2CAN069804, Application for Amend to License NPF-6,increasing Allowable as-found Lift Setting Tolerances on MSSVs & Pressurizer Safety Valves

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Application for Amend to License NPF-6,increasing Allowable as-found Lift Setting Tolerances on MSSVs & Pressurizer Safety Valves
ML20236G483
Person / Time
Site: Arkansas Nuclear Entergy icon.png
Issue date: 06/29/1998
From: Hutchinson C
ENTERGY OPERATIONS, INC.
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20236G486 List:
References
2CAN069804, 2CAN69804, NUDOCS 9807060265
Download: ML20236G483 (13)
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Text

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M Entargy Operations, Inc.

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[Of ib 72801 Tel 501 W,81388 C. Randy Hutchinson WT PrM&fil (WaW6 A*O June 29,1998 2CAN069804 U. S. Nuclear Regulatory Commission Document Control Desk Mail Station OPI-17 Washington, DC 20555

Subject:

Arkansas Nuclear One - Unit 2 Docket Nos. 50-368 License Nos. NPF-6 Technical Specification Change Request Concerning the Main Steam Safety Valves and Pressurizer Safety Valves Gentlemen:

Attached for your review and approval is a proposed Arkansas Nuclear One-Unit 2 (ANO-2)

Technical Specification (TS) amendment. The proposed amendment increases the allowable as-found lift setting tolerances on the main steam safety valves (MSSVs) and pressurizer safety valves. The increase in the as-found tolerance requires a reduction in the linear power level-high trip setpoint associated with three inoperable MSSVs. In addition, an NRC-approved small break loss of coolant accident analytical methodology will be utilized to develop the core operating limits report listed in TS 6.9.5. Similar TS changes have been approved for use at Waterford 3 and other facilities.

The proposed change has been evaluated in accordance with 10CFR50.91(a)(1) using criteria in 10CFR50.92(c) and it has been determined that this change involves no significant hazards

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considerations. The bases for these determinations are included in the attached submittal. /

This change will reduce plant personnel exposure to radiological hazards associated with valve testing and adjustment.

Although the circumstances of this amendment request are neither exigent or emerency in nature, Entergy Operations requests the effective date for this change to be 30 days before tnc

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next refueling outage, currently scheduled to begin on January 8,1999, so that these changes can be implemented for use during valve testing to be performed during this outage.

9807060265 990629 PDR ADOCK 05000360 P PDR

U. S. NRC June 29,1998 2CAN069804 Page 2 l Very truly yours, .

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

SUBSCRIBED AND SWORN TO before me a Notary Public in and for #1dits County and the State of Arkansas, this /'l>#Nday of doo ,1998. '

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ash M9/4dkff7v Notary Public /

My Commission Empires 7/8ur//####

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U. S. NRC June 29,1998 ,

2CAN069804 Page 3 i

f 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. William D. Reckley NRR Project Manager Region IV/ANO-1 & 2 l l

U. S. Nuclear Regulatory Commission i NRR Mail Stop 13-H-3 j One White Flint North i 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

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ATTACHMENT T_OQ 2CAN069804 PROPOSED TECHNICAL SPECIFICATION I

AND RESPECTIVE SAFETY ANALYSES IN THE MATTER OF AMENDING LICENSE NO. NPF-6 ENTERGY OPERATIONS. INC.

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

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, Attachment to 2CANM9804 Page1of9 DESCRIPTION OF PROPOSED CIIANGES The as-found lia setting tolerance for the ANO-2 main steam safety valves (MSSVs) and pressurizer safety valves (PSVs) will be increased The proposed increase in the lift setting l tolerance is contingent upon a reduction in a linear power level-high supoint and use of the latest small break loss of coolant accident (SBLOCA) methodology for development of the Core Operating Limits Report (COLR). The specific changes to the Technical Specihcations (tsp) are as follows:

. TS Table 3.7-5, " Steam Line Safety Valves", has been revised to increase the as-found lift setting tolerance from +1% / -3% to

  • 3%. In addition, the " Orifice Size" column, which specifies design requirements, is deleted from the table in accordance with NUREG-1432, Standard Technical Specifications Combustion Engineering Plants, Revision 1, dated April,1995.

. TS 3.4.3, " Reactor Coolant System Safety Valves - Operating" has been revised to increase the as-found lift setting tolerance from +1, -3% to 3%.

