Text

TS Change Request 154 to Licenses DPR-24 & DPR-27,changing TS Sections 15.3.3, Instrumentation Sys & 15.4.1, Operational Safety Review to Incorporate Items,Identified by Comparing Safety Analysis in FSAR & LCOs
	ML20125D553
Person / Time
Site:	Point Beach
Issue date:	12/10/1992
From:	Link B; WISCONSIN ELECTRIC POWER CO.
To:	; NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20125D555	List: ML20125D553; ML20125D556;
References
	CON-NRC-92-141 VPNPD-92-364, NUDOCS 9212150273
	Download: ML20125D553 (25)
	v • d • e

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VPf1PD-92-364 10 CFR 50.4 NRC-92-141 30 CFR 50.90 December 10, 1992 Docurr nt Control Desk U.S. 11UCLEAR REGULATORY COMMISSION Mail Station P1-137 Washington, DC 20055 contlemen:

D_0_Cl%TF 50-2M._AliD_10_:191 IEC11RLCAL_EfLCIElCATIONS .CilhtLGX_ggnJMT_DA tiQDlfICAIl0lls TO TEritti1CE_SFECIFICATIQ1LG SlgTLol!S 15. 3. S htiD_21,_bl ITLllT_13MC]I NUCLl:AILELAl[T, UlLITS 1 AllLL2 In accordance with the requirements of 10 CFR 50.4 and 50.90, Wisconsin Electric Power Company (Licensee) hereby requests amendments to Facility Operating Licenses DPR-24 and DPR-27 for Point Beach Nuclear Plant, Units 1 and 2 respectively, to incorporate changes to the plant Technical Specifications.

The proposed changes will make additions to various tables in Sections 15.3.5, " Instrumentation System," and 15.4.1, " Operational Safety Review," and reformat Table 15.4.1-1," Minimum Frequencies for Checks, Calibrations, and Test of Instrument Channels."

Marked-up Tech.tical Specifications pages, a safety evaluation, and the determination of no significant hazards are enclosed.

DESCRIPTION OF CURIENT LICENSE CONDITION Section 15.3.5, " Instrumentation System" contains requirements for the instrumentation and safety cir( its required to ensure react <>r safety and provide for the automatic initiation of the Engineered Safety Features. The tables included in this section are:

Table 15.3.5-1, " Engineered Safety Features Initiation Instrument Setting Limits" Table 15.3.5-2, " Instrument Operation Conditions for Reactor Trip" T> 15.3.5-3, " Emergency Cooling" Table 15.3.5-4, " Inst.ument Operating Conditions for Isolation Functions" Table 15.3.5-5, " Instrument Operating Condit tons for Indications" 9212150273 921210 PDR ADOCK 05000266 O Of P PDR t wM1m otuiwaww /km nymuw .p

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Section 15.4.1, " Operational Safety Review," specifies the surveillance requirements for items directly related to safety limits and limiting conditions for operation. The tables included in this section are Table 15.4.1-1, " Minimum Frequennies for Checks, Calibrations, and Test of Instrument Channels" Table 15.4.1-2, " Minimum Frequencies for Equipment and Sampling Tests" l DESCRIPTION OF PROPOSED CHANGES ;

This Technical Specification change request proposes to modify Sections 15.3.5 and 15.4.1 to incorporate items identified by comparing the safety analyses in the Point Beach Nuclear Plant Final Safety Analysis Report (FSAR) and the Limiting conditions for Operation (LCO) and survoillance sections of the current Technical Specifications. The items being added are required by the safety analyses for indication or propet functioning of systems for accident mitigation, but are not currently included in the Point Beach Technical Specifications. The proposed changes are as follows:

1. Modifications to Tables 15.3.5-2,-15.3.5-3, and 15.3.5-4 are proposed to remove Column 4,'" Minimum Degree of '

Redundancy," from the tables. Additionally, the definition of degrees of redundancy in Specification 15.1.C.1, as well as the reference to minicum degree of redundancy in Specification 15.3.5.C, will also be removed since they;will no longer be applicable. To ensure that the minimum degree of redundancy is maintaine'l should an instrument channel be declared inoperable, the following endnote will be added to the three tarles:

"'.f a channel is determined to-be inoperable,-

resulting in one less than the total number of channels being operable, power operatic 1 may continue if the following conditions are met:

1. The minimum number of operable channels is still satisfied.

2. The affected channel is placed in trip within-I hour."

This proposed. note is consistent with Standard Technical Specification requirement .

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Document Control Desk ly December 10, 1992 Page 3

2. Modifications to the operator action column of Tables 15.3.5-2, 15.3.5-3, and 15.3.5-4 are proposed. This column currently requires the operator, in most canos, to " Maintain hot shut-down" or " Maintain <50% of rated power." The proposed revi-sion of this column will provide more specific dire: tion to the operator. The column will be modified to read, "Be in hot shutdown in 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months " or "Do <50% of rated power within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ."

3. Additions to Table 15.3.5-2, " Instrument Operation Conditions for Reactor Trip," are proposed. The turbine autostop oil pressure, turbine stop valve position, reactor coolant pump breaker open, and Safety Injection reactor trip features will be added to ensure that the table contains all the features installed at Point Beach that actuate a reactor trip.

Surveillance requirements for those four items are also proposed for addition to Table 15.4.1-1, " Minimum Frequencies for Checks, Calibrations, and Tests of Instrument Channels."

4. The channel requirements, bypass conditions, and associated operator actions for the reactor trip breakers are proposed for addition to Table 15.3.5 1. The proposed bypass conditions and required operator actions are as follows:

"When at power, one channel may be bypassed for up to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months previded that the other channel is operable. When the plant is shutdown and rod withdrawal is possible, restore the inoperable channel to operabic status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or open the Reactor Trip Braakers within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ."

5. A change to the title of Table 15.3.5-3 is praposed. The title will be changed from " Emergency Cooling" to " Engineered Safety Features." This new title will more accurately reflect the contents of the tabic.

6. Item 3.b of Table 15.3.5-3, " Trip of Both Main Feedpumps Starts Motor Driven Pumps," is one of the Anticipated Transient Without Scram Mitigating System Actuation Circuitry (AMSAC) features installed at Point Beach. However, it is the cnly feature included in the current Technical Specifications.

Since this feature is not included in the Standard Technical Specifications and because it does not provide a safety-related function, we propose to remove this. item from the Technical Specifications.

