Text

Proposed Tech Specs Re MSSVs & Pressurizer Safety Valves
	ML20236G491
Person / Time
Site:	Arkansas Nuclear
Issue date:	06/29/1998
From:	; ENTERGY OPERATIONS, INC.
To:	;
Shared Package
ML20236G486	List: 2CAN069804, Application for Amend to License NPF-6,increasing Allowable as-found Lift Setting Tolerances on MSSVs & Pressurizer Safety Valves; ML20236G491;
References
	NUDOCS 9807060270
	Download: ML20236G491 (14)
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PROPOSED TECHNICAL SPECIFICATION CHANGES l

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l 9907060270 990629' PDR ADOCK 05000368 P

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l REACTOR COOLANT SYSTEM l

SAFETY VALVES - SHUTDOWN LIMITING CONDITION FOR OPERATION l

3.4.2 A minimum of one pressurizer code safety valve shall be OPERABLE l

with a lift setting of 2500 PSIA i 3%*.

l l

APPLICABILITY:

MODES 4 and 5.

ACTION:

With no pressurizer code safety valve OPERABLE, immediately suspend all operations involving positive reactivity changes and place an OPERABLE shutdown cooling loop into operation.

SURVEILLANCE REQUIREMENTS 4.4.2 No additional Surveillance Requirements other than those required by Specification 4.0.5.

The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.

If found outside of a + 1% tolerance band, the setting shall be adjusted to within + 1% of the lift setting shown.

ARKANSAS - UNIT 2 3/4 4-3 Amendment No. 449,444,

1 1

REACTOR COOLANT SYSTEM SAFETY VALVES - OPERATING

]

LIMITING CO'NDITION FOR OPERATION 3.4.3 All pressurizer code safety valves shall be OPERABLE with a lift setting 2500 psia i 3%*.

l APPLICABILITY: MODES 1, 2 and 3.

ACTION:

With one pressurizer code safety valve inoperable, either restore a.

the inoperable valve to OPERABLE status within 15 minutes or be in HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b.

The provisions of specification 3.0.4 may be suspended for one valve at a time for up to 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months for entry into and during i

operation in MODE 3 for the purpose of setting the pressurizer code safety valves under ambient (hot) conditions provided a preliminary cold setting was made prior to hestup.

SURVEILLANCE REQUIREMENTS 4.4.3 No additional Surveillance Requirements other than those required by Specification 4.0.5.

l

The lift setting pressure shall correspond to ambient conditions of the j

valve at nominal operating temperature and pressure.

If found outside of a i 1% tolerance band, the setting shall be adjusted to within i 1% of the lift setting shown.

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ARKANSAS - UNIT 2 3/4 4-4 Amendment No. 46,-1-14,

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3/4.7 PLANT SYSTEMS BASES 3/4.7.1 TURBINE CYCLE 3/4.7.1.1 SAFETY CYCLES The OPERABILITY of the main steam line code safety valves ensures that the secondary system pressure will be limited to within 110% of its design pressure of 1100 psia during the most severe anticipated system operational transient.

The maximum relieving capacity is associated with a turbine trip from 100% RATED THERMAL POWER coincident with an assumed loss of condenser heat sink (i.e., no steam bypass to the condenser).

The specified valve lift settings and relieving capacities are in accordance with the requirements of Section III of the ASME Boiler and Pressure Code, 1971 Edition. The "as-found" requirements are consistent with Section XI of the ASME Boiler and Pressure Vessel Code, 1986 Edition, and Addenda through 1987. The MSSV rated capacity passes the full steam flow at 102% RATED THERMAL POWER (100%

+ 2% for instrument error) with the valves fully open. A minimum of 2 OPERABLE safety valves per steam generator ensures that sufficient relieving capacity is available for removing decay heat.

STARTUP and/or POWER OPERATION is allowable with safety valves inoperable within the limitations of the ACTION requirements on the basis of the reduction in secondary system steam flow and THERMAL POWER required by the reduced reactor trip settings of the Power Level-High channels.