. TS Table 3.7-1, " Maximum Allowable Linear Power Level-High Trip Setpoint With Inoperable Steam Line Safety Valves During Operation With Both Steam Generators" has been revised. The maximum allowable linear power level-high trip setpoint with three inoperable safety valves per steam generator is reduced from 45 per cent to 36 per cent of rated thermal power.

  • The bases for TS 3.7-1 have been revised to reflect the changes to TS Table 3.7.5 and Table 3.7-1, and to incorporate other improvements in the information presented. The bases now reflect that the linear power level-high trip setpoint values linked to the number ofinoperable MSSVs are selected based on the most conservative of two methods for determination of the trip setpoints. In addition, discussion of the MSSV rated capacity is revised.
  • TS 6.9.5.1, which lists the analytical methods used to determine the core operating limits addressed by the individual TS, has been revised to include topical report CENPD-137, Supplement 2-P-A, " Calculative Methods For The C-E Small Break LOCA Evaluation Model". The current references in TS 6.9.5.1 have been renumbered, as necessary.

Minor editorial changes to the affected TS pages have also been incorporated to assure consistency in the presentation of TS requirements. Other than tiie change to the bases for TS 3.7.1, the bases for other TS changes are unaffected.

. Attachment to 2CAN069804 Page 2 of 9 BACKGROUND Code of Federal Regulations,10CFR50.55a, Codes and standards, invokes the American Society of Mechanical Engineers- (ASME),Section XI pressure relief valve testing -

requirements for MSSVs and PSVs at ANO-2. These valves provide assurance that pressure transients resulting' from operation of the secondary steam system and the reactor coolant system will not exceed 110% of design pressure. To assure operational readiness of these valves,' ASME code criteria requires periodic surveillance testing of certain valves each outage.

ASME recognizes the potential for setpoint (referred to as " lift setting") drift in pressure relief valves; i.e., ASME/ ANSI OM Part 1 (1987) requires that a valve be repaired or replaced and the cause of failure be determined and corrected only if the valve varies from its set pressure by 3% or greater. Current ANO-2 TS as-found lift setting tolerances for these valves of +1%,

-3% are more stringent than the ASME allowed tolerance. The current as-found tolerance for lift settings is often exceeded during surveillance testing, usually in the plus (+) direction.

Additional testing of other valves, beyond that normally required by the ASME Section XI surveillance testing program must then be performed, impacting plant outage schedules and resulting in additional radiation exposure to plant personnel. The proposed TS changes will also reduce the number of Licensee Event Reports (LERs) for "as-found" lift pressure settings being out of tolerance.

The changes to the linear power level-high trip setpoint and incorporation of the most current small break loss of coolant accident methodology are necessary due to the revised as-found lift setting tolerances. The increased tolerance requires a revision of t.nalytical assumptions

.and operational parameters regarding maximum design pressure. Tne proposed changes, which are consistent with ASME c. ode criteria, will not change the as-left lift setting tolerance.

- During testing, any valve found to be outside of a

  • 1 % tolerance band will be reset to within
  • 1% of the lift setting specified by the TS. The safety analyses supporting these changes

- have been performed with a 3% tolerance added to the TS lift setting pressures. This approach significantly bounds any instrument uncertainty linked to the test equipment and the

  • 1% as-left lift setting.

DISCUSSION OF CHANGE Anticipated operational occurrences (AOOs), accident events, and containment events documented in the ANO-2 Safety Analysis Report (SAR) were reviewed to determine whether they are affected by an increase in the MSSV and PSV lifi e.iting tolerances. This l analysis of operational transients and postulated accidents assuming a

and PSV lift setting tolerance resulted in the conclusion that turbine cycle and RCS pressures are within acceptable limits (i.e., <110% of the steam generator and reactor coolant system desiga pressure). However, the results of the analysis reveal that peak pressures are slightly

> higher than - the analysis of record currently documented in the SAR. The 'RCS

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. Attachment to

. 2CAN069804 Page 3 of 9 overpressurization analysis and steam generator overpressurization analysis results are

' impacted. This proposed amendment impacts five TS requirements and one TS bases.