7. Additions to Table 15.3.5-4, " Instrument 7perating Conditions for Isolation Functions" are proposed. The instrument operating conditions for high steam generator water level and safety injection are being added because they cause the feedwater system to be isolated following a steam line rupture

4 Document Control Desk December 10, 1992 i f Page 4 i or steam generator tube rupture. A surveillance to test the logic for high steam generator level is also proposed for l

addition as part of note 17 in Table 15.4.1-1. This addition would require a monthly logic test to be performed. This addition was previously proposed in Technical Specification l Change Request 140, dated July 31, 1990, but is being '

resubmitted in this change for completeness. !

8. A change to the titic of Table 15.3.5-5 la proposed. The titic will be changed from " Instrument Operating Conditions for l

Indications" to " Instrument Operating Conditions for Post-Accident Monitoring Instrumentation." This naw title will more accurately reflect the contents of this table.

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9. The addition of the following explanatory note to Tablo 15.3.5-5 19 proposed:

"The channel requirements in this table refer only to that portion of the instrument channel required for post-accident monitoring. The applicable channels are listed in FSAR Table 7.7-2."

This note is being added to provide sufficient guidance con-corning the requirements of Table 15.3.5-5. This guidance l

1s being provided because several of the instruments included in this table, such as steam generator level and containment pressure provide inputs into the Reactor Protection and Engineered Safeguard Features circuitry. Therefore, the required number of operable channels specified in Tables 15.3.5-2 and 15.3.5-3 may be different than the requirements of Table 15.3.5-5. This note will ensure that ,

the indication portion of the 2nstrument channel is governed by the requirements of Table 15.3.5-5 and that the protection

-portion of the instrument channel is governed by the require-ments of Tables 15.3.5-2 and 15.3.5-3. The associated FSAR table will be incorporated in the 1993 FSAR revision.

10. The removal of Item 6 in Table 15.3.5-5, " Control Rod Misalignment as Monitored by On-line Computer," and subsequent placement of it in Item 19 of Table 15.4.1-1, " Analog Rod Position," is proposed. The actions taken by the operators will not be affected by this change. This change will also ensure that only instrumentation associated with post accident monitoring is included in Table 15.3.5-5.

11. The addition of the instrument operating conditions for the following post-accident monitoring instrumentation to Table 15.3.5-5 is proposed: refueling water storage tank level; RCS wide and narrow range pressure; RCS hot and cold leg temperature; pressurizer level; containment wide, inter-mediate, and low range pressure; condensate storage tank-4

4 Document Control Desk (y December 10, 1992 Page 5 level; steam generator wide and narrow range level; steam generator pressure; and containment isolation valve position indication. These additions are propcsed to make the table as complete as possible by including Type A and the safety-significant Category 1, Hon-Type A post-accident monitoring instrumentation identified in our response to Regulatory Guide 1.97. Additionally, Item 5 of Table 15.3.5-5,

" Auxiliary Feedwater Flowrate," is being split up into two l separate items, "AFW Pump Discharge Flowrate" and "AFW to Steam Generator Flowrate," to clear up some existing contusion concerning the requirements for this item.

12. Modification of Items 3 and 4 of Table 15.3.5-5, " Safety Valve Position Indicator"~cnd " Reactor Coolant System Subcooling" respectively, is proposed. Modifications have been completed which provide two reactor coolant system safety valvo position indicators per valve and two channels of reactor coolant i system subcooling. Therefore, Column 1 of Table 15.3.5-5 for i Items 3 and 4 will be updated to list two channels por valve and two channels, respectively. This addition was previously proposed in Technical Specification Change Requast 138, dated March 30, 1990, but is being resubmitted in this change request for completeness.

13. Reformatting Table 15.4.1-1, " Minimum Frequencies for Checkn, Calibrations, and Tests of Instrument channels" is proposed an a means to make tha table more user friendly and easier to read. The new table will remove the remarks column from the body of the tabic and place the information within the channel description column or at the end of the table in the forn of end notes. These end notes will be referenced within the body of the table by their associated surveillances. The new table will also have a column titled " Plant Canditions When Required" that lists all of the plant conditions during which a given surveillance requirement is applicable.

14. The addition of calibration requirements to Items 1, 2, and 3 of Table 15.4.1-1 is proposed. These calibrations will be performed on the source range, intermediate range, and power range nuclear instruments. These calibrations are currently performed each refueling interval. Therefore, this addition simply makes the performance of these calibrations a Technical Specification requirement. The proposed calibration interval is consistent with Standard Technical Specification requirements.

15. The addition of surveillance requirements to test the ability I to isolate the feedwater system following a safety injection l is proposed. A safety injection signal isolates the feedwater l system by tripping the main feed pumps and shutting the main

! feed regulating valves. The proposed additions to Item 17 of t -

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Document Control Desk i December 10, 1992

/ Page G Table 15.4.1-1 will require these functions to be tested every refueling interval. This test interval is consistent with what is currently being performed at Point Beach Nuclear Plant.

16. The clarification of the calibration and test requirements for the different turbine overspeed trip functions is proposed.

The proposed revision to Item 43 of Table 15.4.1-1 will list each trip function separately for completeness. The current calibration und test intervals will not be changed.

17. Since the condensate storage tanks'are the initial suction source for the auxiliary feedwater pumps, the addition of Item 24 of Table 15.4.1-1 is proposed. This addition will consist of a shift check and a refueling interval calibration requirement for the level instrumentation. The calibration will also be performed on the low level alarm associated with this instrumentation. These surveillance intervals are consistent with those applied to existing level inLtrumentation installed at Point Beach.

18. The addition of a weekly requirement to verify prcper breaker alignment and that the 120 Vac inotrument buses are energized is being proposed. These instrument buses provide reliabic -

power to the reactor protection system. This requirement will be included as Item 14 of' Table 15.4.1-1. The proposed check interval is consistent with Standard Technical Specification requirements.

19. Should the rod insertion limit alarm be determined to be inoperable, there needs to be some method to ensure that the rod insortion' limits are not being violated. To support this, the addition of Note (8) to Item 19.of Table 15.4.1-1 is proposed. This note will state, " Verify that the associated rod insertion limit is not being violated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months whenever the rod insertion limit alarm for a control bank is inoperable."

20. The source and intermediate range. nuclear instrument channels are currently required to be checked during each shift that the instruments are in service. For clarity, a modification is proposed to require this shift surveillance to be performed whenever the instrumentation is not blocken. The. source range' nuclear instrument channels are blocked when 1 of 2 intermediate range channels' exceed 10* amperes. The intermediate range nuclear instrument channels are blocked when 2 of 4 power range channels exceed 10% of full power.