The reactor trip setpoints are determined from the most conservative value obtained from two methods. The first method calculates setpoints based on the reduction in relieving capacity with the number of MSSVs inoperable and are derived as follows:

SP = (X) - (Y) x ( Z% )

l X

where:

1 Reduced reactor trip setpoint in percent of RATED SP

=

THERMAL POWER Total naximum safety valve relieving capacity of 102% power l

4 Z%

=

steam flow Total relieving capacity of all safety valves per steam X

=

line in lbs/ hour Y

Total maximum relieving capacity of the inoperable safety

=

valve (s) in lbs/ hour.

In each case, the valves with the greatest relieving capacity were assumed inoperable.

In the second method, the setpoint is determined from the Loss of Condenser vacuum (LOCV) event, which is the most restricting of all design basis events that affect peak RCS and peak secondary pressures. With + 3% MSSV tolerance, the LOCV event is analyzed with 1, 2, or 3 inoperable MSSVs on any steam generator, to determine the allowable initial power levels (and the trip setpoints) that yield acceptable results.

For each of the inoperable MSSV conditions (1, 2, or 3 inoperable) in Table 3.7-1, the more conservative trip setpoint from the two analytical methods is selected.

l ARFANSAS - UNIT 2 B 3/4 7-1 Amendment No. MG,MG,MB, L__________-____

e ADMINISTRATIVE CONTROL CORE OPERATING LIMITS REPORT

10) " Calculative Methods for the CE Small Break LOCA Evaluation Model,"

CENPD-137, Supplement 2-P-A, dated April, 1998 (Methodology for Specification 3.1.1.4 for MTC, 3.2.1 for Linear Heat Rate, 3.2.3 for Azimuthal Power Tilt, and 3.2.7 for ASI).

11) "CESEC-Digital Simulation of a Combustion Engineering Nuclear Steam l

Supply System," December 1981 (Methodology for Specifications 3.1.1.1 and 3.1.1.2 for Shutdown Margin, 3.1.1.4 for MTC, 3.1.3.1 for CEA Position, 3.1.3.6 for Regulating CEA and Group P Insertion Limits, and 3.2.4.b for DNBR Margin).

12) " Technical Manual for the CENTS Code," CENPD 282-P-A, February 1991 l

(Methodology for Specifications 3.1.1.1 and 3.1.1.2 for Shutdown Margin, 3.1.1.4 for MTC, 3.1.3.1 for CEA Position, 3.1.3.6 for Regulating and Group P Insertion Limits, and 3.2.4.b for DNBR Margin.

13) Letter:

0.D. Parr (NRC) to F.M. Stern (CE), dated June 13, 1975 l

(NRC Staff Review of the Combustion Engineering ECCS Evaluation Model). NRC approval for 6.9.5.1.4, 6.9.5.1.5, and 6.9.5.1.8 methodologies.

14) Letter:

0.D. Parr (NRC) to A.E. Scherer (CE), dated December 9, 1975 l

(NRC Staff Review of the Proposed Combustion Engineering ECCS Evaluation Model changes). NRC approval for 6.9.5.1.6 methodology.

15) Letter:

K. Kniel (NRC) to A.E. Scherer (CE), dated September 27, 1977 l (Evaluation of Topical Reports CENPD-133, Supplement 3-P and CENPD-137, Supplement 1-P).

NRC approval for 6.9.5.1.9 methodology.

16) Letter:

2CNA038403, dated March 20, 1984, J.R. Miller (NRC) to l

J.M. Griffin (AP&L), "CESEC code Verification." NRC approval for 6.9.5.1.11 methodology.

l 6.9.5.2 The core operating limits shall be determined so that all applicable limits (e.g. fuel thermal-mechanical limits, core thermal-hydraulic lindts, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.

l 6.9.5.3 The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

I l

I ARKANSAS - UNIT 2 6-21a Amendment No. M7, M4, M9, 7-9, Wa, N- __ _ _ ____ _ _.

