " prior to entering MODE 2 or 3 from MODE 4. if primary containment was de inerted while in MODE 4. if not performed within the previous 92 days" is based on engineering judgment and is considered reasonable in view of the inaccessibility of the devices and other administrative controls ensuring that O I device misalignment is an unlikely possibility. ldg Required Action D.2 is modified by three Notes. Note 1 applies to valves and blind flanges located in high radiation areas and allows them to be verified by use of administrative means. Allowing verification by administrative means is considered acceptable, since access to these areas is typically restricted. Therefore, the probability of misalignment of these valves, once they have been verified to be in the proper position, is low. Note 2 applies to isolation devices that are locked, sealed, or otherwise secured in position and allows these devices to be verified closed by use of administrative means. Allowing j verification by administrative means is considered l

acceptable since the function of locking, sealing. or ;

securing components is to ensure that these devices are not !

inadvertently repositioned. Note 3 states that verification that the penetration is isolated applies only to penetration flow paths isolated to restore leakage within limits.

l FERMI UNIT 2 Page B 3.6 22(1) (INSERT) Revision 13. 08/06/99l l

PCIVs B 3.6.1.3

>t Insert B 3.6.1.3 5b l

. . For MSIV leakage, an 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Completion Time is allowed.

The Completion Time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months for MSIV leakage allows a period of time to restore the MSIVs to OPERABLE status given the fact that the MSIV closure will result in isolation of the main stea;n line(s) and potential for plant shutdown.

The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Completion Time for purge valve leakage is acceptable con:idaririg thE puise valves remain ci6 sed so A that a gross breach of the containment does not exist. The d 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time for leakage on hydrostatically D tested lines on a closed system is acceptable based on the available water seal expected to remain as a gaseous fission product boundary during the accident and the associated closed system. The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time for EFCV leakage v is acceptable based on the instrument and small pipe diameter of the penetration (hence, reliability) to act as a penetration isolation boundary.

I, FERMI UNIT - 2 Page B 3.6 22(2) (INSERT) Revision 13 08/06/99

DISCUSSION OF CHANGES ITS: SECTION 3.6.1.7 REACTOR BUILDING TO SUPPRESSION CHAl6ER VACUUM BREAKERS ADMINISTRATIVE A.1 In the conversion of the Fermi 2 current Technical Specifications (CTS) to the proposed plant specific Improved Technical Specifications (ITS), certain wording preferences or conventions are adopted which do not result in technical changes (either actual or interpretational). Editorial changes, reformatting, and revised numbering are adopted to make the ITS consistent with the Boiling Water Reactor (BWR) Standard Technical Specifications l NUREG 1433 Rev. 1.

A.2 CTS 3.6.4.2 details that reactor building to suppression chamber vacuum breakers are to be~~0p5Fibli and closed." ITS LC0 3.6.1.7 simply requires Operability. Since Operability addresses the entire safety related function, and is adequately detailed in the Bases, elimination of the explicit statement to be closed (as a subset of Operable) is administrative, with no impact on safety.

A.3 Not used. lS Tv A.4 Not used. l8 A.5 Not used.

Qv A.6 Not used. l9 E

l FERMI UNIT 2 1 REVISION 13. 08/06/99l

Suppression Chamber to Drywell Vacuum Breakers 3.6.1.8 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.8.1 --- - --

- NOTES - - -- - --

1. Not required to be met for vacuum breakers that are open during Surve111ances.

2. Not required to be met for vacuum breakers open when performing their intended function.

Verify each vacuum breaker is closed. 7 days l SR 3.6.1.8.2 Perform a functional test of each vacuum Prior to breaker, entering MODE 2 or 3 from MODE 4 if not performed in the previous 92 days 6.NQ Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after any discharge of steam to the suppression chamber from the SRVs SR 3.6.1.8.3 Verify the opening setpoint of each 18 months vacuum breaker is s 0.5 psid.

l FERMI - UNIT 2 3.6-21 Revision 13 08/06/99

l i

Suppression Chamber-to-Drywell Vacuum Breakers B 3.6.1.8 BASES APPLICABILITY (continued)

O breakers, therefore, are required to be OPERABLE in MODES 1.

l

-l 2, and 3. when the Suppression Pool Spray System is required I

to be OPERABLE, to mitigate the effects of inadvertent actuation of the Suppression Pool Spray System.

Also, in MODES 1, 2, and 3, a DBA could result in excessive negative differential pressure across the drywell wall, caused by the rapid depressurization of the drywell. The event that results in the limiting rapid de)ressurization of the drywell is the primary system rupture t1at purges the drywell of air and fills the drywell free airspace with steam. Subsequent condensation of the steam would result in i depressurization of the drywell. The limiting pressure and temperature of the primary system prior to a DBA occur in MODES 1, 2, and 3.

In MODES 4 and 5 the probability and consequences of these events are reduced by the pressure and temperature limitations in these MODES: therefore, maintaining suppression chamber to.drywell vacuum breakers OPERABLE is not required in MODE 4 or 5.

ACTIONS /L1 gl With one of the vacuum breakers inoperable for opening (e.g., the vacuum breaker is not open and may be stuck closed or not within its opening setpoint limit, so that it would not function as designed during an event that depressurized the drywell), the remaining eleven OPERABLE vacuum breakers are capable of providing the vacuum relief function. However, overall system reliability is reduced, and since normal periodic functional testing of the vacuum breakers is deferred to MODE 4 (SR 3.6.1.8.2), additional undetected failures could result in an excessive suppression chamber to drywell differential pressure during a DBA.

Therefore, with one vacuum breaker inoperable, 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is allowed to restore the inoperable vacuum breaker to OPERABLE status. The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time is considered i acceptable due to the low probability of an event in which i the remaining vacuum breaker capability would not be adequate.

l h

l FERMI UNIT 2 B 3.6.1.8- 3 Revision 13. 08/06/99 i

l l

I l

Suppression Chamber to Drywell Vacuum Breakers B 3.6.1.8 BASES 1

SURVEILLANCE REQUIREMENTS (continued)

If position indication appears reliable (dual or open indication while torus to drywell differential pressure is steady at 0 psid). and indicates open, the alternate methods outlined in the TRM can prove the indication to be in error and the vacuum breaker closed. However. in this case the vacuum breaker is assumed open until otherwise proved to i satisfy the leakage test, and this confirmation must be l performed within the Technical Specification 3.6.1.8 l Required Action B.1 Completion Time of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . The 7 day Frequency is based on engineering judgment, is considered adequate in view of other indications of vacuum breaker status available to operations personnel, and has been shown i to be acceptable through operating experience.

{

Notes 1 and 2 are added to this SR which allows suppression chamber to drywell vacuum breakers opened in conjunction !

with the performance of a Surveillance or open while performing their intended function to not be considered as '

failing this SR. These periods do not represent inoperable vacuum breakers.

SR 3.6.1.8.2

! Each vacuum breaker must be cycled to ensure that it opens adequately to perform its design function and returns to the fully closed position. This ensures that the safety analysis assumptions are valid. The Frequency of " prior to entering MODE 2 cr 3 from MODE 4 if not performed in the previous 92 days" is based u)on the demonstrated reliability of the vacuum breakers and t1e pote?ial for the test to result in a stuck open vacuum breaker, which could be caused by a failure of the pneumatically operated test mechanism.

Since the vacuum breaker is inaccessible in MODES 1, 2. and

3. test induced inoperability would result in a forced shutdown of the unit. In addition, there exists substantial redundancy in that 4 vacuum breakers must fail to o>en before the safety function is lost. In addition, t11s functional test is required within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after a discharge of steam to the suppression chamber from the safety / relief valves, j FERMI - UNIT 2 B 3.6.1.8 - 5 Revision 13 08/06/99

l 1

(

Suppression Chamber-to-Drywell Vacuum Breakers 3.6.1.8 2 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.1.8.2 Perform m ;::.;w,a vacuum functional test of each M {t).L,.'4.\.h.l h breaker.

ale

[t-forkM/ erin 3 M o06 2 or3 Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

& M 0 0 6 Li lf no+ Pb MA after any t l'n % pruiov5 72.{aV5 J $5$8,]*the

( suppression chamber from OP. I ..

._ ,Rvs l

l thin 12 ours ollowi an operat n that caus any o the acuum br kers o n

l k

SR 3.6.1.8.3 Verifv the opening setpoint of each I g8gnonths f;r'r;Gvacuumbreakerissg.5gpsid.

l l

l 7"I/4 5 3.6-28 A*4-h s*/vify5 l L .

Suppression Chamber-to-Drywell Vacuum Breakers B 3.6.1.8

,- BASES (continued)

S

}

ACTIONS L1 With one of thefsecr17e'dlvacuum breakers inoperable for opening (e.g., the vacuum breaker is not open and may be l

Of.'l _ --

~

stuck closed or not within its opening setpoint limit, so Mg ] that it would not function as designed during an event that e gUC depressurized the drywell), the remaining {.a.1 htf0PERABLE 9

> M { r,in.C6 # .L {esgd vacuum breakers are capable of providing the vacuum relief E

p.c60p rt "# prs / function. However, overall system reliability is reduced h 22M S fi T ? #311270 '" M ?f th ~2 ITU feriD e'#

4e red to tAODc S:9-fould result in an excessive suppression chamber-AAAN,U to-drywell differential pressure during a DBA. Therefore,

[of6 d*be g .6

g, g.%

3 with one of th [n h G 7:7'i-i vacuum breakerf inoperable, 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is allowed to restore :t h .;; r : .. the

@Hgedf g [gc, inoperable vacuum breakerg to OPERABLE status,:: t h ; p h .;

p r Mitt- : r: --- 4:* rt .;ith ti.;. .... ; O. 1 A n yn h:1: r:b:S. The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time is considered ;

acceptable due to the low probability of an event in which 1 the remaining vacuum breaker capability would not be I adequate.

l l

L1 An open vacuum breaker: allows communication between the drywell and suppressiori: chamber airspace, and, as a result,

,T ( p.firmafim o f ur there is the potential!for suppression chamber N overpressurization due.'to this bypass leakage if a LOCA were 1 i

CIM'd.sfalvs /ould t to occur. Therefore,gthe open vacuum breaker must be m

/ Q(ows pyocedwas a5 closed / * ' ^ is allowed to close the vacuum breaker M l due to the low probabi'ity of. an event gM in fu. 82.seJ primary containment. f f v. uum breaker position indication i that would oressurire .

.{pr .5/' 3. (../ K'< /)

1s nat4eliab 6 an alt ate method of ye'ifying r tha tfie y vp fun bre rs are e sed is to veri ( that a dif

),N2hosCCWfrfiel f f[0.5) d between the ppression cha er. an tial N. f ,4ressur dryw is maint ned for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months r35 fired 2 hou completion Ti thout makeup /. The is considereradeau to j) g qs.l .[Qerform this est.

C.1 and C l If the inoperable suppression chamber-to-drywell vacuum breaker cannot be closed or restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To (continued)

'*Rji STF B 3.6-51 "r; ;, 04/07/n -

h(3

/fevf

i Suppression Chamber-to-Drywell Vacuum Breakers B 3.6.1.8 BASES ACTIONS C.1 and C.2 (continued) I achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and to MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

I SURVE!LLANCE SR 3.5.1.8.1 REQUIREMENTS Each vacuum breaker is verified closed to ensure that this [',

potential large bypass leakage path is not present. This ,

survelliance is performed by observing the vacuum breaker L.,t position indication or by verifying that a differential g 3,s. ,g.

pressure of nd drywell is m$0.S$ psid aintained forbetween the suppression I hour without makeup. ch e day Frequency is based on engineering judgment, is considered adequate in view of other indications of vacuum breaker pl status available to operations personnel, and has been shown to be 1 - <a...~

acceptable

-- <. .uthrough operati.n.g

..a - u. experience.x '. ". :.... \

l Wes; sar t FREE ~EEEE9EG'sE97E M'T PJ rewvew g ,_,,,,,,,,.-onny

% ,,6. ,6 .-..,u,.,anet,emer

a. u. ---......e.; .. a

.. "--' h ;. '".; 7;.:d.

dd_

2 _ts,:I and to

_. added & this arc.SR which allows suppression chamber-to-drywell vacuum breakers opened in conjunction with the p,3 performance of a Surveillance not be considered as failing this SR. These periods ' :;n'r., vec. . in .L . .

tre ;ntr:1hd h p1=t ;rr rn and_do not represent , -

inoperable vacuum breakers.

h AQ deQ SR thyr ioWded FuqCAion 3.6.1.B.2 }

Each{riantsETvacuum breaker must be cycled to ensure that P3 it opens adequately to perform its design function and l

~ returns to the fully closed position. This ensures that the safety analysis assumptions are valid. Tr;;;=:y

. Tn$ed '

--+ gy,..e!".jn;*vdr;;1:;;d, );..{... :... .J?e

d.,

1p.jemngrvur.m S J.C. I,6-2. --,---.6. ni >- .-- -.-- .. -. 6 - i .r -

9Mitys. * ? ! I;, T. . ......, .. i_ ... t; ;n ;'. d; (continued)

SWR /4 STS - B 3.6-52 5-" ?, ^'/07/95-l L.

l SCIVs i

B 3.6.4.2 BASES APPLICABLE SAFETY ANALYSES (continued) containment performs no active function in response to

.either of these limiting events, but the boundary established by SCIVs is required to ensure that leakage from the primary containment is processed by the Standby Gas Treatment (SGT) System before being released to the environment.

Maintaining SCIVs OPERABLE with isolation times within limits ensures that fission products will remain trapped inside secondary containment so that they can be treated by the SGT System prior to discharge to the environment.

SCIVs satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

n LC0 SCIVs form a part of the secondary containment boundary, b The SCIV safety function is related to control of offsite radiation releases resulting from DBAs.

yl The power operated automatic isolation valves are considered OPERABLE when their isolation times are within limits and the valves actuate on an automatic isolation signal. The valves covered by this LCO. along with their associated stroke times are listed in Reference 3.

O The normally closed isolation valves or blind flanges are d considered OPERABLE when manual valves and blind flanges are closed, or open in accordance with appropriate

{ administrative controls. These passive isolation valves or devices are listed in plant procedures.

t.PPLICABILITY In MODES 1, 2. and 3. a DBA could lead to a fission produrt release to the primary containment that leaks to the secondary containment. Therefore, the OPERABILITY of SCIVs is required.

In MODES 4 and 5. the probability and consequences of inese events are reduced due to pressure and temperaturc limitations in these MODES. Therefore. maintaising SCIVs OPERABLE is not required in MODE 4 or 5. except for other l FERMI - UNIT 2 B 3.6.4.2-2 Revision 13. 08/06/99

SCIVs B 3.6.4.2 BASES APPLICABLE SAFETY ANALYSES estabitshed by SCIVs is required to ensure that leakage from the primary containment is processed by the Standby Gas (continued) Treatment (SGT) System before being released to the environment.

Maintaining SCIVs OPERABLE with isolation times within !

t limits ensures tha fission products will remain trapped I inside'ffeconda containment so that they can be treated by the SGT System prior to discharge to the environment.

[.b -

SCIVs satisfy Criterion 0 3 of t%=C Telicy 4tet. d. t

. gpi -@( ,

LCO SCIVs form a part of the y' secondary)' containment boundary.

The SCIV safety function is related to control of offsite radiation releases resultino f s.

Mba The power operated soation'vaivesareconsideredOPERABLEl when their isolation times are within limits and the valves actuate on an automatic isolation signal. The valves covered by this LCO, along with their associated stroke _

times, are listed in Reference 3.

gggg fgg h

~

The normally closed isolation valves or blind flan s are

.g considered OPERABLE when manual valveshre clos 'd open in h

accordance wi,th appropriate administrative cont _,

f. .n . .... ...

s . .-

k bpositich.andblindflanos are in ol ace M. - m n m i,hese passive isolation valves or devices are listed in Re'..- l l# $ Y ?

APPLICABILITY In MODES 1, 2, and 3, a DBA could lead to a fission product release to the primary containment that leaks to the

$econdary)< containment. Therefore, the OPERASILITY of SCIVs is required.

In MODES 4 and 5, 'the probability and consequences of these events are reduced due to pressure and temperature' limitations in these MODES. Therefore, maintaining SCIVs OPERABLE is not required in MODE 4 or 5, except for other situations under which significant radioactive releases can be postulated, such as during operations with a potential ,

for draining the reactor vessel (OPDRVs), during CORE '

(continued)

Lap his '

B 3.6-103 Rev4;-04/07/95

~

Rcv a UI Ren y Ret / [ I

L

! JUSTIFICATION FOR DIFFERENCES FROM NUREG 1433 ITS: SECTION 3.6.4.2 SCIVs

-r)

NON BRACKETED PLANT SPECIFIC CHANGES l P.1 Not used.

L P.2 Not used.

l

! P.3 Not used.

I P.4' The Bases Background reference to DBAs is general - not specific only to LOCAs, which is what Reference 1 entails. Eliminating reference to the UFSAR Section for LOCA safety analyses will have l no impact on the Bases content or understanding.

l P.5 The Bases discussion of normally closed SCIVs is modified. This i editorial preference is based on incomplete and misleading discussion of these valves. This does not modify the requirements k or the interpretation of those requirements. j i The LCO Bases are intended to provide a few details regarding O l

! OPERABILITY. These details are not all encompassing, but only 1 4 m serve to outline salient features (in this case of SCIV i i OPERABILITY). The paragraph discussing normally closed SCIVs l

attempts to define OPERABILITY of these valves, but makes I l statements that are not true in all cases. The Fermi Unit 2 ' ,

I design does not have lines with normally closed SCIVs where the !

SCIV is an automatic isolation valve (and if a normally open valve was closed, it is NOT required to be " deactivated and secured in their closed position" to be OPERABLE).

l l P.6 The reference to the NRC Policy Statement has been replaced'with a more appropriate reference to the Improved Technical Specification

" split" criteria found in 10 CFR 50.36(c)(2)(ii).

e-FERMI UNIT 2 1 REVISION 13, 08/06/99l 1-

l INSERT TIIIS PAGE IN FRONT OF YOL,UME 8 '

Volume 8 SECTION 3.7 Remove Replace 3.7.1 CTS M/U (3/4 7-1) pg i of 2 3.7.1 CTS M/U (3/4 7-1) pg 1 of 2 Rev 13 3.7.1 DOCS pg 1 Rev 0 3.7.1 DOCS pg i Rev 13 3.7.1 DOCS pg 2 Rev 0 3.7.1 DOCS pg 2 Rev 13 3.7.2 CTS M/U (3/4 7-5) pg 4 of 9 3.7.2 CTS M/U (3/4 7-5) pg 4 of 9 Rev 13 3.7.2 DOCS pg 6 Rev 0 3.7.2 DOCS pg 6 Rev 13 3.7.4 ITS pg 3.7-13 Rev 0 3.7.4 ITS pg 3.7-13 Rev 13 B 3.7.4 ITS pg B 3.7.4-5 Rev 0 B 3.7.4 ITS pg B 3.7.4-5 Rev i3 3.7.4 CTS M/U (3/4 7-9) pg 3 of 3 3.7.4 CTS M/U (3/4 7-9) pg 3 of 3 Rev 13 3.7.4 DOCS pg 2 Rev 0 3.7.4 DOCS pg 2 Rev 13 3.7.4 NUREG M/U pg 3.7-15 3.7.4 NUREG M/U pg 3.7-15 Rev 13 B 3.7.4 NUREG M/U pg B 3.7-29 B 3.7.4 NUREG M/U pg B 3.7-29 Rev 13 1

Rev 13 08/06/99 l

l .

i speci,%6on 2 7.I I~ 3/4.7 PLANT SYSTEMS

\ -

3/4.7.1 SERVICE WATER SYSTEMS RESIDUAL HEAT REMOVAL SERVICE WATER SYSTEM [.l

[fMITING CONDTTTON FOR OpFpATION sr7-rl-t At least the followingb::19%esidual heat removal _ service LCOl 3, /, water (RHRSW) system subsystems.ynn eacn sutisysteltfTb1fipriseo of:

a. Two OPERABLE JtHR$W pumps', and / LA'l

b. A PERABLFMlow patif capable,cf taking su tion fro/the /

associated ultimate heat sink and transf' erring tJfe waterArough one RHT heat eichan shall be OPERABLE:

a. In OPERATIONAL CONDITIONS 1, 2, and 3. two subsy, stems.

rb. Jn'vFE NATCONDI IDN5 4 an .5, the subsystem {s ssociated ems andj d onentsdeouired JPERABLE / ,3 t [ withSee fications 4.4.9.1, M 4.9.2. 3.9".11.1. d 3.9.fl._2 APPlfCABILITY: OPERATIONAL CONDITIONS 1, 2, 3, (', 2nd E ,3 ACTION:

a. In OPERATIONAL CONDITION 1, 2, or 3:

With one RHRSW pump inoperable, restore the inoperable pump gg/ h:'

to OPERABLE status within 30 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN A c.7/oa E within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

~

Qa8 h- With one RHRSW pump in each subsystem inoperable, restore at least one inoperable pump to OPERABLE status within 7 days j ypfJg or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

7C/ay5 L,f M With one RHR$W subsystem otherwis inoperable, restore the

/ Cb/J C inoperable suosystem to OPERABL status with at least one

~

OPERABLE pump within 02 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in at least HOT SHUTDOWN C.[lOAl 6 within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the followin'g 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , gggp g With both RHRSW subsystems otherwise inoperable, restore at least one subsystem to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in C/M COLD SHUTDOWN

within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ,

b. In OPERATIONAL CONDITION 3 h with the RHRSW subsystem (s), whichj

~

s [c7/Orj(shjp is associ ed with an RHR loop require'd OPERABLE by Specification A2 3.4.9.1 o .Lp inoperable, declare the associated RHR loop / /P .) -

NOI6 inoperable and take the _ ACTION required by Specification 3.4.9.1 j-

@ s-e pp+*a t>l e). t k

M.I

Whene r Doth RSW s stems tr/inopera , if unabl o attain LD 5 Na ecuir y this) (TION. m ~ ain react coolant ;perature as as attica y use oValternate eat removal ethods.

FERMI - UNIT 2 3/4 7-1

,A

\

PAGE l 0F 02 Ral l3

1

)

DISCUSSION OF CHANGES

)

ITS: SECTION 3.7.1 - RESIDUAL HEAT REMOVAL SERVICE WATER (RHRSW) SYSTEM I ADMINISTRATIVE A.1 In the conversion of the Fermi 2 current Technical Specifications (CTS) to the proposed plant specific Improved Technical Specifications (ITS). certain wording preferences or conventions are adopted which do not result in technical changes (either actual or interpretational). Editorial changes, ' eformatting, r and revised numbering are adopted to make the ITS consistent with the Boiling Water Reactor (BWR) Standard Technical Specifications NUREG-1433, Rev. 1.

g A.2 Not used.

l5-A.3 CTS 3.7.1.1 LCO, Applicability, Action b, and Action c require the RHRSW System to be Operable in non operating Modes and require RHR SDC be declared inoperable when RHRSW is inoperable. ITS 3.7.1 only requires that the RHRSW System be Operable in Modes 1, 2, and 3. This is acceptable because in Modes 4 and 5 the RHRSW System is a support system for RHR SDC. The ITS definition of Operability of RHR SDC will require the Operability of the appropriate support systems, including RHRSW components needed for RHR to perform its function. Therefore. these requirements are adequately defined in the ITS definition of Operability and the explicit requirements are not necessary. The CTS Actions for inoperability of these systems in these shutdown Modes, each direct that the supported equipment be declared inoperable and the applicable Actions for the inoperable supported systems be taken.

These CTS Actions are consistent with allowing application of the definition of Operability to apply. Therefore, this change introduces no technical change, and is considered administrative. l This change results in a presentation of the RHRSW system '

requirements that is consistent with NUREG-1433.

A.4 CTS 4.7.1.1 requires a monthly RHRSW System valve lineup to " verify each valve is in its correct position." ITS clarifies this requirement by adding "or can be aligned to the correct position."

Since RHRSW is a manually actuated and operated system, " correct l

position" can be misleading. The intent of " correct" is to allow for '

certain manual valve realignments that may be required on manual initiation, to be acceptable configurations for standby Operability. ,

This intent is supported by the NUREG 1433 clarifying wording.

Therefore this clarification is considered administrative only.

i FERMI UNIT 2 1 REVISION 13, 08/06/99l

DISCUSSION OF CHANGES ITS: SECTION 3.7.1 - RESIDUAL HEAT REMOVAL SERVICE WATER (RHRSW) SY TECHNICAL CHANGES - MORE RESTRICTIVE M.1 CTS 3.7.1.1 footnote "*". is deleted in ITS 3.7.1. The CTS footnote provides unnecessary duplication of the Actions required I by ITS 3.4.8. "RHR Shutdown Cooling System - Hot Shutdown."

l contains no additional restrictions on the operation of the plant, and could be interpreted as a relaxation of the requirements to {

achieve Mode 4 (Cold Shutdown). The ITS 3.4.8 Action to be in T l Mode 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months adequately prescribes tne requirement to y l

make efforts to " maintain reactor coolant temperature as low as g practical ." If conditions are such that Mode 4 cannot be attained, the Actions remain in effect, requiring that efforts to reach Mode 4 continue. Elimination of the potential relaxation is a more restrictive change with no impact on safety.

l TECHNICAL CHANGES LESS RESTRICTIVE

" Generic" LA.1 CTS 3.7.1.1 requires that two RHRSW subsystems be Operable and I defines what equipment comprises an RHRSW subsystem. ITS 3.7.1 requires that two RHRSW subsystems are Operable, but does not identify what equipment is required for an RHRSW subsystem: these details are relocated to the Bases of ITS 3.7.1. This is acceptable because the identification of the specific equipment does not impact the ITS requirement to maintain two RHRSW subsystems Operable. Therefore, this information can be adequately defined and controlled in the ITS Bases which requires change control ir eaccordance with ITS 5.5.10. Bases Control Program. These details are not required to be in the ITS to provide adequate protection of the public health and safety because these details do not impact the requirement to maintain the equipment Operable.

l FERMI - UNIT 2 2 REVISION 13. 08/06/99l l

Specification 3.7.2

< Page Removed in Rev 13 >

l 1

l l

i 1

W*

3jy 7-5 PAGE Y OF 09 I?W )3

DISCUSSION OF CHANGES ITS: SECTION 3.7.2 EECW / EESW SYSTEM AND UHS y

TECHNICAL CHANGES - LESS RESTRICTIVE

" Specific" L.1 CTS 3.8.2.1, DC Sources. Actions for inoperable battery chargers are modified by a footnote allowing a 16 hour1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months delay "for battery chargers made inopa able due to loss of EECW cooling." Similarly, CTS 3.8.3.1, Onsite Power Distribution Systems, for inoperable distribution busses are modified by a footnote allowing a 16 hour1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months delay "for AC distribution system components made inoperable due to loss of EECW cooling." The ITS, in accordance with LC0 3.0.6, allows the Actions of EECW to suffice when the sole cause of the supported systems inoperability is the inoperability of the EECW system: without invoking the Actions for the inoperable battery charger, or inoperable power distribution systems. This provides a restoration time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months for loss of EECW cooling to one division of battery chargers, and one division of AC power ,

distribution. This extension to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is acceptable based on the following factors:

a. Normally operating non safety cooling is also provided to I 3 maintain the battery and battery chargers, and AC distribution system, functional.

b. Additional alternate cooling mechanisms may be able to be placed in service in the unlikely event of loss of normal cooling in conjunction with loss of EECW.

c. The battery and the AC distribution system remain capable of providing their design support function for some period following loss of area cooling.

1

d. Should the entire DC division or AC distribution division fail, the remaining DC division and AC distribution division are sufficient to support the minimum required capabilities ,

following an accident.

Based on the above, the extension of the restoration time to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months does not represent a significant impact on safety.

FERMI UNIT 2 6 REVISION 13. 08/06/99l

Control Center AC System 3.7.4 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME E. Two control center AC - ----- - -NOTE----- --- -

subsystems inoperable LC0 3.0.3 is not applicable.

during movement of --- - - -- --- ------ - -

irradiated fuel assemblies in the E.1 Suspend movement of Immediately secondary containment. irradiated fuel ,

during CORE assemblies in the ALTERATIONS. or during secondary OPDRVs. containment.

M E.2 Suspend CORE Immediately ALTERATIONS.

M E.3 Initiate actions to Immediately suspend OPDRVs.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY a SR 3.7.4.1 Verify the control room air temperature is 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months tr s 95'F.

= -

l l FERMI UNIT 2 3.7 13 Revision 13 08/06/99

Control Center AC System B 3.7.4 l

i BASES ACTIONS (continued)

E.1. E.2. and E.3 l

The Required Actions of Condition E are modified by a Note indicating that LC0 3.0.3 does not apply. If moving irradiated fuel assemblies while.in MODE 1, 2, or 3, the fuel movement is independent of reactor operations.

Therefore, inability to suspend movement of irradiated fuel assemblies is not a sufficient reason to require a reactor shutdown.

During movement of irradiated fuel assemblies in the secondary containment, during CORE ALTERATIONS, or during OPDRVs. with two control center AC subsystems inoperable, action must be taken immediately to suspend activities that present a potential for releasing radioactivity that might require isolation of the control room. This places the unit in a condition that minimizes risk.

If applicable, CORE ALTERATIONS and handling of irradiated fuel in the secondary containment must be suspended immediately. Suspension of these activities shall not preclude completion of movement of a component to a safe position. Also, if applicable, actions must be initiated immediately to suspend OPDRVs to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until the OPDRVs are suspended.

SURVEILLANCE SR 3.7.4.1 REQUIREMENTS This SR verifies that the heat removal capability of the system is sufficient to remove the control room heat load.

[ The SR consis+.s of a verification of the control room temperature. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months F.equency is appropriate since significant degradation of the Control Center AC System is not expected over this time period.

l REFERENCES 1. UFSAR, Section 6.4.

2. UFSAR Section 9.4.1.

l l

l FERMI - UNIT 2 B 3. 7.4 -5 Revision 13. 08/06/99 '

kC IP~ICA77bN 3.7.Ll

, CMo see spe44cah 3,7.'3 )

PLANT SYSTEMS h/so ss.r Specjg ca % g,5, ~j )

$URVEILLANCE REOUIREMENTS 4.7.2.1 The control room emergency filtration systes shall be demonstrated OPERABLE:

a. At least onceisper temperature less12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months than by verifyig'that or equal to F. the control room air

b. fat least once per 31 days by:

1

1. Initiating fan operation from the control room with each subsystem, establishing flow for at least 15 minutes through g the HEPA filters and charcoal adsorbers.

Spec;ficaNm 2. Verifying flow through the HEPA filters and charcoal adsorbers for at least 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months with the associated emergency makeup 3'7*3 inlet air heater OPERA 8LE. The subsystem used to estabitsh the 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months of flow through the HEPA filters and charcoal adsorbers shall be staggered such that each subsystem is utilized at least once per 62 days.

c.fAtleastonceper18monthsor(1)afteranystructuralmaintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire, or chemical release in any venttiation zone communicating with the system by:

1. Verifying that the system satisfies the in place penetration testing acceptance criteria of less than 1.0% and uses the test procedure guidance in Regulatory Positions C.5.a. C.5.c and

'{ C.S.d of Regulatory Guide 1.52, Revision 2, March 1978, while

' operating the. systes at a flow rate of 1800 cfm a 105 through the makeup filter and 3000 cfm a 105 through the recirculation filter.

2. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in M- accordance with Regulatory Position C.6.b of Regulatory Guide qNMg 1.52, Revision 2. March 1978, shows a methyl iodide penetration of less than 1.0% when tested at a temperature of 30*C and at a g$,~) relative humidity of 705 in accordance with ASTM D38031989 g with a 2 inch bed for the emergency makeup filter train; and a g 4 inch bed for the emergency recirculation air filter train. ,

3. Verifying a systes flow rate of 3000 cfm i 10% during systes operation when tested in accordan:a with ANSI H510 1980.

d. After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.5.b of Regulatory Guide 1.52 Revision 2, March 1978, shows a methyl iodide penetration of less than 1.0% when tested at a ?

temperature of 30*C and at a relative hustdtty of 70% in accordance with ASTM 03803-1989 with a 2 inch bed for the emergency makeup air filter train; and a 4 inch bed for the emergency recirculation air (filtertrain.

u FERMI - UNIT 2 3/4 7-9 Amendment No. EJ. EE, 110 PAGE b _

0F 03 gev f3

l DISCUSSION OF CHANGES ITS: 'SECTION 3.7.4 - CONTROL CENTER AIR CONDITIONING (AC) SYSTEM i ,-.,

TECHNICAL CHANGES - MORE RESTRICTIVE M.1 Not used.

i l

TECHNICAL CHANGES - LESS RESTRICTIVE

Generic" l LA.1 Not used.

c r-FERMI - UNIT 2 2 REVISION 13. 08/06/99l L

1 i

$tontrol com- System (e.,% c 3.7

's ACTIONS (continued)

J CONDITION REQUIRED ACTION COMPLETION TIME E. TwoAcontro C$r NOTE

<g,7,zAcWd subsystems inoperable LCO 3.0.3 is not applicable.

during movement of irradiated fuel assemblies in the E.1 Suspend movement of Imediately

(secondary}c irradiated fuel containment, during CORE ALTERATIONS, or assemblies in the fsecondary}c h 787 M ** C' during OPDRVs. containment.

M E.2 Suspend CORE Imediately ALTERATIONS.