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21. Item 1 of Table 15.4.1-1, " Nuclear Power Range," currently specify a monthly check and quarterly calibration requirement to, " compare incore and excore axial flux difference.

Recalibrete if the absolute difference is greater than or equal to 3 percent" and " upper and lower chambers for axial offset." A revision is proposed to clarify these surveillance requirements. The proposed specification would require a month).y check be performed, during power operation, to,

" Compare the results of the incore detector measurements to NIS axial flux difference." This comparison would be performed using the moveable incore detector system. This specification would additionally require a calibration be performed if the absolute difference is 23 percent.

22. Table 15.4.1-7 currently contains a surveillance interval of

" Prior to aach startup if not done previous week." This has caused some confusion because startup is not defined in the Point Beach Technical Specifications. Therefore, a revision to this surveillance interval is proposed. The proposed revision will read, " Prior to reactor criticality if not performed during the previous week."

23. Table 15.4.1-1 currently contcins a note that states, "Not required during periods of refueling shutdown, but must be performed prior to starting up if it has not been performed during the previous surveillance period." For the same reason as discussed in item 23, a revision to this note is proposed.

The proposed revision will read, "Not required during periods of refueling shutdown, but must be performed prior to reactor criticality if it has not been performed during the previous surveillance period."

24. Table 15.4.1-1 currently nt.ains a note that states, "Not required during periods of c3 fueling shutdown, but must be performed prior to starting up if it has not been performed during the previous surveillance period. Tests of the permissive and low power trip b. stable setpoints which cannot be done during power operations shall be conducted prior to startup if not done in the previous two weeks." For the same reason as discussed in Item 23, a revision to this ncte is proposed. The proposed revision will read, " Tests of permissive and low power trip bistable setpoints which cannot be donc during power operations shall be conducted prior to reactor criticality if not done in the previous two weeks."

= 25. Item 32 of Table 15.4.1-1 currently requires the overpressure mitigating system to be checked on a shift basis. Since this system is not always in operation, the addition of a note to the " Check" column for this item is proposed. This note will state, "A shift check is required when the reactor coolant system is not open to the atmosphere and the reactor coolant

Document Control Desk December 10, 3992 Page 8 cystem temperature is less than the minimum temperature for the in-service pressure test as specified in TS Figure 15.3.1-1."

This will require a shift check to be performed only when the overpressure mitigating system is required to be operable, as stated in Specification 15.3.15.A.1. A similar change was proposed in Technical Specification change Request 145, dated May 30, 1991, but is being included for completeness.

26. Item 20 of Table 15.4.1-1, " Auxiliary Feedwater Flow Rate,"

currently requires the flowrate indication to be checked at each unit startup and shutdown. A modification to the check requirement for this item is proposed. The new requirement would state, "An AFW flow path to each steam gene. hall be demonstrated operabic, following each cold shuto. of greater than 30 days, prior to entering power operation, by verifying AFW flow to each steam generator." This proposed requirement is consistent with Standard Technical Specification requirements.

27. Section 15.5.4 of the Technical Specifications requirec that the spent fuel storage pool be filled with borated water at a concentration of at least 1800 ppm. There is currently no surveillance requirement for level verification. Therefore, the addition of a weekly requirement to physically verify watcr level is proposed. This requirement will be added to Item 7 of Table 15.4.1-2, " Minimum Frequencies for Equipment and Sampaing Tests."

28. In order to ensure that sufficient shutdown margin exists prior to commencing power operation following a refueling shutdown, the addition of a requirement to perform a measurement of control rc.d worth is proposed. This requirement will be added to Item 10 of Table 15.4.1-2 and will be required to be performed following each refueling shutdown prior to commencing power operation.

29. The atmospheric steam dumps are safety-related components used to cool down the reactor coolant system in preparation for going on residual heat removal cooling. Therefore, the addition of a requirement to cycle each atmospheric steam dump on a quarterly basis is proposed. This requirement will ensure the operability of these components. This requirement will be added as Item 28 of Table 15.4.1-2.

30. The crossover steam dump system serves to remove the steam energy from the turbine in the event of a unit trip to prevent an overspeeding event from occurring. Therefore, the addition of a requirement to verify the operability of each crossover steam dump valve by performing a complete valve cycle on a quarterly basis is proposed. This requirement will be ad9ed as Item 29 of Table 15.4.1-2.

Document Control Desk g December 10, 1992 Y Page 9

31. Section 15.3.1.A.6 of the Technical Specifications requires that at least 100KW of prassurizer heaters be available to ensure operability of the pressurizer during steady state ;

power operation. Therefore, the addition of a quarterly requirement to verify that at least 100KW of heaters are available is proposed. This will be accomplished by measuring the circuit current. This reg'2irement will be added as Item 30 of Table 15.4.1-2.

32. The charging pumps are used to malutnin reactor coolant system inventory in the event of a small break loss of coolant accident. Therefore, the addition of a quarterly requirement I to verify the operability of the charging pumps is proposed.

The operability of the pumps will be verified as part of the In-Service Test program. This requirement will be added as Item 31 of Table 15.4.1-2.

33. The potential dilution in progress alarm is used to warn operators of a dilution event when the plant is in cold shutdown. Therefore, 'ho addition of a requirement to verify the operability of thi. alarm prior to placin.3 the plant in a cold shutdown condition is proposed. This requirement will be added as Item 32 of Tablo 15.4.1-2.

34. Iten 10 of Table 15.4.1-2 requires the partial movement of all rods every two weeks, except during periods of retuoling shutdown. However, during hot and cold shutdown conditions, it may not be possible to meet this requirement. In order to correct this inconsistency, a change to this requirement is proposed. This change will require the partial movement of all rods to be performed overy two weeks except when the reactor is subcritical.

DAjlIF AND JUSTIFICATLQlf In a letter subuitted to the NRC on December 3, 1991, we committed to conduct a review of the safety analyses in the Point Beach Nuclear Plant Final Safety Analysis Report (FSAR) against the Limiting Conditions for Operation (LCO) section and surveillance section of the Point Beach Nuclear Plant Technical Specifications.

This review was conducted to determine if any systems, components, or functions taken credit for in our accident analyses should be added to the Technical Specifications. This review identified numerous items that should possess Technical Specification LCOs and surveillances.

This Technical Specification change request only addresses those l items associated with instrumentation. The remaining items l identified during this review process will be addressed in futurc l

Document Control Desk l Decembir 10, 1992 Pago 30 change roguests. Tho justification for each of the items included in the " Description of Propos6u Changes" section of this letter is located in the attached safety evaluation.