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MARKUP OF CURRENT ANO-2 TECHNICAL SPECIFICATIONS l

(FORINFO ONLY) i l

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_______.___'.U.__

REACTOR COOLANT SYSTEM j

SAFETY VALVES - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.2 A minimum of one pressurizer code safety valve shall be OPERABLE i

with a lift setting of 2500 PSIA U +b -3%*.

ll APPLICABILITY:

MODES 4 and 5.

ACTION:

With no pressurizer code safety valve OPERABLE, immediately suspend all operations involving positive reactivity changes and place an OPERABLE shutdown cooling loop into operation.

SURVEILLANCE REQUIREMENTS 1

4.4.2 No additional Surveillance Requirements other than those required i

by Specification 4.0.5.

l l

The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.

If found outside of

{

a i 1% tolerance band, the setting shall be adjusted to within i 1% of the j

lift setting shown.

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ARKANSAS - UNIT 2 3/4 4-3 Amendment No. 449,444, I

l REACTOR COOLANT SYSTEM l

SAFETY VALVES - OPERATING i

LIMITING CONDITION FOR OPERATION

\\

3.4.3 All pressurizer code safety valves shall be OPERABLE with a lift setting 2500 psia W M %#.

l APPLICABILITY: MODES 1, 2 and 3.

ACTION:

i a.

With one pressurizer code safety valve inoperable, either restore the inoperable valve to OPERABLE status within 15 minutes or be in HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b.

The provisions of specification 3.0.4 may be suspended for one valve at a time for up to 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months for entry into and during operation in MODE 3 for the purpose of setting the pressurizer code safety valves under ambient (hot) conditions provided a preliminary cold setting was made prior to heatup.

SURVEILLANCE REQUIREMENTS 4.4.3 No additional Surveillance Requirements other than those required by l

Specification 4.0.5.

i

The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.

If found outside of a 1 1% tolerance band, the setting shall be adjusted to within i 1% of the lift setting shown.

I i

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I ARKANSAS - UNIT 2 3/4 4-4 Amendment No. M,4%,

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3/4.7 PLANT SYSTEMS BASES 3/4.7.1 TURBINE CYCLE 3/4.7.1.1 SAFETY CYCLES The OPERABILITY of the main steam line code safety valves ensures that the secondary system pressure will be limited to within 110% of its design pressure of 1100 psia during the most severe anticipated system operational transient.

The maximum relieving capacity is associated with a turbine trip from 100% RATED THERMAL POWER coincident with an assumed loss of condenser heat sink (i.e.,

no steam bypass to the condenser).

The specified valve lift settings and relieving capacities are in accordance with the requirements of Section III of the ASME Boiler and Pressure Code, 1971 Edition. The "as-found" requirements are consistence:n ::v tive with :::pect to l

Section XI of the ASME Boiler and Pressure Vessel Code, 1986 Edition, and Addenda through 1987. Tne MSSV rated caoacity casses the full steam flow at 102% RATED Thermal Power (100% + 2% for instrument error) with the valves fully open.The t;t:1 ::li;ving ;;;;;ity f:: 11 MSSV; i: 11,000,000 lb/hr.

Th;.'J:0 2 full p;x::

(100S: :t::= fi:= plu: 2S f:: ::n::: :ti;= i: 12,050,000 15/hr. Th:::f:::, the t:t:1 ::li:ving ::p::ity f th: MSSV: 1: 110.22S ;f the 102S p:x:: :te:: flex th:: ugh ::ch : in-;t::= h::d::. A minimum of 2 OPERABLE safety valves per steam generator ensures that sufficient relieving capacity is available for removing decay heat.

STARTUP and/or POWER OPERATION is allowable with safety valves inoperable within the limitations of the ACTION requirements on the basis of the reduction in secondary system steam flow and THERMAL POWER required by the reduced reactor trip settings of the Power Level-High channels. The reactor trio setooints are determined from the most conservative value obtained from two methods.