3.7.2. Actk C.

M E.3 Initiate actions to Immediately suspend OPDRVs. / N Q3,7 2. Ac to,on C. Ej l

~ 1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.h.I' Verify 29 # ::tr;l .. x f 7; L;y;ta in t ' - - b ' 's 'oy- - ^ -ww- -s susw -'

- - -y-s 'b s'es m a a nse,= w" .'.w==

I.S' - the '

k m endrat aw a;< k.epnatwe i2. hows is 4 95'f. -

~- -

s.;R/; STS - 3.7-15 -R=, ;, 0?/^7/o5 ea

r OscW jControl4een.AC) B3.7%System [

I

~; aASES I ACTIONS E.1. E.2. and E.3 (continued)

I require isolation of the control room. This places the unit in a condition that minimizes risk.

If applicable, CORE ALTERATIONS and handling of irradiated fuel in the } secondary) containment must be suspended immediately. Suspension of these activities shall not preclude completion of movement of a component to a safe position. Also, if applicable, actions must be initiated )

, innediately to suspend OPDRVs to minimize the probability of i l

a vessel draindown and subsequent potential for fission I

I product release. Actions must continue until the OPDRVs are suspended, SURVEILLANCE SR 3.7 1 1 b i l )

REQUIREMENTS '

This SR verifies that the heat removal capability of the veti[rcaflexed system is sufficient to remove the control room heat lead , ,

-fW. colhel (tbn4

,bZ.. .'.I."._. 6n; p7..;...{, yg7..(The SR, consist,s of a p

44wr</db N I Frequencyisappropriatesincesignificantdeg..dationof ra l gg the peri (od.ControlRpen.AC{Systemisnotexpectedoverthistime s' -

RGnN )

REFERENCES 1. uFSAR,Section(6.43 X W SA Ai S ec fien 9. 'l . l l

l l

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""/i ST! B 3.7-29 4ev 1, 04/07/05--

Rev 0 L

INSERT THIS PAGE IN FRONT OF VOLUME 10 l

Volume 10 SECTION 3.9,3.10,4.0 & 5.0 Remove Replace

3.9.8 ITS pg 3.9-12 Rev 4 3.9.8 ITS pg 3.912 Rev 13 3.9.8 ITS pg 3.9-14 Rev 0 3.9.8 ITS pg 3.9-14 Rev 13 l 3.9.8 NUREG M/U pg 3.9-14 Rev 4 3.9.8 NUREG M/U pg 3.9-14 Rev 13 l 3.9.8 NUREG M/U pg 3.9-16 3.9.8 NUREG M/U pg 3.916 Rev 13 1

3.10.5 ITS pg 3.10-13 Rev 0 3.10.5 ITS pg 3.10-13 Rev 13 l 3.10.7 ITS pg 3.10-18 Rev 0 3.10.7 ITS pg 3.10-18 Rev 13 3.10.7 ITS pg 3.10-20 Rev 0 3.10.7 ITS pg 3.10-20 Rev 13 I B 3.10.7 ITS pg B 3.10.7-5 Rev 0 B 3.10.7 ITS pg B 3.10.7-5 Rev 13 3.10.7 NUREG M/U pg 3.10 20 3.10.7 NUREG M/U pg 3.10-20 Rev 13 3.10.7 NUREG M/U pg 3.10-22 3.10.7 NUREG M/U pg 3.10-22 Rev 13 B 3.10.7 NUREG M/U pg B 3.10-37 B 3.10.7 NUREG M/U pg B 3.10-37 Rev 13 l 5.2 ITS pg 5.0-3 Rev 0 5.2 ITS pg 5.0-3 Rev 13 5.2 CTS M/U (6-5) pg 3 of 5 5.2 CTS M/U (6-5) pg 3 of 5 Rev 13 j 5.2 DOCS pg 2 Rev 0 5.2 DOCS pg 2 Rev 13 5.2 NUREG M/U pg 5.0-2 (Insert) 5.2 NUREG M/Uyg 5.0-2 (Insert) Rev 13 5.5 CTS M/U (3/4 6-1) pg 4 of 24 5.5 CTS M/U N/461)pg 4 of 24 Rev 13 5.5 CTS M/U (3/4 6-2) pg 5 of 24 5.5 CTS M/U (3/4 6-2) pg 5 of 24 Rev 13 1

5.5 CTS M/U (3/4 6-3) pg 6 of 24 5.5 CTS M/U (3/4 6-3) pg 6 of 24 Rev 13 l 5.5 CTS M/U (3/4 6-9) pg 8 of 24 5.5 CTS M/U (3/4 6-9) pg 8 of 24 Rev 13 5.5 CTS M/U (3/4 6-11) pg 9 of 24 5.5 CTS M/U (3/4 6-11) pg 9 of 24 Rev 13 l

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Rev 13 08/06/99

RHR-Low Water Level 3.9.8 3.9 REFUELING OPERATIONS 3.9.8 Residual Heat Removal (RHR)-Low Water Level h l LC0 3.9.8 Two RHR shutdown cooling subsystems shall be OPERABLE. and, with no recirculation pump in operation, one RHR shutdown cooling subsystem shall be in operation.

............................N0TES ----- - -- --- -- ------

1. The required operating RHR shutdown cooling subsystem may be removed from operation for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period.

2. One RHR shutdown cooling subsystem may be inoperable

..... .$.!S.. I. .!.S$ $$$.S . !!...!"!!!". >

9 APPLICABILITY: MODE 5 with irradiated fuel in the reactor pressure vessel (RPV). the water level < 20 ft 6 inches above the top of L

h the RPV flange, and heat losses to ambient not greater than or equal to heat input to reactor coolant.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or two required A.1 Verify an alternate I hour RHR shutdown cooling method of decay heat subsystems inoperable, removal is available AND for each inoperable required RHR shutdown Once per cooling subsystem. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter (continued) l FERMI UNIT 2 3.9-12 Revision 13. 08/06/99

RHR-Low Water Level 3.9.8 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.8.1 Verify one RHR shutdown cooling subsystem 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months l or recirculation pump is operating.

SR 3.9.8.2 Verify each RHR shutdown cooling subsystem 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is capable of decay heat removal.

1 9

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i l FERMI - UNIT 2 3.9-14 Revision 13. 08/06/99 I J

RHR-Low Water Level 3.9.)/

. 3.9 REFUELING OPERATIONS 3.9p Residual Heat Removal (RHR)-Low Water Level _

S , Wi4, no recirculation pump

_ in ee<^ab, LCO 3.9./

S Two one RHRRHR shutdown shutdown coolingshall cooling subsystem subsystems be in operation. snail De UPLKAB

_.._ NE -NOT --

h _. The

_____ .___ operating shutdown cooling subsystem may be----------------

required

_ removed from operation for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period.

( ~

D t UTE~5~dTonis

~ ~

9.ll Ik

+o was e m.Mol "Asak [so4nx,3, sYs lem 5~5~~'~n~o}~enob ll@

APPLICABILITY: MODE 5 with irradiated fuel in the reactor pressure vessel Si"((Aq (RPV (RPV ancthe 1ang water level <Q[22Ln 2o H.,(,above e n c. AEJB the top of the4m hee + loues h ambo ear not greew &n ACTIONS Gr Ig5*/ 5 heal moat & reack tooted ; $E' CONDITION REQUIRED ACTION COMPLETION TIME A. One or two required A.1 Verify an alternate I hour RHR shutdown cooling method of decay heat 3, 9 .11, 2 3 subsystems inoperable, removal is available AND Acliew a.

s for each inoperable required RHR shutdown Once per cooling subsystem. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter B. Required Action and 8.1 Initiate action to Inunediately associated Completion restore Time of Condition A containm)entto$econdaryy not met. OPERABLE status. DOC At)

AN.Q (continued) l BWR/4 STS 3.9-14 Rev 1, 04/07/95 ggi P. ;\

Raly [

i

RHR-Low Water Level l

3.9.)F (C

SURVEILLANCE REOUIREMENTS SURVEILLANCE FREQUENCY SR 3.9. 1 Verify one RHR shutdown cooling subsystem 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is operating. ( q,q,gt,1,g) g # recircatahm pm _- _-

3 & 3.4.S 2; ifenify each RItR shu+dovin Coolio,3 a han 0PI subsys4em is capebte of decay tua+ <wovat.

2 s

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l BWR/4 STS 3.9-16 Rev 1, 04/07/95 ReV 13

I Single CRD Removal-Refueling 3.10.5 3.10 SPECIAL OPERATIONS 3.10.5 Single Control Rod Drive (CRD) Removal-Refueling LC0 3.10.5 The requirements of LC0 3.3.1.1 " Reactor Protection System (RPS) Instrumentation": LC0 3.3.8.2 " Reactor Protection System (RPS) Electric Power Monitoring": and LC0 3.9.5.

" Control Rod OPERABILITY-Refueling," may be suspended in MODE 5 to allow withdrawal of a single control rod, and -

subsequent removal of the associated CRD from a core cell I containing one or more fuel assemblies, provided the following requirements are met:

I a. All other control rods are fully inserted: and l b. All other control rods in a five by five array centered on the withdrawn control rod are disarmed: at which time LCO 3.1.1, " SHUTDOWN MARGIN (SDM)," MODE 5 rec uirements may be changed to allow the single control roc withdrawn to be assumed to be the highest worth control rod.

ANQ In conjunction with a. and b. above, the requirements of l LC0 3.9.1, " Refueling Equipment Interlocks": LC0 3.9.2 I

" Refuel Position One Rod Out Interlock": and LC0 3.9.4

" Control Rod Position Indication": may be suspended provided the following requirements are met:

c. No other CORE ALTERATIONS are in progress: and l l d. A control rod withdrawal block is inserted.

APPLICABILITY: MODE 5 with LC0 3.9.5 not met.

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l FERMI UNIT 2 3.10 13 Revision 13 08/06/99 u

t SDM Test-Refueling l

3.10.7 l l

3.10 SPECIAL OPERATIONS 3.10.7 SHUTDOWN MARGIN (SDM) Test-Refueling LC0 3.10.7 The reactor mode switch position specified in Table 1.11 for MODE 5 may be changed to include the startup/ hot standby position, and operation considered not to be in MODE 2. to allow SDM testing, provided the following requirements are met:

a. LC0 3.3.1.1 " Reactor Protection System A Instrumentation." MODE 2 requirements for Functions 2.a.

F 2.d. and 2.e of Table 3.3.1.1-1:

b. 1. LCO 3.3.2.1 " Control Rod Block Instrumentation."

MODE 2 requirements for Function 2 of Table 3.3.2.1-1, with the prescribed withdrawal sequence requirements of SR 3.3.2.1.7 changed to require the control rod sequence to conform to the SDM test sequence.

M

2. Conformance to the approved control rod sequence for the SDM test is verified by a second licensed operator or other qualified member of the technical staff:

c. Each withdrawn control rod shall be coupled to the associated CRD:

d. All control rod withdrawals during local critical testing shall be made in notch out mode:

e. No other CORE ALTERATIONS are in progress; and ;

f. CRD charging water header pressure a 940 psig.

APPLICABILITY: MODE 5 with the reactor mode switch in startup/ hot standby '

position.

l FERMI UNIT 2 3.10-18 Revision 13. 08/06/99

SDM Test-Refueling l l

l 3.10.7 l

. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.10.7.1 Perform the MODE 2 applicable SRs for LC0 According to 3.3.1.1. Functions 2.a. 2.d. and 2.e of the applicable

@l Table 3.3.1.1-1. SRs l SR 3.10.7.2 ---- ---- --- -----NOTE-- ---- --- --- ---

l Not required to be met if SR 3.10.7.3 i satisfied.

Perform the MODE 2 applicable SRs for According to LC0 3.3.2.1. Function 2 of Table 3.3.2.1-1. the applicable SRs l

SR 3.10.7.3 -- ---- ...-

-- NOTE- -------- ---- - -

Not required to be met if SR 3.10.7.2 j satisfied.

Verify movement of control rods is in During control compliance with the approved control rod rod movement l

sequence for the SDM test by a second I

licensed o>erator or other qualified member of the tec1nical staff.

i l

SR 3.10.7.4 Verify no other CORE ALTERATIONS are in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months l progress. I (continued) l l FERMI UNIT 2 3.10 20 Revision 13. 08/06/99 i

L :

SDM Test-Refueling B 3.10.7 BASES ACTIONS (continued)

The allowed Completion Times are reasonable, considering the '

small number of allowed ino>erable control rods, and provide time to insert and disarm tie control rods in an orderly manner and without challenging plant systems.

Condition A is modified by a Note allowing separate Condition entry for each uncoupled control rod. This is acceptable since the Required Actions for this Condition provide appropriate compensatory actions for each uncoupled control rod. Complying with the Required Actions may allow for continued operation. Subsequent uncoupled control rods are governed by subsequent entry into the Condition and application of the Required Actions.

!L1 With one or more of the requirements of this LC0 not met for reasons other than an uncoupled control rod, the testing should be immediately stopped by placing the reactor mode switch in the shutdown or refuel position. This results in a condition that is consistent with the requirements for MODE 5 where the provisions of this Special Operations LC0 are no longer required.

SURVEILLANCE SR 3.10.7.1 REQUIREMENTS ,

Performance of the applicable SRs for LC0 3.3.1.1. Functions l 2.a. 2.d. and 2.e will ensure that the reactor is operated within the bounds of the safety analysis.

SR 3.10.7.1. SR 3.10.7.2. and SR 3.10.7.3 hl LC0 3.3.1.1. Functions 2.a. 2.d. and 2.e. made applicable in this Special Operations LC0. are required to have applicable Surveillances met to establish that this Special Operations l LC0 is being met. However, the control rod withdrawal sequences during the SDM tests may be enforced by the RWM ,

(LC0 3.3.2.1. Function 2. MODE 2 requirements) or by a i second licensed operator or other qualified member of the technical staff. As noted, either the applicable SRs for '

l the RWM (LC0 3.3.2.1) must be satisfied according to the applicable Frequencies (SR 3.10.7.2), or the proper movement of control rods must be verified (SR 3.10.7.3). This latter l FERMI - UNIT 2 B 3.10.7 -5 Revision 13. 08/06/99

SDM Test-Refueling 3.10 3.10 SPECIAL OPERATIONS

\'

j QCTS h 3.10 HUTDOWN MARGIN (SDM) Test-Refueling LCO 3.10 The reactor mode switch position specified in Table 1.1-1 for MODE 5 may be changed to include the startup/ hot s position, and operation considered not to be in MODE 2, to allow SDM testing, provided the following requirements are met:

a. LCO 3.3.1.1, " Reactor Protection System Instrumentation," MODE 2 requirements for Functions 2.a and 2.e of Table 3.3.1.1-1;

b. 1, LC0 3.3.2.1, " Control Rod Block Instrumentat

MODE 2 requirements for Function 2 of eresenbed f,Q Table sequence3.3.2.1-1, requirements ofwith the i=E:d SR 3.3.2.1. changedpe:iti...

to , withdrawa require the control rod sequence to enform to the SDM test sequence, 7

E

2. Conformance to the approved control rod sequence for the SDM test is verified by a second licensed operator or other qualified member of the technical 3ao.3.b) staff;

c. Each withdrawn control rod shall be coupled to the J.(.3,GD associated CRD; 4 f,gro4c.g

p co d.

criHed

All control rod withdrawalsTduring :t :f ::;;:n: (3.to.3.C.)

hl0 ymtr:1 nd sing shall be made in notch out mode; d e. No other CORE ALTERATIONS are in progress; and [3.1o.3, d) f.

CRDchargingwaterheaderpressure2p40ppsig. g g, APPLICABILITY: MODE 5 with the reactor mode switch in startup/ hot standby position. 3 to.h O'#/' STS- 3.10-20 ":V 1, 00/07/05 R6V13

i t SDM Test-Refueling l

3.10 l

< %. SURVEILLANCE REQUIREMENTS i

SURVEILLANCE FREQUENCY j SR 3.10 1 Perform the MODE 2 applicable SRs for LCO According to 3.3.1.1, Functions 2 2.dpof Table the applicable f 3.3.1.1-1. SRs SR 3.10. 2 --- ---

NOTE -- - ----

Not required to be met if SR 3.10.M.3 satisfied. .t0,3,b)

Perform the MODE 2 applicable SRs for According to LC0 3.3.2.1, Function 2 of Table 3.3.2.1-1. the applicable SRs SR 3.10 3 ----- --NOTE------- -----

Not' required to be met if SR 3.10.V.2 satisfied.

Verify movement of control rods is in During control

's compliance with the approved control rod rod movement sequence for the SOM test by a second licensed operator or other qualified member of the technical staff.

SR 3.10 .4 Verify no other CORE ALTERATIONS are in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months progress. \,fo,3.h

/

4 (continued)

BWR/4 STS 3.10-22 Rev 1, 04/07/95 i l l

F'

(

SDM Test-Refueling B3.10.K 7

., BASES (continued) i SURVEILLANCE SR 3.10. 2 1 3 Z.d, on REQUIREMENTS Perfo e f the applicable SRs for LCO 3.3.1.1, Functions I 2.a - will ensure that the reactor is operated within the M of the safety analysis. ii l SR 3.10 1. SR 3.10 2. and SR 3.10.[ 3 LC0 3.3.1.1, Functions 2.a and 2. , de applicable in this I

Special Operations LCO, are requfred to have applicable !

l Survalliances met to establish that this Special Operations LCD is being met. However, the control rod withdrawal sequences during the SOM tests may be enforced by the RidM l

(LC0 3.3.2.1, Function 2, MODE 2 requirements) or by a second licensed operator or other qualified member of the i

1 technical staff. As noted, either the applicable SRs for sfied according to the theRlel(LC03.3.2.1)mustbesatJ.2),orthepro applicable Frequencies (SR 3.10J .

of control rods must be verified (SR 3.10.F.3). perThis movement latter verification (i.e., SR 3.10,#'/3) must be performed during control rod movement to prevent deviations from the' specified sequence. These surveillances provide adequate i i

assurance that the specified test sequence is being followed. l l

f.(p

-w onpurunakaunw t- tf r cr-tr;k --"pis) ;

' Periodic verification d R -f '

e teliM ty thi; .'0^ will ensure that the reactor is operated within the bounds of the safety analysis. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is intended to provide appropriate assurance that each. operating hi'^ 6 aware of and verif compliance with these Special Operations LC0 requirements.

SR 3.10 5 f1*

i Coupling verification is performed to ensure the control rod !

is connected to the control rod drive mechanism and will '

perform its intended function when necessary. The verification is required to be performed any time a control g rod is withdrawn to the " full out' notch position, w' prior ~

to d;;hr'"; *" --tr;I .-M 0";"O'Ipsfter work on the control rod or CRD System that couldjaffect coupling. This p,6 w n u m .:,i y n G r tu uqos e of perfe, mig thh SpesM

_- __ nh%

continue d)

C^/4 ais B 3.10-37 L 1, G /e7/^h w

E Organization 5.2 5.2 Organization (continued) l 5.2.2 Unit Staff The unit staff organization shall include the following:

a. At least two non licensed operators shall be assigned while k operating in MODE 1. 2. or 3 and at least one non licensed operator shall be assigned whenever the reactor contains fuel.

b. At least one licensed Reactor Operator (RO) shall be present in the control room when fuel is in the reactor. In addition, while the unit is in MODE 1, 2. or 3. at least one licensed Senior Reactor Operator (SRO) shall be present in the control room,

c. Shift crew composition may be less than the minimum requirement of 10 CFR 50.54(m)(2)(i) and 5.2.2.a and 5.2.2.9 for a period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in order to accommodate unexpected absence of on duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements.

d. A Radiation Protection Technician shall be on site when fuel is in the reactor. The position may be vacant for not more than 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in order to provide for unexpected absence, provided immediate action is taken to fill the required position.

e. Administrative procedures shall be developed and implemented to limit the working hours of unit staff who perform safety related functions (e.g., licensed SR0s, licensed R0s, radiation protection technicians, auxiliary operators. and key maintenance personnel). The controls shall include guidelines on working hours that ensure that adequate shift coverage is maintained without routine heavy use of overtime for individuals.

Any deviation from the established guidelines shall be authorized in advance by the Plant Manager or designee, in accordance with approved administrative procedures. or by higher levels of management. in accordance with established procedures and with documentation of the basis for granting the deviation.

(continued) j FERMI UNIT 2 5.0 3 Revision 13. 08/06/99 i

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spec.;Rcdion 5 2

$0 SQ b h* CCt h(TM Ts w .,

TAB P 6.7.?-1

~

MINIMUM T Cpr M OMPOSIT W.

\

POSITION

/ NU

/OF 1

/DUALS R xED 10

/

t. POSill /

.h [ DNDIT! , 2, or DNDITION or 5 l

' S 1 1

(/ NASS '

No 5,1.1.a. --=-- viiponaopo3 ~ ~~ 2 l' ' ~ ~ ~ I M.I J g ,g,2 .] q_.(TA ._# ~~None ~ -

1 3

Ta __ NOTATIO*

j NS$ -

Nuclear 5 t Superv r with a 5 l

or Operator cense l NAS - Nuclear ssistant ift Supervis with a Seni Operator licen 0 -

Nuc ar Superv .ing Operator ith an Opera or licens NPD0/ NAP .

N lear Pow Plant Operat or Nuclear sistant P er lant Ocera or f.1*1 3 STA - ,

Shift Technical Advisor

'g g, ( fF r *(fer the kirei n,--6 4 4 herymr] tne shift __ crew composition may be tme

~

less than the minimum requirements er in : wr for a period of time not

~

to exceed crew members 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months provided in order to accommodate immediate action is takenunexpected to restore thiaosence shift crewof on-cuty snift g' [

composition to within the minimum reouirementtAt ia e 6. 1. Inis o

l During any absence of the Nuclear Shift Supervisor from the control room while the unit is in OPERATIONAL CONDITION 1. 2 or 3, an individual (other than the g,6 Shift Technical Advisor) with a valid Senior Operator license shall be

designated to assume the control room command function. During any absence of

$fDh,cagp, the Nuclear Shift Supervisor from the control room while the unit ts in OPERATIONAL CONDITION 4 or 5, an individual with a valid Senior Operator l

51 15c'a5' ar oo'r* tor 15cea5' 5ha11 be desisaat'd to a55u== the contra' roo=

command function.

l 1

l FERMI - UNIT 2 6-5

~

PAGE O 0F 05 Reva

l l

l DISCUSSION OF CHANGES l ITS: SECTION 5.2 - ORGANIZATION l

Operators, whose specific titles are replaced with the NUREG-1433 I presentation of the position name; refer also to DOC LA.3) have h i

been transferred from CTS Table 6.2.2-1 to ITS 5.2.2.a. In

) addition, ITS 5.1.2 contains requirements for the control room command function, ITS 5.2.2.c contains the minimum requirements l for licensed Reactor Operators and Senior Reactor Operators to be present in the control room. The relocated details are not required in the ITS to provide adequate protection of the public health and safety since these details have no impact on plant operation. safety system Operability, or other safety related function.

1 LA.2 CTS 6.2.2.d requires the proper unit staffing and specifically i requires all Core Alterations to be supervised by a licensed l Senior Reactor Operator (SRO) or Senior Operator Limited to Fuel Handling. These CTS requirements are contained in 10 CFR 50.54 (m)(2)(iv) and do not need to be repeated in the ITS. As stated above, these requirements are specified in 10 CFR 50.54, and therefore, cannot be changed by Detroit Edison without prior i approval from the NRC. The qualification of personnel observing or directly controlling the Core Alterations is also specified in UFSAR Section 13.1.2.5. This provides an equivalent level of I regulatory control and is an administrative change with no impact i on safety. These details are not required in the ITS to provide adequate protection of the public health and safety since the requirements for all Core Alterations to be supervised by a licensed SR0 or Senior Operator Limited to Fuel Handling are still maintained.

LA.3 CTS 6.2.1.b, 6.2.2.g. Table 6.2.2 1, 6.2.4.1. and 6.8.6 details '

the plant specific titles for various positions. Some of these specific titles continue to be utilized in ITS Section 5.0. but lh are modified by a Note: ITS 5.0 Note, which allows revisions to be made to these titles while considering the ITS titles to continue to represent the same organizational position. (In the case of the CTS titles NPP0 and NAPP0, the specific title is replaced with the position name.) The relationship between the ;

ITS titles (or organizational position) and any variations in actual plant specific titles will be located in the UFSAR.

Changes to the UFSAR are controlled in accordance with 10 CFR 50.59. This detail is not required in the ITS to provide adequate protection of the public health and safety since these details l

have no impact on plant operation, safety system Operability, or other safety related function.

FERMI - UNIT 2 2 REVISION 13. 08/06/99l

Organization 5.2 Insert 5.2-2 At least two non licensed operators shall be assigned while operating in MODE 1. 2. or 3 and at least one non licensed operator shall be assigned whenever the reactor contains fuel.

FERMI - UNIT 2 Page 5.0 2 (Insert) REVISION 13. 08/06/99l

Seeer'A c* YI*^ f (fl{ cosec Sfe ct fica YIDA S'S

h

,,$cehecs7/C0 IM

i CONTAINMENT SYSTEMS

]/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PR ! MARY CONTAINMENT h PRIMARY CON"AINMENT INTEGRITY l 7

tiMITING CONDITION FOR OPERATION 3.6.1.1 PRIMARY CONTAINMENT INTEGRITY shall be maintained.

t APPLICABILITY: OPERATIONAL CONDITIONS 1, 2* and 3.

o a.CI.ID.N:

4 Without PRIMARY CONTAINMENT INTEGRITY, restore PRIMARY CONTAINMENT INTEGRITY k within I hour or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in G COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

so g SURVEftlANCE RE0VIREMENTS O

Y b4.6.1.1 PRIMARY CONTAINMENT INTEGRITY shall be demonstrated:

a. ffte acn closing each penetrat suo. lect to ly exc the primary ontainment air eks, if opene followino B testig,

~

Ay .

A B test, by I ak rate testino he se31s with (fat P a , 56.5 5 5.r2 b/ psig, and veritying that wnen the measured leakage rate for these g, g, gt ,f , g seals is added to the leakage rates determined pursuant to Surveillance Requirement 4.6.1.2.b for all other Type B and C penetrations, the combined leakage ra.te is less than or equal to 0.60 La -

b. ast once per 31 days by verifying that all primary containment penetrations except those inside the containment or in locked high radiation areas (listed in Table 4.6.1.1-1) not capable of being closed by OPERABLE containment automatic 5*C isolation valves and required to be closed during accident conditions are closed by locked closed valves, blank flanges, or yp:dr.cafs.on deactivated automatic valves secured in position, except for 8 valves that are open under administrative control as permitted by 3,6 , [,3 -

Specification 3.6.3. l

1. Valves, flanges, and deactivated automatic valves which are located inside the containment, and are locked, sealed or otherwise secured in the closed position shall be verified closed during each COLD SHUTDOWN except such verification need not be performed when the primary containment has not been deinerted since the last verification or more often than once per 92 days.

See Special Test Exception 3.10.1.

FERMI UNIT 2 3/4 6-1 Amendment No. /7, 102 PAGE k 0F 24 RdI3 L .

Spe cF,'ca1/ori 5, S~

(p(u see .pe cWic.fim 3, Cel.3 ') l fANTAINMENT SYSTEMS PRIMARY CONTAINMENT LEAKAGE #.

/) [/)/fo set fcc[$ce[t'#,4 00./ * / )

LIMITING CONDITION FOR OPERATTON 4.C.I.2- Primary containment leakage rates shall be limited.to:

5 4.8L d,l a.

f fal2.0- An overall integrated leakage rate of less than or equal to: L, i 0.5 percent by weight of the containment air per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at Pa .

5.'5.12 6 56.5 psig.

g*g'g'j* b. A combined leakage rate of less than or equal to 0.60 L for primary containment penetrations and primary containmen! isolation valves subject to Type B and C tests when pressurized to P, in accordance with the Primary Containment Leak _ age Rate Testing Program described in Soecification 6.8.5.c.f ex pt for main st m L line ipolatton valves * #d primary containdien solation valv twhichfare hydrostaticaY y tested. / N'I 'h bt [c. *Less than or equal to 100 scf per hour for all four main steam j lines when tested at 25.0 psig.

. d. A combined leakage rate of less than or equal to 5 gpm for all -

containment isolation valves in hydrostatically tested lines which C[ penetrate the primary containment, when tested at 1.10 Pa , 62.2 psi 9

,M

- e. Less than or equal to I gpm times the number of valves per u penetration not to exceed 3 gpm per penetration for any line Y* penetrating 62.2 psig. containment and hydrostatically tested at 1.10 Pa.

N APPLICABILITY: When PRIMARY CONTAINMENT INTEGRITY is required per 1 5pecification 3.6.1.1.

ACTION:

Wita:

64. i2. d.1 a.

The measured overall integrated primary containment leakage rate exceeding 0.75 L., or

b. The measured com5tned leakage rate for primary containment 6 54Mel penetrations and primary containment isolation valves subject to Type B and C tests in accordance with the Primary Containment

~v Leakage Rate Testing Program. Jepcept for myin steam sine upation/

ves" 3Ma primary gentainment isolatinn/ valves which art / /

h drostnically tesfed.fexceeding 0.60 La, or

~

(c. ine measured leakage rate exceeding 100 scf per hour for all four see main steam lines, or

. - d. The measured combined leakage rate for all containment isolation

)D C# ' p"" f'," ^

valves in hydrostatically tested lines which penetrate the primary 3,6,l.3 i e. containment exceedins 5 spm. or The leakage rate of any hydrostatically tested line penetrating ~

primary containment exceeding I gpm per isolation valve times the set C number of containment isolation valves per penetration or greater than 3 gpm per penetration, S rior to increasing reactor coolant system temperature above 200*F, restore:

\p 3e6 ce l,t f gcaf,ona. The overall integrated leakage rate (s) to less than or equal to 0.75 La, and Get {'Exem tion to Appendix J of 10 CFR Part 50.

Sf ectfica1r on .

3 . G , I. 'l FERMI - UNIT 2 3/4 6-2 Amendment No. Jg2,108 PAGE $ OF 24 gg

Spcclft' cad 0A 5 $^

(ttha seespeconcdion L' ! N CONTAINMENT SYSTEMS

, jgfg f,*y f,6 , /,3 LIMITING CONDTTTON FOR OPERATION (Centinued) i ACTION: (Continued)

b. The combined leakage rate for primary containment penetrations and '

g,g, g , f , g primary containment isolation valves subject to Type B and C tests in accordance with the Primary Containment Leakage Rate Testing lQg,f 1

Program,ryxcept tontainment f orvalvn isolatWn pain which steamIre sine ysosauon valves' hyda"+?+4"11y aryr prima [ry Iae+=d tests 16 less than or equal to 0.60 La, and g M. The leakage rate to less than or equal to 100 scf per hour for all four main steam lines, and Sg d. The combined leakage rate for all containment isolation valves in M8 ..

hydrostatically tested lines which penetrate the primary 3.(,.l.3 h )J containment to less than or equal to 5 gpm, and l e. The leakage rate of any hydrostatically tested line penetrating primary containment to less than 1 gpm per isolation valve times the number of containment isolation valves per penetration or less than 3 gpm per penetration.

SURVEfttANCE REOUTREMENTS _.

4.6.1.2 Perform required primary containment leakage rate testing in y accordance with the Primary Containment Leakage Rate Program described in Specification 6.8.5.g.**

Sco TPlei(TC4 Non

%. I .I l

l f*Exemp on to Apifendix J of 10 FR Part 50 T LR'I h I

- Exc pt for LPCI Loop A an B Injection Isolation alves, which are h rostatically tested _accordance with Speci cation 4.4.3.2.2 i lieu

( this requirement. /

FERMI - UNIT 2 3/4 6-3 Amendment No. J92,108 PAGE /0 0F 24 Rev'I3

Spec 11Ficofion 5, 5~ {

(/ffsa .see &ecWical Ion 3 6,l.2 CONTAINMENT SYSTEMS SURVElltANCE RE0VIREMENTS h .6.1.3 Each primary containment air lock shall be demonstrated OPERABLE:

a. Witninf days following/tacn closing, except when the air / lock 13 bgi nn An e n.t sa- mn1+4n14 antriet. then at Yeart once nor 10 r'a ve .)

verifying seal leakage rate less than or equal to 5 scf per 6,6,12..d/2 g hour when the gap between the door seals is pressurized to Pa

56.5 psig, i

I

b. Prio'r~ o establishing PR RY CONTAINMENT *!

loc as been opened d ing periods when ntainment inteor yw EGRITY when the r inev ranoired. The demonstration chall/ verify a seal leakage rate less than or equal to 5 scf per hour when the gap between the door

,_ seals is pressurized to P Quwd py C.t tc QMI , 56.5 n.s u psig.junie3ar un u . a . i .3.ct2.;Ine gir loca i

c. By conducting an overall air lock leakage test at P , 56.5 psig, s.s.n.A.t, t ;ad %;gi'ria9 that th' av'ran air i c' k rat' $a withia g g f1. Prior o initial fuel loadin and at 30 months

i ervals ther fter, , i

2. P or to establishing P RY CONTAINMENT INT RITY when the ir lock has been open during periods whe containment integrity was not re tred, if maintenance hich could affect the leak ti integrity of the d rs has been T [ perforined since t last successful tes pursuant to M ecification 4. .l.3.c.1. f bE At least once per 6 months by verifying that only one door in each gg gg,g. air lock can be opened at a time.**

5J .G . I,2.

1 1

p f.T,l'L.e.

The provisions of Specification 4.0.2 are not applicable.

Eg e(Icab

,(, .1,t )**Except OPERABILITY thatwhenthe theinner primarydoor needisnot containment be provided inerted, opened thattothe verify inner interlock door interlock is tested within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months after the primary containment has ;

Qendeinerted.

_. FERMI - UNIT 2 3/4 6 9 Amendment No.108 Corrected August 19. 19 %

PAGE N OF 24 g (3

F Q ectYicati on F,5-l # # #' h8C'YiCAIlanS'6,lel)

CONTAINMENT SYSTEMS

%. PRIMARY CONTAINMENT STRUCTURAL INTEGRTTY 7 LTMITING CONDITION FOR OPERATION 6.1.5 The structural integrity of the primary containment shall be maintained at a level consistent with the acceptance criteria in Specification

[ 4.6.1.5.1.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.