It hau been determined that the proposed amendments do not involvo a significant hazards consideration, authorize a significant change in the types or total amounts of any offluent release, or result in !

any significant increaro in individual or cumulative occupational l oxposuro. Wo, therefore, conclude that the proposed amendmonto moet the requirements of 10 CFR 51.22 (c) (9) and that an l environmental impact statement or negativo declaration and environmental impact appraisal nood not be prepared.

In summary, the proposed changes contained in this packago will, when approved, result in an improvement in our Technical Specifications. For this reason, wt. request that you proccss this <

change at the earliest opportunity. We also request that at least sixty days be allowed following issuance for implomontation because of' the large number of procedural upgrades that will be required to support t ho amendments.

Pleans contact us if there are any qt. cations.

Sincerely,

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Bob Link Vice President Nuclear Power FDP/jg Enclctures cc: NHC Regional Administrator NRC Resident Inspector Public Servico Comminaion of Wisconsin Subscribed and syqrn b9fore me on this lCM day of Lj u m M yt_ 1992.

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j IECijfilfqAL SPECIFICATIONS CHANGE REOUEST 154 EAFETY EVALUATION l

IHIR9DMCTIRif Wisconsin Electric Power Company (Licensco) is applying for amendments to Facility Operating Licensos DPR-24 and DPR-27 for Point Beach Nuclear Plant, Units 1 and 2. The amendments propose to make additions to various tables in Technical Specification Sections 15.3.5, " Instrumentation," and 15.4.1, " Operational Pafety Review," and reformat Table 15.4.1-1, " Minimum Frequencies ,

for Checks, Calibrations, and Test of Instrument Channels." ,

EVALUATION This Technical Specification change request is bojng submitted to incorporato items identiflod during a compariron of the safety analysos in the Point Beach Nuclear Plant Final Safety Analysis Report and the curront Technical Specifications.

The minimum degree of redundancy is defined as the difference between the number of operable channels and the minimum number of channels which, when tripped, will causo an automatic shutdown. The reason for specifying a minimum degroo of redundancy is to ensure that the reactor-protection system has the ability to moot single failure critoria.- Should a channel bo declared inoperable, resulting in one loss than the total number of channels being operable, the channel will be placed in trip.

This causos the inoperabio channel to provido an input signal to the reactor protection system, ensuring that-the minimum degree of redundancy is maintained. Currently thoro is no time requiremont specified in the Technical. Specifications for operating in this configuration. This revision will require the operator to place an inoperable channel in trip within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months of datormining that a channel is inoperable. This timo limit-is consistent with Standard-Techni:21 Specification requirements.

Since this requiremont is being added, there is consequently no nood to include the minimum degroo of redundancy columns in Tables 15.3.5-2, 15.3.5-3, and 15.3.5-4.

Currently, in Tables 15.3.5-2, 15.3.5-3, and 15.3.5-4 the t

operator action columns specify " Maintain hot shutdown," " Hot Shutdown," or " Maintain <50% of rated power." These action statements are sufficle;.c if the plant is shutdown when operator action ic required, but they are inadequato if the plant is operating at power. To correct this problem, the throo state-monts above will be replaced by "Be in 1.at shutdown in 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months " and "Be <50% of rated power within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ." These statomonts L provide specific instructions to the operator. The O-and-4 hour

timo limits are based on the amount of time it would take to reduce turbine from 100% load to the required load without 1

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excoeding the 15% per nour limit. This will allow a controlled reduction of power to take place without challenging any plant systems.

The addition of the four reactor trip features to Table 15.3.5-2 is being performed so that the table contains all the features that actuate a reactor trip at Point Beach. Item 11 of Table 15.3.5-2, namely "Turbino Trip," is being split up to list the requirerents for both the turbine aute.stop oil pressure and turbine stnp valve position reactor trips. The required operator action for both features will be identical to current require-mentn. The safety injection reactor trip is being added as Item 15. The number of channels, minimum number of channels, and required operator action are identical to those specified in Table 15.3.5-3, Item 1. The reactor coolant pump breaker open reactor trip will be separated into two parts because the channel requirements change based on power level. Both channels are required to initiate a trip when less than 50% power while only one channel is required when greater than 50% power. The operator actions for these four cases are identical to the actions for the remaining items in Tabla 16.3.5-2. Monthly test requirements, except during refueling shutdown, for the safety injection and turbine trip features are also being added to Table 15.4.1-1. These requirements are identical with what is currently being performed at Point Beach Nuclear Plant.

Currently, no LCO exists in the point Beach Nuclear Plant Technical Specifications for the reactor trip breakers. Point Beach has two trains of reactor t71p breakers for each unit with only one channel required to initiate a reactor trip. . The proposed operator action, should either channel be declared inoperable, is to be in hot shutdown within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . This requirement is identical to the other reactor trip features included in Table 15.3.5-2, and is consistent with Standard Technical Specification requirements.

A bypass condition for the reactor trip breakers is also proposed. This proposed bypass condition states that, "When at power, one channel may ba bypassed for up to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months provided that the other channel is operable. When the plant is shutdown and rod withdrawal is possible, restore the inoperable channel to operable status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or open the F2 actor Trip breakers l

within ' hour." The proposed operating conditions and required operator actions are identical to the requirements of the Standard Technical Specifications with the exception of the 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period allo 9ed. The Standard Technical Specifications only allows 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for testing We request an 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period to ,

allow us sufficient time to perform required reactor protection system logic testing and reactor trip breaker vporability testing. Likely, it would not be possible to perform this testing in the two-hour period specified by the Standard l

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g Technical Specifications. Additionally, this 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period would provide us with a short period of time to perform required maintenance on the breakers.

The current point Beach Technical opecifications provide the operating conditions for the instrument channels which start the motor-driven auxiliary foodwater pumps upon a trip of both main feedwater pumps. This feature was installed, as part of the Anticipated Transient Without Scram Mitigating System Actuation '

Circuitry (AMSAC), to meet the requirements of 10 CFR 50.62.

AMSAC features are not required to be Technical Specification items. For this reason, we propose to remove this line item from Table 15.3.5-3. Removing this line item from the Technical >

Specifications will not leesen the maintenance of, and opera-tional requirements for tne AMSAC system because administrative controls will be in place to meintain the system.