The first method calculates setooints based on the reduction in relievina capacity with the number of MSSVs inocerable and are derived as follows:Th ::::::: trip ;;tp; int

d::ti:== ::: d:: ires := th fell;uing b:::::

SP = (X) - (Y) x (,g11!.22%,)

l X

where SP Reduced reactor trip setpoint in percent of RATED

=

THERMAL POWER Total maximum safety valve relieving capacity of 102% power j

1111.22%

=

steam flow X

.=

Total relieving capacity of all safety valves per steam line in lbs/ hour Y

Total maxiuma relieving capacity of the inoperable safety

=

valve (s) in lbs/ hour.

In each case, the valves with the greatest relieving capacity were assumed inoperable.

In the second method, the setooint is determined from the Loss of Condenser Vacuum (LOCV) event, which is the most restricting of all desian basis events that affect ceak RCS and Deak secondarv Dressures. With + 3% MSSV tolerance, the LOCV event,Ja_An.alyzed with 1, 2 or 3 inocerable MSSVs on any steam cenerator, to determine the allowable initial power levels (and the trio setooints) that vield acceptable results.

For each of the inoperable MSSV conditions (1, 2,

or 3 inoperable) in Table 3.7-1, the more conservative trio setooint from the two analvtical methods is selected.

ARKANSAS - UNIT 2 B 3/4 7-1 Amendment No. 444,-160,444, i

A

l l

l

~

ADMINISTRATIVE CONTROL l

CORE OPERATING LIMITS REPORT

10) " Calculative Methods for the CE Small Break LOCA Evaluation Model,"

CENPD-137, Supplement 2-P-A, dated April, 1998. (Methodoloav for Joecification 3.1.1.4 for MTC, 3.2.1 for Linear Heat Rate, 3.2.3 for Azimuthal Power Tilt, and 3.2.7 for ASI).

110) "CESEC-Digital Simulation of a Combustion Engineering Nuclear Steam l

Supply System," December 1981 (Methodology for Specifications 3.1.1.1 and 3.1.1.2 for Shutdown Margin, 3.1.1.4 for MTC, 3.1.3.1 for CEA Position, 3.1.3.6 for Regulating CEA and Group P Insertion Limits, and 3.2.4.b for DNBR Margin).

121) " Technical Manual for the CENTS Code," CENPD 282-P-A, February 1991 l

(Methodology for Specifications 3.1.1.1 and 3.1.1.2 for Shutdown Margin, 3.1.1.4 for MTC, 3.1.3.1 for CEA Position, 3.1.3.6 for Regulating and Group P Insertion Limits, and 3.2.4.b for DNBR Margin.

1 133) Letter:

0.D.

Parr (NRC) to F.M. Stern (CE), dated June 13, 1975 l

l (NRC Staff Review of the Combustion Engineering ECCS Evaluation Model). NRC approval for 6.9.5.1.4, 6.9.5.1.5, and 6.9.5.1.8 i

methodologies.

l 114) Letter:

0.D.

Parr (NRC) to A.E. Scherer (CE), dated December 9, 1975 l

(NRC Staff Review of the Proposed Combustion Engineering ECCS Evaluation Model changes). NRC approval for 6.9.5.1.6 methodology.

I 154) Letter:

K.

Kniel (NRC) to A.E. Scherer (CE), dated September 27, 1977 l

(Evaluation of Topical Reports CENPD-133, Supplement 3-P and CENPD-137, Supplement 1-P).

NRC approval for 6.9.5.1.9 methodology.

If6) Letter:

2CNA038403, dated March 20, 1984, J.R. Miller (NRC) to l

l J.M. Griffin (AP&L), "CESEC Code Verification." NRC approval i

for 6.9.5.1.110 methodology.

l 6.9.5.2 The core operating limits shall be determined so that all applicable limits (e.g. fuel thermal-mechanical lindts, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.

6.9.5.3 The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance to the NRC Document control Desk with copies to the Regional Administrator and Resident Inspector.

ARKANSAS - UNIT 2 6-21a Amendment No. -1M,M4,M9, M9, M3, E

__

ML20236G491

Text

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