2CAN099703, Application for Amend to License NPF-6,modifying RPS & ESFAS Trip Setpoint & Allowable Values for SG Low Pressure

Application for Amend to License NPF-6,modifying RPS & ESFAS Trip Setpoint & Allowable Values for SG Low Pressure

ML20211C624
Person / Time
Site:	Arkansas Nuclear
Issue date:	09/23/1997
From:	Hutchinson C ENTERGY OPERATIONS, INC.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20211C628	List: 2CAN099703, Application for Amend to License NPF-6,modifying RPS & ESFAS Trip Setpoint & Allowable Values for SG Low Pressure ML20211C644
References
2CAN099703, 2CAN99703, GL-93-08, GL-93-8, NUDOCS 9709260258
Download: ML20211C624 (9)

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Ta 501-858 4888 C. Randy Hutchinson we het September 23,1997 *==

2CAN099703 U. S. Nuclear Regulatory Commission Document Control Desk Mail Station OPI-17 Washington, DC 20555

Subject:

Arkansas Nuclear One - Unit 2 Docket No. 50-368 License No. NPF-6 Technical Specification Change Request Concerning Main Steam Isolation Signal Setpoint Reduction and Relocation of Response Time Information

. Gentlemen:

Attached for your review and approval is a proposed amendment to Arkansas Nuclear One

. (ANO) Unit 2 Technical Specifications. The proposed amendment modifies the Reactor Protective System (RPS) and Engineered Safety Features Actuation System (ESFAS) trip setpoint and allowable valaes for steam generator low pressure. This amendment request is necessary due to the secondary system pressure reduction that has occurred due to the reduction in the Reactor Coolant System hot leg temperature and the removal of steam generator tubes from service due to service induced degradation.

, This amendment request also contains the relocation of the RPS and ESFAS response time

. tables from the Technical Specifications to the Safety Analysis Report. This proposed change adopts the Technical Specification "line-item improvement" as recommended in NRC Generic Letter 93-08, " Relocation of Technical Specification Tables of Instrument Response Time Limits," dated December 29,1993.

As required by Generic Letter 93-08, ANO will confirm that the plant procedures for response time testing includes acceptance criteria that reflects the RPS and ESFAS response time limits from the information that is being relocated. This confirmation will be complete prior to the relocation of the associated response time information. The associated RPS and ESFAS response time limits will be in the next Safety Analysis Report update that is currently scheduled for December 1997.

The proposed change has been evaluated in accordance with 10 CFR 50.91(a)(1) using criteria in 10 CFR 50.92(c) and it has been determined that this change involves no significant hazards considerations. The bases for these determinations are included in the attached submittal. .-  !

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U. S. NRC September 23,1996 2CAN099703 Page 2 Entergy Operations requests that the effective date for this change be within 30 days of issuance and prior to the completion of the planned mid-cycle outage currently scheduled for February 1998.

Very tmly yo rs,

.p w CRH/rde Attachments To the best of my knowledge and belief, the statements contained in this submittal are true.

SUBSCRIBED AND SWORN TO before me a Notary ublic in ud for [#tWL, County and the State of Arkansas, this #Nday of M !d b ,1997. /

r 1 d // AS$Ni Notary PGblic (7 My Commission Empires //7A/ //,1000 y

cc: Mr. Ellis W. Merschoff Regional Administrator U. S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-8064 ,

NRC Senior Resident Inspector Arkansas Nuclear One P.O. Box 310 London, AR 72847 Mr. George Kalman NRR Project Manager Region IV/ANO-1 & 2 U. S. Nuclear Regulatory Commission NRR Mail Stop 13-H-3 One White Flint North 11555 Rockville Pike Rockville, MD 20852 Mr. David D. Snellings Director, Division of Radiation Control and Emergency Management Arkansas Department ofHealth 4815 West Markham Street Little Rock, AR 72205 i

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ATTACHMENT 2CAN099703 PROPOSED TECHNICAL SPECIFICATION RESPECTIVE SAFETY ANALYSES IN TTE MATTER OF AMENDING LICENSE NO. NPF-6 ENTERGY OPERATIONS. INC.

ARKANSAS NUCLEAR ONE. UNIT TWO DOCKET NO. 50-368 4

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I Attachment to 2CAN099703 Page1of6  !

DESCRIPTION OF PROPOSED CHANGES e Tecanical Specification (TS) Table 2.2-1 has been modified by the reduction of the trip setpoint and allowable values for Steam Generator Pressure - Low from 2 751 to 2 712 psia and from 2 729.613 to 2 699.6 psia respectively, e TS 3/4.3.1 was modified consistent with Generic Letter 93-08, " Relocation of Technical Specification Tables ofInstrument Response Time Limits." These modif: cations include the removal of the reference in TS 3.3.1.1 to the response times on Table 3.3-2. A reference in TS 4.3.1.1.3 was added to indicate that the neutron detectors are exempt from response time testing.