Table 15.3.5-4 contains functional units that cause the isolation of a given system. Should a Safety Injection or nigh steam generat c water level occur, the feedwater system will be auto-matically isolated by tripping the associated main feed pumps and shutting the associated main feed regulating valves. Currently, instrument operating conditions for these features are not included in Table 15.3.5-4. The required operator actions in these two cases will be identical to the actions already existing for the remaining items in Table 15.3.5+4. The operator actions require the plant to be placed in hot shutdown within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , should the minimum number of operable channels not be satisfied.

Since the operating conditions for feedwater isclation are proposed for addition to Tab?e 15.3.5-4, an associated surveillance is proposed for addition 'n Table 15.4.1-1, Item 8.

This surveillance would require a monthly test of the logic for high steam generator level. This r,urveillance would satisfy a commitment made in our March 20, 1990, response to Generic Letter 89-19, " Safety Implications of Control Systems in LWR Nuclear Power Plants." The high level bistables are already tested monthly, so this addition only makes the performance of this test a Technical Specification requirement. This uur-veillance was previously propor in Technical Specification Change Request 140, dated July .4, 1990, but is being resubmitted in this change for completeness. Additionally, a requirement to perform a refueling interval test of the feedwater isolation feature following a safety Injection signal is being added as Item 17 of Table 16.4.1-1.

Table 15.3.5-5 lists the instrument operating conditions for the post-accident monitoring instrumentation installed at Point Beach Nuclear Plant. This instrumentation ir used to provide the primary information required to permit the control room operator to take specific manually-controlled actions for which no auto--

matic control is provided, and is required for safety systems to 3

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j accompliuh their safety functions for design basis accident events. In addition to the instrumentation currently listed in Table 15.3.5-5, instrumentation for 13 additional indications has becn identified as being necescary for post-accident monitoring.

The required operator actions for all of the additional instrumentation, with the exception of containment isolation valvo position indication, will allow 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months to rostore operability or require the plant to bo in hot shutdown within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . These actions are consistant with the opurats; actions for the other instrumentation in Tabin 15.3.5-5 and with Standard Technical Specification requirements. The proposed operator actions for containment isolation 'taiva position indication will allow 7 days to restore operability of the shut position indication or require the affected valvo to be closed or the plant to be placed in hot shutdown within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . This requirement is consistent with the requirement for Item 3 of Table 15.3.5-5, " Safety Valve Position Indicator." The requiro-ment for operability of only the shut indication is because that is the required valvo position following a containment isolation signal. These additions will onsure that all Type A and safety-significant Category 1, non-Type A instrumentation identified during our response to Regulatory Guido 1.97 are included. The only Category 1, non-Type A instrumentation not being added to Tablo 15.3.5-5 is reactor coolant system radioactivity concon-tration monitors.

Peactor coolant system radiohetivity concentration is not being included because the Tnalysis of grab samples is adequate to detect a fuel cladding breach. Additionally, the postsaccident sampling system and various area monitors serve as backups and will provido acceptable instrumentation for this item.

" Auxiliary Teodwater Flow Rate" is bein; split ip into two different items, namely, "AFW Pump Discharge Flowrnte" and "AFW to Steam Generator Flowrate." This is being done to clear up come confusion concerning which channels are required to monitor flowrate to the steam generators. There are 3 channels that monitor auxilia>y feedwater pump discharge flowrate and 2 channels that monitor auxiliary foodwater to steam generator flowrate. The operator action statement is being modiflad for "AFW Pump Discharge Flowrato" to road, "If the minimum number of I

AFW Pump Dischargo Flowrate channels requirdd to provide indica-tion of AFW flow to both steam generators cannot be rostored to an operable status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , be in hot shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ." Thn operator action statement for "AFW to Steam Generator Flowrate" is being tuodified to read, "If operability cannot be restored within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , be in hot shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ."

An explanatory note will be added to Table 15.3.5-5 that states, l "The minimum number of operable channels for AFW Pump Dischargo

( Flowrate is the number of AFW Pump Discharge Flowrato channels, 4

1 i

j in conjunction with the number of 6perable AFW to Steam Generator Flowrato channels, required to provide indication of AFW flow to both steam generators." This note is being included because the minimuin required channels for AFW pump Discharge Flowrate will change based upon the number of operable AFW to Steam Generator Flowrate channels. The operator actions for both of those items will ensure that the ability to monitor AFW flow to both steam generators is available. The associated operator actions are also consistent with Stsndard Technical Spec'.fication require-monts. This proposed revision will additionally allow an existing Technical Specification interpretation to be removed from the Point Beach Duty and Call Superintendent Handbook.

In addition to adding the 13 new items to Table 15.3.5-5, a note is being added to this tabic that states, "The channel require-monts in this table refer only to that portion of the instrument channel required for post accident monitoring. The apolicable channels are listed in FSAR Table 7.7-2." This note is being added to provide sufficient guidance concerning the requirements of this table. This guidance is being provided because several of the instruments included in this tabic, such as steam generator water level and containment pressure provide inputs into the reactor protection and Engineered Safety Features circuitry. Therefore, the required number of operable channels specified in Tables 15.3.5-2 and 15.3.5-3 f'r these items may be different than the requirements of Table 15.3.5-5. This note will ensure that the indication portion of the instrument channel is guverned by the requirements of Table 15.3.5-5, and that the protection portion of the instrument is governed by the require-ments of Tables 15.3.5-2 and 15.3.5-3. The actual table listing all of the applicable post-accident instrumentation, by instru-ment tag number, will be added to the point seach Nuclear Plant Final Safety Analysis Report as part of the 1993 annual revision.

Items 3 and 4 of Table 15.3.5-5, " Safety Valve Position Indication" and " Reactor Coolant System Subcooling," are also being revised. Modifications to both of these systems have been completed which now provide 2 safety valve position indicator channels per valve and 2 channels of reactor coolant system subcooling. In order to ensure that Table 15.3.5-5 is accurate, Column 1 for items 3 and 4 will be changed to reflect the correct number of available channels. These changes were originally submitted in Technical Specification change Request 138, dated March 30, 1990, but are being resubmitted in this change for completeness.

Calibrations of the source range, intermediate range, and power range nuclear instruments are currently being performed each refueling interval not to exceed eighteen months. The safety significance of ensuring that these instruments are properly calibrated warrants their addition to Table 15.4.1-1 of the Technical Specifications. The requirement to calibrate these 5

j instruments each refueling interval is connistent with the guidance provided in Standard Technical Specifications and with current Point Beach Nuclear Plant practices.