2 L ACTION:

0 With the structural integrity of the primary containment not conforming to the h above requirements, restore the structural integrity to within the limits within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in

[ COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , w

SURVEill ANCE REOUTREMENTS M . 0.1. 5. IT structural integrity of the xposed accessible interior nd f exterior s aces of the primary contain nt shall be determined duri the shutdown r each Type A containment 1 age rate test by a visual i spection of thcs surfaces. This inspection s 11 be performed prior to th Type A contat nt leakage fthen t Type A test ifrate thetest and d ing two other refueling out es before intery for the Type A test has be extended to

,' 10 ars to verify no apparent c nges in appearance or other normal de adation.

'l'

.6.1.5.2 Reoort s Any abn al degradation of the prima containment structure detected during e above required inspections shall be reported i a Special Report to the ission pursuant to Specifi tion 6.9.2 within days. This report shal include a description of the ondition of the structure, the inspec on procedure, the inspection riteria, and the corrective actions t en. f f

LAI

~

FERMI UNIT 2 3/4 6-11 Amendment No.108 Corrected August 19. 1996 O

PAGE OF 24

[g1/13

INSERT THIS PAGE IN FRONT OF VOLUME 11 !

Volume 11: CTS MARKUP COMPILATION-Remove Replace 6-5 (5.2 CTS M/U) pg 3 of 5 6-5 (5.2 CTS M/U) pg 3 of 5 Rev 13 3/44-1(3.4.1 CTS M/U) pg 2 of 6 Rev 10 3/44-1(3.4.1 CTS M/U) pg 2 of 6 Rev 13 3/4 6-1 (5.5 CTS M/U) pg 4 of 24 3/4 6-1 (5.5 CTS M/U) pg 4 of 24 Rev 13 3/4 6-2 (5.5 CTS M/U) pg 5 of 24 3/4 6-2 (5.5 CTS M/U) pg 5 of 24 Rev 13

]

3/4 6-3 (5.5 CTS M/U) pg 6 of 24 3/4 6-3 (5.5 CTS M/U) pg 6 of 24 Rev 13 3/4 6-9 (5.5 CTS M/U)pg 8 of 24 3/4 6-9 (5.5 CTS M/U) pg 8 of 24 Rev 13 3/4 6-l1 (5.5 CTS M/U) pg 9 of 24 3/4 6-11 (5.5 CTS M/U) pg 9 of 24 Rev 13 3/4 614 (3.6.1.3 CTS M/U) pg 6 of 9 Rev 11 3/4 6-14 (3.6.1.3 CTS M/U) pg 6 of 9 Rev 13 3/4 6-20 (3.6.1.3 CTS M/U) pg 7 of 9 Rev Sa 3/4 6-20 (3.6.1.3 CTS M/U) pg 7 of 9 Rev 13 3/47-1(3.7.1 CTS M/U) pg 1 of 2 3/47-1(3.7.1 CTS M/U) pg 1 of 2 Rev 13 3/4 7-5 (3.7.2 CTS M/U) pg 4 of 9 3/4 7-5 (3.7.2 CTS M/U) pg 4 of 9 Rev 13 3/4 7-9 (3.7.4 CTS M/U) pg 3 of 3 3/4 7-9 (3.7.4 CTS M/U) pg 3 of 3 Rev 13 1

l

)

i Rev 13 08/06/99 1

1

speelRcdion 5 2

[Gf60 S64.

det M S'

-s.,

TABLP 6.7.7-1

/ .

MINIMUMJfif FT CAE MPOSIT POSITION

/ NU

/OF IN

/

DUALS R ED TO POSITI j

h.h [ ONDITI , 2, or DNDITION or 5 f

5 N

{

g,77..$.~ ' r (NPPD/kA 2 g 5

e!

7,- __: <Ta $~ ~ ~ - 3 ~~ - - g,57, -

~

3 TA __ NOTaff0 NSS -

Nuclear 5 t Superv r with a S for Operator cense NAS - Nuclear ssistant ift Supervis with a Sent Operator licen 0 -

Nuc ar Superv' ing Operator 4th an Oper r licens NPP0/ NAP -

N lear Pow Plant Operat or Nuclear sistant P er Iant Oeera or b '1, $ STA - ,

Shift Techntcal AcVisor c

s 4 ( 3e41er tne wirr1.ar h hervma the shift crew composition may be em-less than tne minimum requirements:cY ::::: _

-u fer a period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months inimmediate order to action accommocate unexpected aosence of on-outy shift Q,[

crew members provided is taken to restore the_ shift crew composition to within the minimum recuirementLAt la* e o.' c-1. Ints o __

During any absence of the Nuclear Shift Supervisor from the control room while the unit is in OPERATIONAL EONDITION 1, 2 or 3, an individual (other tnan the g6 shift Technical Advisor) with a valid Senior Doerator license shall be a designated to assume the control room command function. During any absence of

$[N'{#CSg the Nuclear Shift Supervisor from the control room wnile the unit is in OPERATIONAL CONDITION 4 or 5, an individual with a valid Senior Operator 5,\ license or Operator license shall be cesignated to assume the control room co*nmand function.

FERMI - UNIT 2 6-5 O

~

PAGE OF 05 mwa

l l

l gper.:FicmiarJ 3. 4.1 l- 3 /4. 4 REACTOR COOLANT SYSTEM (ga, su. spcikcgHon 5410) i I

3/44.1 RECTRCULATION SYSTEM RECIRCULAT10N LOOPS <l (IMITING CONDITION FOR OPERATION 3.t. . - Two reactor coolant system recirculation loops shall be in operation.

APPLICABf t fTY: OPERATIONAL CONDITIONS I and 2*.

ACTION:

a. With one reactor coolant system recirculation loop not in operation:

tio M .i 1. Within 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months s: LA.I -

tJSTF '

7 P1 the individual re culation pump flow c s

op atino recirculatio pumointheManualmod_poro11er'forthe]

[

bg gy,g,g b) Reduce THERMAL POWER to less than or equal to 67.2% of RATED THERMAL POWER.

l

) Limit the spee_ d[f the operating recirc[ation pump to les/'thaQ or eoual to 75% of rated pump speed s

^

d) inc ase Ine MininurycRITICAL PDWLK MA147 L th value for singy loop operation reg /re(nLVM) hafety L1 on )tol d by Specifica

.1.2. f (hbD: MLHGt i MCPA *, Lw 5 #l b l l I'Y'I'b'I-h i g -

l -

e) Change the Average Power Range Monitor (APRM) Simulated T rmal,,

@ M '!'g'3 Power - Upscale Flow Biased Scram fard M "u:L 'mr icirrt ts/

MS Allowable Values to those applicable for single retirculationl ~ l' loop operation per Specifications 2.2.1 and 3.3.6.

l f) Perform Surveille.nce Requirement 4.4.1.1.4 if THERMAL POWER is f,gg f;a h less than or equal to 30% of RATED THERMAL POWER or the g ,,o f recirculation loop flow in the operating loop is less than or

{

u equal to 50% of rated loop flow.

-2. OU.: H -- 6 " it S::: "" ;;;;n;;,,n

. win tne u m n ..- . 2 .

b.

With no reactor coolant system recirculation loop in operation while in NMD OPERATIONAL CONDITION 1, immediately place the Reactor Mode Switch in the SHUTDOWN position.

c.

With no reactor coolant system recirculation loops in operation, while in {

h C.TlD td C,. OPERATIONAL CONDITION 2, initiate measures to plact the unit in at least !

HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . ,

See ectalle/ Exception 3./0.4.]

I e

FERMI - UNIT 2 3/4 4 1

~

Amendment No. EJ,64,ES,E3, E7,JES, l 122 PAGE 2. OF 06 b Rev10 t i

Spe c e'fr e s y,'o n C, g (Also se e pS ectYicafuan 3.6,l.I)

_~ [Also .sce pS ee; fica m., 5.G.I.3)

CONTAINMENT SYSTEMS 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT K PRIMARY CONTAINMENT INTEGRITY l 7 tIMITING CCNDITION FOR OPERATION Y 3.6.1.1 PRIMARY CONTAINMENT INTEGRITY shall be maintained.

M t ApptICABILTTY: OPERATIONAL CONDITIONS 1, 2* and 3.

3 i ACTION:

Without PRIMARY CONTAINMENT INTEGRITY, restore PRIMARY CONTAINMENT INTEGRITY N [ within I hour or be in at least NOT SHUTDOWN within G COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , es b

g SURVEfttANCE REOUTREMENTS

$ \

Y b 4.6.1.1 PRIMARY CONTAINMENT INTEGPITY shall be demonstrated:

a. /Lfte ach closing each penetrati suDJect to ly B testtr)(,7 exc the primary ontainment ai eks, if opene following Av.nd A B test, by 1 ak rate testino he seals with dat P a, 56.5 S.S.r2 b/ . psig, and verifying that when the measured leakage rate for these f, g ,,t,A . : seals is added to the leakage rates determined pursuant to Surveillance Requirement 4.6.1.2.b for all other Type B and C penetrations, the combined leakage ra.te is less than or equal to 0.60 La-

b. ast once per 31 days by verifying that all primary containment penetrations except those inside the containment or in locked high radiation areas (listed in Table 4.6.1.1 1) not capable of being closed by OPERABLE containment automatic sec isolation valves and required to be closed during accident conditions are closed by locked closed valves, blank flanges, or deactivated automatic valves secured in position, except for yfM 8 p(Cafs.on valves that are open unoer administrative control as permitted by l 3,6,[,3 Specification 3.6.3.

f 1. Valves, flanges, and deactivated automatic valves which are located inside the containment, and are locked, sealed or otherwise secured in the closed position shall be verified closed during each COLD SHUTDOWN except such verification need not be performed when the primary containment has not been deinerted since the last verification or more often than once per 92 days.

See Special Test Exception 3.10.1.

FERMI - UNIT 2 3/4 6-1 Amendment No. /9, 102 PAGE k 0F 24 RdI3

Specif,*cdion 5, S~

(71/u sec .Goe cifo*c efio n 3 C l*0 CONTAINMENT SYSTEMS PRIMARY CONTAINMENT LEAKAGE .! [/J/fo Sec fer/$ce((04 00 */. / }

tIMITING CONDIT10N FOR OPERATION 4.C.1.2 Primary containment leakage rates shall be limited to:

S S. R.d,j a. An overall integrated leakage rate of less than or equal to: L E*f l2* C- 0.5 percent by weight of the containment air per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at Pai,

5. s.'r2. . b 56.5 psig.

E'g'q'J* b. A t.ombined leakage rate of less than or equal to 0.60 L for primary containment penetrations and primary containmenk isolation valves subject to Type B and C tests when pressurized to P, in j accordance with the Primary Containment Leakage Rate Testing Program described in Specification 6.8.5.c.f ex pt for main st m -

l line ifolation valves * #d primary containnien solation valv iwhich/are hydrostaticaY v tested. / NI'h' i

I bb

.j [c. *Less than or equal to 100 scf per hour for all four main steam lines when tested at 25.0 psig.

d. A combined leakage rate of less than or ecual to 5 gpm for all

k. containment isolation valves in hydrostatically tested lines which G[ penetrate the primary containment, when tested at 1.10 Pa , 62.2 psig.

,M

. e. Less than or equal to 1 gpm times the number of valves per u penetration not to exceed 3 gpm per penetration for any line m

Y penetrating containment and hydrostatically tested at 1.10 Pa .

62.2 psig.

' APPLICABILITY: When PRIMARY CONTAINMENT INTEGRITY is required per 1 Specification 3.6.1.1.

ACTION:

htn:

5.s.12. d.1 a. The measured overall integrated primary containment leakage rate exceeding 0.75 L , or

b. The measured como,ined leakage rate for primary containment 6 5 G.M penetrations and primary containment isolation valves subject to Type B and C tests in accordance with the Primary Containment Leakage Rate Testing Program,Jefcept for myin steam sine minationj (vjrives= yia primary pontainmenr isolation / valves which are/ /

1 h or (c. Vtydrostrtically Ine measured leakage tested.Texceeding rate exceeding 100 sc 0.60 La,f per hour for all four se e, main steam lines, or

. . d. The measured combined leakage rate for all containment isolation

)D ' p' # " f '," ^

3,6,13 j e. valves in hydrostatically tested lines which penetrate the primary l containment exceeding 5 gpm, or The leakage rate of any hydrostatically tested line penetrating '

primary containment exceeding I spm per isolation valve times the set, C number of containment isolation vif ves per penetration or greater than 3 gpm per penetration,

P e c s hedeon rior to a.

increasing reactor coolant system temperature above 200*F, restore:

\ 3 6.l,! The 0.75 overall La, andintegrated leakage rate (s) to less than or equal to G r e- (ExemtiontoAppendixJof10CFRPart50.

Sp ecif:/ca fr o n 3 . C . I. 3 FERMI - UNIT 2 3/4 6-2 Amendment No. Jp2,108 PAGE $ OF 24 g3

f1CC j l l'C& YI o O E (k/.so .See. Qtco*fiCA Non 5' b

l '

CONTAINMENT SYSTEMS

, ,5 jgjp f/,, f,6 , /,3

..g (1MITING CONDTTION FOR OPERATION fCertinuedi a

. ACTION: (Continued)

b. The combined leakage rate for primary containment penetrations and 5,g, g, f , g primary containment isolation valves subject to Type B and C tests ,

in accordance with the Primary Containment Leakage Rate Testing g,(

Program,iyxcept for pain neam sine ysoasuon valves" arut primarn ,

containmarn t isolatMn valves which Ira hyd~ * + > + M =11y I= e + =d J tests t6 less than or equal to 0.60 La, and l g !

i

c. The leakage rate to less than or equal to 100 scf per hour for all four main steam lines, and gg d. The combined leakage rate for all containment isolation valves in k",8. Mb )$ hydrostatically tested lines which penetrate the primary containment to less than or equal to 5 gpm, and 3 (,.l.3 i

e. The leakage rate of any hydrostatically tested line penetrating primary containment to less than I gpm per isolation valve times

( the number of containment isolation valves per penetration or less '

i

( than 3 gpm per penetration.

SURVEILLANCE REOUTREMENTS '

4.6.1.2 Perfortn required primary containment leakage rate testing in y accordance with the Primary Containment Leakage Rate Program described in Specification 6.8.5.g.**

\ Seb

!Ifulfit480n

% .I.I /

Exemp on to Apliendix J of 10 FR Part 50 ]

LR,1 ' I h

- Ext pt for LPCI Loop A an B Injection Isolation alves, which are h restatically tested _accordance with Speci cation 4.4.3.2.2 i lieu

( this requirement. / ,

FERMI - UNIT 2 3/463 Amencment No. JS2,10B PAGE /0 0F 24 gev' B

I i

Spec ificafior' 5' 5~ \

CONTAINMENT SYSTEMS

(/f{so .se e Qecifica% 3*SoI*2) j SURVEILLANCE RE0VIREMENTS l .cah .6.1.3 Each primary containment air lock shall be demonstrated OPERABLE:

3.(,.ia

a. Witn1n days followin aCn Closing, eXCe

/,,g, l h heina ne.a sa, ..1+4n1 when the at lock iM ant ries. then at pact once eer 0 davt) l, verifying seal leakage rate less than or equal to 5 scf per g, g , f 2, ,y, g hour when the gap between the door seals is pressurized to Pa

56.5 psig. l t l

b. #P Nor o establishing PR RY CONTAINMENT EGRITY when the r loc as been opened d ing periods when ntainment inteer yw i

<nnv renuired. The d onstration shall> verify a seal leakage rate less than or equal to 5 scf per hour when the gap between the door 3_ seals is pressurized to P,, 56.5 psig.juniepr tne pt soca _i hurd nfr ,t tc 0- @rIewswien as.o.a.3.ctz.;

c. By conducting an overall air lock leakage test at P , 56.5 psig, 6 5.G.A.2., i. @M,7'7i"9"'t****"**Ic'"'*'"'**i'**"

g. g f 1. Priorfter, o initial fuel loadin /and at 30 months

i ervals ] ,

ther

2. P or to establishing P RY CONTAINMENT INT uRITY when the ir lock has been open during periods whe containment integrity was not re tred, if maintenance hich could affect the leek ti integrity of the d rs has been T performed since t last successful tes pursuant to k cification 4. 1.3.c.1. f E

At least once per 6 months by verifying that only one door in each Speciftent,en

'ir I*ck c'" b' P'"'d 't 8 t i"' "

cJ .G, , I,2 p- f. F,12.. e.

The provisions of Specification 4.0.2 are not applicable, f $c % f"Except that the inner door need not be opened to verify interlock g ,(, . ,1 ) OPERABILITY when the primary containment is inerted, provided that the inner door interlock is tested within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months after the primary containment has Qendeinerted.

_ FERMI - UNIT 2 3/4 6 9 Amendment No.108 Corrected August 19. 19 %

PAGE N 0F _

24 gp

.) GCI'fiC0f** O A S' T (glp see Gecifica f,4 E.G,I I)

CONTAINMENT SYSTEMS

%. PRIMARY CONTAINMENT STRUCTURAL INTEGRITY 7 LIMITING CONDITION FOR OPERATION 6.1.5 The structural integrity of the primary containment shall be maintained at a level consistent with the acceptance criteria in Specification

! 4.6.1.5.1.

APPLICABILfTY: OPERATIONAL CONDITIONS 1, 2, and 3.

2 k ACTION:

0 With the structural integrity of the primary containment not conforming to the h above requirements, restore the structural integrity to within the limits within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in

[ COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

l SURVEiltANCE REOUTREMENTS 1

^4 structural integrity of the xposed accessible interior nd f ex.C.;.5.M terior s aces of the primary contain nt shall be determined duri the shutdown r each Type A containment I age rate test by a visual i spection of thos surfaces. This inspection s 11 be performed prior to th Type A i contai nt leakage rate test and d ing two other refueling out es before the n t Type A test if the intery for the Type A test has be extended to i' 10 ars to verify no apparent c nges in appearance or other norwal de adation.

.6.1.5.2 Reoorts Any abn al degradation of the prima containment structure detected during e above required inspections shall be reported i a Special Report to the ission pursuant to Specifi tion 6.9.2 within days. This report shal include a description of th ondition of the structure, the inspec on procedure, the inspection riteria, and the corrective actions t en. f f

LB. /

FERMI - UNIT 2 3/4 6-11 Amendment No.108 Corrected Augus: 19. 1996 O

PAGE OF 24

[g1/13

Spec PtcM t M $ bal. 3

/%

) CONTAINMENT SYSTEMS DRYWELL AND SUPPRESSION CHAMBER PURGE SYSTEM LIMITING CONDITION FOR OPERATION j sg g,g,3,( 3.6.1.8 The drywell and suppression chamber purge system G ....n, G.u..,

gOE ^^ ' :h, : ' ?' '--' -4 may be in operation with the supply and exhaust i-isolation valves in one supply line and one exhaust line open for inerting, I deinertin or pressure control. Nitrogen VENTJNG/ makeup and pressure cont etMA' "*'4""Me*R'"'" "M!! sd" "" """ @rol

, APPLICABILITY: OPERATIONAL CONDITIONS 1, g2 and 3. j l

\

bC.11QN; ADe AcitoNS Nntt y Acnons NOTE 3- >

. [.g gs

'WX A^9 A c n e M S el0 1 E 3 > '

a. With a orywell and suppression chamber purge system supply and/or

'aA N "A exhaust isolation valve open, except as permitted above, close the valve (s) or otherwise isolate the penetration (s) within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or AC N b be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD h SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ,

b. With a drywell and suppression chamber purge system supply and/or h"2 htta) D exhaust isolation valve (s) with resilient material seals having a measured leakage rate exceeding the limit of Specification T

g* 4.6.1.8.2, restore the inoperable valve (s) to OPERABLE status 'h dp or c

g within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT SHUTDOWN within the next 12 hCil0d b hours and in COLD SH DOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . lje eq

001 Aemia B/ ' L.2 ~k t _)

SURVEfttANCE REOUTREMENTS i N

64 % 13 I 4.6.1.8.1 [Before being opened ' r purge / vent operation b u h SETS) the D

/JOTG - drywell ana suppression enamDer purge supply and exhaust butterfly isolation 7 valves shall be verified not to have been open for purge / vent operatio R

'-~ " ' - - '

l -l 4

thr - ' "" '- - " - ^ ^ ' - - -

At ast once pe [92 sl a io or e h ? . . . . . , ; ;; .'.

bk NO'4 ' 4.6 .

l0 . :..:. , ::25 00 ' :h, -d ::'. :: n d drywell and suppression chamber purge supply and exhaust isolation valve with resilient material seals shall be demonstrated OP_ERABLE by verifying that_the measured leakage rate 6 9 e " 2-J

{ .;... ; u r.2 y ': : :::r----- =

34 MJ 3a[ A Primary containment nitrogen VENTING and pressure control is permitted ' I through the 1-inch valves :nd i: n:t dje+'a '" 9' 'r" : "Or 255 d;~ I W , E

,N

~

FIF.".! - U!i!T 2 2/4 6-14 Amendment No. 53 ktN 0 lB 0F 09 PAGE ((f

1 l Spec.nricarsva 3.G.t3 m

]

CONTAINMENT SYSTEMS 3/4.6.3 PRIMARY CONTAINMENT ISOLATION VALVES hel tIMITING CONDITION FOR OPERATION *

[r eAcbrbwMdy-fo-Svilpr.aseckw4u Was breakW)5

_.3--..Each 3.,6.

3 p

. ,rimary3 .. .

conta,inment i solati on valve 6-d --"* -- * -r + r---' * '-- l

_<__o

. ... shall be OPERABLE.** 3 APPLICABILITY: i do OPERATIONAL

-044: When reguiN CONDITIONS 1, d b msho*MI'**2, 3 and 3h00 l,.

Med ACTION: g g, gag 7 ig MO MM a. With oneror morniof the primary containment isolation valves *2 NOTES 3/f inoperable.fmaintain at iusst one isolation valve UWKABLE in eacy (affectea onnetration that is open an$within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months either:

C -- m ,. 1 ._ ,_,_m . - - - .

1

2. Isolate each affected penetration by use of at least one 8eguird AcYiDo deactivated automatic valve secured in the isolated d' g,l position,* or J .

3. Isolate each affected penetration by use of at least one E T ocke closed manual vulva a- h1=nk finana *

GFcher,k veW,withttun'aw suore ,o

! l. . I 1.

Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 'd AC'IIod E and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . g b With one or more of the reactor' instrumentation line excess flow ACmou c-30 check vaives inoperabie, operation may continue ,..~ m._

c' c - - -- -' ' . _ .

_. ^ ^t :::lic;bMrovided that within 4 g g,.7 rs either: ng {

fl . ine inoperaose vaive is returned to OPERABLE status, arr ) f-0 i2. The instrument line is isolated [and the associatea j h ACTION S #PTE .3 Instrument is ceclarea inoperaDie.

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

h00 Acmed Ch lJ Acno#J NofE l alsolation valves closed to satisfy these requirements may be reopened on an inttraittent basis under administrative control.

gag 14.

48:'-d ;r m!:) closed valves may be opened on an intermittent basis under j!d

-5R 34.131 administrattve ontrol. , 1 SR 34.613 , ,/

Q FERMI - UNIT 2 3/4 6-20 Amendment No. E3, 102 h lb PAGE 7 0F 09 W"I

L i l

! l Specti%6oa R 7.I 3/4.7 PLANT SYSTEMS

-( ' 3/4.7.1 SERVICE WATER SYSTEMS l

RESIDUAL HE AT REMOVAL SERVICE WATER SYSTEM [. l LIMITING CONDTTION FOR OpFRATION EMri At least the following 6b::rr@esidual heat removal __ service (I l LC O) g, 7, water (RHR5W) system subsystems.ynn eacn subs'ystem"Tb1fiprisco of:

a.

b.

Two OPERABLE RHR5W pumps', and A PERABL

/

low patif capablefef [ suttion /from[the taking I

L

associat ultimafe heat sink and trangerring tKe waterArough

! one RHR' heat richan shall be OPERABLE:

! a. In OPERATIONAL CONDITIONS 1 2 and 3. two subsystems, f

Jn wtRAi tuRAL CONDTJTO 4 an ,5 the subsystem ssociated' rb.

/withJyffems and c'mponentsdeouired Specifications 0PERABLa J.4.9.1, K 4.9.2, 3.9,.11.1.

/

3.9.U .

3 t

APPLICABillTY: OPERATIONAL CONDITIONS 1, 2, 3, Qnd 5 3 l ACTION:

a. In OPERATIONAL CONDITION 1, 2, or 3:

R With one RHRSW pump inoperable, restore the inoperable pump f cyg / to OPERABLE status within 30 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN A cr/oa E within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

gof; 8 k With one RHRSW pump in each subsystem inoperable, restore at least one inoperable pump to OPERABLE status within 7 days J g,,ppfJ g' or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Ar" With one RHR5W subsystem otherwis inoperable, restore the CNO/J C inoperable suosystem to _0PERABL status with at least one i OPERABLE pump within 02 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the l

d C.[lO/J E followin'g 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

/gpg p g With both RHR5W subsystems otherwise inoperable, restore at least one suosystem to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or be in at least HDT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in NCNON b COLD SHUTDOWN

within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ,

b. In OPERATIONAL CONDITION 3 h with the RHRSW subsystem (s), which),

[c7/On(shjp is associ ed with an RHR loop require'd OPERABLE by Specification A (g

s 3.4.9.1 c ' W i inoperable, declare the associated RHR loop / d' ~

d076 inoperable and take the_ ACTION required by Specification 3.4.9.l g gh e--as-epp+u-at>1 e].

I l <

i m M.l e

'Whene r both HRSW s stems tr /inopera [ if unabl o attai g lD g l 5 Na eQuir y this,Ad 10N. m ain react coolantyperatureas 3 l w as actica y use ova,lternate eat removal ethods.

l l FERMI - UNIT 2 3/4 7-1 1

1 f( l t

PAGE / OF 02 Ral I3 l.

1 Specification 3.7.2

\

< Page Removed in Rev 13 > j l

e r

3/ q 7-5' PAGE V 0F 09 RW '3

SA tiFsCA773rd 8,7,L[

, CUso se. spu!.4cah 3.7 3 )

i PLANT SYSTERS 0/so se.L S p ecj f c a % g , 5 ~) )

SURVEILLANCE RE001REwENTS 4.7.2.1 The control room emergency filtration system shall be demonstrated OPERA 8LE:

a. At least once per less12than hours by verifylg'that F. the control room air temperature-is or equal to l

b. fat least once per 31 days by:

i

1. Initiating fan operation free the control room with each subsystem, estabitshing flow for at least 15 minutes through g the HEPA filters and charcoal adsorbers.

Spec;fraNm 2. Verifying flow through the HEPA filters and charcoal adsorbers for at least 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months with the associated emergency makeup 3'73 inlet air heater OPERA 8LE. The subsystem used to establish the 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months of flow through the HEPA filters and charcoal adsorbers shall be staggered such that each subsystem is d ilized at least once per 62 days.

c. f At least once per 18 months or (1) after any structural maintenance on the HEPA fliter or charcoal adsorber housings, or (2) fo11 ewing painting, fire, or chemical release in any ventilation zone communicating with the system by:

1. Verifying that the system satisfies the in place penetration testing acceptance criteria of less than 1.0% and uses the test procedure guidance in Regulatory Positions C.5.a. C.5.c, and

'( C.S.d of Regulatory Guide 1.52. Revision 2 March 1978, while operating the systes at a flow rate of 1800 cfs a 105 through the makeup filter and 3000 cfm a 10% through the recirculation f11ter.

2. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in

$d accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2. March 1978, shows a methyl iodide penetration qNg Q g of less than 1.0% when tested at a temperature of 30*C and at a r*1=tive hu=idity of 705 in accordance with ASTM o3:03 1989 65 ~) with a 2 inch bed for the emergency makeup filter train; and a 4 inch bed for the emergency recirculation air filter train.

3. Verifying a system flow rate of 3000 cfm i 10% during system operation when tested in accordan:a with ANSI h510 1980,

d. After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows a methyl iodide penetration of less than 1.0% when tested at a j temperature of 30*C and at a relative humidity of 70% in accordance g with ASTM D3803-1989 with a 2 inch bed for the emergency makeup air i filter train; and a 4 inch bed for the emergency recirculation air (flitertrain.

FERMI - UNIT 2 3/4 7 9 Amendment No. EJ. ES, 110 PAGE b 0F 03 Rev 6

INSERT THIS PAGE IN FRONT OF VOLUME 12 Volume 12: IMPROVED TECHNICAL SPECIFICATIONS Remove Replace 1.1 ITS pg 1.1-3 Rev 0 1.1 ITS pg 1.13 Rev 13 3.3.3.2 ITS pg 3.3-30 Rev 0 3.3.3.2 ITS pg 3.3-30 Rev 13 3.3.6.3 ITS pg 3.3-65 Rev 6 3.3.6.3 ITS pg 3.3-65 Rev 13 3.3.6.3 ITS pg 3.3-67 Rev 6 3.3.6.3 ITS pg 3.3-67 Rev 13 3.4.1 ITS pg 3.4-1 Rev 10 3.4.1 ITS pg 3.4-1 Rev 13 3.4.1 ITS pg 3.4-2 Rev 2 3.4.1 ITS pg 3.4-2 Rev 13 3.4.1 ITS pg 3.4-3 Rev 2 3.4.1 ITS pg 3.4-3 Rev 13 3.4.1 ITS pg 3.4-4 Rev 13 3.6.1.3 ITS pg 3.6-11 Rev i1 3.6.1.3 ITS pg 3.6-11 Rev 13 3.6.1.3 ITS pg 3.6-12 Rev 11 3.6.1.3 ITS pg 3.6-12 Rev 13 3.6.1.3 ITS pg 3.613 Rev Sa 3.6.1.3 ITS pg 3.6-13 Rev 13 3.6.1.3 ITS pg 3.6-13a Rev Sa 3.6.1.3 ITS pg 3.6-14 Rev 13 3.61.3 ITS pg 3.6-15 Rev Sa 3.6.1.3 ITS pg 3.6-15 Rev 13 3.6.1.3 ITS pg 3.6-16 Rev 13 3.6.1.3 ITS pg 3.6-17 Rev 13 3.6.1.8 ITS pg 3.6-21 Rev 0 3.6.1.8 ITS pg 3.6-2i Rev 13 l 3.7.4 ITS pg 3.713 Rev 0 3.7.4 ITS pg 3.713 Rev 13 3.9.8 ITS pg 3.9-12 Rev 4 3.9.8 ITS pg 3.9-12 Rev i3 3.9.8 ITS pg 3.9-14 Rev 0 3.9.8 ITS pg 3.9-14 Rev 13 3.10.5 ITS pc 3.10-13 Rev 0 3.10.5 ITS pg 3.10-13 Rev 13 3.10.7 ITS pg 3.10-18 Rev 0 3.10.7 ITS pg 3.10-18 Rev 13 3.10.7 ITS pg 3.10-20 Rev 0 3.10.7 ITS pg 3.10-20 Rev 13 5.2 ITS pg 5.0-3 Rev 0 5.2 ITS pg 5.0-3 Rev 13 l

l l

l Rev 13 08/06/99

Definitions 1.1

) 1.1 Definitions (continued)

CORE OPERATING LIMITS The COLR is the unit specific document that REPORT (COLR) provides cycle specific parameter limits for the current reload cycle. These cycle specific limits ,

shall be determined for each reload cycle in J accordance with Specification 5.6.5. Plant )

operation within these limits is addressed in individual Specifications.

DOSE EQUIVALENT I-131 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries/ gram) that alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131. I-132, 1 133. 1-134, and 1 135 actually present. The thyroid dose I conversion factors used for this calculation shall

@ be those listed in Table III of TID-14844 AEC.1962, " Calculation of Distance Factors for Power and Test Reactor Sites."

EMERGENCY CORE COOLING The ECCS RESPONSE TIME shall be that time interval SYSTEM (ECCS) RESPONSE from when the monitored )arameter exceeds its ECCS TIME initiation setpoint at tw channel sensor until !

l the ECCS equipment is capable of performing its I safety function (i.e., the valves travel to their 1 required positions, pump discharge pressures reach I their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may !

be measured by means of any series of sequential. l overlapping, or total steps so that the entire response time is measured.

ISOLATION SYSTEM The ISOLATION SYSTEM RESPONSE TIME shall be that RESPONSE TIME time interval from when the monitored parameter exceeds its isolation initiation setpoint at the channel sensor until the isolation valves travel to their required positions. Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential.

overlapping, or total steps so that the entire response time is measured.

I

~

(continued) l FERMI - UNIT 2 1.1-3 Revision 13. 08/06/99

Remote Shutdown System 3.3.3.2 m

Table 3.3.3.21 (page 1 of 1)

Remote Shutdown System Instrumentation l INSTRUMENT FUNCTION

1. Reactor Vessel Pressure

2. Reactor Vessel Water Level

3. Suppression Chamber Water Temperature

4. Drywell Pressure

5. RHR Heat Exchanger Discharge Flow

6. RCIC Flow l CONTROL FUNCTION

1. Control Rod Drive Pump A

2. Control Rod Drive Pump B

3. RHR Valve E1150 F009

4. RHR Valve E1150 F008

5. RHR Valve E1150 F006A

6. Recirc Pump A Valve B3105 F023A

7. Main Steam Line (D) Relief Valve B2104 F013A

8. Main Steam Line (C) Relief Valve B2104 F0138

9. RHR Valve E1150 F015A

10. RHR Valve E1150 F017A

11. RHR Valve E1150 F004A

12. RHR Pump A

13. RHR Valve E1150-F024A

14. RHR Valve E1150 F023A

15. RHR Valve E1150 F048 A

16. RHR Valve E1150 F06fA

17. RHR Service Water Pump A

18. RHR Service: Water Pump C

19. Cooling Tower Fan A

20. Cooling Tower Fan C

21. RCIC Valve E5150 F059

22. RCIC Valve E5150 F045

23. RCIC Initiate

24. Division II DC Transfer

25. B0P Transfer

26. Division I DC Transfer

27. Division I AC Transfer

28. Swing Bus Transfer l

1 l.