• Table 3.3-2, " Reactor Protective Instrumentation Response Times" will be relocated from the TS to the Safety Analysis Report (SAR).

• Figure 3.3-1, "CPC Penalty Vs Effective RTD Time Constant" which is referred to by Table 3.3-2 will be relocated to the SAR.

• TS 3.3.2.1 was modified by the removal of the reference to the response times on Table 3.3 5.

• Table 3.3' 4, Functional Units 4.b and 8 e were modified by the reduction of the trip setpoint and allowable values for Steam Generator Pressure - Low from 2 751 to 2 712 psia and from 2 729.613 to 2 699.6 psie. respectively.

• Table 3.3-5, " Engineered Safety Features Response Times," will be relocated from the TS to the SAR.

• TS Figure 3.6-1, " Containment Internal Pressure VS. Containment Average Air Temperature," has been modified to reflect the new initial conditions for the containment design base accidents (DBAs),

e The last paragraph discussing the RTD response time constants on bases page B 3/4 3 ivill be relocated to the SAR along with the associated Figure 3.3-1. A sentence was also added to the same page indicating the location of the response time tables.

BACKGROUND FOR THE LOW SG PRESSURE SETPOINT REDUCTION In an attempt to inhihit the steam generator (SG) secondary side stress corrosion cracking, one of the measures tak:n was a reduction in the reactor coolant system (RCS) hot leg temperature.

This reduction la the RCS hot leg temperature resulted in a reduced steam generator secondary side steam pressure.

ANO-2 has an active damage mechanism affecting the steam generator tubing which requires the repair or the removal of tubes from service when they meet the repair criteria. A reduction in the heat transfer surface area occurs each time a SG tube is removed from service. The reduced heat

Attachment to 2CAN099703 i Page 2 of 6 l transfer surface area requires an increased differential temperature across the remaining SG tubes.

This increased-differential temperature results in a reduced SG secondary side temperature

j. because the primary side temperature is essentially constant at full power. Since the secondary

side of the SG operates in a saturated condition, a red'iction in the secondary side temperature l- produces a reduction in the SG steam pressure. The low. 3 pressure has reduced the operating margin E ween the full power SG pressure and the Pla.. Protection System (PPS) setpoints based on low SG pressure.

As the operating margin was reduced, an effort was under taken to evaluate the SG Pressure -

Low setpoints to see if they could be reduced. Based on this review, the secondary system pipe

, break safety analyses were reanalynd. This effort included the removal of unnecessary analysis conservatisms resulting in a significant reduction in the associated setpoints. In order to increase g the operating margin, this amendment request is being submitted to reduce the SG Pressure - Low l setpoints in the TS. The secondary system pipe break safety analyses have been reanalyzed and is discussed in detail in Attachment 1 of this amendment request.

f The calculations associated with these proposed setpoints are done in accordance with the j guidelines set forth in the Instrument Loop Error and Setpoint Methodology Manual, Design l Guide IDG-001-0. This methodology is the same as that used in past calculations reviewed by the j staff for previous- ANO-2 TS amendments that have been approved. These include the j replacement of the narrow range containment piessure transmitters issued as amendment 137 and

the pressurizer pressure setpoint reduction effort issued as amendment 138 to the ANO-2 TS.

j Although ANO is not committed to strict compliance with the Instrument Society of America

- Standard, ISA-S67.04-1982, "Setpoints for Nuclear Safety Related Instrumentation Used in

. Nuclear Power Plants," the requirements of S67.04 were used as a guide for IDG-001-0 and is an i integral part of the ANO-2 setpoint program.

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The allowable containment atmospheric conditions are defined by the following considerations;

1) inadvertent containment spray actuation, 2) Emergency Core Cooling System (ECCS) analysis, j 3) peak contamment pressure analysis for Main Steam Line Break (MSLB) and Large Break Loss l of Coolant Accident (LBLOCA). The inadvertent spray actuation defines the lower right portion l of the pressure and temperature limits for the containment operating envelope on TS Figure 3.6-1.

l These lower pressure and temperature limits prevent the containment building from exceeding the i design negative pressure differential of 5.0 psig on an inadvertent containment spray system l actuation. The lower left portion of the limits are defined by the ECCS analysis in accordance

with 10 CFR 50.46. These lower limits are not modified by this amendment request which

eliminated the need to reanalyze the inadvertent spray actuation.

i-l The LBLOCA and MSLB evenu defme the upper pressure and temperature limits for the i containment operating envelope en TS Figure 3.6-1. These limits ensure the containment design pressure of 54 psig will not be exceeded during design basis conditions. The LBLOCA was not i reanalyzed as a result of this effort. However, the MSLB accident was reanalyzed as part of tds effort. The MSLB reanalysis is discussed in detail in Attachment 1 of this amendment request.

l Attachment 1 also discusses the initial atmospheric conditions for the containment. The required l' modi 6 cations to the containment operating envelope are a result of these assumptions and are depicted on TS Figure 3.6-1, 4

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Attachment to 2CAN099703 Page 3 of 6 The proposed value for the low SG pressure setpoints are based on 30% tube plugging and a 10%

reduction in the RCS design flow. The proposed values should be adequate to account for the projected SG tube plugging until the SGs are replaced. The ANO-2 SGs are scheduled for replacement in the fall of 2000.