The source and intermediate range nuclear instruments are currently required to be checked on a shift basis whenever tia

nstruments are in service. For clarity, a change is proposed that will require this shift check .o be performed whenever the instrumentation is not blocked. This revision will not change any actions taken at Point Beach Nuclear Plant, but it will make it clear when this shift check is required. The source range nuclear instrument channels are blocked when 1 of 2 intermediate range channels exceed 10* amperes. The intermediate range nuc1 car instrument channels can be blocked when 2 of 4 power range channels exceed 10% of full power.

Currently, a core flux map is performed at Point Deach every month when the reactor is critical. This flux map is used to confirm that core hot channel limits are being satisfied.

Additionally, the incore measurements obtained during this flux mapping are compared to the excore channels. The excore channels provide input for axial flux difference. If an cbsolute difference 23 percent is present, a recalibrntion is performed.

The proposed change would revise Item 1 of Table 15.4.1-1 to more clearly describa the surveillances performed at Point Beach. The proposed change would require a monthly check to, " compare resulte of the incore detector measurements to NIS axial flux difference." If the absolute difference is 23 percent, a recalibration will still be performed. The existing quarterly requirement is not being included in the change becaure the proposed surveillance frequency will ensure that the difference between the incore and excore vsadings remains small. This proposed change will not change the way that Point Beach is operated.

One of the many events generated by a safety inje?, tion signal is the isolation of the main feedwater system. This isolation is accomplished by tripping both main feed pumps and by shutting both main feed regslating valves. To ensure the operability of this feature, a refueling interval test requirement will be added to Table 15.4.1-1. This test interval is consistent with current Point Beach surveillance requirements for other Safety Injection features.

Currently, Table 15.4.1-1 requires the turbine overspeed trip features to be tested monthly, except when in refueling shutdown, and calibratad each refueling interval. In order to be complete, both the Independent overspeed Protection system and overspeed block trip features will be listed separately under the general l

6 l

heading of " Turbine overspeed-Trips" in Item 43. The existing surveillance requirements will not be changed and will be the same for both overspeed trip features.

The condensate storage tanks are the initial suction sourcas for the auxiliary feed water pumps. For this reason, a shift check cf condensate storage tank level will be added to Tablo 15.4.1-1. 1 The level instrument will also be required to be calibrated each I refueling interval not to exceed eighteen months. A test requirement is not being added oecause condensate storage tank j level does not provide any input into a protective system. While i the Standard Technical Specifications do not provide any guidance ,

for this specific level instrument, the proposed check and cali- i bration requirements are consistent with the Standard Technical i specification requirements for other level instruments.

The 120 Volt instrument buses provide reliable-power to the reactor protection system. For this reason, a weekly check will j be added to Table 25.4.1-1. This check will verify proper breaker alignment and bus voltages. This check requirement is consistent with the requirements in Standard Technical Specifications.

The rod insertion limit alarm is used to ensure that control rods ate sufficiently withdrawn to ensure that adequate shutdown i reactivity is inserted by a reactor trip should~one be required.

Should this alarm become inoperable, there are currently no compensatory measures included in the Technical Specif3ca- !

tions. For this reason, a note is being added to Item 19 of l Table 15.4.1-1, " Analog Rod Position," that states " Verify that the associated rod insertion limit is not being violated at least once por 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months whenever the rod insertion limit alarm for a control bank is inoperable." This nota will ensure that an adequate verification of control rod position is being performed.

This requirement is consistent with Standard Technical-Specification requirements.

Table.15.4.1-1 currently contains a surveillance requirement,

" Prior to startup if not done previous week." This requirement has caused some confusion because startup is not defined in the Peint Beach Technical Specifications. An evaluation of this requirement determined that the associated surveillances are.all required to be performed prior to reactor criticality.. There-fore, this surycillanca requirement is being modified to read,

" Prior to reactor criticality if not performed during the previous veek."

Table 15.4.1-1 currently contains a note that states, "Not required during periods of refueling shutdown, but must be performed prior to starting up if it has not been performed ,

during the previous surveillance period."- For the same reason as discussed above, this note is being modified to read, "Not 7

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t required during periods of refueling shutdown, but must be performed prior to reactor criticality if it has not been performed during the previous surveillance pcriod."

Table 15.4.1-1 currently contains a note that states, "Not required during periods of refueling shutdown, but must be performed prjor to starting up if it has not been performed during the previous surveillance period. Tests of the permissive and low power trip bistable setpoints which cannot be done during power operations shall be conducted prior to startup if not done in the previous two weeks." For the same reason as discussed above, this note is being modified to read, " Tests of permissive and low power trip bistable setpoints whicn cannot be done during power operations chall be conducted prior to reactor criticality if not done in the previous two weeks."

Table 15.4.1-1 currently requires the overpressure mitigating system to be checked on a shift basis. Since this system is not '

always in operation, a shift check should not be required. To >

clarify this requirement, s note is being added to the table that states, "A shift check is required when the reactor coolant system is not open to the a:mouphere and the reactor coolant system temperature is less than the minimum temperature for the inservice pressure test as specified in TS Figure 15.3.1-1."

This will require the shift check to be made whenever the overpressure mitigatipg system is raquired to.be operable, as stated in Specification 15.3.15.A,1. This addition was previously proposed in Technical Specification Chaage Request 145, dated May 30, 1991, but is being included in this change for completeness.

Item 20 of Table 15.4.1-1, " Auxiliary Feedwater Flow Rate,"

currently requires the flowrate indication to be checked at each unit startup and shutdown This item was added to the Technical Specifications in respot- to NUREG-0737, "TMI Action Plans." We propose to clarify the requirement to state, "An AFW flow path to each steam generator shall be demonstrated operable, following each cold shutdown of greater'than 30 days, prior to entering power operation, by vor.ifying AFW flow to each steam generator."

This revision will reduce the number of times that AFW flowrate is verified. This proposed requirement is consistent with Standard Technical Specification requirements.

Section 15.5.4 of the current Technical Specifications requires that the spent fuel storage pool be filled with borated water at a concentration'of at least 1800 ppm. The boron concentration of the pool is currently checked monthly, but there is no surveil-j lance for pool level. For this reason, a requirement to physi-l cally check pool level weekly is being added to Table 15.4.1-2.

l This weekly interval is consistent with Standard Technical .

i Specification requirements. !

8

i In order to ensure that sufficient shutdown margin exists prior to cor:moncing power operation following a refueling shutdown, the Technical Services Group at Point Beach Nuclear Plant currently

- performs a measurement of control rod worth. This measurement is performed following reactor criticality and prior to power operation. Because of the safety significance of this measure-ment, a requirement to perform the mqasurement is being added to Table 15.4.1-2. This measurement will be required following reactor criticality and prior to power operation duri.;g the plant startup following each refueling shutdown. This is consistent with current Point Beach Nuclear Plant requirements.