! FERMI - UNIT 2 3.3 30 Revision 13. 08/06/99

[

l 1

l LLS Instrumentation 3.3.6.3 3.3 INSTRUMENTATION 3.3.6.3 Low Low Set (LLS) Instrumentation LC0 3.3.6.3 The LLS valve instrumentation for each Function in Table 3.3.6.3-1 shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One LLS valve A.1 Restore channel (s) to 14 days inoperable due to OPERABLE status, inoperable channel (s).

B. -- - -- -NOTE -- ----- B.1 Restore one tailpipe 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

, Separate Condition pressure switch for

,, entry is allowed for 11 OPERABLE SRVs, each SRV. including one in each

@, / ...................... Division for an '

OPERABLE SRV in the One or more safety / lowest setpoint t<s relief valves (SRVs) group, to OPERABLE I with one or more status.

d Function 3 channel (s) inoperable. AND

............N0TE - - -- --

LC0 3.0.4 is not applicable.

B.2 Restore both tailpipe Prior to pressure switches for entering H0DE 2 G 11 OPERABLE SRVs.

including 4 of 5 or 3 from H0DE 4 OPERABLE SRVs with the lowest relief l setpoints, to OPERABLE status.

(continued) l FERMI - UNIT 2 3.3 65 Revision 13, 08/06/99

Recirculation Loops Operating 3.4.1 l

3.4 REACTUR COOLANT SYSTEM (RCS) 3.4.1 Recirculation Loops Operating f

LC0 3.4.1 The reactor core shall not exhibit core thermal hydraulic instability or operate in the " Scram" or " Exit" Regions. l rl M

a. Two recirculation looa's with matched recirculation loop D jet pump flows shall ae in operation:

W L gg

b. One recirculation loop may be in operation provided the following limits are applied when the associated LC0 is g

applicable:

1. LC0 3.2.1. " AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)." single loop operation limits specified in the COLR:

2. LC0 3.2.2. " MINIMUM CRITICAL POWER RATIO (MCPR)."

single loop operation limits specified in the COLR:

1

! 3. LCO 3.3.1.1. " Reactor Protection System (RPS) l k

,4 Instrumentation." Function 2.b (Average Power Range Monitors Simulated Thermal Power-Upscale) Allowable o Value of Table 3.3.1.11 is reset for single loop q operation, when in MODE 1: and

4. THERMAL POWER is s 67.2% RTP.

............................N0TE -- - -- - - - -

ll Application of the required limitations for single loop operation may be delayed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after transition from two recirculation loop operations to single recirculation loop operation, l

............................................................ l 1

1 l

APPLICABILITY: MODES I and 2. l l

l l FERMI UNIT 2 3.4-1 Revision 13 08/06/99 l

o

Recirculation Loops Operating 3.4.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME l

A. Recirculation jet pump A.1 Declare recirculation 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months I loop flow mismatch not loop with lower flow:

within limits. "not in operation."

B. Reactor core o wrating ......... .-N0TE..-..........

g in the " Exit"ilegion. Restart of an idle recirculation loop or resetting a recirculation m flow limiter is not allowed. '

N B.1 Initiate action to Immediately insert control rods j l yl or increase core flow to restore operation i outside the " Exit" l \l 1 Region.

l l

i l C. No recirculation loops- C.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months operating while in MODE 2.

(continued) l l

l j FERMI UNIT 2 3.4 2 Revision 13. 08/06/99 V

i l

Recirculation Loops Operating 3.4.1 s

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. No recirculation loops D.1 Place the reactor Immediately operating while in mode switch in the MODE 1. shutdown position.

2 Reactor core operating Gl in the " Scram" Region.

d a

Core thermal hydraulic (l instability evidenced.

I SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.1.1 -- -

-- NOTE - -- - - - -

Only re uired to be performed when operati g in the " Stability Awareness" 3lN Region.

........................................... l Verify the reactor core is not exhibiting 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months core thermal hydraulic instability.

(continued) !

g

Recirculation Loops Operating 3.4.1 SURVEILLANCE FREQUENCY SR 3.4.1.2 -- - -. ....-

NOTE -- - - -

, Not required to be erformed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months l after both recircul tion loops are in l operation.

Verify recirculation loop jet pg flow 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months mismatch with both recirculation loops in operation is:

a. s 10% of rated core flow when operating at < 70% of rated core flow; and

b. s 5% of rated core flow when operating at = 70% of rated core flow.

1 i

f(l FERMI UNIT 2 3.4 4 Revision 13. 08/06/99 l

I PCIVs 3.6.1.3

) ACTIONS (continued) i CONDITION REQUIRED ACTION COMPLETION TIME D. One or more D.1 Restore leakage rates 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for j penetration flow 3aths to within limit. leakage on with one or more SCIVs hydrostatically I N inoperable due to tested line !

secondary containment without a closed O bypass leakage rate, system MSIV leakage rate, T

b purge valve leakage E rate, hydrostatically l tested line leakage 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for rate, or EFCV leakage secondary f

( rate not within limit. containment bypass leakage M

8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months for MSIV nlg leakage I,p 1 M

h t 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for

% purge valve leakage M

O -

72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months for leakage on p hydrostatically l w tested line on a d closed system and EFCV leakage i

M

(

ig l (continued) i l

l .

l FERMI UNIT 2 3.6-11 Revision 13, 08/06/99

I i

l PCIVs 3.6.1.3 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. (continued) . - - ----- NOTES -- - -- - -

l 1. Isolation devices in high ,

I radiation areas may be verified by use of administrative means.

2. Isolation devices that I

are locked, sealed, or h otherwise secured may be verified by use of administrative means. <

3. Only applicable to penetration flow paths isolated to restore leakage to within limits.

D.2 Verify the affected Once per 31 days penetration flow path for isolation is isolated. devices outside primary containment 8!Q Prior to entering MODE 2 or 3 from MODE 4, if primary containment was de inerted while in MODE 4. if I

not performed within the previous 92 days, for isolation devices inside primary containment (continued) l FERMI' UNIT 2 3.6 12 Revision 13 08/06/99

I PCIVs 3.6.1.3 '

ACTIONS (continued) i CONDITION REQUIRED ACTION COMPLETION TIME i

ll E. Required Action and associated Completion E.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

, Time of Condition A. 8@

n B. C. or D not met in g MODE 1. 2. or 3. E.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months b

l Dl F. Required Action and associated Completion F.1 Initiate action to isolate RHR Shutdown Immediately 1 Time of Condition A. Cooling System.

l B. C. or D not met for l RfR SDC PCIV(s) @

required to be il OPERABLE during MODE 4 F.2 Initiate action to Immediately or 5. restore valve (s) to OPERABLE status.

l w*

k FERMI UNIT 2 3.6 13 Revision 13. 08/06/99

n PCIVs 3.6.1.3 l

l l ) SURVEILLANCE REQUIREMENTS 1

SURVEILLANCE FREQUENCY SR 3.6.1.3.1 --------------.-- NOTE-- - - ----- - --

Qs Not required to be met when the isolation valves for one purge or containment pressure control supply line and one purge or containment pressure control exhaust line are open for inerting, de-

, V inerting, pressure control. ALARA or air quality considerations for personnel entry, or Surveillances that require the valves to be open.

d Verify each drywell and suppression chamber purge system and containment 31 days pressure control isolation valve is closed.

SR 3.6.1.3.2 --- -

- --.-..- NOTES- --- -- --- --- -

1. Valves and blind flanges in high radiation areas may be verified by l

use of administrative means.

2. Not required to be met for PCIVs that are open under administrative controls.

m Verify each primary containment isolation 31 days t manual valve and blind flange that is Us' located outside primary containment and

& is not locked, sealed or otherwise b secured and is requi.ed to be closed V during accident conditions is closed.

(continued) l l FERMI - UNIT 2 3.6 14 Revision 13. 08/06/99

F~

PCIVs 3.6.1.3

' ') SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.1.3.3 -------- - - --NOTES --- ------ ----

1. Valves and blind flanges in high radiation areas may be verified by use of administrative means.

2. Not required to be met for PCIVs that are open under administrative controls.

k '

Verify each primary containment isolation Prior to n - manual valve and blind flange that is entering MODE 2 located inside primary containment and is or 3 from b< not locked. sealed. or otherwise secured MODE 4 if Pa and is required to be closed during primary t accident conditions is closed. containment was L de-inerted while in MODE 4. if not performed within the previous 92 days SR 3.6.1.3.4 Verify continuity of the traversing 31 days incore probe (TIP) shear isolation valve explosive charge.

A SR 3.6.1.3.5 Verify the isolation time of each power In accordance dl operated automatic PCIV, except for. with the h MSIVs. is within limits. Inservice C( Testing Program (continued) l FERMI - Uti1T 2 3.6 15 Revision 13. 08/06/99

L PCIVs l 3.6.1.3 i SURVEILLANCE REQUIREENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.1.3.6 Perform leakage rate testing for each 184 days primary containment purge valve with resilient seals. AN_Q Once within 92 days after opening the valve SR 3.6.1.3.7 Verify the isolation time of each MSIV is In accordance

= 3 seconds and s 5 seconds. with the Inservice Testing Program SR 3.6.1.3.8 Verify each automatic PCIV actuates to 18 months the isolation position on an actual or simulated isolation signal.

SR 3.6.1.3.9 Verify each reactor instrumentation line 18 months EFCV actuates on a simulated instrument line break to restrict flow.

SR 3.6.1.3.10 Remove and test the explosive squib from 18 months on a each shear isolation valve of the TIP STAGGERED TEST System. BASIS (continued) i I

l FERMI - UNIT 2 3.6 16 Revision 13. 08/06/99 V

PCIVs 3.6.1.3

) SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY O

j SR 3.6.1.3.11 Verify the combined leakage rate for all In accordance l secondary containment bypass leakage with the Y paths that are not provided with a seal Primary system is s 0.04 L, when pressurized to Containment

= 56.5 psig. Leakage Rate

& Testing Program and Inservice fl Testing Program e

SR 3.6.1.3.12 Verify combined MSIV leakage rate for all In accordance four main steam lines is s 100 scfh when with the tested at = 25 psig. Primary Containment Leakage Rate Testing Program SR 3.6.1.3.13 ---- ----- -

---NOTE --- -- -- - --

l Only required to be met in MODES 1. 2 and 3.

n .........................................

d 1 hl Verify combined leakage rate through In accordance hydrostatically tested lines that with the )

penetrate the primary containment is Primary l within limits. Containment Leakage Rate Testing Program l FERMI - UNIT 2 3.6 17 Revision 13. 08/06/99

1 l Suppression Chamber-to Drywell Vacuum Breakers 3.6.1.8 l

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.8.1 NOTES - ---------------

1. Not required to be met for vacuum breakers that are open during I Surveillances. I

2. Not required to be met for vacuum i breakers open when performing their l intended function. !

Verify each vacuum breaker is closed. 7 days l SR 3.6.1.8.2 Perform a functional test of each vacuum Prior to breaker. entering MODE 2 or 3 from H0DE 4 if not j performed in the previous 92 days MD Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after any i discharge of steam to the suppression chamber from I the SRVs SR 3.6.1.8.3 Verify the opening setpoint of each 18 months vacuum breaker is s 0.5 psid.

l FERMI - UNIT 2 3.6 21 Revision 13. 08/06/99 i

3 l

l Control Center AC System 3.7.4 {

ACTIONS (continued) l CONDITION REQUIRED ACTION COMPLETION TIME 4 E. Two control center AC ------ --- NOTE --- -- - --

subsystems inoperable LC0 3.0.3 is not applicable.

during movement of ---------- ----- - -- -- - --

irradiated fuel assemblies in the E.1 Suspend movement of Immediately secondary containment. irradiated fuel during CORE assemblies in the }

ALTERATIONS or during secondary j OPDRVs. containment.

ANQ E.2 Suspend CORE Immediately ALTERATIONS.

AND E.3 Initiate actions to Immediately suspend OPDRVs. .

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.4.1 Verify the control room air temperature is 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months s 95'F. i 1

1 l FERMI - UNIT 2 3.7 13 Revision 13. 08/06/99 l I

l

l I

i I

RHR-Low Water Level 3.9.8 3.9 REFUELING OPERATIONS i

3.9.8 Residual Heat Removal (RHR)-Low Water Level l

{

h l LCO 3.9.8 Two RHR shutdown cooling subsystems shall be OPERABLE, and, with no recirculation pump in operation, one RHR shutdown cooling subsystem shall be in operation.

{

............................N0TES -- .-- ---- ---- -- - l

1. The required operating RHR shutdown cooling subsystem I may be removed from operation for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months per !

8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period. !

I

2. One RHR shutdown cooling subsystem may be inoperable i h ..... $5.$.!S. . $$!. $. .$$ $$!"!$.$[. $!!...!"!!!'.

9 APPLICABILITY: MODE 5 with irradiated fuel in the reactor pressure vessel (RPV). the water level < 20 ft 6 inches above the top of I L

h the RPV flange, and heat losses to ambient not greater !

than or equal to heat input to reactor coolant. l ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME

)

A. One or two required A.1 Verify an alternate 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months RHR shutdown cooling method of decay heat subsystems inoperable. removal is available 6ND for each inoperable required RHR shutdown Once per cooling subsystem. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter (continued) l FERMI UNIT 2 3.9-12 Revision 13 08/06/99 l l

I RHR-Low Water Level 3.9.8 l

SURVEILLANCE REQUIREMENTS !

SURVEILLANCE FREQUENCY l

l SR 3.9.8.1 Verify one RHR shutdown cooling subsystem 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months pl or recirculation pump is operating.

i l

l SR 3.9.8.2 Verify each RHR shutdown cooling subsystem 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months l 1s capable of decay heat removal.

l l FERMI - UNIT 2 3.9 14 Revision 13. 08/06/99

Single CRD Removal-Refueling 3.10.5 3.10 SPECIAL OPERATIONS 3.10.5 Single Control Rod Drive (CRD) Removal-Refueling LC0 3.10.5 The requirements of LC0 3.3.1.1. " Reactor Protection System (RPS) Instrumentation": LCO 3.3.8.2. " Reactor Protection System (RPS) Electric Power Monitoring": and LC0 3.9.5

" Control Rod OPERABILITY-Refueling." may be suspended in MODE 5 to allow withdrawal of a single control rod, and subsequent removal of the associated CRD from a core cell containing one or more fuel assemblies, provided the following requirements are met:

4

a. All other control rods are fully inserted: and i

b. All other control rods in a five by five array centered on the withdrawn control rod are disarmed: at which time LCO 3.1.1. " SHUTDOWN MARdIN (SDM)." MODE 5 recuirements f may be changed to allow the single control roc withdrawn to be assumed to be the highest worth control rod.

AND In conjunction with a. and b. above. the re LC0 3.9.1, " Refueling Equipment Interlocks"quirements

LC0 3.9.2 of

" Refuel Position One Rod Out Interlock": and LC0 3.9.4

" Control Rod Position Indication": may be suspended provided the following requirements are met:

c. No other CORE ALTERATIONS are in progress: and I d. A control rod withdrawal block is inserted.

APPLICABILITY: MODE 5 with LC0 3.9.5 not met.

[ FERMI UNIT 2 3.I0 13 Revision 13 08/06/99

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t

i 1

l SDM Test-Refueling 3.10.7 3.10 SPECIAL OPERATIONS 3.10.7 SHUTDOWN MARGIN (SDM) Test-Refueling l

i l LC0 3.10.7 The reactor mode switch position specified in Table 1.1-1 for MODE 5 may be changed to include the startup/ hot standby position, and operation considered not to be in MODE 2. to allow SDM testing, provided the following requirements are j met:

a. LC0 3.3.1.1. " Reactor Protection System 3 Instrumentation." MODE 2 requirements for Functions 2.a.

F 2.d. and 2.e of Table 3.3.1.1-1:

l

b. 1. LC0 3.3.2.1 " Control Rod Block Instrumentation."

MODE 2 requirements for Function 2 of Table 3.3.2.1-1. with the prescribed withdrawal l sequence requirements of SR 3.3.2.1.7 changed to l require the control rod sequence to conform to the SDM test sequence.

2

2. Conformance to the approved control rod sequence for i the SDM test is verified by a second licensed i

operator or other qualified member of the technical staff:

l

c. Each withdrawn control rod shall be coupled to the associated CRD:

d. All control rod withdrawals during local critical l

testing shall be made in notch out mode: l

e. No other CORE ALTERATIONS are in progress; and j

f. CRD charging water header pressure a 940 psig.

i l

APPLICABILITY: MODE 5 with the reactor mode switch in startup/ hot standby position.

l l

l l FERMI - UNIT 2 3.10 18 Revision 13. 08/06/99

1

)

SDM Test-Refueling 3.10.7

. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY

. SR 3.10.7.1 Perform the MODE 2 applicable SRs for LC0 According to

[1)

W l 3.3.1.1. Functions 2.a. 2.d. and 2.e of the applicable Table 3.3.1.1 1. SRs I

SR 3.10.7.2 -----

-- NOTE-- - ---------- -

i Not required to be met if SR 3.10.7.3 satisfied.

1 Perform the MODE 2 applicable SRs for According to LC0 3.3.2.1. Function 2 of Table 3.3.2.1-1. the applicable SRs ;

I l

I SR 3.10.7.3 -

- - - -------- NOTE- - - - --- - -- -

Not required to be met if SR 3.10.7.2 satisfied.

Verify movement of control rods is in During control compliance with the approved control rod rod movement sequence for the SDM test by a second licensed o>erator or other qualified member of the tec1nical staff.

SR 3.10.7.4 Verify no other CORE ALTERATIONS are in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months progress.

(continued) t l l FERMI UNIT 2 3.10 20 Revision 13. 08/06/99

F~

Organization 5.2 5.2 Organization (continued) 5.2.2 Unit Staff The unit staff organization shall include the following:

a. At least two non licensed operators shall be assigned while Jr operating in MODE 1. 2. or 3 and at least one non licensed operator shall be assigned whenever the reactor contains fuel.

b. At least one licensed Reactor Operator (RO) shall be present in the control room when fuel is in the reactor. In addition, while the unit is in MODE 1, 2. or 3. at least one licensed Senior Reactor Operator (SRO) shall be present in the control room.

c. Shift crew composition may be less than the minimum requirement of 10 CFR 50.54(m)(2)(1) and 5.2.2.a and 5.2.2.9 for a period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in order to accommodate unexpected absence of on duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements.

d. A Radiation Protection Technician shall be on site when fuel is in the reactor. The position may be vacant for not more than 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , in order to provide for unexpected absence.

provided immediate action is taken to fill the required position.

e. Administrative procedures shall be developed and implemented to limit the working hours of unit staff who perform safety related functions (e.g., licensed SR0s. licensed R0s. l radiation protection technicians, auxiliary operators, and i key maintenance personnel). The controls shall include l guidelines on working hours that ensure that adequate shift l coverage is maintained without routine heavy use of overtime for individuals.

' Any deviation from the established guidelines shall be authorized in advance by the Plant Manager or designee. in accordance with approved administrative procedures, or by higher levels of management, in accordance with established procedures and with documentation of the basis for granting the deviation.

(continued) l FERMI - UNIT 2 5.0 3 Revision 13 08/06/99 1

i INSERT THIS PAGE IN FRONT OF VOLUME 13 l

l Volume 13: IMPROVED TECHNICAL SPECIFICATIONS BASES Remove Replace B 3.1.2 ITS pg B 3.1.2-2 Rev 0 B 3.1.2 ITS pg B 3.1.2-2 Rev 13 B 3.1.7 ITS pg B 3.1.7-6 Rev 0 B 3.1.7 ITS pg B 3.1.7-6 Rev 13 B 3.1.7 ITS pg B 3.1.7-7 Rev 13 l B 3.3.1.1 ITS pg B 3.3.1.1-24 Rev 0 B 3.3.1.1 ITS pg B 3.3.1.1-24 Rev 13 B 3.4.1 ITS pg B 3.4.1-3 Rev 2 B 3.4.1 ITS pg B 3.4.1-3 Rev 13 B 3.4.1 ITS pg B 3.4.1-4 Rev 10 B 3.4.1 ITS pg B 3.4.1-4 Rev 13 B 3.4.1 ITS pg B 3.4.1-5 Rev 0 B 3.4.1 ITS pg B 3.4.1-5 Rev 13 B 3.4.1 ITS pg B 3.4.1-6 Rev 2 B 3.4.1 ITS pg B 3.4.1-6 Rev 13 ,

B 3.4.1 ITS pg B 3.4.1-7 Rev 2 B 3.4.1 ITS pg B 3.4.1-7 Rev 13 B 3.4.1 ITS pg B 3.4.1-8 Rev 2 B 3.4.1 ITS pg B 3.4.1-8 Rev 13 B 3.4.1 ITS pg B 3.4.1-9 Rev 2 B 3.4.1 ITS pg B 3.4.1-9 Rev 13 l B 3.4.1 ITS pg B 3.4.1-10 Rev 2 B 3.4.1 ITS pg B 3.4.1-10 Rev 13 l

B 3.6.1.3 ITS pg B 3.6.1.3-9 Rev i1 B 3.6.1.3 ITS pg B 3.6.1.3-9 Rev 13 B 3.6.1.3 ITS pg B 3.6.1.3-10 Rev Sa B 3.6.1.3 ITS pg B 3.6.1.310 Rev 13 B 3.6.1.3 ITS pg B 3.6.1.3-11 Rev Sa B 3.6.1.3 ITS pg B 3.6.1.3-11 Rev 13 B 3.6.1.3 ITS pg B 3.6.1.3-12 Rev Sa B 3.6.1.3 ITS pg B 3.6.1.3-12 Rev 13 B 3.6.1.3 ITS pg B 3.6.1.3-13 Rev Sa B 3.6.1.3 ITS pg B 3.6.1.3-13 Rev 13 B 3.6.1.3 ITS pg B 3.6.1.3-14 Rev Sa B 3.6.1.3 ITS pg B 3.6.1.3-14 Rev 13 B 3.6.1.3 ITS pg B 3.6.1.3-15 Rev Sa B 3.6.1.3 ITS pg B 3.6.1.3-15 Rev 13 B 3.6.1.3 ITS pg B 3.6.1.3-16 Rev Sa B 3.6.1.3 ITS pg B 3.6.1.316 Rev 13 l B 3.6.1.3 ITS pg B 3.6.1.3-17 Rev Sa B 3.6.1.3 ITS pg B 3.6.1.3-17 Rev 13 B 3.6.1.3 ITS pg B 3.6.1.3-18 Rev 13 B 3.6.1.8 ITS pg B 3.6.1.8-3 Rev i1 B 3.6.1.8 ITS pg B 3.6.1.8-3 Rev 13 B 3.6.1.8 ITS pg B 3.6.1.8-5 Rev 5 B 3.6.1.8 ITS pg B 3.6.1.8 5 Rev 13 B 3.6.4.2 ITS pg B 3.6.4.2-2 Rev 1i B 3.6.4.2 ITS pg B 3.6.4.2-2 Rev 13 B 3.7.4 ITS pg B 3.7.4-5 Rev 0 B 3.7.4 ITS pg B 3.7.4-5 Rev 13 B 3.10.7 ITS pg B 3.10.7-5 Rev 3 B 3.10.7 ITS pg B 3.10.7-5 Rev 13 l

l Rev 13 08/06/99 l

Reactivity Anomalies B 3.1.2 l l l BASES l

BACKGROUND (continued) l l The predicted core reactivity is calculated by a 3D core simulator code as a function of cycle exposure. This calculation is performed for projected operating states and conditions throughout the cycle. The core reactivity is determined from actual plant conditions and is then compared to the predicted value for the cycle exposure.

1 APPLICABLE Accurate prediction of core reactivity is either an explicit SAFETY ANALYSES or implicit assumption in the accident analysis evaluations (Ref. 2). In particular. SDM and reactivity transients.

such as control rod withdrawal accidents or rod drop accidents, are very sensitive to accurate prediction of core l

reactivity. These accident analysis evaluations rely on !

computer codes that have been qualified against available i test data, operating plant data, and analytical benchmarks. !

Monitoring reactivity anomaly provides additional assurance that the nuclear methods provide an accurate representation of the core reactivity.

The comparison between measured and predicted initial core reactivity provides a normalization for the calculational models used to predict core reactivity. If the measured and predicted reactivity for identical core conditions at B0C do ;

not reasonably agree, then the assumptions used in the i reload cycle design analysis or the calculation models used to predict reactivity may not be accurate. If reasonable agreement between measured and predicted core reactivity exists at BOC. then the prediction may be normalized to the measured value. Thereafter, any significant deviations in the measured reactivity from the predicted reactivity that develop during fuel depletion may be an indication that the assumptions of the DBA and transient analyses are no longer valid, or that an unexpected change in core conditions has occurred.

Reactivity anomalies satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LC0 The reactivity anomaly limit is established to ensure plant operation is maintained within the assumptions of the safety analyses. Large differences between monitored and predicted l FERMI UNIT 2 B 3.1.2 - 2 Revision 13 08/06/99

[

SLC System B 3.1.7

) BASES SURVEILLANCE REQUIREENTS (continued) tested every 36 months at alternating 18 month intervals.

The Surveillance may be performed in separate steps to prevent injecting boron into the RPV. An acceptable method for verifying flow from the pump to the RPV is to pump demineralized water from a test tank through one SLC subsystem and into the RPV. The 18 month Frequency is based on the need to perform this Surveillance under the l conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown these components usually pass the Surveillance when performed at the 18 month Frequency:

.therefore, the Frecuency was concluded to be acceptable from a reliability stancpoint.

Demonstrating that all piping between the boron solution storage tank and the explosive valve is unblocked ensures that there is a functioning flow path for injecting the sodium pentaborate solution. An acceptable method for g verifying that the suction piping is unblocked is to pump from the storage tank to the test tank (this is followed by draining and flushing the piping with demineralized water).

The 18 month Frequency is acceptable since there is a low probability that the subject piping will be blocked due to precipitation of the boron from solution in the piping. .

This is especially true in light of the temperature !'

verification of this piping required by SR 3.1.7.3.

' However, if, in performing SR 3.1.7.3. it is determined that the temperature of this piping has fallen below the specified minimum. SR 3.1.7.9 must be performed once within l 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the piping temperature is restored to a 48'F.

SR 3.1.7.10 Enriched sodium pentaborate solution is made by mixing granular, enriched sodium pentaborate with water. Isotopic )

tests on the granular sodium pentaborate to verify the actual B 10 enrichment must be performed prior to addition to the SLC tank in order to ensure that the proper B 10 atom percentage is being used.

i l FERMI UNIT 2 B 3.1.7 - 6 Revision 13. 08/06/99 a

SLC System B 3.1.7 BASES-REFERENCES 1. 10 CFR.50.62.

2. UFSAR. Section 4.5.2.4.

l l

t r

f.

g l FERMI UNIT 2 B 3.1.7 - 7 Revision 13. 08/06/99 i

RPS Instrumentation ,

B 3.3.1.1 l 1

l BASES ACTIONS (continued)

[L1 Required Action D.1 directs entry into the appropriate Condition referenced in Table 3.3.1.1-1. The applicable Condition specified in the Table is Function and MODE or other specified condition dependent and may change as the Required Action of a previous Condition is completed. Each time an inoperable channel has not met any Required Action of Condition A B, or C and the associated Com)letion Time has expired. Condition D will be entered for t1at channel and provides for transfer to the appropriate subsequent Condition.

E.1. F.1. G.I. H.1. and H.2 If the channel (s) is not restored to OPERABLE status or placed in trip (or the associated trip system placed in trip) within the allowed Completion Time, the plant must be

) laced in a MODE or other specified condition in which the l _C0 does not apply. Alternately, for Condition H, the MSLs may be isolated (Required Action H.1), and, if allowed l (i.e.. plant safety analysis and minimal steam flow in MODE 2 allows operation with the MSLs isolated), operation :

with the MSLs isolated may continue. Isolating the MSLs conservatively accomplishes the safety function of the l i

inoperable channel. The allowed Completion Times are I reasonable, based on operating experience, to reach the '

specified condition from full power conditions in an orderly manner and without challenging plant systems. In addition, the Completion Time of Required Action E.1 is consistent with the Completion Time provided in LC0 3.2.2. " MINIMUM CRITICAL POWER RATIO (MCPR)."

i l FERMI - UNIT 2 B 3.3.1.1 - 24 Revision 13 08/06/99

Recirculation Loops Operating B 3.4.1

) BASES APPLICABLE The operation of the Reactor Coolant Recirculation System is SAFETY ANALYSES an initial condition assumed in the design basis loss of coolant &ccident (LOCA) (Ref.1). During a LOCA caused by a recirculation loop pipe break, the intact loop is assumed to provide coolant flow during the first few seconds of the l accident. The initial core flow decrease is rapid because l

the recirculation pump in the broken loop ceases to ) ump reactor coolant to the vessel almost immediately. T1e pump in the intact loop coasts down relatively slowly. This pump coastdown governs the core flow response for the next several seconds until the jet pump suction is uncovered.

The analyses assume that both loops are operating at the same flow prior to the accident. However, the LOCA analysis was reviewed for the case with a flow mismatch between the I two loops, with the pi with the higher flow. peWhile breakthe assumed to be inand the core loop flow coastdown response are potentially more severe in this assumed case (since the intact loop starts at a lower flow rate and the core response is the same as if both loops were operating at a lower flow rate), a small mismatch has been determined to be acceptable based on engineering judgement. The recirculation system is also assumed to have sufficient flow coastdown characteristics to maintain fuel thermal margins during abnormal operational transients (Ref. 2), which are analyzed in Chapter 15 of the UFSAR.

A plant specific LOCA analysis has been performed assuming <

only one operating recirculation loop. This analysis has demonstrated that, in the event of a LOCA caused by a pipe break in the operating recirculation loop, the Emergency Core Cooling System response will provide adequate core cooling provided the APLHGR requirements are modified accordingly (Ref. 3).

The transient analyses of Chapter 15 of the UFSAR have also 1 been performed for single recirculation loop operation (Ref. 3) and demonstrate sufficient flow coastdown characteristics to maintain fuel thermal margins during the p abnormal operational transients analyzed provided the MCPR requirements are modified. During single recirculation loop operation, modification to the Reactor Protection System (RPS) average power range monitor (APRM) instrument setpoints is also required to account for the different relationships between recirculation drive flow and reactor I. FERMI - UNIT 2 B 3.4.1 -3 Revision 13, 08/06/99

Recirculation Loops Operating B 3.4.1 BASES APPLICABLE SAFETY ANALYSIS (continued)

I core flow. The APLHGR and MCPR setpoints for single loop o>eration are specified in the COLR. The APRM Simulated T1ermal Power - Upscale setpoint is in LC0 3.3.1.1. " Reactor l Protection System (RPS) Instrumentation."

Thermal-hydraulic stability analysis (Ref. 5) has concluded that procedures for detecting and suppressing power oscillations that might be induced by a thermal hydraulic instability are necessary to provide reasonable assurance that the requirements of Reference 4 are satisfied.

Recirculation loops operating satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). '

LC0 Two recirculation loops are required to be in operation with their flows matched within the limits specified in SR 3.4.1.2 to ensure that during a LOCA caused by a creak of the piping of one recirculation loop the assumptions of the LOCA analysis are satisfied. With the limits specified in SR 3.4.1.2 not met, the recirculation loop with the lower flow must be considered not in operation. With only one

recirculation loop in operation, modifications to the required APLHGR limits (LC0 3.2.1. " AVERAGE PLANAR LINEAR i HEAT GENERATION RATE (APLHGR)"). MCPR limits (LC0 3.2.2. I m " MINIMUM CRITICAL POWER RATIO (MCPR)"). APRM Simulated R. Thermal Power-Upscale setpoint (LCO 3.3.1.1) and limitation 2 on liiERMAL POWER may be applied to allow continued operation 6 consistent with the assumptions of the safety analysis.

p) Operations that exhibit core thermal hydraulic instability m are not permitted. Additionally, in order to avoid N potential power oscillations due to thermal-hydraulic g instability, operation at certain combinations of power and flow are not permitted. These restricted power and flow k*' regions are referred to as the " Scram" and " Exit" regions C and are defined by Bases Figure B 3.4.1-1.

i I

A Note is provided to allow 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months following the transition to single loop operation from two loop operation to establish the applicable limitations in accordance with the single loop analysis. The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months period is sufficient to make the adjustment given the relatively small change required. This transition only results in applying the new single-loop allowable values to APRM OPERABILITY. Any ARPM l

l FERMI - UNIT 2 B 3.4.1 -4 Revision 13. 08/06/99

l l

Recirculation Loops Operating B 3.4.1 l

BASES LC0 (continued) non compliance with the required allowable value after this 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months allowance, results in ACTIONS of LC0 3.3.1.1 being entered: no ACTION of LCO 3.4.1 would apply. Similarly, any operation with APLHGR or MCPR out of limits results in the ACTIONS of LCO 3.2.1 or LCO 3.2.2 being entered: no ACTION of LCO 3.4.1 would apply.

APPLICABILITY In MODES 1 and 2, requirements for operation of the Reactor Coolant Recirculation System are necessary since there is considerable energy in the reactor core and the limiting design basis transients and accidents are assumed to occur.