BACKGROUND FOR THE RELOCATION OF THE RESPONSE TIME TABLES Generic Letter 93-08 was issued by the NRC to provide guidance for utilities requesting a license amendments to relocate the instrumentation time response tables for the Reactor Protective System (RPS) and Engineered Safety Features Actuation System (ESFAS) from the TS to the

SAR. One of the components of this TS amendment request is the relocation of the associated response time infonnation for the RPS and ESFAS consistent with the requirements of Generic

] Letter 93-08 The response time information relocated to the SAR will in turn be controlled in accordance with the requirements of Section 50.59 of Title 10 of the Code of Federal Regulations, " Changes, tests and experiments." 10 CFR 50.59 provides criteria to determine when changes planned by a licensee require pnu Commission approval in the form of a license amendment in order tc address any unreviewed safety questions. The staff concluded that 10 CFR 50.36 does not require the response time tables to be retained in TSs and has issued Generic Letter 93-08 as a

! lir4 item improvement to allow their removal. The response times are considered to be an operational detail related to the licensee's safety analyses which are adequately controlled by the requirements of 10 CFR 50.59. The staff has previously concluded the continued processing of license amendments related to revisions of the affected instrument or component response times, where the revisions to those requirements do not involve an unreviewed safety question under 10

CFR 50.59, would afford no significant benefit with regard to protecting the public health and safety.

Response time testing will still be required by the ANO-2 TS after the relocation of the associated response time information in this amendment request. Specification 4.3.1.11 will continue to require RPS response time testing and specification 4.3.2.1.3 will continue to require the ESFAS

, response time testing. Relocating the response time information for the RPS and ESFAS from the

! TS to the SAR will not alter these surveillance requirements. The SAR will now address the response time limits for the RPS and ESFAS instruments including those channels for which the response time limit is indicated as "N/A"; that is, a response time limit is not applicable. The SAR will also clarify response time limits where footnotes are included in the tables that describe how those limits are applied.

The relocation of these tables will eliminate the need for ANO and the Staff to process TS changes to revise these tables due to the present and future analysis updates.

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. Attachment 13 2CAN099703 j Page 4 cf 6 f DISCUSSION OF PROPOSED CHANGES I

l TS Table 2.21 and Table 3.3-4 (Functional Units 4.b and 8.e) were modified to reflect the l proposed trip setpoint and allowable value for Steam Generator Pressure . Low from 2 751 to i 2 712 psia and firom 2 729.613 to 2 699.6 psia respectively. hae revised setpoints are a result

! of the recently completed secondary system pipe break reanalyzes and the associated calculations

! presented in the ANO 2 PPS Trip Setpoint Calculation Upgrade Project. The secondary system l pipe break reanalyzes are discussed in detail in Attachment 1 of this amendment request.

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! The modifications to TS 3/4.3.1.1 include the removal of the reference to the response times on j Table 3.3 2. Speci6 cation 4.3.1.1.3 was modified to indicate that the neutron detectors are

! exempt from response time testing. The neutron detectors are currently exempted from response i time testing by the first note on Table 3.3 2. Both of these changes are consistent with the

! requirements in Generic Letter 93-08.

Table 3.3 2 and Figure 3.31 are proposed to be relocated from the TS to the SAR. The only TS i - reference to Figure 3.3 1 is on Table 3.3 2 and the bases for Specification 3/4.3.1 Both of these i

! references are being relocated to the SAR by this amendment request. The Generic Letter uses I the Westinghouse standard TS as its model. The Westinghouse plants do not have coie i prc' ? ion calculators that would require the use of the RTD response times on Figure 3.31.

ThL egure. was not included in the latest revision of NUREG 1432, " Standard Technical 4

Specifications for Combustion Engineering Plants" (ISTS) where it would remain if it were j required to be retained in the TS. This information is being relocated to the SAR consistent with j the requirements of Generic Letter 93-08 and the ISTS.

s i TS 3.3.2.1 was modified by the removal of the reference to the response times on Table 3.3 5.