The atmospheric steam dumps are used to cool down the reactor coolant system in preparation for going on residual heat removal cooling. They are currently tested on a quarterly basis as part i of the Point Beach In-Cervice Test program. This testing is ;

performed to meet the requirements of ASME Boiler and Pressure <

Vessel Code,Section XI, " Rules for In-Service Inspection of Nuclear Power Plant Components." The addition of this identical requirement to Table 15.4.1-2 will ensure that the requirements of bection XI are delineated in the Technical Specifications for these valves.

The crossover steam dump system serves to remove steam energy from the main turbine in the event of a unit trip. This ensures that a turbine overspeeding event does not occur. Currently, the operability of the crossover steam dump valves is verified quarterly by the performance of a periodic check. Because of the importance of this system in preventing a turbine from over-L speeding, this quarterly test requirement will be added to 1 Table 15.4.1-2. Performing this quarterly operability test will ensure that the crossover steam dump system is tested at the same frequency as ASME Section XI components Section 15.3.1.A.6 requires that at least 100KW of pressuriger

heaters are available. This amount is required to ensure the

! operability of the pressurizer. Currently, there is no sur-

! veillance requirement in Technical Specifications concerning the <

heaters. For this reason, a quarterly requirement is being added to Table 15.4.1-2. The capacity of the pressurizer heaters shall bo /erified by measuring their circuit current. This surveil-lance and testing interval are consistent with Standard Technical Specification requirements.-

The charging pumps are used to maintain reactor coolant system inventory in the event of a small break loss of coolant accident.

There are currently no Technical Specification surveillance .,'

requirements to determine the operability of these pumps. How-ever, a functional flow test is performed quarterly as part of the Point Beach In-Service Test program in order-to meet the-requirements of ASME Boiler and Pressure Vessel Code,Section XI, hRules for In-Service Inspection of Nuclear Power Plant Compo-9 m~---r.w- --we-,,m _...w - . . - . _+_m-._.+s--e ve -- _w wte-w a.. 4 .+,ww. v . -- e4 e--,,_m%m+-wwwm.,-.-.ww,.,,,,,-wrwvy,=rv,vv,-,,- v,,v.g.,, . . ~ , w,,-y

nonts." An addition will be made to Table 15.4.1-2, Item 31 to incorporate the existing surveillance requiremonto into the Technical Specifications. ,

The Point Beach Final Safety Analysis Roport accident analysis identifies the potential dilution in progress alarm as the >

primary featuro protecting the plant when in cold shutdown. The operability of this alarm is currently required, by proceduro, to be veriflod prior to placing the plant in a cold shutdown condi-tion. Because this alarm is a primary protective featuro, the requiremont to verify the operability of this alarm prior to entoring cold shutdown is being added to Tablo 15.4.1-2, Item _32.

This verification will onsure that a unit is protected against a Loactivity addition ovent when in cold shutdown. No other sur-veillancos are proposed because this feature providos only the control room alarm.

Item 10 of Table 15.4.1-2 currently requires the partial movement !

of all rods overy two wooks, except during periods of refueling shutdown. Ilowever, during both hot and cold shutdown conditions, it may not be possible to moot this requiremont. Thorofore, in order to correct this discropancy, the frequency of this surveil-lance is being changed. The now test frequency proposed will bo overy two wooks ok apt when the reactor is subcritical.

The remaining changes are all administrative in nature. The title.of Table 15.3.5-3 is being changed from " Emergency Cooling" to "Engincored Safeguard Features" and the title of Table 15.3.5-5 is being changed from " Instrument operating Conditions for Indications" to Instrument Operating Conditions-for post Accident Monitoring Instrumentation." Those changes are ,

being made to accurately reflect the contents of their respective tables.

One surveillance requiremont, namely the monitoring for control-rod misalignment using the on-lino computer is boing moved from ,

Table 15.3.5-S to Tablo-15.4.1-1. The existing surveillance ,

requiromonts are not being changed. The location of the surveillance is being changed to removo-it from the post-accident monitoring instrumentation table and place it in Item 19 in Table 15.4.1-1, " Analog Rod Position."

Finally, in an attempt to make the table easier to road and moro-user friendly, Table 15.4.1-1 is being reformatted. The existing surveillance requirements, except for those specifically addressed by this change request, are not being changed. The -

major changes to this table consist of the following: ,

1. The " Remarks" column is being removed from the body of the ,

, table. The existing romarks will be incorporated as part of L each individual channel description or they will bo . included as end notes following the table. '

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2. A " Plant Conditions When Required" column is being added to the body of the tablo. This column will specify the piant conditions, as defined in Technical Specifications Section ,.

15.1, " Definitions," during which a specific surveillance '

requirement is applicable. These requirements are consistent with existing specifications.

3. The items are being rearranged within the table.in order to place them in a more logical order.

CONCLUSIONS In summary, the above proposed changes to the Point Beach Nuclear Pla.t Technical Specifications are being made in an attempt to make the document more complete and user friendly. Numerous requirements are being added. The time intervals specified for -

these additional Technical Specification requirements are either consistent with Standard Technical Specifications requirements ,

or, if they are already being performed at Point Beach, with the established surveillance requirements. Therefore, the addition of these new LCow and surveillances, as well an the proposed modifications to existing requirements, will ensure and enhance the continued safe operation of Point Beach.

}

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L.

1 TECHNICAL SPECIFICATION CHANGF J QQEST 154 e

"NO SIGNIFICANT HAZARDS CG JIDERATION" In accordance with the enquirements of 10 CFR 50.91(a), Wisconsin Electric Power Company (Licennee) has evaluated the proposed changes against the standards of 10 CFR 50.92 and has determined '

that the operation of Point Beach Nuc1 car Plant, Units 1 and 2 in accordance with the proposed amendments does not present a signi--

ficant hazards consideration. The analysis of the requirements of 10 CFR 50.92 and the basis for this conclusion are as follows: '

1. Operation of this facility under the proposed Tech-nical Specification changes will not ottata _ a signi-ficant increase in the probability or consequennts of an accident provicualy evaluated. These pro-posed changes will make additionc to various tables in Sections 15.3.5 and 15.4.1, and reformat Table 15.4.1-1. These proposed changes will add additional requirements to the Technical Specifica-tions, making the document acoc restrictive. The majority of the items being added to the Technical Specifications are already being performed at Point Beach. They are being mado Technical Spccification requirements because of their safety significance.