In MODES 3, 4, and 5. the consequences of an accident are I reduced and the coastdown characteristics of the recirculation loops are not important. In addition, sufficient power to create power oscillations that threaten fuel design limits does not exist.

l ACTIONS A_J With the requirements for matched recirculation loop flow not met, the recirculation loops must be restored to l operation with matched flows within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or the loop with the lower flow must be considered not in operation. Should a LOCA occur with one recirculation loop at a significantly lower flow than the other loop, the core flow coastdown and ,

l resultant core response may not be bounded by the LOCA analyses. Therefore, only a limited time is allowed to restore the loop to operating status.

Alternatively, if the single loop requirements of the LC0 are applied to RPS setpoints, operation with only one recirculation loop would satisfy the requirements of the LC0 and the initial conditions of the accident sequence.

The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time is based on the low probability of an accident occurring during this time period, on a reasonable time to complete the Required Action, and on frequent core monitoring by operators allowing abrupt changes in core flow conditions to be quickly detected.

l FERMI - UNIT 2 B 3.4.1 - 5 Revision 13. 08/06/99

Recirculation Loops Operating B 3.4.1 BASES ACTIONS (continued)

This Required Action does not require tripping the recirculation pump in the lowest flow loop when the mismatch between total jet pump flows of the two loops is greater than the required limits. However, in cases where large flow mismatches occur, low flow or reverse flow can occur in the low flow loop jet pumps, causing vibration of the jet pumps. If zero or reverse flow is detected, the condition should be alleviated by changing pump speeds to re-establish forward flow or by tripping the pump.

IL1 l When operating in the " Exit" region (refer to Figure l B 3.4.11), the potential for thermal-hydraulic n instabilities is increased and sufficient margin may not be g available for operator response to sup>ress potential power oscillations. Therefore, action must >e initiated immediately to restore operation outside of the " Exit" 5

region. Control rod insertion and/or core flow increases

( are designated as the means to accomplish this objective.

Required Action B.1 is modified by a Note that precludes core flow increases by restart of an idle recirculation loop, or by resetting a recirculation flow limiter. Core flow increases by these means would not support timely completion of the action to restore operation outside the )

" Exit" Region.

M With no recirculation loops in operation in MODE 2 the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . In this condition, the recirculation loops are not required to be operating because of the reduced severity of DBAs and minimal dependence on the recirculation loop coastdown characteristics. The allowed Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is reasonable, based on operating experience, to reach MODE 3 from HODE 2 conditions in an orderly manner and without challenging plant systems.

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j FERMI - UNIT 2 B 3.4.1 - 6 Revision 13. 08/06/99 l

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l Recirculation Loops Operating l B 3.4.1 l '

BASES ACTIONS (continued) l

[L.1 If operating with no recirculation pumps in operation in uY MODE 1 or operating in the " Scram" region (refer to Bases Figure B 3.4.11), or if core thermal-hydraulic instability is detected, then unacceptable power oscillations may result. Therefore, the reactor mode switch must be immediately placed in the shutdown position to terminate the Nl potential for unacceptable power oscillations. )

' % Thermal hydraulic instability is evidenced by a sustained i increase in APRM or LPRM peak to peak noise level reaching 2 A or more times its initial level and occurring with a N characteristic period of less than 3 seconds.

If entry into this condition is an unavoidable and well

%y known consequence of an event, early initiation of the l L Required Action is appropriate. Also it is recognized that i during certain abnormal conditions, it may become

% operationally necessary to enter the " Scram" or " Exit" !

region for the purpose of: 1) protecting plant equipment, which if it were to fail could impact plant safety, or

2) protecting a safety or fuel operating limit. In these cases, the appropriate actions for the region entered would be performed as required.

These requirements are consistent with References 5 and 6.

SURVEILLANCE SR 3.4.1.1 REQUIREMENTS This SR provides frequent periodic monitoring for core thermal hydraulic instability by monitoring APRM and LPRM signals for a sustained increase in APRM or LPRM peak to m peak noise level reaching 2 or more times its initial level N and occurring with a characteristic period of less than 3 i seconds. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Frequency is based on the small !

l potential for core thermal hydraulic oscillations to occur !

outside the " Scram" or " Exit" regions. Therefore, frequent

( monitoring of the APRM and LPRM signals is appropriate when operating in the " Stability Awareness" region.

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Recirculation Loops Operating B 3.4.1 BASES SURVEILLANCE REQUIREMENTS (continued)

This SR is modified by a Note that states performance is only required when operating in the " Stability Awareness" region (refer to Bases Figure B 3.4.1-1) (i.e.. in the power to flow region that is near regions of higher probability for core thermal hydraulic instabilities). This a' is acceptable because outside the " Stability Awareness" d region, power and flow conditions are such that sufficient margin exists to the potential for core thermal hydraulic instability to allow routine core monitoring. Any unanticipated entry into the " Stability Awareness" region would require immediate verification of core stability since D the Surveillance would not be current.

SR 3.4.1.2 This SR ensures the recirculation loops are within the allowable limits for mismatch. At low core flow (i.e..

< 70% of rated core flow). the MCPR requirements provide larger margins to the fuel cladding integrity Safety Limit such that the potential adverse effect of early boiling transition during a LOCA is reduced. A larger flow mismatch can therefore be allowed when core flow is < 70% of rated core flow. The recirculation loop jet pump flow, as used in this Surveillance, is the summation of the flows from all of the jet pumps associated with a single recirculation loop.

The mismatch is measured in terms of percent of rated core flow. If the flow mismatch exceeds tie specified limits, the loop with the lower flow is considered "not in operation". The SR is not required when both loops are not in operation since the mismatch limits are meaningless during single loop or natural circulation operation. The Surveillance must be performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after both locos are in operation. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency is consistent {

l withtheSurveillanceFrequencyforjetpumpOPERABILITY l verification and has been shown by operating experience to be adequate to detect off normal jet pump loop flows in a timely manner.

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h j FERMI UNIT 2 B 3.4.1 - 8 Revision 13. 08/06/99 l

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Recirculation Loops Operating B 3.4.1 BASES 1

REFERENCES 1. UFSAR. Section 6.3.3.

2. NEDE 23785 P A. " SAFER /GES1R Models for the Evaluation of the Loss of-Coolant Accident." Revision 1. October 1984.

3. m E 56-0386. " Fermi 2 Single Loop Operation Analysis."

Rev.1. April 1987, and NEDC 32313 P. "Enrico Fermi Energy Center Unit 2 Single Loop Operation." September 1994.

4. 10 CFR 50. Appendix A. GDC 12.

5. NRC Generic Letter 94-02. "Long Term Solutions and Upgrade of Interim Operating Recommendations for Thermal Hydraulic Instabilities in Boiling Water Reactors." July 1994.

6. BWROG Letter 94078. "BWR Owners' Group Guidelines for Interim Corrective Action." June 1994.

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l FERMI - UNIT 2 B 3.4.1 - 9 Revision 13. 08/06/99

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! Recirculation Loops Operating B 3.4.1

.s BASES l

l THERMAL POWER vs CORE FLOW Figure B 3.4.1-1 1

IT 2 B 3.4.1 - 10 Revision 13. 08/06/99

! l t .

I PCIVs B 3.6.1.3

(

l BASES l

l ACTIONS (continued)

D.1 and D.2 I With one or more penetration flow paths with one or more PCIVs ineperable due to secondary containment bypass leakage

^ rate (SR 3.6.1.3.11). MSIV leakage rate (SR 3.6.1.3.12),

d purge valve leakage rate (SR 3.6.1.3.6), hydrostatically tested line leakage rate (SR 3.6.1.3.13), or EFCV leakage rate (SR 3.6.1.3.9) not within limit. the assumptions of the - I safety analysis may not be met. Therefore, the leakage must be restored to within limit. Restoration can be

\ accomplished by repairing the leaking PCIV(s) (and exiting Condition D).

The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Completion Time for leakage un hydrostatically tested lines and for secondary containment bypass leakage is k reasonable considering the time required to restore the l o '

leakage by isolating the penetration and tne relative importance of leakage to the overall containment function.

For MSIV leakage, an 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Completion Time is allowed. The A Completion Time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months for MSIV leakage allows a period d of time to restore the MSIVs to OPERABLE status given the h fact that the MSIV closure will result in isolation of the A main steam line(s) and potential for plant shutdown. The 24 W hour Completion Time for purge valve leakage is acceptable T considering the purge valves remain closed so that a gross breach of the containment does not exist. The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time for leakage on hydrostatically tested lines on a closed system is acceptable based on the available water seal expected to remain as a gaseous fission product boundary during the accident and the associated closed system. The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time for EFCV leakage is 1

acce) table based on the instrument and small pipe diameter h of t1e penetration (hence, reliability) to act as a penetration isolation boundary and small pipe diameter of the affected penetrations.

Alternately, the leakage can be restored to within limit by isolating the penetration that caused the limit to be exceeded by use of one closed and de activated automatic valve, closed manual valve. or blind flange. When a penetration is isolated, the leakage rate for the isolated penetration is assumed to be the actual pathway leakage through the isolation device. If.two isolation devices are used to isolate the penetration, the leakage rate is assumed to be the lesser actual pathway leakage of the two devices.

'$l In this case (isolation of the affected penetration), the l

1 l FERMI UNIT 2 B 3.6.1.3 - 9 Revision 13. 08/06/99

PCIVs B 3.6.1.3 BASES ACTIONS (continued) leaking PCIV(s) remain inoperable due to leakage and Condition D remains applicable. Required Action D.2 must also be performed to verify the penetration is isolated on a periodic basis. This is necessary to ensure that primary p containment penetrations required to be isolated following an accident are isolated. The Completion Time of "once per 31 days for isolation devices outside primary containment" is appropriate because the devices are operated under administrative controls and the probability of their misalignment is low. For the devices inside primary containment, the time period specified " prior to entering MODE 2 or 3 from MODE 4. if primary containment was de inerted while in MODE 4. if not performed within the previous 92 days" is based on engineering judgment and is considered reasonable in view of the inaccessibility of the devices and other administrative controls ensuring that device misalignment is an unlikely possibility.

Required Action D.2 is modified by three Notes. Note 1 applies to valves and blind flanges located in high radiation areas and allows them to be verified by use of administrative means. Allowing verification by administrative means is considered acceptable, since access to these areas is typically restricted. Therefore, the

)robability of misalignment of these valves, once they have 3een verified to be in the pro)er position, is low. Note 2 ap) lies to isolation devices t1at are locked, sealed, or otlerwise secured in position and allows these devices to be verified closed by use of administrative means. Allowing verification by administrative means is considered acceptable since the function of locking sealing, or securing components is to ensure that these devices are not inadvertently repositioned. Note 3 states that verification that the penetration is isolated a-) plies only to penetration flow paths isolated to restore lea (age within limits.

I FERMI UNIT 2 B 3.6.1.3 - 10 Revision 13. 08/06/99 l

PCIVs B 3.6.1.3 BASES ACTIONS (continued) 3l E.1 and E.2 If any Required Action and associated Completion Time cannot n be met in MODE 1. 2. or 3. the plant must be brought to a MODE in which the LC0 does not apply. To achieve this bd status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and to MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating y experience. to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

l El If any Required Action and associated Completion Time cannot be met. the unit must be placed in a condition in which the LC0 does not apply. Action must be immediately initiated to isolate the RHR Shutdown Cooling System. However, if the shutdown cooling function is needed to provide core cooling.

these Required Actions allow the penetration flow path to remain unisolated provided action is immediately initiated to restore the valve to OPERABLE status or to isolate the RHR Shutdown Cooling System (i.e., provide alternate decay heat removal capabilities so the penetration flow path can be isolated). Actions must continue until the valve is restored to OPERABLE status or the RHR Shutdown Cooling System is isolated.

SURVEILLANCE SR 3.6.1.3.1 REQUIREMENTS This SR ensures that the drywell and suppression chamber purge system isolation valves (6 inch.10 inch. 20 inch. and h 24 inch) and the containment pressure control valves (1 inch) are closed as required or, if open, open for an allowable reason. If a purge or containment pressure N control valve is o)en in violation of this SR, the valve is D considered inopera)le. If the inoperable valve is not otherwise known to have excessive leakage when closed it is D not considered to have leakage outside of limits. Primary y

hl containment purge and containment pressure control valves are only required to be closed in MODES 1, 2. and 3 (i.e..

no isolation instrumentation functions of LC0 3.3.6.1 are dI required to be OPERABLE for isolation of these valves outside of MODES 1, 2. and 3). If a LOCA inside primary l FERMI - UNIT 2 B 3.6.1.3 - 11 Revision 13 08/06/99

i PCIVs j B 3.6.1.3 i BASES SURVEILLANCE REQUIREMENTS (continued) containment occurs in these MODES, the purge valves may not be capable of closing before the pressure pulse affects systems downstream of the purge valves. At other times m (e.g., during handling of irradiated fuel).. pressurization ,

O concerns are not present and the purge and containment pressure control valves are allowed to be open. The SR is modified by a Note stating that the SR is not required to be. i met when the purge or containment pressure control valves

'l are open for the stated reasons. The Note states that these valves may be opened for inerting, de inerting, pressure i control ALARA or air quality considerations for personnel '

entry, or Surveillances that require the valves to be open.

The purge valves (6 inch.10 inch, 20 inch, and 24 inch) and h the containment pressure control valves (1 inch) are capable m of closing in the environment following a LOCA. Therefore, g these valves are allowed to be open for limited periods of 1 time. The 31 day Frecuency is consistent with other PCIV I requirements discussec in SR 3.6.1.3.2.

SR 3.6.1.3.2 This SR verifies that each primary containment isolation manual valve and blind flange that is located outside i primary containment and is not locked, sealed, or otherwise secured and is required to be closed during accident conditions is closed. The SR helps to ensure that post accident leakage of radioactive fluids or gases outside the primary containment boundary is within design limits.

This SR does not require any testing or valve manipulation.

Rather, it involves verification that those PCIVs outside primary containment, and capable of being mispositioned, are in the correct position. Since verification of valve position for PCIVs outside primary containment is relatively easy, the 31 day Frequency was chosen to provide added assurance that the PCIVs are in the correct positions.

Two Notes have been added to this SR. The first Note allows valves and blind flanges located in high radiation areas to be verified by use of administrative controls. Allowing verification by administrative controls is considered l acceptable since access to these areas is typically restricted during MODES 1, 2, and 3 for ALARA reasons.

Therefore, the )robability of misalignment of these PCIVs, once they have >een verified to be in the proper position, is low. A second Note has been included to clarify that plFERMI-UNIT 2 B 3.6.1.3 - 12 Revision 13. 08/06/99 1

e I PCIVs B 3.6.1.3 4

BASES l

SURVEILLANCE REQUIREMENTS (continued)

PCIVs that are open under administrative controls are not required to meet the SR during the time that the PCIVs are i open. This SR does not apply to valves that are locked. ]

sealed, or otherwise secured in the closed position since y these were verified to be in the correct position upon locking, sealing, or securing.

SR 3.6.1.3.3 gq This SR verifies that each primary containment isolation manual valve and blind flange that is located inside primary bl containment and is not locked, sealed, or otherwise secured and is required to be closed during accident conditions is closed. The SR helps to ensure that post accident leakage of radioactive fluids or gases outside the primary b containment boundary is within design limits. For PCIVs 1

inside primary containment, the Frequency defined as " prior to entering MODE 2 or 3 from MODE 4 if primary containment was de inerted while in MODE 4. if not performed within the ,

previous 92 days" is appropriate since these PCIVs are operated under administrative controls and the probability 4 of their misalignment is low. This SR does not apply to i valves that are locked, sealed. or otherwise secured in the i closed position since these were verified to be in the correct position upon locking, sealing, or securing. l Two Notes have been added to this SR. The first Note allows '

valves and blind flanges located in high radiation areas to be verified by use of administrative controls. Allowing verification by administrative controls is considered acceptable since the primary containment is inerted and access to these areas is typically restricted during MODES 1. 2, and 3 for ALARA reasons. Therefore, the probability of misalignment of these PCIVs, once they have been verified to be in their proper position, is low. A second Note has been included to clarify that PCIVs that are o>en under administrative controls are not required to meet t1e SR during the time that the PCIVs are open.

l FERMI UNIT 2 B 3.6.1.3 - 13 Revision 13, 08/06/99

I f PCIVs i B 3.6.1.3 BASES SURVEILLANCE REQUIREMENTS (continued) 1' l SR 3.6.1.3.4 l

l The traversing incore probe (TIP) shear isolation valves are I

actuated by explosive charges. Surveillance of explosive charge continuity provides assurance that TIP valves will' I actuate when required. Other administrative controls, such as those that limit the shelf life of the explosive charges, l must be followed. The 31 day Frequency is based on operating experience that has demonstrated the reliability of the explosive charge continuity.

SR 3.6.1.3.5 V Verifying the isolation time of each power operated automatic PCIV is within limits is required to demonstrate OPERABILITY. MSIVs may be excluded from this SR since MSIV full closure isolation time is demonstrated by SR 3.6.1.3.7. l The isolation time test ensures that the valve will isolate i in a time period less than or equal to that assumed in the A safety analyses. The isolation time and Frequency of this 4 SR are in accordance with the requirements of the Inservice h Testing Program.

SR 3.6.1.3.6 For rimary containment urge valves with resilient seals (6 i ch.10 inch. 20 inc and 24 inch), additional leakage rate testing beyond the test requirements of 10 CFR 50.

J Appendix J. Option B (Ref. 3), is required to ensure 1 4 OPERABILITY. This will ensure that leakage is s 0.05 L when tested at P. Operating experierr,e has demonstrated that this type oI seal has the potential to degrade in a i shorter time period than do other. seal types. Based on this '

observation and the importance of maintaining this ,

penetration leak tight (due to the direct path between 1 primary containment and the environment), a Frequency of 184 days was established.

Additionally, this SR must be performed once within 92 days after opening the valve. The 92 day Frequency was chosen recognizing that cycling the valve could introduce additional seal degradation (beyond that which occurs to a valve that has not been opened). Thus, decreasing the interval (from 184 days) is a prudent measure after a valve has been opened.

l l FERMI UNIT 2 B 3.6.1.3 - 14 Revision 13. 08/06/99 l

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PCIVs B 3.6.1.3 BASES SURVEILLANCE REQUIREMENTS (continued)

The primary containment purge valves are only required to meet leakage rate testing requirements in MODES 1. 2. and 3.

(i.e.. no isolation instrumentation functions of LC0 3.3.6.1 are required to be OPERABLE for purge system isolation outside of MODES 1. 2. and 3). If a LOCA inside primary containment occurs in these MODES. purge valve leakage must be minimized to ensure offsite radiological release is within limits. At other times (e.g.. during handling of irradiated fuel). pressurization concerns are not present and the purge valves are not required to meet any specific leakage criteria.

i SR 3.6.1.3.7 i Verifying that the isolation time of each MSIV is within the I

s)ecified limits is required to demonstrate OPERABILITY. J T1e isolation time test ensures that the MSIV will isolate -

1 in a time period that does not exceed the times assumed in the DBA analyses. This ensures that the calculated ,

radiological consequences of these events remain within i 10 CFR 100 limits. The minimum stroke time ensures that 4 isolation does not result in a pressure spike more rapid than assumed in the transient analyses. The Frequency of this SR is in accordance with the requirements of the Inservice Testing Program.

1 SR 3.6.1.3.8 Automatic PCIVs close on a primary containment isolation signal to prevent leakage of radioactive material from o primary containment following a DBA. This SR ensures that each automatic PCIV will actuate to its isolation position bt on a primary containment isolation signal. The LOGIC SYSTEM h.l FUNCTIONAL TEST in SR 3.3.6.1.5 overlaps this SR to provide Db complete testing of the safety function. The 18 month Frequency was developed considering it is prudent that this Surveillance be performed only during a unit outage since isolation of )enetrations would eliminate cooling water flow and disrupt t1e normal operation of many critical components. Operating ex)erience has shown that these components usually pass t11s Surveillance when performed at the 18 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

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[] FERMI-UNIT 2 B 3.6.1.3 - 15 Revision 13. 08/06/99

PCIVs B 3.6.1.3 BASES SURVEILLANCE REQUIREMENTS (continued) l SR 3.6.1.3.9 This SR requires a demonstration that each reactor instrumentation line excess flow check valve (EFCV) is OPERABLE by verifying that the valve restricts flow on a simulated instrument line break. This SR provides assurance that the instrumentation line EFCVs will perform so that predicted radiological consequences will not be exceeded during the postulated instrument line break event evaluated in Reference 5. The 18 month Frequency is based on the typical performance of this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown that these components usually pass this Surveillance when performed at the 18 month Frequency Therefore, the Frequency was concluded to be acceptabie from a reliability standpoint.

SR 3.6.1.3.10 The TIP shear isolation valves are actuated by explosive charges. An in place functional test is not possible with this design. The explosive squib is removed and tested to provide assurance that the valves will actuate when required. The replacement charge for the explosive squib shall be from the same manufactured batch as the one fired or from another batch that has been certified by having one of the batch successfully fired. No squib will remain in service beyond the expiration of its shelf life or its operating life. The Frequency of 18 months on a STAGGERED TEST BASIS is considered adequate given the administrative controls on replacement charges and the frequent checks of circuit continuity (SR 3.6.1.3.4).

SR 3.6.1.3.11 This SR ensures that the leakage rate of secondary containment bypass leakage paths is less than the specified leakage rate. This provides assurance that the assumptions i

in the radiological evaluations of Reference 1 are met. The leakage rate of each by) ass leakage path is assumed to be the maximum pathway leacage (leakage through the worse of OQl the two isolation valves) unless the penetration is isolated by use of one closed and de activated automatic valve, closed manual valve, or blind flange. In this case, the

, g l FERMI UNIT 2 B 3.6.1.3 - 16 Revision 13. 08/06/99 l

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I

I PCIVs B 3.6.1.3 BASES l SURVEILLANCE REQUIREMENTS (continued) leakage rate of the isolated bypass leakage path is assumed a to be the actual pathway leakage through the isolation device. If both isolation valves in the penetration are bd I

closed, the actual leakage rate is the lesser leakage rate of the two valves. The frequency is required by the Primary Containment Leakage Rate Testing Program. This SR simply D imposes additional acceptance criteria. Additionally, some secondary containment bypass paths (refer to UFSAR 6.2.1.2.2.3) use non PCIVs and therefore are not addressed by the testing Frequency of 10 CFR 50. Appendix J. testing.

To address the testing for these valves, the Frequency also 1 includes a requirement to be in accordance with the l Inservice Testing Program.

' Secondary containment bypass leakage is also considered part I of L,.

SR 3.6.1.3.12 ,

The analyses in References 1 and 4 are based on leakage that is less than the specified leakage rate. Leakage through all four main steam lines must be s 100 scfh when tested at a Pi (25 )sig). This ensures that MSIV leakage is properly accountec for to assure safety analysis assumptions.

regarding the MSIV-LCS ability to provide a positive pressure seal between MSIVs. remain valid. This leakage test is performed in lieu of 10 CFR 50. Appendix J. Type C test requirements, based on an exemption to 10 CFR 50 Appendix J. As such, this leakage is not combined with the Type B and C leakage rate totals. The Frequency is required by the Primary Containment Leakage Rate Testing Program.

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PCIVs B 3.6.1.3 _

l l- BASES l

SURVEILLANCE REQUIREMENTS (continued)

SR 3.6.1.3.13 Surveillance of hydrostatically tested lines provides assurance that the calculation assumptions of Reference 2 p are met. The acceptance criteria for the' combined leakage of all hydrostatically tested lines is 1 gpm times the V

b number of valves per penetration, not to exceed 3 gpm. when I

(= 62.2 psig) . Additionally, a combined tested leakageatrate 1.1lim P.it of s 5 gpm when tested at 1.1 P

(= 62.2 psig) is applied for all hydrostatically tested l PCIVs that )enetrate containment. The combined leakage l rates must )e demonstrated in accordance with the leakage rate test Frequency re Rate Testing Program. quired by Primary Containment Leakage This SR has been modified by a Note that states that these valves are only required to meet the combined leakage rate !

in MODES 1. 2, and 3. since this is when the Reactor Coolant System is pressurized and primary containment is required.

In some instances, the valves are required to be capable of automatically closing during MODES other than MODES 1. 2 and 3. However, specific leakage limits are not applicable in these other MODES or conditions.

REFERENCES 1. UFSAR. Chapter 15.

2. UFSAR. Table 6.2 2.

3. 10 CFR 50. Appendix J. Option B.

4. UFSAR. Section 6.2.

5. UFSAR. Section 15.6.2. <

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hFERMI-UNIT 2 B 3.6.1.3 - 18 Revision 13. 08/06/99 l

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Suppression Chamber-to Drywell Vacuum Breakers B 3.6.1.8 BASES

! APPLICABILITY (continued)

O breakers, therefore, are required to be OPERABLE in H0 DES 1.

2. and 3, when the Suppression Pool Spray System is required to be OPERABLE to mitigate the effects of inadvertent L

%@ l actuation of the Suppression Pool Spray System.

Also. in MODES 1, 2. and 3, a DBA could result in excessive negative differential pressure across the drywell wall, caused by the rapid depressurization of the drywell. The event that results in the limiting rapid dearessurization of the drywell is the primary system rupture tlat purges the drywell of air and fills the drywell free airspace with steam. Subsequent condensation of the steam would result in depressurization of the drywell. The limiting pressure and temperature of the primary system prior to a DBA occur in MODES 1, 2. and 3.

In H0 DES 4 and 5. the probability and consequences of these events are reduced by the pressure and temperature limitations in these MODES: therefore, maintaining suppression chamber to drywell vacuum breakers OPERABLE is not required in MODE 4 or 5. l ACTIONS /L1 gl With one of the vacuum breakers inoperable for opening 1 (e.g., the vacuum breaker is not open and may be stuck closed or not within its opening setpoint limit, so that it would not function as designed during an event that depressurized the drywell), the remaining eleven OPERABLE vacuum breakers are capable of providing the vacuum relief function. However, overall system reliability is reduced, and since normal periodic functional testing of the vacuum breakers is deferred to H0DE 4 (SR 3.6.L8.2), additional undetected failures could result in an excessive suppression chamber-to drywell differential pressure during a DBA.

Therefore, with one vacuum breaker inoperable, 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is allowed to restore the inoperable vacuum breaker to OPERABLE status. The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time is considered acceptable due to the low probability of an event in which the remaining vacuum breaker capability would not be adequate.

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Suppression Chamber to Drywell Vacuum Breakers 8 3.6.1.8 l BASES SURVEILLANCE REQUIREMENTS (continued)

If position indication appears reliable (dual or open indication while torus to drywell differential pressure is steady at 0 psid), and indicates open, the alternate methods I outlined in the TRH can rove the indication to be in error I and the vacuum breaker c osed. However, in this case the l vacuum breaker is assumed open until otherwise proved to l satisfy the leakage test, and this confirmation must be performed within the Technical Specification 3.6.1.8 Required Action B.1 Completion Time of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . The 7 day Frequency is based on engineering judgment. is considered adequate in view of other indications of vacuum breaker status available to operations personnel, and has been shown to be acceptable through operating experience.

Notes 1 and 2 are added to this SR which allows suppression l chamber to drywell vacuum breakers opened in conjunction 1 with the performance of a Surveillance or open while i performing their intended function to not be considered as 1 failing this SR. These periods do not represent inoperable l vacuum breakers. j SR 3.6.1.8.2 I j

l Each vacuum breaker must be cycled to ensure that it opens I

adequately to perform its design function and returns to the fully closed position. This ensures that the safety analysis assumptions are valid. The Frequency of " prior to entering MODE 2 or 3 from MODE 4 :f not performed in the previous 92 days" is based upon the demonstrated reliability l of the vacuum breakers and the potential for the test to result in a stuck open vacuum breaker, which could be caused by a failure of the pneumatically operated test mechanism.

Since the vacuum breaker is inaccessible in MODES 1. 2. and

, 3. test induced inoperability would result in a forced shutdown of the unit. In addition there exists substantial I

redundancy in that 4 vacuum breakers must fail to open before the safety function is lost. In addition, this functional test is required within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after a discharge of steam to the suppression chamber from the I safety / relief valves.

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l FERMI UNIT 2 B 3.6.1.8- 5 Revision 13 08/06/99

SCIVs B 3.6.4.2 BASES APPLICABLE SAFETY ANALYSES (continued) containment performs no active function in response to either of these limiting events, but the boundary established by SCIVs is required to ensure that leakage from the primary containment is processed by the Standby Gas Treatment (SGT) System before being released to the environment.

Maintaining SCIVs OPERABLE with isolation times within limits ensures that fission products will remain trapped inside secondary containment so that they can be treated by the SGT System prior to discharge to the environment.

SCIVs satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

m LC0 SCIVs form a part of the secondary containment boundary.

b The SCIV safety function is related to control of offsite radiation releases resulting from DBAs.

yl The power operated automatic isolation valves are considered OPERABLE when their isolation times are within limits and the valves actuate on an automatic isolation signal. The valves covered by this LCO along with their associated stroke times, are listed in Reference 3.

O The normally closed isolation valves or blind flanges are i d considered OPERABLE when manual valves and blind flanges are closed, or open in accordance with appropriate

( administrative controls. These passive isolation valves or devices are listed in plant procedures.

l APPLICABILITY In MODES 1. 2. and 3. a DBA could lead to a fissian product l release to the primary containment that leaks to the secondary containment. Therefore, the OPERABILITY of SCIVs 1s required. l In MODES 4 and 5. the probability and consequences of these events are reduced due to pressure and temperature limitations in these MODES. Therefore, maintaining SCIVs l OPERABLE is not required in MODE 4 or 5. except for other l

\-

l FERMI UNIT 2 B 3.6.4.2-2 Revision 13. 08/06/99 L

1 Control Center AC System B 3.7.4 BASES ACTIONS (continued)

E.1. E.2. and E.3 The Required Actions of Condition E are modified by a Note indicating that LC0 3.0.3 does not apply. If moving irradiated fuel assemblies while in MODE 1, 2. or 3. the fuel movement is independent of reactor operations.

Therefore, inability to suspend movement of irradiated fuel assemblies is not a sufficient reason to require a reactor shutdown.

During movement of irradiated fuel assemblies in the i secondary containment, during CORE ALTERATIONS. or during OPDRVs. with two control center AC subsystems inoperable.

action must be taken immediately to suspend activities that present a potential for releasing radioactivity that might require isolation of the control room. This places the unit in a condition that minimizes risk.

If applicable. CORE ALTERATIONS and handling of irradiated fuel in the secondary containment must be suspended l immediately. Suspension of these activities shall not preclude completion of moverrent of a component to a safe position. Also, if applicable. actions must be initiated immediately to suspend OPDRVs to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until the OPDRVs are suspended.

SURVEILLANCE SR 3.7.4.1 REQUIREMENTS This SR verifies that the heat removal capability of the system is sufficient to remove the control room heat load.

The SR consis+.s of a verification of the control room temperature. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months F.equency is appropriate since significant degradation of the Control Center AC System is not expected over this time period.

REFERENCES 1. UFSAR. Section 6.4.

2. UFSAR. Section 9.4.1.

! FERMI - UNIT 2 83.7.4-5 Revision 13 08/06/99

SDM Test-Refueling B 3.10.7 BASES ACTIONS (continued)

The allowed Completion Times are reasonable, considering the small number of allowed ino>erable control rods, and provide time to insert and disarm t1e control rods in an orderly manner and without challenging plant systems.

Condition A is modified by a Note allowing separate Condition entry for each uncoupled control rod. This is acceptable since the Required Actions for this Condition provide appropriate compensatory actions for each uncoupled control rod. Complying with the Required Actions may allow for continued operation. Subsequent uncoupled control rods are governed by subsequent entry into the Condition and application of the Required Actions.

B.d With one or more of the requirements of this LC0 not met for reasons other than an uncoupled control rod, the testing should be immediately stopped by placing the reactor mode switch in the shutdown or refuel position. This results in a condition that is consistent with the requirements for MODE 5 where the provisions of this Special Operations LC0 are no longer required.

SURVEILLANCE SB 3.10.7.1 REQUIREMENTS Performance of the applicable SRs for LC0 3.3.1.1. Functions 2.a. 2.d. and 2.e will ensure that the reactor is operated l

l within the bounds of the safety analysis.

SR 3.10.7.1. SR 3.10.7.2. and SR 3.10.7.3 hl LC0 3.3.1.1. Functions 2.a. 2.d. and 2.e. made applicable in this Special Operations LC0. are required to have applicable Surveillances met to establish that this Special Operations LCO is being met. However, the control rod withdrawal sequences during the SDM tests may be enforced by the RWM (LC0 3.3.2.1. Function 2. MODE 2 requirements) or by a l

second licensed operator or other qualified member of the technical staff. As noted. either the applicable SRs for the RWM (LC0 3.3.2.1) must be satisfied according to the applicable Frequencies (SR 3.10.7.2) or the proper movement l of control rods must be verified (SR 3.10.7.3). This latter l FERMI - UNIT 2 B 3.10.7 - 5 Revision 13 08/06/99 1

ATTACHMENT 3 TO NRC-99-0089 LR DESIGNATED NO SIGNIFICANT HAZARDS CONSIDERATION EVALUATIONS 1

1 I

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NO SIGNIFICANT HAZARDS EVALUATION ITS: SECTION 3.3.1.1 - kPS INSTRUMENTATION I

( TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED l

(Specification 3.3.1.1 "LR.1" Labeled Conraents/ Discussions)

Detroit Edison has evaluated the proposed Technical Specification change identified as "Less Restrictive - Removed" I in accordance with the criteria specified by 10 CFR 50.92 and has determined that the proposed changes do not involve a significant hazards consideration.

The bases for the determination that the proposed change does not involve a significant hazards consideration is an evaluation j of these changes against each of the criteria in 10 CFR 50.92.