I Specification 4.3.2.1.3 will continue to require the ESFAS response time testing at the current l frequency. . Relocating the response time information for the ESFAS from the TS to the SAR will 1

not alter these surveillance requirements. This change is consistent with the requirements of

Generic Letter 93-08.

i Table 3.3-5 is proposed to be relocated from the TS to the SAR. Specification 4.3.2.1.3 will i continue to require the ESFAS response tirae testing at the current frequency. Relocating the I response time information for the ESFAS from the TS to the SAR will not alter these surveillance requirements. This change is consistent with the requirements of Generic Letter 93-08.

The Containment Intemal Pressure VS, Containment Average Air Temperature graph (TS Figure 3.6-1) has been updated to reflect the new initial conditions for the containment DBAs.

- AtMehment I to this amendment request discusses the changes to the containment initial conditions in detailjustifying the changes to this figure, y The bases for specifications 3/4.3.1 and 3/4.3.2 was modified by the relocation of the paragraph discussing the RTD constants associated with Figure 3.3-1. This information is being relocated to the SAR consistent with the requirements of Geraic Letter 93-08 and the ISTS.

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Attachment to 2CAN099703 Page 5 cf 6 DETER 14tNATION OF NO SIGNIFICANT HAZARD 3 CONSIDERATION An evaluation of the proposed change has been performed in accordance with 10 CFR 50.91(a)(1) regarding no signl8 cant hazards considerations using the standards in 10 CFR 50.92(c). A discussion of these standards as they reiste to this amendment request follows:

Cdtedon 1 - Does Not lavolve a Sigalficant Imeresse la the Probability or Consequences of an Accident Previously Evaluated.

l The proposed changes included in this amendment request do not affect the accident initiators in

) any of the a 6 dents previously evaluated. The proposed trip setpoints and allowable values for Steam Generator Pressure Low are being reduced by this proposed amendment request. This change is necessary to increase the operating margin between the full power _ steam generator pressure and these setpoints. The change should reduce the probability of an inadvertent Main Steam Isolation Signal (MSIS) from occurring at power since it will increase the operating space between the operating pressure and the setpoints. Therefore, this amendment request will not increase the probability of any accident previously evaluated.

The secondary system pipe break safety analyses were reanalyzed for the Steam Generator Pressure - Low setpoint reduction effort. "l ais effort included the removal of unnecessary analysis conservatisms resulting in a signincant reduction in the associated setpoints. The proposed changes do not involve any change to the configuration or method of operation of any plant equipment used to mitigate the consequences of an accident. The previously evaluated accidents which were determined to be impacted by this setpoint change were evaluated with no significant increase in the consequences.

This amemiment request contains the relocation of the Reactor Protective System (RPS) and Engineered Safety Features Actuation System (ESFAS) response time information from the Technical Specifications (TS) to the Safety Analysis Repart. This proposed change adopts the TS "line-item improvement" as recommended in NRC Generic Letter 93-08, " Relocation of Technical Specification Tables ofinstrument Response Time Limits," dated December 29,1993. The NRC has concluded that 10 CFR 50.36 does not require the response time tables to be retained in TSs and has issued Generic Letter 93-08 as a line item improvement to allow their removal. Response time testing will still be required by the ANO 2 TS after the relocation of the associated response time information in this amendment request. Relocating the response time information for the RPS and ESFAS from the TS to the SAR will not alter these surveillance requirements.

Therefore, the relocated response time portion of this amendment request is considered administrative in nature and will not affect the probability or consequences of any accident previously evaluated.

Therefore, this change does Del involve a significant increase in the probability or consequences of any accident previously evaluated.

Attachtnent to 2CAN099703 Page 6 of 6 Criterion 2 Does Not Create the Possibility of a New or Different Kind of Accident from any Previously Evaluated.

The proposed changes do not involve any physical modifications (i.e. new systems, n(w components, etc.) to the plant. The proposed changes do not involve any change to the configuration or method of operation of any plant equipment used to mitigste the consequences of an accident. The results of the accident reanalyzes suggest no different phenomena or plant behavior than previously considered. The Steam Generator Pressure Low setpnint change does not create any new or different system actuations or interactions than evaluated previously. The relocated response time portion of this amendment request is considered administrative in nature and is not considered an accident initiator Therefore, this change does nel create the possibility of a new or different kind of accident from any previously evaluated.

l Criterion 3 - Does Not involve a Significant Reduction in the Margin of Safety.

The accidents which were determined to be impacted by the Steam Generator Pressure Low setpoint change were evaluated to ensure acceptable results are maintained. The instrument error calculations supporting the lower Steam Generator Pressure Low setpoint and allowable values will ensure the present accident analysis assumptions are still maintained. The methodology used to determine the instrument loop errors and uncertainties is the same as that used in previous amendment requests that have been reviewed and approved by the NRC, Based on these evaluations, the proposed changes do agl involve a significant reduction in a margin of safety.

Therefore, based upon the reasoning presented sove and the previous discussion of the amendment request, Entergy Operations has determined that the requested change does act involve a significant hazards consideration.

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