Therefore, the addition of those items will not change the operation of Point Bcach. It will, however, put more stringent controls in place to ensure that the 1 associated requirements are met.

This change also proposes to' add requirements to the Technical Specifications that are not currently being performed at Point Beach. Their addition can only improve the overall operation of Point Beach. The surveillances and surveillance frequencies for these additional items are consistent with requirements '

existing in the Standard Technical ~ Specifications or existing Point Beach Technical Specifications.

, This change also proposes removing the operating conditions for the motor-driven AFW pump actuation.

L following the trip of both main feed pumps from i Table 15.3.5-3. This AMSAC (ATWS Mitigating System Actuation Circuitry) feature is being removed because it does not provide a safety-related function. This ,

system is required by 10 CFR 50.62. Administrative controls will be put in place priorito the implemen-tation of this change to ensure that the requirements for AMSAC are maintained. This change also prc)oses to reduce the surveillance frequency for verification of l AFW flowrate. The current surveillance requires AFW flowrate to be verified during each unit startup and shutdown. The proposed requirement would require this w:- .- --. . .-- - - - - -- - - - - - - - - - -

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i flowrate verification to be performed prior to enter- ,

ing power operation, following each cold shutdown of i greater than 30 days. Although this proposed sur-veillance interval is less frequent than what is currently required, it is consistent with Standard Technical Specification requirements.

The remaining changes are all administrative in nature.

The only significant administrative item is the

reformatting of Table 15.4.1-1, " Minimum Frequeycles for Checks, Calibrations, and Tests of Instrument -

Channels." The reformatting of Table 15.4.1-1 will not modify any existing surveillance requirements. j However, it will revise the table to make it more, user friendly and easier to read. This will enhance the useability of this table and ensure that all of ti.e required surveillances are clearly identified. <

There is no physical change to the facility, its systems, or its operation. Thus, an increased probability or consequences of an accident previously evaluated cannot occur.

2. Operation of this facility under the proposed Technical Specification changes will not create the possibility of a new or different kind of accident from any accident previously evaluated. These proposed changes will make additions to various tables in Sections 15.3.5 and 15.4.1, and reformat Table 15.4.1-1. These proposed changas will y add additional requirements to the Technical Specifications, making the document more restrictive. The majority of the items being added to the Technical Specifications ars 4 already being performed at Point Beach. They are being made Technical Specification requirements because of their safety significance. Therefore, the addition of these items will not change the way Point Beach is operated. It will, how-ever, place more stringent controls in place to ensure that the associated requirements are' mot.

This change also proposes to add requirements to the Tech-nical Specifications that are not currently being performed at Point Beach. Their addition can only improve the over-

.all operation of Point Beach. The surveillances and sur-veillance frequencies for these items are consistent with requirements existing in the Standard Technical Specifica-tions or existing Point Beach Technical Specifications.

This change also proposes removing the operating conditions for the motor-driven AFW pump actuation following the trip of both-main feed pumps from Table 15.3.5-3. This AMSAC (ATHS Mitigating System Actuation Circuitry) feature is being removed because it does not provide a' safety-related ,

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function. This system is required.by 10 CFR 50.62.

Administrative controls.will-be. put in place prior to the implementation of_this change to ensure that the require-ments for AMSAC are maintained. This change also proposes to reduce the surveillance frequency for verification t" AFW flowrate. The current surveillance requires AFW flowra to be verified during each unit startup and shutdown.. The proposed requirement would require this flowrate verifi-cation to be performed prior to entering power operation, following each cold shutdown of greater than 30 days.

Although this proposed surveillance interval is less frequent than what is currertly required,_ it is consistent with Standard Technical Specification requirements.

The remaining changes are till administrative in nature.

The only significant administrative item is the reformatting of Table 15.4.1-1, " Minimum Frequencies for Checks, Cali-brations, and Tests of Inntrument Channels." The refor-matting of Table 15.4.1-1 will not modify any existing surveillance requirements. It will revise the table to make it more user friendly and easier to' read. This will enhance the use-ability of this table and ensure that all of the required surveillances are clearly identified.

There is no physical change to the facility, its systems, or its operation. Thus, a new or different kind of accident-cannot occur.

3. Operation of this facility under the proposed Technical Specification changes will not create a significant reduction : a margin of safety. These proposed changes will make additions to various tables in Sections 15.3.5 and 15.4.1, and refornat Tablo 15.4.1-1. These proposed-changes will add additional requirements to the Technical specifications, making the document more restrictive. The majority of the iteme being added to the Technical Speci-ficatione are already being performed at Point Beach. They are being made Technical Specification requirements because of their safety significance. Therefore, the addition of~

these items will not change the overall operation of Point Beach. It will, however, place more stringent controls in place to ensure that the associated requirements are met. ,

This change also proposes to add requirements to the Tech-nical Specifications that are not currently being performed at Point Beach. Their addition can only improva the overall operation of Point Beach.- The surveillances and surveil-lance frequencies for these items are consistent with requirements existing in the Standard Technical Specifica-tions or_ existing Point Beach Technical Specification requirements.

3 s

This change also proposes removing the operating conditions for the motor-driven AFW pump actuation following the trip-of both main feed pumps from_ Table 15.3.5-3. This AMSAC (ATWS Mitigating System Actuetion Circuitry) feats e is being r: moved because it does not provide a safety-related function. -This system is required by 10 CFR 50.62.

Administrative controls will be pu' in place prior to the implementation of this change to ensure that the require-ments for AMSAC are maintained. This change also proposes to reduce the surveillance fre.pumcy for verification of AFW flowrate. The current surveiLlur i requires AFW flowrate to-be verified during each unit " .stup and shutdown. The pro-posed requit -+ would require this flowrate verification to be performed prior to entering power operation, fo] lowing each cold shutdown of greater than 30 days. Although this proposed surveillance interval is less frequent than what is currently required, it is consistert with Standard Technical Specification requirements.

The remaining changes are all administrative in nature. The only significant. administrative item is the reformatting of Table 15.4.1-1, " Minimum Frequencies for Checks, Calibra-tions, and Tests of Instrument Channels." The reformatting

, of Table 15.4.1-1 will not modify any existing surveillance l requirenients . It will revise the table to make it more user friendly and easier to read. This will enhance the useabil-ity of this table and eneure that all of the required sur-veillances are clearly identified.

There is no physical change to the facility,'its systems, or its operation. Thus, a significant reduction in a margin of safety cannot occur. In fact, the proposed additions may result in an increased margin of safety.

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