The criteria and the conclusions of the evaluation are presented below.

l

1. Does the change involve a significant increase in the probability or consequences of an accident previously '

evaluated?

l The proposed change removes from the Fermi 2 Technical Specifications the action requirement to lock the mode switch in the shutdown position. This change relocates informational or implementing details that are not required l to be under regulatory control (e.g. , Technical Specification amendment or 10 CFR 50.59). The requirements l being removed from Fermi 2 Technical Specifications are not l necessarily being deleted or changed at this time, but are simply eliminated from 10 CFR 50.59 or 10 CFR 50.92 change controls. Reduction of the level of regulatory control on the identified requirements is appropriate since the requirement for all control rods to remain inserted continues to be maintained in the Technical Specifications.

Given the remaining Technical Specification requirements, the proposed change in the level of regulatory control has no impact on the probability or the consequences of an accident previously evaluated, therefore, these changes have no impact on the probability or consequences of an accident previously evaluated.

I FERMI - UNIT 2 1 REVISION 13, 8/6/99

NO SIGNIFICANT HAZARDS EVALUATION ITS: SECTION 3.3.1.1 - RPS. INSTRUMENTATION TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED (Specification 3.3.1.1 "LR.1" Labeled Commsents/ Discussions)

2. Does the change create the possibility of a new or different-kind of accident from any accident previously evaluated?

The proposed changes will not involve any physical changes to plant systems, structures, or components (SSC), or the manner in which these SSC are operated, maintained, modified, tested, or inspected. The proposed changes will not impose or eliminate any requirements. Therefore, these changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction in a margin of safety?

The margin of safety as defined in the bases of any Technical Specification is not reduced. The requirements being removed from Technical Specifications, eliminate regulatory control of changes to this requirement (e.g.,

Technical Specification amendment or 10 CFR 50.59) that are not necessary to provide adequate protection of the public health and safety since the ITS continue to impose the appropriate requirements and limitations.

Removal of these items from Technical Specifications eliminates the requirement for NRC review and approval of

revisions in accordance with 10 CFR 50.92. Elimination of this administrative process does not have a margin of safety that can be evaluated. However, the proposed changes continue to provide a level of detail that is consistent with the BWR Standard Technical Specification, NUREG-1433, Rev 1, which was approved by the NRC. Revising the Technical Specifications to reflect the approved level of detail entails no significant reduction in the margin of safety.

FERMI - UNIT 2 2 REVISION 13, 8/6/99

NO SIGNIFICANT HAZARDS EVALUATION ITS: SECTION 3.3.1.1 - RPS INSTRUMENTATION TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED (Specification 3.3.1.1 "LR.3" Labeled Comments / Discussions)

Detroit Edison has evaluated the proposed Technical Specification change identified as "Less Restrictive - Removed" in accordance with the criteria specified by 10 CFR 50.92 and has determined that the proposed changes do not involve a significant hazards consideration.

The bases for the determination that the proposed change does not involve a significant hazards consideration is an evaluation of these changes against each of the criteria in 10 CFR 50.92.

The criteria and the conclusions of the evaluation are presented below.

1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?

The proposed change removes from the Fermi 2 Technical Specifications the acceptance criteria for instrumentation overlaps (SRM to IRM and IRM to APRM) of at least 1/2 ;

decade. This change relocates informational or

{

implementing details that are not required to be under i regulatory control (e.g., Technical Specification amendment or 10 CFR 50.59) . The requirements being removed from l Fermi 2 Technical Specifications are not necessarily being l deleted or changed at this time, but are simply eliminated from 10 CFR 50.59 or 10 CFR 50.92 change controls.

Reduction of the level of regulatory control on the identified requirements is appropriate since the requirement to verify overlap is maintained and the response of multiple monitoring channels is available to provide indication that neutron instrumentation is functioning properly. Given the remaining Technical Specification requirements, the proposed change in the level of regulatory control has no impact on the probability or the consequences of an accident previously evaluated, therefore, these changes have no impact on the probability or consequences of an accident previously evaluated.

FERMI - UNIT 2 3 REVISION 13, 8/6/99

NO SIGNIFICANT HAZARDS EVALUATION ITS: SECTION 3.3.1.1 - RPS ZNSTRUMENTATION TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED (Specification 3.3.1.1 "LR.3" Labeled Comments / Discussions)

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed changes will not involve any physical changes to plant systems, structures, or components (SSC), or the manner in which these SSC are operated, maintained, modified, tested, or inspected. The proposed changes will not impose or eliminate any requirements. Therefore, these changes do not create the possibility of a new or different j kind of accident from any accident previously evaluated. l l

3. Does this change involve a significant reduction in a margin of safety? i The margin of safety as defined in the bases of any l

Technical Specification is not reduced. The requirements I being removed from Technical Specifications, eliminate regulatory control of changes to this requirement (e.g.,

Technical Specification amenchnent or 10 CFR 50.59) that are !

not necessary to provide adequate protection of the public l health and safety since the ITS continue to impose the !

appropriate requirements and limitations.

Removal of these items from Technical Specifications eliminates the requirement for NRC review and approval of revisions in accordance with 10 CFR 50.92. Elimination of this administrative process does not have a margin of safety that can be evaluated. However, the proposed changes continue to provide a level of detail that is consistent with the BWR Standard Technical Specification, NUREG-1433, Rev 1, which was approved by the NRC. Revising the Technical Specifications to reflect the approved level of detail entails no significant reduction in the margin of safety.

FERMI - UNIT 2 4 REVISION 13, 8/6/99

I NO SIGNIFICANT HAZARUS EVALUATION ITS: SECTION 3.3.1.2 - SRM INSTRUMENTATION TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED

! (Specification 3.3.1.2 "LR.1" Labeled Comments / Discussions)

Detroit Edison has evaluated the proposed Technical Specification change identified as "Less Restrictive - Removed" l in accordance with the criteria specified by 10 CFR 50.92 and has determined that the proposed changes do not involve a significant hazards consideration.

The bases for the determination that the proposed change does i not involve a significant hazards consideration is an evaluation l of these changes against each of the criteria in 10 CFR 50.92.

The criteria and the conclusions of the evaluation are presented below.

1. Does the change involve a significant increase in the probability or consequences of an accidant previously l evaluated?

l The proposed change removes from the Fermi 2 Technical Specifications details concerning the positioning of the SRM detectors. This change relocates informational or implementing details that are not required to be under regulatory control (e.g., Technical Specification amendment or 10 CFR 50.59). The requirements being removed from Fermi 2 Technical Specifications are not necessarily being deleted or changed at this time, but are simply eliminated I from 10 CFR 50.59 or 10 CFR 50.92 change controls.

Reduction of the level of regulatory control on the identified requirements is appropriate since the Technical Specifications contain sufficient requirements for SRM Operability. Given the remaining Technical Specification requirements, the proposed change in the level of regulatory control has no impact on the probability or the j consequences of an accident previously evaluated, l therefore, these changes have no impact on the probability or consequences of an accident previously evaluated.

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FERMI - UNIT 2 1 REVISION 13, 8/6/99

e NO SIGNIFICANT HAZARDS EVALUATION ITS: SECTION 3.3.1.2 - SRM INSTRUMENTATION TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED (Specification 3.3.1.2 "LR.1" Labeled Comments / Discussions)

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed changes will not involve any physical changes to plant systems, structures, or components (SSC), or the manner in which these SSC are operated, maintained, modified, tested, or inspected. The proposed changes will not impose or eliminate any requirements. Therefore, these changes do'not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction in a margin of safety?

The margin of safety as defined in the bases of any Technical Specification is not reduced. The requirements being removed from Technical Specifications, eliminate regulatory control of changes to this requirement (e.g.,

Technical Specification amendment or 10 CFR 50.59) that are not necessary to provide adequate protection of the public health and safety since the ITS continue to impose the appropriate requirements and limitations.

Removal of these items from Technical Specifications eliminates the requirement for NRC review and approval of revisions in accordance with 10 CFR 50.92. Elimination of this administrative process does not have a margin of safety that can be evaluated. However, the proposed changes continue to provide a level of detail that is consistent with the BWR Standard Technical Specification, NUREG-1433, Rev 1, which was approved by the NRC. Revising the Technical Specifications to reflect the approved level of detail entails no significant reduction in the margin of safety.

FERMI - UNIT 2 2 REVISION 13, 8/6/99

NO SIGNIFICANT HAEARDS EVALUATION ITS: SECTION 3.3.2.1 - CONTROL ROD BLOCK INSTRUMENTATION TECHNICAL' CHANGES - LESS RESTRICTIVE - REMOVED (Specification 3.3.2.1 "LR.1 " Labeled Coannents/ Discussions)

Detroit Edison has evaluated the proposed Technicc.1 Specification change identified as "Less Restrictive - Removed" in accordance with the criteria specified by 10 CFR 50.92 and has determined that the proposed changes do not involve a significant hazards consideration.

The bases for the determination that the proposed change does not involve a significant hazards consideratzon is an evaluation of these changes against each of the criteria in 10 CFR 50.92.

The criteria and the conclusions of the evaluation are presented below.

1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?

The proposed change removes from the Fermi 2 Technical Specifications the specific location of an individual allowed to verify control rod movement in the event of an inoperable RNM. This change relocates informational or implementing details that are not required to be under ;

regulatory control (e.g., Technical Specification amendment or 10 CFR 50.59). The requirements being removed from Fermi 2 Technical Specifications are not necessarily being deleted or changed at this time, but are simply eliminated from 10 CFR 50.59 or 10 CFR 50.92 change controls.

Reduction of the level of regulatory control on the identified requirements is appropriate since the requirement for verification by a second individual continues to be required by the Technical Specifications. j Given the remaining Technical Specification requirements, '

the proposed change in the level of regulatory control has no impact on the probability or the consequences of an accident previously evaluated, therefore, these changes :

have no impact on the probability or consequences of an accident previously evaluated.

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l FERMI - UNIT 2 1 REVISION 13, 8/6/99 1

E l NO SIGNIFICANT HAZARDS EVALUATION ITS: SECTION 3.3.2.1 - CONTROL ROD BLOCK INSTRUMENTATION TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED (Specification 3.3.2.1 "LR.1" Labeled Comments / Discussions)

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed changes will not involve any physical changes to plant systems, structures, or components (SSC), or the manner in which these SSC are operated, maintained, modified, tested, or inspected. The proposed changes will not impose or eliminate any requirements. Therefore, these changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction in a margin of safety?

l The margin of safety as defined in the bases of any Technical Specification is not reduced. The requirements l

being removed from Technical Specifications, eliminate regulatory control of changes to this requirement (e.g.,

Technical Specification amendment or 10 CFR 50.59) that are not necessary to provide adequate protection of the public health and safety since the ITS continue to impose the l appropriate requiremente and limitations.

Removal of these items from Technical Specifications eliminates the requirement for NRC review and approval of l

revisions in accordance with 10 CFR 50.92. Elimination of

( this administrative process does not have a margin of safety that can be evaluated. However, the proposed changes continue to provide a level of detail that is consistent with the.BWR Standard Techdical Specification, NUREG-1433, Rev 1, which was approved by the NRC. Revising I the Technical Specifications to reflect the approved level of detail entails no significant reduction in the margin of safety.

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l FERMI - UNIT 2 2 REVISION 13, 8/6/99 l

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L NO SIGNIFICANT HAZARDS EVALUATION ITS: SECTION 3.3.3.1 - PAM INSTRUMENTATION TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED (Specification 3.3.3.1 "LR.1" Labeled Connaents/ Discussions)

Detroit Edison has evaluated the proposed Technical

! Specification change identified as "Less Restrictive - Removed" in accordance with the criteria specified by 10 CFR 50.92 and '

has determined that the proposed changes do not involve a I significant hazards consideration.

l l The bases for the determination that the proposed change does not involve a significant hazards consideration is an evaluation l of these changes against each of the criteria in 10 CFR 50.92.

The criteria and the conclusions of the evaluation are presented below.

1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?

The proposed change removes from the Fermi 2 Technical l Specifications the requirement to implement preplanned I

alternative methods of monitoring parameters within 72

. hours with the number of Operable channels is less than the 1

minimum required. This change relocates informational or implementing details that are not required to be under regulatory control (e.g., Technical Specification amendment or_10 CFR 50.59) . The requirements being removed from Fermi 2 Technical Specifications are not necessarily being deleted or changed at this time, but are simply eliminated from 10 CFR 50.59 or 10 CFR 50.92 change controls.

Reduction of the level of regulatory control on the identified requirements is appropriate since the requirement to restore the inoperable channel within 7 days remains in Technical Specifications. Given the remaining Technical Specification requirements, the proposed change in the level of regulatory control has no impact on the ;

probability or the consequences of an accident previously !

evaluated, therefore, these changes have no impact on the !

probability or consequences of an accident previously evaluated.

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FERMI - UNIT 2 1 REVISION 13, 8/6/99 L

m NO SIGNIFICANT HAZARDS EVALUATION ITS: SECTION 3.3.3.1 - PAM INSTRUMENTATION TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED (Specification 3.3.3.1 "LR.1" Labeled Comments / Discussions)

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed changes will not involve any physical changes

, to plant systems, structures, or components (SSC), or the

! manner in which these SSC are operated, maintained, modified, tested, or inspected. The proposed changes will not impose or eliminate any requirements. Therefore, these changes do not create the possibility of a new or different i kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction in a margin of safety?

The margin of safety as defined in the bases of any Technical Specification is not reduced. The requirements being removed from Technical Specifications, eliminate l

regulatory control of changes to this requirement (e.g.,

Technical Specification amendment or 10 CFR 50.59) that are not necessary to provide adequate protection of the public health and safety since the ITS continue to impose the appropriate requirements and limitations.

l Removal of these items from Technical Specifications eliminates the requirement for NRC review and approval of revisions in accordance with 10 CFR 50.92. Elimination of this administrative process does not have a margin of safety that can be evaluated. However, the proposed changes continue to provide a level of detail that is I consistent with the BWR Standard Technical Specification, j NUREG-1433, Rev 1, which was approved by the NRC. Revising the Technical Specifications to reflect the approved level of detail entails no significant reduction in the margin of l safety.

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l FERMI - UNIT 2 2 REVISION 13, 8/6/99

I-NO SIGNIFICANT HAZARDS EVALUATION ITS: SECTION 3.3.3.1 - PAM INSTRUMENTATION

-TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED (Specification 3.3.3.1 "LR.2" Labeled Comunents/ Discussions)

Detroit Edison has evaluated the proposed Technical Specification change identified as "Less Restrictive - Removed" in accordance with the criteria specified by 10 CFR 50.92 and has determined that the proposed changes do not involve a significant hazards consideration.

The bases for the determination that the proposed change does not involve a significant hazards consideration is an evaluation of these changes against each of the criteria in 10 CFR 50.92.

The criteria and the conclusions of the evaluation are presented below.

1. Does the. change involve a significant increase in the probability or consequences of an accident previously evaluated?

The proposed change removes from the Fermi 2 Technical Specifications details for performing Channel Calibrations.

This change relocates informational or implementing details that are not required to be under regulatory control (e.g.,

Technical Specification amendment or 10 CFR 50.59) . The requirements being removed from Fermi 2 Technical Specifications are not necessarily being deleted or changed at this time, but are simply eliminated from 10 CFR 50.59 or 10 CFR 50.92 change controls. Reduction of the level of regulatory control on the identified requirements is appropriate since the Technical Specifications continue to require the performance of the Channel Calibrations. Given the remaining Technical Specification requirements, the proposed change in the level of regulatory control has no

-impact on the probability or the consequences of an accident previously evaluated, therefore, these changes ,

have no impact on the probability or consequences of an '

accident previously evaluated.

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FERMI - UNIT 2 3 REVISION 13, 8/6/99 l

NO SIGNIFICANT HAZARDS EVALUATION ITS: SECTION 3.3.3.1 - PAM INSTRUMENTATION TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED (Specification 3.3.3.1 "LR.2" Labeled Comments / Discussions)

2. Does the change create the possibility of a new or l

different kind of accident from any accident previously evaluated?

The proposed changes will not involve any physical changes j l

l to plant systems, structures, or components (SSC), or the I

manner in which these SSC are operated, maintained, modified, tested, or-inspected. The proposed changes will l not impose or eliminate any requirements. Therefore, these changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction in a margin of safety?

The margin of safety as defined in the bases of any Technical Specification is not reduced. The requirements being removed from Technical Specifications, eliminate regulatory control of changes to this requirement (e.g., i Technical Specification amendment or 10 CFR 50.59) that are

not necessary to provide adequate protection of the public health and safety since the ITS continue to 2.mpose the i

appropriate requirements and limitations.

1 Removal of these items from Technical Specifications i eliminates the requirement for NRC review and approval of l revisions in accordance with 10 CFR 50.92. Elimination of I this administrative process does not have a margin of l safety that can be evaluated. However, the proposed I changes continue to provide a level of detail that is consistent with the BWR Standard Technical Specification, NUREG-1433, Rev 1, which was approved by the NRC. Revising the Technical Specifications to reflect the approved level of detail entails no significant reduction in the margin of safety.

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l FERMI - UNIT 2 4 REVISION 13, 8/6/99

NO SIGNIFICANT HAEARDS EVALUATION ITS: SECTION 3.3.5.1 - ECCS INSTRUMENTATION TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED (Specification 3.3.5.1 "LR.1" Labeled Comments / Discussions)

Detroit Edison has evaluated the proposed Technical Specification change identified as "Less Restrictive - Removed" in accordance with the criteria specified by 10 CFR 50.92 and has determined that the proposed changes do not involve a significant hazards consideration.

]

! The bases for the determination that the proposed change does not involve a significant hazards consideration is an evaluation of these changes against each of the criteria in 10 CFR 50.92.

l The criteria and the conclusions of the evaluation are presented below.

1. Does the change involve a significant increase in the l probability or consequences of an accident previously I evaluated?

l The proposed change removes from the Fermi 2 Technical Specifications descriptive methods for performing l calibration testing. This change relocates informational or implementing details that are not required to be under ,

regulatory control (e.g., Technical Specification amendment i or 10 CFR 50.59) . The requirements being removed from Fermi 2 Technical Specifications are not necessarily being deleted or changed at this time, but are simply eliminated from 10 CFR 50.59 or 10 CFR 50.92 change controls.

Reduction of the level of regulatory control on the identified requirements is appropriate since the allowable value and channel calibration requirements remain in the l Technical Specifications. Given the remaining Technical Specification requirements, the proposed change in the level of regulatory control has no impact on the probability or the consequences of an accident previously I

evaluated, therefore, these changes have no impact on the probability or consequences of an accident previously evaluated.

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l FERMI - UNIT 2 1 REVISION 13, 8/6/99 l

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i NO SIGNIFICANT HAZARDS EVALUATION ITS: SECTION 3.3.5.1 - ECCS INSTRUMENTATION TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED (Specification 3.3.5.1 "LR.1" Labeled Comments / Discussions)

2. Does the change create the possibility of a new or different' kind of accident from any accident previously evaluated?

J l

The proposed changes will not involve any physical changes l to plant systems, structures, or components (SSC), or the 1

]

manner in which these SSC are operated, maintained, modified, tested, or inspected. The proposed changes will l not impose or eliminate any requirements. Therefore, these I changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

l l 3. Does this change involve a significant reduction in a margin of safety?

l The margin of safety as defined in the bases of any Technical Specification is not reduced. The requirements l

being removed from Technical Specifications, eliminate regulatory control of changes to this requirement (e . g . ,

Technical Specification amendment or 10 CFR 50.59) that are not necessary to provide adequate protection of the public l health and safety since the ITS continue to impose the l appropriate requirements and limitations.

Removal of these items from Technical Specifications I eliminates the requirement for NRC review and approval of j revisions in accordance with 10 CFR 50.92. Elimination of j l this administrative process does not have a margin of safety that can be evaluated. However, the proposed j changes continue to provide a level of detail that is consistent with the BWR Standard Technical Specification, l NUREG-1433, Rev 1, which was approved by the NRC. Revising l l

the Technical Specifications to reflect the approved level i of detail entails no significant reduction in the margin of l safety.

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FERMI - UNIT 2 2 REVISION 13, 8/6/99 l

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3 NO SIGNIFICANT HAZARDS EVALUATION ITS: SECTION 3.4.3 - SAFETY RELIEF VALVES (SRVs)

TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED (Specification 3.4.3 "LR.1" T*Mled Comments / Discussions)

Detroit Edison has evaluated the proposed Technical Specification change identified as "Less Restrictive - Removed" in accordance with the criteria specified by 10 CFR 50.92 and has determined that the proposed changes do not involve a significant hazards consideration.

The bases for the determination that the proposed change does not involve a significant hazards consideration is an evaluation of these changes against each of the criteria in 10 CFR 50.92.

The criteria and the conclusions of the evaluation are presented below.

1. Does the change involve a significant increase in the probability or consequences of an accident previously !

evaluated?

The proposed change removes frosa the Fermi 2 Technical Specifications details for the testing schedule for SRV set pressure testing. This change relocates informational or implementing details that are not required to be under regulatory control (e.g., Technical Specification amendment or 10 CFR 50.59) . The requirements being removed from Fermi 2 Technical Specifications are not necessarily being deleted or changed at this time, but are simply eliminated from 10 CFR 50.59 or 10 CFR 50.92 change controls.

Reduction of the level of regulatory control on the identified requirements is appropriate since the operability of the SRVs and the testing of the'SRVs in accordance with the IST program continues to be required by the Technical Specifications. Given the remaining l Technical Specification requirements, the proposed change l in the level of regulatory control has no impact on the probability or the consequences of an accident previously j evaluated, therefore, these changes have no impact on the !

probability or consequences of an accident previously l evaluated.

l l

I FERMI - UNIT 2 1 REVISION 13, 8/6/99

e ,

NO SIGNIFICANT HAEARDS EVALUATION ITS: SECTION 3.4.3 - SAFETY RELIEF VALVES (SRVs)

TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED (Specification 3.4.3 "LR.1" Labeled Comments / Discussions)

2. Does the change create the possibility of a new or different kind of accident frosa any accident previously evaluated?

The proposed changes will not involve any physical changes to plant systems, structures, or components (SSC), or the

~

manner in which these CSC are operated, maintained, modified, tested, or inspected. The proposed changes will not impose or eliminate any requirements. Therefore, these changes do not create the possibility of a new or different kind of accident frosa any accident previously evaluated.

3. Does this change involve a significant reduction in a margin of safety?

The margin of safety as defined in the bases of any Technical Specification is not reduced. The requirements being removed from Technical Specifications, eliminate regulatory control of changes to this requirement (e.g.,

Technical Specification amendment or 10 CFR 50.59) that are not necessary to provide adequate protection of the public health and safety since the ITS continue to impose the appropriate requirements and limitations.

Removal of these items from Technical Specifications eliminates the requirement for NRC review and approval of revisions in accordance with 10 CFR 50.92. Elimination of this administrative process does not have a margin of safety that can be evaluated. However, the proposed changes continue to provide a level of detail that is consistent with the BWR Standard Technical Specification, NUREG-1433, Rev 1, which was approved by the NRC. Revising the Technical Specifications to reflect the approved level of detail entails no significant reduction in the margin of safety.

FERMI - UNIT 2 2 REVISION 13, 8/6/99

NO SIGNIFICANT HAZARDS EVALUATION ITS: SECTION 3.4.5 - RCS PRESSURE ISOLATION VALVE (PIV) LEAKAGE TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED (Specification 3.4.5 "LR.1 " Labeled Comuments/ Discussions)

Detroit Edison has evaluated the proposed Technical Specification change identified as "Less Restrictive - Removed" in accordance with the criteria specified by 10 CFR 50.92 and has determined that the proposed changes do not involve a significant hazards consideration.

The bases for the determination that the proposed change does t not involve a significant hazards consideration is an evaluation of these changes against each of the criteria in 10 CFR 50.92.

The criteria and the conclusions of the evaluation are presented below.

1. Does the change involve a significant increase in th'a probability or consequences of an accident previously evaluated?

The proposed change removes from the Fermi 2 Technical Specifications requirementa concerning the PIV leakage pressure monitors. This change relocates informational or implementing details that are not required to be under regulatory control (e.g., Technical Specification amendment or 10 CFR 50.59). The requirements being removed from Fermi 2 Technical Specifications are not necessarily being deleted or changed at this time, but are simply eliminated from 10 CFR 50.59 or 10 CFR 50.92 change controls.

Reduction of the level of regulatory control on the identified requirements is appropriate since the leakage pressure monitors performs an alarm-only function which does not relate directly to system Operability requirements or analysis assumptions. The requirement to meet RCS and PIV leakage limits remains in the Technical Specifications.

Given the remaining Technical Specification requirements, the proposed change in the level of regulatory control has no impact on the probability or the consequences of an accident previously evaluated, therefore, these changes j have no impact on.the probability or consequences of an ;

accident previously evaluated. '

l l

1 FERMI - UNIT 2 1 REVISION 13, 8/6/99

l l

NO SIGNIFICANT HAZARDS EVALUATION ITS: SECTION 3.4.5 - RCS PRESSURE ISOLATION VALVE (PIV) LEAKAGE TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED (Specification 3.4.5 "LR.1" Labeled Cr==mnts/ Discussions)

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed changes will not involve any physical changes to plant systems, structures, or components (SSC), or the manner in which these SSC are operated, maintained, modified, tested, or inspected. The proposed changes will not impose or eliminate any requirements. Therefore, these changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction in a margin of safety?

The margin of safety as defined in the bases of any Technical Specification is not reduced. The requirements being removed from Technical Specifications, eliminate regulatory control of changes to this requirement (e.g.,

Technical Specification amendment or 10 CFR 50.59) that are not necessary to provide adequate protection of the public health and safety since the ITS continue to impose the appropriate requirements and limitations.

Removal of these items from Technical Specifications eliminates the requirement for NRC review and approval of revisions in accordance with-10 CFR 50.92. Elimination of this administrative process does not have a margin of safety that can be evaluated. However, the proposed changes continue to provide a level of detail that is consistent with the BWR Standard Technical Specification, NUREG-1433, Rev 1, which was approved by the NRC. Revising the Technical Specifications to reflect the approved level of detail entails no significant reduction in the margin of safety.

FERMI - UNIT 2 2 REVISION 13, 8/6/99

NO SIGNIFICANT HAZARDS EVALUATION ITS: SECTION 3.4.5 - RCS PRESSURE ISOLATION VALVE (PIV) LEAKAGE TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED (Specification 3.4.5 "LR.2" Labela'd Comments / Discussions)

Detroit Edison has evaluated the proposed Technical Specification change identified as "Less Restrictive - Removed" in accordance with the criteria specified by 10 CFR 50.92 and has determined that the proposed changes do not involve a !

significant hazards consideration.

l The bases for the determination that the proposed change does not involve a significant hazards consideration is an evaluation of these changes against each of the criteria in 10 CFR 50.92. l The criteria and the conclusions of the evaluation are presented below.

1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?

The proposed change removes from the Fermi 2 Technical Specifications post-maintenance leakage rate testing requirements for PIVs. This change relocates informational or implementing details that are not required to be under regulatory control (e.g., Technical Specification amendment or 10 CFR 50.59) . The requirements being removed from Fermi 2 Technical Specifications are not necessarily being deleted or changed at this time, but are simply eliminated from 10 CFR 50.59 or 10 CFR 50.92 change controls.

Reduction of the level of regulatory control on the identified requirements is appropriate since PIV l Operability continues to be required by the Technical Specifications. Given the remaining Technical Specification requirements, the proposed change in the ;

level of regulatory control has no impact on the probability or the consequences of an accident previously l evaluated, therefore, these changes have no impact on the probability or consequences of an accident previously evaluated.

1 i

1 FERMI - UNIT 2 3 REVISION 13, 8/6/99 l

L NO SIGNIFICANT HAZARDS EVALUATION ITS: . SECTION 3. 4.5 - RCS PRESSURE ISOIATION VALVE (PIV) LEAKAGE TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED (Specification 3.4.5 "LR.2" Labeled Comments / Discussions) l

2. Does the change create the possibility of a new or different kind of accident from any accident previously ]

evaluated?

The proposed changes will not involve any physical changes {

to plant systems, structures, or components (SSC), or the '

manner in which these SSC are operated, maintained, modified, tested, or inspected. The proposed changes will not impose or eliminate any requirements. Therefore, these changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction in a margin of safety?

The margin of safety as defined in the bases of any Technical Specification is not reduced. The requirements being removed from Technical Specifications, eliminate regulatory control of changes to this requirement (e.g.,

Technical Specification amendment or 10 CFR 50.59) that are not necessary to provide adequate protection of the public health and safety since the ITS continue to impose the appropriate requirements and limitations.

Removal of these items from Technical Specifications eliminates the requirement for NRC review and approval of revisions in accordance with 10 CFR 50.92. Elimination of this administrative process does not have a margin of l safety that can be evaluated. However, the proposed changes continue to provide a level of detail that is 1 consistent with the BWR Standard Technical Specification, NUREG-1433, Rev 1, which was approved by the NRC. Revising the Technical Specifications to reflect the approved level !

of detail entails no significant reduction in the margin of '

safety.

l l

i FERMI - UNIT 2 4 REVISION 13, 8/6/99

I NO SIGNIFICANT HAZARDS EVALUATION ITS: SECTION 3.4.8 - RER SHUTDOWN COOLING SYSTEM - HOT SHUTDOWN TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED JSpecification 3.4.8 "LR.1" Labeled Comments / Discussions)

Detroit Edison has evaluated the proposed Technical Specification change identified as "Less Restrictive - Removed" in accordance with the criteria specified by 10 CFR 50.92 and has determined that the proposed changes do not involve a significant hazards consideration.

The bases for the determination that the proposed change does i not involve a significant hazards consideration is an evaluation of these changes against each of the criteria in 10 CFR 50.92.

The criteria and the conclusions of the evaluation are presented below.

1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?

The proposed change removes from the Fermi 2 Technical Specifications a periodic non-specific verification of system Operability. This change relocates informational or implementing details that are not required to be under regulatory control (e.g. , Technical Specification amendment or 10 CFR 50.59) . The requirements being removed from Fermi 2 Technical Specifications are not necessarily being deleted or changed at this time, but are simply eliminated from 10 CFR 50.59 or 10 CFR 50.92 change controls.

Reduction of the level of regulatory control on the identified requirements is appropriate since tracking of j system Operability status is an on going activity and RER '

Shutdown Cooling system Operability continues to remain a Technical Specification requirement. Given the remaining Technical Specification requirements, the proposed change in the level of regulatory control has no impact on the probability or the consequences of an accident previously evaluated, therefore, these changes have no impact on the probability or consequences of an accident previously i evaluated.

l l

l FERMI - UNIT 2 1 REVISION 13, 8/6/99 r

J

NO SIGNIFICANT HAZARDS EVALUATION l

ITS: SECTION 3.4.8 - RHR SHUTDONN COOLING SYSTEM - HOT SHUTDONN TECHNICAL CHANGES - LESS RESTRICTIVE-- REMOVED

- (Specification 3. 4. 8 "LR.1" Labeled Comunents/ Discussions)

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed changes will not involve any physical changes to plant systems, structures, or components (SSC), or the manner'in which these SSC are operated, maintained, modified, tested, or inspected. The proposed changes will not unpose or eliminate any requirements. Therefore, these changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Dces this change involve a significant reduction in a margin of safety?

The margin of safety as defined in the bases of any Technical Specification is not reduced. The requirements being removed from Technical Specifications, eliminate regulatory control of changes to this requirement (e . g . ,

Technical Specification amendment or 10 CFR 50.59) that are not necessary to provide adequate protection of the public health and safety since the ITS continue to impose the appropriate requirements and limitations.

Removal of these items from Technical Specifications eliminates the requirement for NRC review and approval of revisions in accordance with 10 CFR 50.92. Elimination of this administrative process does not have a margin of safety that can be evaluated. However, the proposed changes continue to provide a level of detail that is consistent with the BWR Standard Technical Specification, NUREG-1433, Rev 1, which was approved by the NRC. Revising l

the Technical Specifications to reflect the approved level of detail entails no significant reduction in the margin of safety.

i 1

l l

FERMI - UNIT 2 2 REVISION 13, 8/6/99 e

l NO SIGNIFICANT HAZARDS EVALUATION ITS: SECTION 3.4.9 - RHR SHUTDOWN COOLING SYSTEM - COLD SHUTDOWN TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED (Specification 3.4.9 "LR.1" Labeled Comments / Discussions)

Detroit Edison has evaluated the proposed Technical Specification change identified as "Less Restrictive - Removed" in accordance with the criteria specified by 10 CFR 50.92 and has detemined that the proposed changes do not involve a significant hazards consideration.

The bases for the determination that the proposed change does not involve a significant_ hazards consideration is an evaluation of-these changes against each of the criteria in 10 CFR 50.92. !

The criteria and the conclusions of the evaluation are presented I below.

1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated? 1 l

i The proposed change removes from the Fermi 2 Technical l Specifications a periodic non-specific verification of system Operability. This change relocates informational or implementing details that are not required to be under regulatory control (e.g., Technical Specification amendment ,

or 10 CFR 50.59). The requirements being removed from )

Fermi ~2 Technical Specifications are not necessarily being l deleted or changed at this time, but are simply eliminated from 10 CFR 50.59 or 10 CFR 50.92 change controls.

Reduction of the level of regulatory control on the i identified requirements is appropriate since tracking of I system Operability status is an on going activity and RHR Shutdown Cooling system Operability continues to remain a Technical Specification requirement. Given the remaining Technical Specification requirements, the proposed change in the level of regulatory control has no impact on the i probability or the consequences of an accident previously l evaluated, therefore, these changes have no impact on the probability or consequences of an accident previously evaluated.

FERMI - UNIT 2 1 REVISION 13, 8/6/99

NO SIGNIFICANT HAZARDS EVALUATION ITS: SECTION 3.4.9 - RHR SHUTDOWN COOLING SYSTEM - COLD SHUTDOWN TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED (Specification 3.4.9 "LR.1" Labeled Comments / Discussions)

~

2. Does the change create the possibility of a new or i different kind of accident from any accident previously

! evaluated?

l The proposed changes will not involve any physical changes l to plant systems, structures, or components (SSC), or the i

manner in which these SSC are operated, maintained, modified,_ tested, or inspected. The proposed changes will not impose or eliminate any requirements. Therefore, these changes do not create the possibility of a new or different ;

kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction in a margin of safety? ;

The margin of safety as defined in the bases of any Technical Specification is not reduced. The requirements i being removed from Technical Specifications, eliminate I regulatory control of changes to this requirement (e.g.,

Technical Specification amendment or 10 CFR 50.59) that are not necessary to provide adequate protection of the public health and safety since the ITS continue to impose the appropriate requirements and limitations.

Removal of these items from Technical Specifications eliminates the requirement for NRC review and approval of revisions in accordance with 10 CFR 50.92. Elimination of this administrative process does not have a margin of safety that can be evaluated. However, the proposed I changes continue to provide a level of detail that is consistent with the BWR Standard Technical Specification, NUREG-1433, Rev 1, which was approved by the NRC. Revising the Technical Specifications to reflect the approved level of detail entails no significant reduction in the margin of ,

safety. !

i FERMI - UNIT 2 2 REVISION 13, 8/6/99 l

=

NO SIGNIFICANT HAZARDS EVALUATION l ITS: SECTION 3.5.1 - ECCS - Operating TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED (Specification 3.5.1 "LR.1" Labeled Comments / Discussions)

Detroit Edison has evaluated the proposed Technical Specification change identified as "Less Restrictive - Removed" in accordance with the criteria specified by 10 CFR 50.92 and has determined that the proposed changes do not involve a significant hazards consideration.

The bases for the determination that the proposed change does not involve a significant hazards consideration is.an evaluation of these changes against each of the criteria in 10 CFR 50.92.

l The criteria and the conclusions of the evaluation are presented below.

1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?

l The proposed change removes from the Fermi 2 Technical j Specifications requirements related to alarm-only or j indication-only functions This change relocates informational or implementing details that are not required ;

to be under regulatory control (e.g., Technical Specification amendment or 10 CFR 50.59). The requirements being removed from Fermi 2 Technical Specifications are not necessarily being deleted or changed at this time, but are simply eliminated from 10 CFR_50.59 or 10 CFR 50.92 change '

controls. Reduction of the level of regulatory control on the identified requirements is appropriate since these functions do not relate directly to the Operability requirements for the system or analysis assumptions. Given the remaining Technical Specification requirements, the proposed change in the level of regulatory control has no impact on the probability or the consequences of an accident previously evaluated, therefore, these changes have no impact on the probability or consequences of an accident previously evaluated.

FERMI - UNIT 2 1 REVISION 13, 8/6/99

f l

> l NO SIGNIFICANT HAZARDS EVALUATION ITS: SECTION 3.5.1 - ECCS - Operating i TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED l- (Specification 3.5.1 "LR.1" Labeled Comments / Discussions)

2. Does the change create the possibility of a new or i different kind of accident from any accident previously evaluated?

The proposed changes will not involve any physical changes to plant systems, structures,.or components (SSC), or the manner in which these SSC are operated, maintained, modified, tested, or inspected. The proposed changes will not impose or eliminate any requirements. Therefore, these ,

changes do not create the possibility of a new or different i kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction in a ,

margin of safety? l The margin of safety as defined in the bases of any Technical Specification is not reduced. The requirements being removed from Technical Specifications, eliminate regulatory control of changes to this requirement (e.g.,

Technical Specification amendment or 10 CFR 50.59) that are not necessary to provide adequate protection of the public health and safety since the ITS continue to impose the ,

appropriate requirements and limitations. l Removal of these items from Technical Specifications eliminates the requirement for NRC review and approval of revisions in accordance with 10 CFR 50.92. Elimination of this administrative process does not have a margin of safety that can be evaluated. However, the proposed i changes continue to provide a level of detail that is i consistent with the BWR Standard Technical Specification, l NUREG-1433, Rev 1, which was approved by the NRC. Revising the Technical Specifications to reflect the approved level of detail entails no significant reduction in the margin of safety. 1 l

l 1

FERMI - UNIT 2 2 REVISION 13, 8/6/99

NO SIGNIFICANT HAZARDS EVALUATION ITS: SECTION 3.5.2 - ECCS - Shutdown TECHNICAL CHANGES - LES'S RESTRICTIVE - REMOVED (Specification 3.5.2 "LR.1" Labeled Comments / Discussions)

Detroit Edison has evaluated the proposed Technical Specification change identified as "Less Restrictive - Removed" in accordance with the criteria specified by 10 CFR 50.92 and '

has determined that the proposed changes do not involve a )

significant hazards consideration.

.The bases for the determination that the proposed change does 3 not involve a significant hazards consideration is an evaluation j

of these changes against each of the criteria in 10 CFR 50.92. 1 The criteria and the conclusions of the evaluation are presented l below.

1. Does the change involve a significant increase in the l

probability or consequences of an accident previously evaluated?

The proposed change removes from the Fermi 2 Technical Specifications the requirement to lock the mode switch when the Technical Specification require the switch to be placed in the Shutdown or Refuel position. This change relocates informational or implementing details that are not required to-be under regulatory control (e.g., Technical Specification amendment or 10 CFR 50.59). The requirements being removed from Fermi 2 Technical Specifications are not necessarily being deleted or changed at this time, but are simply eliminated from 10 CFR 50.59 or 10 CFR 50.92 change controls. Reduction of the level of regulatory control on the identified requirements is appropriate since because the action to place the mode switch in the proper position remains a Technical Specification requirement. Given the remaining Technical Specification requirements, the proposed change in the level of regulatory control has no ,

impact on the probability or the consequences of an i

accident previously evaluated, therefore, these changes have no impact on the probability or consequences of an )

accident previously evaluated.

FERMI - UNIT 2 1 REVISION 13, 8/6/99

NO SIGNIFICANT HAZARDS EVALUATION ITS: SECTION 3.5.2 - ECCS - Shutdown TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED (Specification 3.5.2 "LR.1" Labeled Comments / Discussions)

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed changes will not involve any physical changes to plant systems, structures, or components (SSC), or the l- manner in which these SSC are operated, maintained, l modified, tested, or inspected. The proposed changes will not impose or eliminate any requirements. Therefore, these changes do not create the possibility of a new or different l kind of accident from any accident previously evaluated.

I

3. Does this change involve a significant reduction in a j margin of safety?

The margin of safety as defined in the bases of any Technical Specification is not reduced. The requirements being removed from Technical Specifications, eliminate regulatory control of changes to this requirement (e.g.,

Technical Specification amendment or 10 CFR 50.59) that are not necessary to provide adequate protection of the public health and safety since the ITS continue to impose the appropriate requirements and limitations. l Removal of these items from Technical Specifications eliminates the requirement for NRC review and approval of j revisions in accordance with 10 CFR 50.92. Elimination of l this administrative process does not have a margin of safety that can he evaluated. -However, the proposed !

changes continuts te provida a level of detail that is '

consistent with the BWR Standard Technical Specification, NUREG-1433, Rev 1, wh3ch was approved by the NRC. Revising the Technical Specifications to reflect the approved level of detail entails no significant reduction in the margin of ;

safety. j i

FERMI - UNIT 2 2 REVISION 13, 8/6/99

7_

NO SIGNIFICANT HAZARDS EVALUATION ITS: SECTION 3.6.1.3 - PCIVs TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED (Specification 3. 6.1.3 "LR.1" Labeled Comments / Discussions)

Detroit Edison has evaluated the proposed Technical Specification change identified as "Less Restrictive - Removed" in accordance'with the criteria specified by 10 CFR 50.92 and l has determined that the proposed changes do not involve a significant hazards consideration.

i i

The bases for the determination that the proposed change does not involve a significant hazards consideration is an evaluation of these changes against each of the criteria in 10 CFR 50.92. t The criteria and the conclusions of the evaluation are presented below.

1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?

The proposed change reroves from the Fermi 2 Technical Specifications the specific requirement to demonstrate Operability of an isolation valve after maintenance or repair. This change relocates informational or implementing details that are not required to be under regulatory control (e.g., Technical Specification amendment or 10 CFR 50.59) . The requirements being removed from e Fermi 2 Technical Specifications are not necessarily being deleted or changed at this time, but are simply eliminated from 10 CFR 50.59 or 10 CFR 50.92 change controls.

Reduction of the level of regulatory control on the identified requirements is appropriate since Operability of the valves is required by the IST program and the Technical Specifications. Given the remaining Technical Specification requirements, the proposed change in the level of regulatory control has no impact on the probability or the consequences of an accident previously evaluated, therefore, these changes have no impact on the probability or consequences of an accident previously evaluated.

(

FERMI - UNIT 2 1 REVISION 13, 8/6/99 l

l NO SIGNIFICANT HAZARDS EVALUATION ITS: SECTION 3.6.1.3 - PCIVs l TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED (Specification 3.6.1.3 "LR.1" Labeled Comments / Discussions)

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed changes will not involve any physical changes to plant systems, structures, or components (SSC), or the manner in which these SSC are operated, maintained, modified, tested, or inspected. The proposed changes will not impose or eliminate any requirements. Therefore, these changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction in a margin of safety?

The margin of safety as defined in the bases of any Technical Specification is not reduced. The requirements being removed from Technical Specifications, eliminate regulatory control of changes to this requirement (e.g.,

Technical Specification amendment or 10 CFR 50.59) that are j not necessary to provide adequate protection of the public health and safety since the ITS continue to impose the appropriate requirements and limitations.

Removal of these items from Technical Specifications eliminates the requirement for NRC review and approval of revisions in accordance with 10 CFR 50.92. Elimination of this administrative process does not have a margin of safety that can be evaluated. However, the proposed changes continue to provide a level of detail that is consistent with the BWR Standard Technical Specification, NUREG-1433, Rev 1, which was approved by the NRC. Revising the Technical Specifications to reflect the approved level of detail entails no significant reduction in the margin of safety.

i l

I 1

FERMI - UNIT 2 2 REVISION 13, 8/6/99 1

1 I

NO SIGNIFICANT HAZARDS EVALUATION ITS: - SECTION 3.6.2.1 - SUPPRESSION POOL AVERAGE TEMPERATURE TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED (Specification 3.6.2.1 "LR.1" Labeled Comments / Discussions)

Detroit Edison has evaluated the proposed Technical Specification change identified as "Less Restrictive - Removed" in accordance with the criteria specified by 10 CFR 50.92 and has determined that the proposed changes do not involve a significant hazards consideration.

The bases for the determination that the proposed change does not involve a significant hazards consideration is an evaluation of these changes against each of the criteria in 10 CFR 50.92.

The criteria and the conclusions of the evaluation are presented below.

1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?

The proposed change removes fram the Fermi 2 Technical Specifications details concerning system operation in response to suppression pool average temperature above j limits. This change relocates informational or implementing details that are not required to be under regulatory control (e.g., Technical Specification amendment j or 10 CFR 50.59) . The requirements being removed from , i Fermi 2 Technical Specifications are not necessarily being deleted or changed at this time, but are simply eliminated from 10 CFR 50.59 or 10 CFR 50.92 change controls.

Reduction of the level of regulatory control on the j identified requirements is appropriate since Technical Specifications continue to require suppression pool temperature limits be maintained below limits. Given the remaining Technical Specification requirements, the proposed change in the level of regulatory control has no impact on the probability or the consequences of an accident previously evaluated, therefore, these changes have no impact on the probability or consequences of an accident previously evaluated.

FERMI - UNIT 2 1 REVISION 13, 8/6/99

^

1 NO SIGNIFICANT HAZARDS EVALUATION ITS: SECTION 3.6.2.1 - SUPPRESSION POOL AVERAGE TEMPERATURE TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED (Specification 3.6.2.1 "LR.1" Labeled Comments / Discussions)

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

j The proposed changes will not involve any physical changes to plant systems, structures, or components (SSC), or the ]

manner in which these SSC are operated, maintained, {

modified, tested, or inspected. The proposed changes will l not impose or eliminate any requirements. Therefore, these i changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction in a margin of safety?

The margin of safety as defined in the bases of any Technical Specification is not reduced. The requirements being removed from Technical Specifications, eliminate regulatory control of changes to this requirement (e.g.,

Technical Specification amendment or 10 CFR 50.59) that are not necessary to provide adequate protection of the public health and safety since the ITS continue to impose the appropriate requirements and limitations.

Removal of these items from Technical Specifications eliminates the requirement for NRC review and approval of revisions in accordance with 10 CFR 50.92. Elimination of this administrative process does not have a margin of safety that-can be evaluated. However, the proposed l changes continue to provide a level of detail that is l consistent with the BWR Standard Technical Specification, l NUREG-1433, Rev 1, which was approved by the NRC. R' wising j the Technical Specifications to reflect the approved level {

of detail entails no significant reduction in the margin of i safety, i i

I l

FERMI.- UNIT 2 2 REVISION 13, 8/6/99 l

NO SIGNIFICANT HAZARDS EVALUATION ITS: SECTION 3.6.4.1 - SECONDARY CONTAINMENT TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED (Specification 3.6.4.1 "LR.1" Labeled Commsents/ Discussions)

I Detroit Edison has evaluated the proposed Technical Specification change identified as "Less Restrictive - Removed" in accordance with the criteria specified by 10 CFR 50.92 and has determined that the proposed changes do not involve a significant hazards consideration. I The bases for the determination that the proposed change does not involve a significant hazards consideration is an evaluation of these changes against each of the criteria in 10 CFR 50.92.

The criteria and the conclusions of the evaluation are presented below.

1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?

The proposed change removes from the Fermi 2 Technical Specifications verification of the initial drawdown flow rate of one SGT subsystem. This change relocates informational or implementing details that are not required to be under regulatory control (e . g. , Technical Specification amendment or 10 CFR 50.59) . The requirements being removed from Fermi 2 Technical Specifications are not necessarily being deleted or changed at this time, but are simply eliminated from 10 CFR 50.59 or 10 CFR 50.92 change controls. Reduction of the level of regulatory control on the identified requirements is appropriate since remaining Technical Specification requirements assure SGT system flows and Secondary Containment inleakage are within appropriate limits. Given the remaining Technical Specification requirements, the proposed change in the level of regulatory control has no impact on the probability or the consequences of an accident previously evaluated, therefore, these changes have no impact on the probability or consequences of an accident previously evaluated. I l

FERMI'- UNIT 2 1 REVISION 13, 8/6/99 l I

r-NO SIGNIFICANT HAZARDS EVALUATION ITS: SECTION 3.6.4.1 - SECONDARY CONTAINMENT TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED (Specification 3.6.4.1 "LR.1" Labeled Comments / Discussions)

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed changes will not involve any physical changes to plant systems, structures, or components (SSC), or the manner in which these SSC are operated, maintained, modified, tested, or inspected. The proposed changes will not impose or eliminate any requirements. Therefore, these changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction in a margin of safety?

The margin of safety as defined in the bases of any Technical Specification is not reduced. The requirements being removed from Technical Specifications, eliminate regulatory control of changes to this requirement (e.g.,

Technical Specification amendment or 10 CFR 50.59) that are not necessary to provide adequate protection of the public health and safety since the ITS continue to impose the appropriate requirements and limitations.

Removal of these items from Technical Specifications eliminates the requirement for NRC review and approval of revisions in accordance with 10 CFR 50.92. Elimination of this administrative process does not have a margin of safety that can be evaluated. However, the proposed changes continue to provide a level of detail that is consistent with the.BWR Standard Technical Specification, NUREG-1433, Rev 1, which was approved by the NRC. Revising the Technical Specifications to reflect the approved level of detail entails no significant reduction in the margin of safety, i

l l

FERMI - UNIT 2 2 REVISION 13, 8/6/99

F l

l NO SIGNIFICANT HAZARDS EVALUATION ITS: SECTION 3.6.4.2 - SCIVs TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED (Specification 3. 6. 4.2 "LR.1" Labeled Comunents/ Discussions)

Detroit Edison has evaluated the proposed Technical Specification change identified as "Less Restrictive - Removed" in accordance with the. criteria specified by 10 CFR 50.92 and has determined that the proposed changes do not involve a significant hazards consideration.

The bases for the determination that the proposed change does not involve a significant hazards consideration is an evaluation of these changes-against each of the criteria in 10 CFR 50.92.

The criteria and the conclusions of the evaluation are presented below.

1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?

The proposed change removes from the Fezini 2 Technical Specifications the specific requirement to demonstrate Operability of an isolation valve after maintenance or repair. This change relocates informational or implementing details that are not required to be under regulatory control (e.g. , Technical Specification amendment or 10 CFR 50.59) . The requirements being removed from Fermi 2 Technical Specifications are not necessarily being deleted or changed at this time, but are simply eliminated ,

from 10 CFR 50.59 or 10 CFR 50.92 change controls.

Reduction of the level of regulatory control on the identified requirements is appropriate since Operability of the valves is required by the IST program and the Technical Specifications. Given the remaining Technical Specification requirements, the proposed change in the level of regulatory control has no impact on the probability or the consequences of an accident previcusly evaluated, therefore, these changes have no impact on the probability or consequences of an accident previously evaluated.

i l

1 FERMI - UNIT 2' 1 REVISION 13, 9/6/99 1 I

NO SIGNIFICANT HAZARDS. EVALUATION ITS: SECTION 3.6.4.2 - SCIVs TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED (Specification 3.6.4.2 "LR.1" Labeled Comunents/ Discussions)

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

l l The proposed changes will not involve any physical changes to plant systems, structures, or components (SSC), or the manner in which these SSC are operated, maintained, modified, tested, or inspected. The proposed changes will not impose or eliminate any requirements. Therefore, these changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction in a margin of safety?

The margin of safety as defined in +he bases of any l Technical Specification is not reducsd. The requirements being removed from Technical Specifications, elininate regulatory control of changes to this requirement (e g.,

Technical Specification amendment er: 10 CFR 50.59) that are not necessary to provide adequate protection of the pt:blic health and safety since the ITS continue to impose the appropriate requirements and limitations.

Removal of these items from Technical Specifications eliminates the requirement for NRC review and approva.1 of revisions in accordance with 10 CFR 50.92. Elimination of this administrative process does not have a margin of safety that can be evaluated. However, the proposed changes continue to provide a level of detail that is consistent with the BWR Standard Technical Specification, NUREG-1433, Rev 1, which was approved by the NRC. Revising the Technical Specifications to reflect the approved level of detail entails no significant reduction in the margin of safety.

FERMI - UNIT 2 2 REVISION 13, 8/6/99

r NO SIGNIFICANT HAZARDS EVALUATION ITS: SECTION 3.8.1 - AC SOURCES - OPERATING TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED (Specification 3.8.1 "LR.1 " Labeled Comunents/ Discussions)

Detroit Edison has evaluated _the proposed Technical Specification change identified as "Less Restrictive - Removed" l in accordance with the criteria specified by 10 CFR 50.92 and I

has determined that the proposed changes do not involve a significant hazards consideration.

The bases for the determination that the proposed change does not involve a significant hazards consideration is an evaluation of these changes against each of the criteria in 10 CFR 50.92.

The criteria and the conclusions of the evaluation are presented below.

1. Does the change involve a significant increase in the probability or consequences of an accident previously

. evaluated?

The proposed change removes from the Fermi 2 Technical Specifications specific details concerning the alignment of EDGs to provide power to their associated emergency buses.

This change relocates informational or implementing details that are not required to be under regulatory control (e.g.,

Technical Specification amendment or 10 CFR 50.59) . The l

requirements being removed from Fermi-2 Technical Specifications are not necessarily being deleted or changed at this time, but are uimply eliminated from 10 CFR 50.59 or 10 CFR 50.92 change controls. Reduction of the level of regulatory control on the identified requirements is appropriate since this alignment is required by the Technical Specification definition of Operability. Given the remaining Technical Specification requirements, the proposed change in the level of regulatory control has no impact on the probability or the consequences of an accident previously evaluated, therefore, these changes have no impact on the probability or consequences of an accident previously evaluated.

l l

l FERMI - UNIT-2 1 REVISION 13, 8/6/99 1

L

t NO SIGNIFICANT HAZARDS EVALUATION ITS: SECTION 3.8.1 - AC SOURCES - OPERATING TECHNICAL CHANGES .LESS RESTRICTIVE - REMOVED (Specification 3.8.1 "LR.1" Labeled Comunents/ Discussions) l l

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed changes will not involve any physical changes to plant systems, structures, or components (SSC), or the manner in which these SSC are operated, maintained, modified, tested, or inspected. The proposed changes will not impose or eliminate any requirements. Therefore, these changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction in a margin of safety?

The margin of safety as defined in the bases of any Technical Specification is not reduced. The requirements being removed from Technical Specifications, eliminate ,

regulatory control of changes to this requirement (e.g.,

Technical Specification amendment or 10 CFR 50.59) that sre not necessary to provide adequate protection of the public health and safety since the ITS continue to impose the appropriate requirements and limitations.

Removal of these items from Technical Specifications eliminates the requirement for NRC review and approval of revisions in accordance with 10 CFR 50.92. Elimination of this administrative process does not have a margin of safety that can be evaluated. However, the proposed changes continue to provide a level of detail that is consistent with the BWR Standard Technical Specification, NUREG-1433, Rev 1, which was approved by the NRC. Revising ,

the Technical Specifications to reflect the approved level of detail entails no significant reduction in the margin of safety.

FERMI - UNIT 2 2 REVISION 13, 8/6/99

NO SIGNIFICANT HAZARDS EVALUATION ITS: SECTION 3.8.1 - AC SOURCES - OPERATING

( TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED (Specification 3.8.1 "LR.2" Labeled Comments / Discussions)

Detroit Edison has evaluated.the proposed Technical Specification change identified as "Less Restrictive - Removed" in accordance with the criteria specified by 10 CFR 50.92 and has determined that the proposed changes do not involve a significant hazards consideration.

The bases for the determination that the proposed change does not involve a significant hazards consideration is an evaluation of these changes against each of the criteria in 10 CFR 50.92.

The criteria and the conclusions of the evaluation are presented below.

1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?

The proposed change removes from the Fermi 2 Technical Specifications the requirement that the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months EDG run be associated with a 10 second start and confirmation of steady state voltage and frequency. This change relocates informational or implementing details that are not required to be under regulatory control (e.g., Technical Specification amendment or 10 CFR 50.59). The requirements being removed frosa Fermi 2 Technical Specifications are not necessarily being deleted or changed at this time, but are !

simply eliminated frosa 10 CFR 50.59 or 10 CFR 50.92 change '

controls. Reduction of the level of regulatory control on the identified requirements is appropriate since the 10 second start, voltage and frequency requirements remain in Technical Specifications in other Surveillance Requirements. Given the remaining Technical Specification l

requirements, the proposed change in the level of regulatory control has no impact on the probability or the :

consequences of an accident previously evaluated, I therefore, these changes have no impact on the probability or consequences of an accident previously evaluated.

1 l

FERMI - UNIT 2 3 REVISION-13, 8/6/99 l l

I l

NO SIGNIFICANT HAZARDS EVALUATION ITS: SECTION 3.8.1 - AC SOURCES - OPERATING TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED (Specification 3.8.1 "LR.2" Labeled Comments / Discussions)

2. Does the change create the possibility of a new or l different kind of accident from any accident previously i evaluated?

The proposed changes will not involve any physical changes to plant systems, structures, or components (SSC), or the manner in which these SSC are operated, maintained, modified, tested, or inspected. The proposed changes will not impose or eliminate any requirements. Therefore, these changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction in a !

margin of safety? I 1

The margin of safety as defined in the bases of any Technical Specification is not reduced. The requirements being removed from Technical Specifications, eliminate l regulatory control of changes to this requirement (e.g., l Technical Specification amendment or 10 CFR 50.59) that are not necessary to provide adequate protection of the public health and safety since the ITS continue to impose the appropriate requirements and limitations.

Removal of these items from Technical Specifications eliminates the requirement for NRC review and approval of revisions in accordance with 10 CFR 50.92. Elimination of this administrative process does not have a margin of safety that can be evaluated. However, the proposed changes continue to provide a level of detail that is consistent with the BWR Standard Technical Specification, NUREG-1433, Rev 1, which was approved by the NRC. Revising the Technical Specifications to reflect the approved level of detail entails no significant reduction in the margin of

, safety.

FERMI - UNIT 2 4 REVISION 13, 8/6/99 l

NO SIGNIFICANT HAZARDS EVALUATION ITS: SECTION 3.8.1 - AC SOURCES - OPERATING TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED (Specification 3.8.1 "LR.3" Labeled Comments / Discussions)

Detroit Edison has evaluated the proposed Technical Specification change identified as "Less Restrictive - Removed" in accordance with the criteria specified by 10 CFR 50.92 and has determined that the proposed changes do not involve a significant hazards consideration.

The bases for the determination that the proposed change does not involve a significant hazards consideration is an evaluation of these changes against each of the criteria in 10 CFR 50.92.

The criteria and the conclusions of the evaluation are presented below.

1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?

The proposed change removes from the Fermi 2 Technical Specifications an explicit requirement to demonstrate EDG Operability after modifications that could affect EDG interdependence. This change relocates informational or implementing details that are not required to be under regulatory control (e.g., Technical Specification amendment or 10 CFR 50.59). The requirements being removed from Fermi 2 Technical Specifications are not necessarily being deleted or changed at this time, but are simply eliminated from 10 CFR 50.59 or 10 CFR 50.92 change controls.

Reduction of the level of regulatory control on the identified requirements is appropriate since remaining Technical Specification requirements, including the definition of Operability, continue to require EDG independence and Operability. Given the remaining Technical Specification requirements, the proposed change in the level of regulatory control has no impact on the probability or the consequences of an accident previously evaluated, therefore, these changes have no impact on the probability or consequences of an accident previously evaluated.

1 FERMI - UNIT 2 5 REVISION 13, 8/6/99

NO SIGNIFICANT HAZARDS EVALUATION ITS: SECTION 3.8.1 - AC SOURCES - OPERATING TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED (Specification 3.8.1 "LR.3" Labeled Comments / Discussions)

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed changes will not involve any physical changes to plant systems, structures, or components (SSC), or the manner in which these SSC are operated, maintained, modified, tested, or inspected. The proposed changes will not impose or eliminate any requirements. Therefore, these changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction in a margin of safety?

The margin of safety as defined in the bases of any Technical Specification is not reduced. The requirements being removed from Technical Specifications, eliminate regulatory control of changes to this requirement (e.g.,

Technical Specification amendment or 10 CFR 50.59) that are not necessary to provide adequate protection of the public health and safety since the ITS continue to impose the appropriate requirements and limitations.

Removal of these items from Technical Specifications eliminates the requirement for NRC review and approval of revisions in accordance with 10 CFR 50.92. Elimination of this administrative process does not have a margin of safety that can be evaluated. However, the proposed changes continue to provide a level of detail that is consistent with the BWR Standard Technical Specification, NUREG-1433, Rev 1, which was approved by the NRC. Revising the Technical Specifications to reflect the approved level of detail entails no significant reduction in the margin of safety.

FERMI - UNIT 2 6 REVISION 13, 8/6/99 o

NO SIGNIFICANT HAEARDS EVALUATION ITS: SECTION 3.9.1 - REFUELING EQUIPMENT INTERLOCKS TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED (Specification 3.9.1 "LR.1" Labeled Comuments/ Discussions)

Detroit Edison has evaluated the proposed Technical Specification change identified as "Less Restrictive - Removed" in accordance with the criteria specified by 10 CFR 50.92 and has determined that.the proposed changes do not involve a significant hazards consideration.

The bases for the determination that the proposed change does not involve a significant hazards consideration'is an evaluation !

of these changes against each of the criteria in 10 CFR 50.92.

The criteria and.the conclusions of the evaluation are presented i below. l

1. Does the change involvs a significant increase in the probability or consequences of an accident previously evaluated?

The proposed change removes from the Fermi 2 Technical Specifications an explicit requirement to functionally test the refueling interlocks following repair, maintenance or replacement. This. change relocates informational or implementing details that are not required to be under regulatory control (e.g., Technical Specification amendment or 10 CFR 50.59) . The requirements being removed from Fermi 2 Technical Specifications are not necessarily being deleted or changed at this time, but are simply eliminated from 10 CFR 50.59 or 10 CFR 50.92 change controls.

Reduction of the level of regulatory control on the ,

identified requirements is appropriate since remaining !

Technical Specification requirements, including the !

definition of Operability, provide requirements for demonstrating the Operability of the refueling interlocks.

Given the remaining Technical Specification requirements, the proposed change in the level of regulatory control has no impact on the probability or the consequences of an accident previously evaluated, therefore, these changes have no impact on the probability or consequences of an accident previously evaluated.

FERNI - UNIT 2 1 REVISION 13, 8/6/99

F. --

NO SIGNIFICANT HAZARDS EVALUATION ITS: SECTION 3.10.1 - INSERVICE LEAK AND HYDROSTATIC TESTING OPERATION l TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED l (Specification 3.10.1 "LR.1" Labeled Comuments/ Discussions)

Detroit Edison has evaluated the proposed Technical Specification change identified as "Less Restrictive - Removed" in accordance with the criteria specified by 10 CFR 50.92 and has determined that the proposed changes do not involve a significant hazards consideration.

The bases for the determination that the proposed change does not involve a significant hazards consideration is an evaluation of these changes against each of the criteria in 10 CFR 50.92.

The criteria and the conclusions of the evaluation are presented below.

1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated? 1 The proposed change removes from the Fermi 2 Technical Specifications explicit restriction on maximum temperature for RPV hydrostatic testing. This change relocates informational or implementing details that are not required to be under regulatory control (e.g., Technical Specification amendment or 10 CFR 50.59) . The requirements being removed from Fermi 2 Technical Specifications are not necessarily being deleted or changed at this time, but are simply eliminated from 10 CFR 50.59 or 10 CFR 50.92 change controls. Reduction of the level of regulatory control on the identified requirements is appropriate since the ;

appropriate Mode 3 LCOs and compensatory actions in the '

Technical Specifications continue to impose the appropriate limitations on the conduct of this special operation.

Given the remaining Technical Specification requirements, the proposed change in the level of regulatory control has no impact on the. probability or the consequences of an accident previously evaluated, therefore, these changes have no impact on the probability or consequences of an accident previously evaluated.

l i

FERMI - UNIT 2 8/6/99 1 REVISION 13, i

i

NO SIGNIFICANT HAZARDS EVALUATION ITS: SECTION 3.10.1 - INSERVICE LEAK AND HYDROSTATIC TESTING OPERATION TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED (Specification 3.10.1 "LR.1" Labeled Comments / Discussions)

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed changes will not involve any physical changes to plant: systems, structures, or components (SSC), or the manner in which these SSC are operated, maintained, modified, tested, or inspected. The proposed changes will not impose or eliminate any requirements. Therefore, these changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction in a margin of safety?

The margin of safety as defined in the bases of any Technical Specification is not reduced. The requirements being removed from Technical Specifications, eliminate regulatory control of changes to this requirement (e.g., j Technical Specification amendment or 10 CFR 50.59) that are not necessary to provide adequate protection of the public health and safety since the ITS continue to impose the appropriate requirements and limitations.

Removal of these items from Technical Specifications eliminates the requirement for NRC review and approval of revisions in accordance with 10 CFR 50.92. Elimination of this administrative process does not have a margin of safety that can be evaluated. However, the proposed changes continue to provide a level of detail that is consistent with the BWR Standard Technical Specification, NUREG-1433, Rev 1, which was approved by the NRC. Revising the Technical Specifications to reflect the approved level of detail entails no significant reduction in the margin of safety, l

I FERMI - UNIT 2 2 REVISION 13, 8/6/99

r NO SIGNIFICANT RAZARDS EVALUATION ITS: SECTION 3.10.2 - REACTOR MODE SWITCH INTERLOCK TESTING TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED (Specification 3.10.2 "LR.1 " Labeled Comunents/ Discussions)

Detroit Edison has evaluated the proposed Technical Specification change identified as "Less Restrictive - Removed" in accordance with the criteria specified by 10 CFR 50.92 and -

has determined that the proposed changes do not involve a significant hazards consideration.

The bases for the determination that the proposed change does not involve a significant hazards consideration is an evaluation of these_ changes against each of the criteria in 10 CFR 50.92.

The criteria and the conclusions of the evaluation are presented below.

1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?

The proposed change removes from the Fermi 2 Technical i specifications the method of verification that all rod are fully inserted in order to perform reactor mode switch interlock testing using this special operation provision.

This change relocates informational or implementing details that are not required to be under regulatory control (e.g.,

Technical Specification amendment or 10 CFR 50.59) . The requirements being removed from Fermi 2 Technical Specifications are not necessarily being deleted or changed at this time, but are simply eliminated from 10 CFR 50.59 or 10 CFR 50.92 change controls. Reduction of the level of regulatory control on the identified requirements is appropriate since the requirement for all rods to remain fully inserted continues to be required by the Technical Specifications. Given the remaining Technical Specification requirements, the proposed change in the level of regulatory control has no impact on the probability or the consequences of an accident previously evaluated, therefore, these changes have no impact on the probability or consequences of an accident previously evaluated.

I FERMI - UNIT 2 1 REVISION 13, 8/6/99

I'.-

NO SIGNIFICANT HAZARDS EVALUATION ITS: SECTION 3.10.2 - REACTOR MODE SWITCH INTERLOCK TESTING TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED (Specification 3.10.2 "LR.1" Labeled Comments / Discussions)

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed changes will not involve any physical changes to plant systems, structures, or components (SSC), or the manner in which these SSC are operated, maintained, modified, tested, or inspected. The proposed changes will not impose or eliminate any requirements. Therefore, these changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction in a margin of safety?

The margin of safety as defined in the bases of any Technical Specification is not reduced. The requirements being removed from Technical Specifications, eliminate regulatory control of changes to this requirement (e.g.,

Technical Specification amendment or 10 CFR 50.59) that are not necessary to provide adequate protection of the public health and safety since the ITS continue to impose the appropriate requirements and limitations.

Removal of these items from Technical Specifications eliminates the requirement for NRC review and approval of revisions in accordance with 10 CFR 50.92. Elimination of this administrative process does not have a margin of safety that can be evaluated. However, the proposed changes continue to provide a level of detail that is consistent with the BWR Standard Technical Specification, NUREG-1433, Rev 1, which was approved by the NRC. Revising the Technical Specifications to reflect the approved level of detail entails no significant reduction in the margin of safety.

FERMI - UNIT 2 2 REVISION 13, 8/6/99

NO SIGNIFICANT HAZARDS EVALUATION ITS: SECTION 3.10.4 - SINGLE CONTROL ROD WITHDRAWAL - COLD SHUTDOWN TECHNICAL CHANGES - LESS RESTRICTIVE --REMOVED (Specification 3.10.4 "LR.1" Labeled Comuments/ Discussions)

Detroit Edison has evaluated the proposed Technical Specification change identified as "Less Restrictive - Removed" in accordance with the criteria specified by 10 CFR 50.92 and has determined that the proposed changes do not involve a significant hazards consideration.

The bases for the determination that the proposed change does not involve a significant hazards consideration is an evaluation of these changes against each of the criteria in 10 CFR 50.92.

The criteria and'the conclusions of the evaluation are presented below.

1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?

The proposed' change removes from the Fermi 2 Technical Specifications the requirement that the reactor mode switch be " locked" in position for this special operation. This change relocates informational or implementing details that are not required to be under regulatory control (e.g.,

Technical Specification amendment or 10 CFR 50.59) . The requirements being removed from Fermi 2 Technical Specifications are not necessarily being deleted or changed at this time, but are simply eliminated from 10 CFR 50.59 or 10 CFR 50.92 change controls. Reduction of the level of regulatory control on the identified requirements is appropriate since moving the mode switch is adequately addressed by other Technical Specifications, including the Mode definitions of ITS Table 1.1-1. Given the remaining Technical' Specification requirements, the proposed change in the level of regulatory control has no impact on the probability or the consequences of an accident previously evaluated, therefore, these changes have no impact on the probability.or consequences of an accident previously evaluated.

i I

i FERMI - UNIT 2 1 REVISION 13, 8/6/99 l l

l L

'NO SIGNIFICANT HAZARDS EVALUATION ITS: SECTION 3.10.4 - SINGLE CONTROL ROD WITHDRAWAL - COLD SHUTDOWN TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED (Specification 3.10.4 "LR.1" Labeled Comments / Discussions)

2. Does the change create the possibili"y of a new or different kind of accident from any accident previously i evaluated?

The proposed changes will not involve any physical changes to plant systems, structures, or components (SSC), or the manner in which these SSC are operated, maintained, modified, tested, or inspected. The proposed changes will l not impose or eliminate any requirements. Therefore, these changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction in a margin of safety?

The margin of safety as defined in the bases of any Technical Specification is not reduced. The requirements being removed from Technical Specifications, eliminste regulatory control of changes to this requirement (e.g.,

Technical Specification amendment or 10 CFR 50.59) that are not necessary to provide adequate protection of the public !

health and safety since'the ITS continue to impose the ;

appropriate requirements and limitations.

Removal of these items from Technical Specifications ;

eliminates the requirement for NRC review and approval of j revisions in accordance with 10 CFR 50.92. Elimination of '

this administrative process does not have a margin of safety that can be evaluated. However, the proposed changes continue to provide a level of detail that is consistent with the BWR Standard Technical Specification, NUREG-1433, Rev 1, which was approved by the NRC. Revising the Technical Specifications to reflect the approved level of detail entails no significant reduction in the margin of safety.

FERMI - UNIT 2 2 REVISION 13, 8/6/99 1

NO SIGNIFICANT HAZARDS EVALUATION ITS: SECTION 3.10.5 - SINGLE CONTROL ROD WITHDRANAL - REFUELING TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED (Specification 3.10.5 "LR.1" Labeled Comments / Discussions)

)

. Detroit Edison has evaluated the proposed Technical Specification change identified as "Less Restrictive - Removed" in accordance with the criteria specified by 10 CFR 50.92 and has determined that the proposed changes do not involve a significant hazards consideration.

The-bases for the determination that the proposed change does not involve a significant hazards consideration is an evaluation of these changes against each of the criteria in 10 CFR 50.92.

The criteria and the conclusions of the evaluation are presented below.

1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?

The proposed change removes from the Fermi 2 Technical Specifications the requirement that the reactor mode switch be " locked" in position for this special operation. This change relocates informational or implementing details that are not required to be under regulatory control (e.g., i Technical Specification amendment or 10 CFR 50.59) . The '

requirements being removed from Fermi 2 Technical Specifications are not necessarily being deleted or changed ;

at this. time, but are simply eliminated from 10 CFR 50.59 i

or 10 CFR 50.92 change controls. Reduction of the level of ;

regulatory control on.the identified requirements is )

appropriate since moving the mode switch is adequately addressed by other Technical Specifications, including the Mode definitions of ITS Table 1.1-1. Given the remaining Technical Specification requirements, the proposed change in the level of regulatory control has no impact on the '

probability or the consequences of an accident previously j

evaluated, therefore, these changes have no impact on the l probability or consequences of an accident previously evaluated.

i f

FERMI - UNIT 2 1 REVISION 13, 8/6/99

NO SIGNIFICANT HAZARDS EVALUATION ITS: SECTION 3.10.5 - SINGLE CONTROL ROD WITHDRAWAL - REFUELING TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED (Specification 3.10.5 "LR.1" Labeled Comments / Discussions)

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed changes will not involve any physical changes to plant systems, structures, or components (GSC), or the manner in which these SSC are operated, maintained, modified, tested, or inspected. The proposed changes will j not impose or eliminate any requirements. Therefore, these changes do not create the possibility of a new or different {

kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction in a margin of safety? )

The margin of safety as defined in the bases of any Technical Specification is not reduced. The requirements being removed from Technical Specifications, eliminate regulatory control of changes to this requirement (e . g . ,

Technical Specification amendment or 10 CFR 50.59) that are not necessary to provide adequate protection of the public health and safety _since the ITS continue to impose the appropriate requirements and limitations.

Removal of these items from Technical Specifications I eliminates the requirement for NRC review and approval of revisions in accordance with 10 CFR 50.92. Elimination of this administrative process does not have a margin of rafety that can be evaluated. However, the proposed

hanges continue to provide a level of detail that is consistent with the BWR Standard Technical Specification, NUREG-1433, Rev 1, which was approved by the NRC. Revising the Technical Specifications to reflect the approved level of detail entails no significant reduction in the margin of i safety.

FERMI - UNIT 2 2 REVISION 13, 8/6/99

F:

p E

NO SIGNIFICANT HAZARDS EVALUATION ITS: SECTION 3.10.6 - MULTIPLE CONTROL ROD WITHDRAWAL -

REFUELING TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED (Specification 3.10.6 "LR.1" Labeled Comunents/ Discussions)

Detroit Edison has evaluated the proposed Technical Specification change identified as "Less Restrictive - Removed" in accordance with the criteria specified by 10 CFR 50.92 and has determined that the proposed changes do not involve a significant hazards consideration.

The bases for the determination that the proposed change does not involve a significant hazards consideration is an evaluation of these changes against each of the criteria in 10 CFR 50.92.

The criteria and_the conclusions of the evaluation are presented below.

1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?

l The proposed change removes from the Fermi 2 Technical

l. Specifications-the requirement that the reactor mode switch be Operable and " locked" in position for this special operation. This change relocates informational or implementing details that are not required to be under

regulatory control (e.g., Technical Specification amendment or 10 CFR 50.59) . The requirements being. removed from Fermi 2 Technical Specifications are not necessarily being i

deleted or changed at this time, but are simply eliminated I from 10 CFR 50.59 or 10 CFR 50.92 change controls.

Reduction of the level of regulatory control on the identified requirements is appropriate since the l Operability of the mode switch and control of the mode switch position is adequately addressed by other Technical Specifications. Given the remaining Technical l Specification requirements, the proposed change in the

! level of regulatory control has no impact on the probability or the consequences of an accident previously evaluated, therefore, these changes have no impact on the probability or consequences of an accident previously evaluated.

l l

FERMI - UNIT 2 1 REVISION 13, 8/6/99 L

m NO SIGNIFICANT HAZARDS EVALUATION ITS: SECTION 3.10.6 - MULTIPLE CONTROL ROD WITHDRAWAL -

REFUELING TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED (Specification 3.10.6 "LR.1" Labeled Comments / Discussions)

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed changes will not involve any physical changes to_ plant systems, structures, or components (SSC), or the manner in which these SSC are operated, maintained, modified, tested, or inspected. The proposed changes will not impose or eliminate any requirements. Therefore, these changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction in a margin of safety?

The margin of safety as defined in the bases of any Technical Specification is not reduced. The requirements being removed from Technical Specifications, eliminate regulatory control of changes to this. requirement (e.g.,

Technical Specification amendment or 10 CFR 50.59) that are not necessary to provide adequate protection of the public health and safety since the ITS continue to impose the appropriate requirements and limitations.

Removal of these items from Technical Specifications eliminates the requirement for NRC review and approval of revisions in accordance with 10 CFR 50.92. Elimination of this administrative process does not have a margin of l safety that can be evaluated. However, the proposed I changes continue to provide a level of detail that is '

consistent with the BWR Standard Technical Specification, NUREG-1433, Rev 1, which was approved by the NRC. Revising the Technical Specifications to reflect the approved level of detail entails no significant reduction in the margin of safety.

I FERMI - UNIT 2 2 REVISION 13, 8/6/99

NO SIGNIFICANT HAZARDS EVALUATION

.ITS: SECTION 3.10.6 - MULTIPLE CONTROL ROD WITHDRAWAL -

REFUELING TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED (Specification 3.10. 6 ' "LR.2" Labeled Comunents/ Discussions)

Detroit. Edison has evaluated the proposed Technical Specification change identified'as "Less-Restrictive - Removed" )

in accordance with the criteria specified by 10 CFR 50.92 and has determined that-the proposed changes do not involve a significant hazards consideration.

The bases for the determination that the proposed change does not involve a significant hasards consideration is an evaluation of these changes against each of the criteria in~10 CFR 50.92.

The criteria and the conclusions of the evaluation are presented below.

1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?

The proposed change removes from the Fermi 2 Technical Specifications the explicit requirement to functionally test the one-rod-out interlock following replacement of a control rod or control rod drive if the function had been bypassed. This change relocates informational or ;

implementing details that are not required to be under i regulatory control (e.g., Technical Specification amendment or 10 CFR 50.59). The requirements being removed from Fermi 2 Technical Specifications are not necessarily being deleted or changed at this time, but are simply eliminated from 10 CFR 50.59 or 10 CFR 50.92 change controls.

Reduction of the level of regulatory control on the identified requirements is appropriate since the requirement to adequately demonstrate Operability of systems and components following replacement is required by the Technical Specifications, including the definition of Operability.- Given the remaining Technical Specification requirements, the proposed change in the level of regulatory control has no impact on the probability or the consequences of an accident previously evaluated, therefore, these changes have no impact on the probability or consequences of an accident previously evaluated.

I l

FERMI - UNIT 2 3 REVISION 13, 8/6/99

NO SIGNIFICANT HAZARDS EVALUATION ITS: SECTION 3.10.6 - MULTIPLE CONTROL ROD WITHDRAWAL -

REPUELING TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED (Specification 3.10.6 "LR.2" Labeled Comments /Discussione)

2. Does the change create the possibility of a new or different kind of accident from any accident previously i evaluated?

The proposed changes will not involve any physical changes to plant systems, structures, or components (SSC), or the manner in which these SSC are operated, maintained, modified, tested, or inspected. The proposed changes will not impose or eliminate any requirements. Therefore, these changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

1

3. !

Does this change involve a significant reduction in a !

margin of safety?

j l

The margin of safety as defined in the bases of any l Technical Specification is not reduced. The requirements j being removed from Technical Specifications, eliminate !

regulatory control of changes to this requirement (e.g.,

Technical Specification amendment or 10 CFR 50.59) that are not necessary to provide adequate protection of the public health and safety since the ITS continue to impose the l appropriate requirements and limitations.

Removal of these items from Technical Specifications eliminates the requirement for NRC review and approval of revisions in accordance with 10 CFR 50.92. Elimination of this administrative process does not have a margin of safety that can be evaluated. However, the proposed changes continue to provide a level of detail that is consistent with the BWR Standard Technical Specification, NUREG-1433, Rev 1, which was approved by the NRC. Revising the Technical Specifications to reflect the approved level of detail entails no significant reduction in the margin of safety.

FERMI - UNIT 2 4 REVISION 13, 8/6/99

NO SIGNIFICANT HAZARDS EVALUATION ITS: SECTION 3.10.7 - SDM TEST - REFUELING TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED (Specification 3.10.7 "LR.1" T= haled Comments / Discussions)

Detroit Edison has evaluated the proposed Technical Specification change identified as "Less Restrictive - Removed" in accordance with the criteria specified by 10 CFR 50.92 and

- has determined that the proposed changes do not involve a

-significant hazards consideration.

The bases for the determination that the proposed change does not involve a significant hazards consideration is an evaluation of these changes against each of the criteria in 10 CFR 50.92.

The criteria and the conclusions of the evaluation are presented below.

1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?

The proposed change removes from the Fermi 2 Technical Specifications an additional step to observe indicated response of the nuclear instrumentation during the control rod coupling surveillance. This change relocates informational or implementing details that are not required to be under regulatory control (e.g., Technical Specification amendment or 10 CFR 50.59) . The requirements being removed from Fermi 2 Technical Specifications are not necessarily being deleted or changed at this time, but are simply eliminated from 10 CFR 50.59 or 10 CFR 50.92 change controls. Reduction of the level of regulatory control on the identified requirements is appropriate since the safety related functional requirement to verify the control rod coupling is still required by the Technical Specifications.

Given the remaining Technical Specification requirements, the proposed change in the level of regulatory control has no impact on the probability or the consequences of an accident previously evaluated, therefore, these changes have no impact on the probability or consequences of an accident previously evaluated.

FERMI - UNIT 2 1 REVISION 13, 8/6/99

7 1

l NO SIGNIFICANT HAZARDS EVALUATION ITS: SECTION 3.10.7 - SDM TEST - REFUELING TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED (Specification 3.10.7 "LR.1" Labeled Comments / Discussions)

2. Does the change create the possibility of a new or l

I different kind of accident from any. accident previously i evaluated?

The proposed changes will not involve any physical changes to plant systems, structures, or components (SSC), or the manner in which.these SSC are operated, maintained, modified, tested, or inspected. The proposed changes will not impose or eliminate any requirements. Therefore, these j changes do not create the possibility of a new or different i kind of accident from any accident previously evaluated. I

3. Does this change involve a significant reduction in a margin of safety?

The margin of safety as defined in the bases of any Technical Specification is not reduced. The requirements I being removed from Technical Specifications, eliminate regulatory control of changes to this requirement (e.g., l Technical Specification amendment or 10 CFR 50.59) that are !

not necessary to provide adequate protection of the public health and safety since the ITS continue to impose the !

appropriate requirements and limitations. '

Removal of these items from Technical Specifications eliminates the requirement for NRC review and approval of revisions in accordance with 10 CFR 50.92. Elimination of I this administrative process does not have a margin of ,

safety that can be evaluated. However, the proposed

}

changes continue to provide a-level of detail that is consistent with the BWR Standard Technical Specification, NUREG-1433, Rev 1, which was approved by the NRC. Revising the Technical Specifications to reflect the approved level of detail entails no significant reduction in the margin of safety, i

l i

I l

FERMI - UNIT 2 2 REVISION 13, 8/6/99

NO SIGNIFICANT HAZARDS EVALUATION i ITS: SECTION 3.9.1 - REFUELING EQUIPMENT INTERLOCKS i

TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED '

(Specification 3.9.1 "LR.1" Labeled Comments / Discussions)

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

1 The proposed changes will not involve any physical changes to plant systems, structures, or components (SSC), or the manner in which these SSC are operated, maintained, l modified, tested, or inspected. The proposed changes will not impose or eliminate any requirements. Therefore, these l changes do not create the possibility of a new or different j kind of accident from any accident previously evaluated. l i

3. Does this change involve a significant reduction in a '

margin of safety?

The margin of safety as defined in the bases of any Technical Specification is not reduced. The requirements ,

being removed from Technical Specifications, eliminate )

regulatory control of_ changes to this requirement (e.g.,

Technical Specification amendment or 10 CFR 50.59) that are not necessary to provide adequate protection of the public j health and safety since the ITS continue to impose the !

appropriate requirements and limitations.

Removal of these items from Technical Specifications eliminates the requirement for NRC review and approval of revisions in accordance with 10 CFR 50.92. Elimination of this administrative process does not have a margin of I safety that can be evaluated. However, the proposed l' changes continue to provide a level of detail that is consistent with the BWR Standard Technical Specification, NUREG-1433, Rev 1, which was approved by the NRC. Revising l the Technical Specifications to reflect the approved level of detail entails no significant reduction in the margin of i safety.

l FERMI - UNIT 2 2 REVISION 13, 8/6/99

NO SIGNIFICANT HAZARDS EVALUATION ITS: SECTION 3.9.2 - REFUEL POSITION ONE-ROD-OUT INTERLOCK TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED (Specification 3.9.2 "LR.1" Labeled Comments / Discussions)

Detroit Edison has evaluated the proposed Technical Specification change identified as "Less Restrictive - Removed" in accordance with the criteria specified by 10 CFR 50.92 and has determined that the proposed changes do not involve a significant hazards consideration.

The bases for the determination that the proposed change does not involve'a significant hazards consideration is an evaluation of these changes against each of the criteria in 10 CFR 50.92.

The criteria and the conclusions of the evaluation are presented below.

1. Does the change involve a significant increase in the l probability or consequences of an accident previously evaluated?

The proposed change removes from the Fermi 2 Technical Specifications the requirement to " lock" the reactor mode switch in position except for the situation when a control rod is withdrawn in accordance with ITS 3.9.2. This change relocates informational or implementing details that are not required to be under regulatory control (e.g.,

Technical Specification amendment or 10 CFR 50.59). The requirements being removed from Fermi 2 Technical Specifications are not necessarily being deleted or changed at this time, but are simply eliminated from 10 CFR 50.59 or 10 CFR 50.92 change controls. Reduction of the level of regulatory control on the identified requirements is appropriate since the Technical Specifications continue to require the Operability of the one-rod-out interlock.

Given the remaining Technical Specification requirements, the proposed change in the level of regulatory control has no impact on the probability or the consequences of an accident previously evaluated, therefore, these changes have no impact on the probability or consequences of an accident previously evaluated.

-FERMI - UNIT 2 1 REVISION 13, 8/6/99

NO SIGNIFICANT RAZARDS EVALUATION j ITS: SECTION 3.9.2 - REFUEL POSITION ONE-ROD-OUT INTERLOCK TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED (Specification 3.9.2 "LR.1" Labeled Comments / Discussions) i

2. Does the change create the possibility of a new or I different kind of accident from any accident previously evaluated?

The proposed changes will not involve any physical changes to plant systems, structures, or components (SSC), or the manner in which these SSC are operated, maintained, modified, tested, or inspected. The proposed changes will not impose or eliminate any requirements. Therefore, these changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction in a margin of safety?

The margin of safety as defined in the bases of any Technical Specification is not reduced. The requirements being removed from Technical Specifications, eliminate regulatory. control of changes to this requirement (e.g.,

Technical Specification amendment or 10 CFR 50.59) that are not necessary to provide adequate protection of the public health and safety since the ITS continue to impose the appropriate requirements and limitations.

Removal of these items from Technical Specifications ~

eliminates the requirement for NRC review and approval of revisions in accordance with 10 CFR 50.92. Elimination of J this administrative process does not have a margin of safety that can be evaluated. However, the proposed changes continue to provide a level of detail that is consistent with the BWR Standard Technical Specification, NUREG-1433, Rev 1, which was approved by the NRC. Revising the Technical Specifications to reflect the approved level j of detail entails no significant reduction in the margin of '

safety.

l FERMI - UNIT 2 2 REVISION 13, 8/6/99

NO SIGNIFICANT HAZARDS EVALUATION ITS: SECTION 3.9.5 - CONTROL ROD OPERABILITY - REFUELING TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED (Specification 3.9.5 "LR.1" Labeled Coasnents/ Discussions)

Detroit Edison has evaluated the proposed Technical Specification change identified as "Less Restrictive - Removed" in accordance with the criteria specified by 10 CFR 50.92 and has determined that the proposed changes do not involve a s significant hazards consideration. d The bases for the determination that the proposed change does not involve a significant hazards consideration is an evaluation of these changes against each of the criteria in 10 CFR 50.92.

The criteria and the conclusions of the evaluation are presented below.

1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?

The proposed change removes from the Fermi 2 Technical Specifications requirements for indication-only and test equipment associated with the control rod scram accumulators. This change relocates informational or implementing details that are not required to be under regulatory control (e.g.,-Technical Specification amendment or 10 CFR 50.59) . The requirements being removed from Fermi 2 Technical Specifications are not necessarily being deleted or changed'at this time, but are simply eliminated from 10 CFR 50.59 or 10 CFR 50.92 change controls.

Reduction of the level of regulatory control on the l

identified requirements is appropriate since the removed functions do not relate directly to the Operability of any systems and components. The Technical Specifications continue to require the Operability of the control scram accumulators. Given the remaining Technical Specification requirements, the proposed change in the level of I regulatory control has no impact on the probability or the l consequences of an accident previously evaluated, i therefore, these changes have no impact on the probability i or consequences of an accident previously evaluated.

l l

l 1

I I

FERMI - UNIT 2 1 REVISION 13, 8/6/99

[

NO SIGNIFICANT HAZARDS EVALUATION ITS: SECTION 3.9.5 - CONTROL ROD OPERABILITY - REFUELING TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED (Specification 3. 9.5 "LR.1" Labeled Comunents/ Discussions)

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed changes will not involve any physical changes to plant systems, structures, or components (SSC), or the manner in which these SSC are operated, maintained, modified, tested, or inspected. The proposed changes will not impose or eliminate any requirements. Therefore, these changes do not create the possibility of a-new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction in a margin of safety?

The margin of safety as defined in the bases of any Technical Specification is not reduced. The requirements being removed frca Technical Specifications, eliminate regulatory control of changes to this requirement (e.g.,

Technical Specification amendment or 10 CFR 50.59) that are not necessary to provide adequate protection of the public health and safety since the ITS continue to impose the appropriate requirements and limitations.

Removal of these items from Technical Specifications eliminates the requirement for NRC review and approval of revisions in accordance with 10 CFR 50.92. Elimination of this administrative process does not have a margin of safety that can be evaluated. However, the proposed changes continue to provide a level of detail that is consistent with the BWR Standard Technical Specification, NUREG-1433, Rev 1, which was approved by the NRC. Revising the Technical Specifications to reflect the approved level of detail entails no significant reduction in the margin of safety.

FERMI - UNIT 2 2 REVISION 13, 8/6/99 1 I

r NO SIGNIFICANT HAZARDS EVALUATION i

ITS: SECTION 5.1 - RESPONSIBILITY TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED (Specification 5.1 "LR.1" Labeled Comments / Discussions)

Detroit Edison has evaluated the proposed Technical Specification change identified as "Less Restrictive - Removed" in accordance with the criteria specified by 10 CFR 50.92 and has determined that the proposed changes do not involve a significant hazards consideration.

The bases for the determination that the proposed change does not involve a significant hazards consideration is an evaluation of'these changes against each of the criteria in 10 CFR 50.92.

The criteria and the conclusions of the evaluation are presented below.

1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?

The proposed change removes from the Fermi 2 Technical Specifications the restriction that the Shift Technical Advisor cannot assume the control room command function in the absence of the Nuclear Shift Supervisor. This change relocates informational or implementing details that are not required to be under regulatory control (e.g.,

Technical Specification amendment or 10 CFR 50.59) . The requirements being removed from Fermi 2 Technical Specifications are not necessarily being deleted or changed at this time, but are simply _ eliminated from 10 CFR 50.59 or 10 CVR 50.92 change controls. Reduction of the level of regulatory control on the identified requirements is appropriate since the Technical Specifications still require the an individual with a Senior Reactor Operators license assume the control room command function. Given the remaining Technical Specification requirements, the proposed change in the level of regulatory control has no impact on the probability or the consequences of an accident previously evaluated, therefore, these changes have no impact on the probability or consequences of an accident previously evaluated.

FERMI --UNIT 2 1 REVISION 13, 8/6/99

NO SIGNIFICANT HAZA* ADS EVALUATION ITS:.SECTION 5.1 - RESPONSIBILITY TECHNICAL - CHANG.tS - LESS RESTRICTIVF. - REMOVED (Specification 0.1 "LR.1" Labeled Comments / Discussions)

2. Does the change create'the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed :hanges will not involve any physical changes to plant systoms, structu:es, or components (SSC), or the manner in which these SSC are operated, maintained, modified, tes ti wi, or inspected. The proposed changes will not impose or aliminate any requirements. Therefore, these changes do not create the possibility of a new or different kind of accident from s.ny accident previously evaluated.

3. Does this change involve a significant reduction in a margin of safety?

The margin of safety as defined in the bases of any Technical Specification is not reduced. The requirements being removed from Technical Specifications, eliminate regulatory control of changes to this requirement (e . g . ,

Technical Specification amendment or 10 CFR 50.59) - that are not necessary to provide adequate protection of the public health and safety since the ITS continue to impose the appropriate requirements and limitations.

Removal of these items from Technical Specifications eliminates the requirement for NRC review and approval of revisions in accordance with 10 CFR 50.92. Elimination of this administrative process does not have a margin of safety that can be evaluated. However, the proposed changes continue to provide a level of detail that is consistent with the BWR Standard Technical Specification, NUREG-1433, Rev 1, which was approved by the NRC. Revising the Technical Specifications to reflect the approved level of detail entails no significant reduction in the margin of safety.

FERMI - UNIT 2 2 REVISION 13, 8/6/99

NO SIGNIFICANT HAZARDS' EVALUATION ITS: SECTION 5.5 - PROGRAMS AND MANUALS TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED (Specification 5.5 "LR.1" Labeled Comments / Discussions)

Detroit Edison has evaluated the proposed Technical Specification change identified as "Less Restrictive - Removed" in accordance with the criteria specified by 10 CFR 50.92 and has determined that the proposed changes do not involve a significant hazards. consideration. I The bases for the determination that the proposed change does not involve a significant hazards consideration is an evaluation of these changes against each of the criteria in 10 CFR 50.92.

The criteria and the conclusions of the evaluation are presented below. l

1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?

1 The proposed change removes from the Fermi 2 Technical Specifications details of containment leak rate testing frequencies, exceptions, exemptions, and inspections. In addition, reporting details are removed from the Technical I Specifications. This change relocates informational or implementing details that are not required to be under regulatory control (e.g., Technical Specification amendment or 10 CFR 50.59) . The requirements being removed from Fermi 2 Technical Specifications are not necessarily being deleted or changed at this time, but are simply eliminated from 10 CFR 50.59 or 10 CFR 50.92 change controls.

Reduction of the level of regulatory control on the identified requirements is appropriate since these details are adequately addressed by the requirements of the Code of Federal Regulations. Given the remaining Technical Specification requirements, the proposed change in the level of regulatory control has no impact on the probability or the consequences of an accident previously evaluated, therefore, these changes have no impact on the probability or consequences of an accident previously evaluated.

. FERMI - UNIT 2 1 REVISION 13, 8/6/99

F-NO SIGNIFICANT HAZARDS EVALUATION ITS: SECTION.-5.5 - PROGRAMS AND MANUALS TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED (Specification 5.5 "LR.1" Labeled Comments / Discussions)

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed changes will not involve any physical changes to plant systems, structures, or components (SSC), or the manner in which these SSC are operated, maintained, modified, tested, or inspected. The proposed changes will not impose or eliminate any requirements. Therefore, these changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction in a margin of safety?

The margin of safety as defined in the bases of any Technical Specification is not reduced. The requirements being removed from Technical Specifications, eliminate regulatory control of changes to this requirement (e.g. ,

Technical Specification amendment or 10 CFR 50.59) that are not necessary to provide adequate protection of the public health and safety since the ITS continue to impose the appropriate requirements and limitations.

Removal of these items from Technical Specifications eliminates the requirement for NRC review and approval of revisions in accordance with 10 CFR 50.92. Elimination of this administrative process does not have a margin of safety that can be evaluated. However, the proposed changes continue to provide a level of detail that is ,

I consistent with the BWR Standard Technical Specification, NUREG-1433, Rev 1, which was approved by the NRC. Revising the Technical Specifications to reflect the approved level of detail entails no significant reduction in the margin of safety.

FERMI - UNIT 2 2 REVISION 13, 8/6/99

NO SIGNIFICANT HAZARDS EVALUATION

'ITS: SECTION 5.5 - PROGRAMS AND MANUALS TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED (Specification 5.5 "LR.1" Labeled Comunents/ Discussions)

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed changes will not involve any physical changes to plant systems, structures, or components (SSC), or the manner in which these SSC are operated, maintained, modified, tested, or inspected. The proposed changes will not impose or eliminate any requirements. Therefore, these changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction in a margin of safety?

The margin of safety as defined in the bases of any Technical Specification is not reduced. The requirements being removed from Technical Specifications, eliminate regulatory control of changes to this requirement (e.g.,

Technical Specification amendment or 10 CFR 50.59) that are not necessary to provide adequate protection of the public health and safety since the ITS continue.to impose the appropriate requirements and limitations.

Removal of these items from Technical Specifications eliminates the requirement for NRC review and approval of revisions in accordance with 10 CFR 50.92. Elimination of this administrative process does not have a margin of safety that can be evaluated. However, the proposed changes continue'to provide a level of detail that is consistent with the BWR Standard Technical Specification, NUREG-1433, Rev 1, which was approved by the NRC. Revising the Technical Specifications to reflect the approved level of detail entails no significant reduction in the margin of safety.

FERMI - UNIT 2 2 REVISION 13, 8/6/99

e NO SIGNIFICANT HAEARDS EVALUATION ITS: SECTION 5.6 - REPORTING REQUIREMENTS TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED (Specification 5.6 "LR.1 " Labeled Comments / Discussions)

Detroit Edison has evaluated the proposed Technical Specification change identified as "Less Restrictive - Removed" in accordance with the criteria specified by 10 CFR 50.92 and has determined that the proposed changes do not involve a significant hazards consideration.

The bases for the determination that the proposed change does not involve a significant hazards consideration is an evaluation of these changes'against each of the criteria in 10 CFR 50.92.

The criteria and the conclusions of the evaluation are presented below.

1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?

The proposed change removes from the Fermi 2 Technical Specifications the submission of reports covering ECCS outage. data and primary coolant specific activity when it has exceeded its limit. This change relocates informational or implementing details that are not required tre be under regulatory control (e.g., Technical Specification amendment or 10 CFR 50.59) . The requirements being removed from Fermi 2 Technical Specifications are not necessarily being deleted or changed at this time, but are simply eliminated from 10 CFR 50.59 or 10 CFR 50.92 change controls. Reduction of the level of regulatory control on the identified requirements is appropriate since the reports perform no compensatory or mitigative functions and any appropriate reporting requirement associated with these items is required by 10 CFR 50.73. Given the remaining Technical Specification requirements, the proposed change ;

in the level of regulatory control has no impact on the I probability or the consequences of an accident previously evaluated, therefore, these changes have no impact on the probability or consequences of an accident previously evaluated.

FERMI - UNIT 2 1 REVISION 13, 8/6/99 ,

NO SIGNIFICANT HAZARDS EVALUATION ITS: SECTION 5.6 - REPORTING REQUIREMENTS TECHNICAL CHANGES - LESS RESTRICTIVE - REMOVED (Specification 5.6 "LR.1" Labeled Comments / Discussions)

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed changes will not involve any physical changes to plant systems, structuren, or ccaponents (SSC), or the manner in which these SSC are operated, maintained, modified, tested, or inspected. The proposed changes will not impose or eliminate any requirements. Therefore, these changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction in a margin of safety?

The margin of safety as defined in the bases of any Technical Specification is not reduced. The requirements being removed from Technical Specifications, eliminate regulatory control of changes to this requirement (e.g.,

Technical Specification amendment or 10 CFR 50.59) that are not necessary to provide adequate protection of the public health and safety since the ITS continue to impose the appropriate requirements and limitations.

Removal of these items from Technical Specifications eliminates the requirement for NRC review and approval of revisions in accordance with 10 CFR 50.92. Elimination of this administrative process does not have a margin of safety that can be evaluated. However, the proposed changes continue to provide a level of detail that is consistent with the BWR Standard Technical Specification, NUREG-1433, Rev 1, which was approved by the NRC. Revising the Technical Specifications to reflect the approved level ,

of detail entails no significant reduction in the margin of safety.

FERMI - UNIT 2 2 REVISION 13, 8/6/99

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ML20211B999

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