Imoscts Due to Revision of MSSV Lin Settinn Tolerance Overpressure protection for the secondary side of the steam generators is provided by spring loaded pressure relief valves. The design capacities of the MSSVs are as shown in SAR Table 10.3-1. Operational parameters for the MSSVs are given in SAR Table 5.5-11. The total capacity of the MSSVs is sufficient to pass greater than 100 percent of the steam flow generated at rated load. When the MSSVs are at full discharge flow, with valve opening characteristics that include 3% tolerance on the liA setting and the associated accumulation, the steam generator pressure must not exceed 110% of design pressure.

Accident events and AOOs pertaining to the MSSV liA setting tolerance revision were evaluated. Based on a review of the SAR, the feedwater line break is the limiting accident event for peak RCS pressure and peak steam generator pressure. This accident event which includes potential effects on RCS pressure is evaluated in the discucsion ofimpacts due to revision of PSV liA setting tolerances. Nse evaluations reveal allowable peak pressures will not be exceeded.

The limiting AOO for peak RCS and steam generator secondary pressure is the loss of condenser vacuum (LOCV) event. The results of the LOCV event analysis indicates the peak secondary pressure is 1195 psia. The peak analytical pressure determined by this analysis is within the ASME acceptance criteria of 110% of design pressure (1210 psia).

Prior to this proposed amendment, new analysis results based on increased steam generator tube plugging and reduced RCS flow were submitted for NRC review. Both the feedwater line break and LOCV analyses were approved by the Staffin Amendments 189 and 190, dated March 12,1998. These amendments approved TS changes to address a reduction in the main steam isolation signal (MSIS) setpoint and a 10% reduction in the RCS flow associated with increased steam generator tube plugging.

The Staft's evaluation of TS Amendment 189, regarding main steam isolation signal setpoint reduction and relocation of response time information, and TS Amendment 190, regarding a '

reduction in RCS flow requirements, approved the use of the topical reports referenced in those amendments. The Staff confirmed that the SER conditions associated with each of the referenced topical reports had been satisfied. Those confirmations are also applicable to this proposed ==aadmaat.

The increase in MSSV liA pressure also adversely impacts the peak clad temperature (PCT) during a SBLOCA event. The increase in MSSV liA pressure results in a higher steam generator pressure and in turn higher RCS pressure during the limiting SBLOCA event. The higher RCS pressure decreases the safety injection flow and increases break flow, resulting in a higher PCT. This event was reanalyzed using the methodology presented in CENPD-137, Supplement 2-P-A, Calculative Methodsfor the ABB CE Small Break Emluation Model l

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Attachment to 2CAN069804 Pcge 4 of 9 (S2M). This methodology was reviewed and approved by the NRC in a letter dated

' December 16,1997. The associated safety evaluation by the Staff approved the methodology without any technical constraints or changes.

Using the S2M evaluation model, a SBLOCA emergency core cooling system performance analysis was performed for a spectrum of reactor coolant pump (RCP) discharge leg breaks 2 2 ranging in size from 0.02 ft to 0.06 ft . The analysis was performed for the equivalent of up to 30% average length steam generator tubes plugged in each steam generator. The limiting break size (i.e., the break size that resulted in the highest PCT) was determined to be the 0.05 2

ft break. Conformance to Criteria 1 through 3 of 10CFR50.46, Acceptance criteriafor emergency core cooling systemsfor light water nuclearpower reactors, is as follows for the limiting break at a linear heat rate of 13.5 kW/ft:

Parameter Criterion Value Peak Clad Temperature F <2200 1798 Maximum Cladding Oxidation, % <l7 4.81 Core-Wide Cladding Oxidation, % <1 < 0.358 In letter 2CANI19610, dated November 24,1996, Entergy Operations requested the approval to use CENPD-137, Supplement 1-P, SBLOCA evaluation model. This request was approved in Amendment 179, dated January 14, 1997. In that request, the PCT for the limiting break was determined to be 2011*F. The currently calculated PCT of 1798 'F is a greater than 50*F difference when compared to the previous limit. Since the Staff has approved the use of a different evaluation model for the SBLOCA analysis, and the currently calculated PCT is less than the previously approved limit, it is Entergy Operations' position that the reporting criteria of 10CFR50.46, to notify the NRC of a greater than 50 'F difference in analysis results, does not apply to the differences between the two evaluations.

CENPD-137, Supplement 2-P-A, a calculative methodology for SBLOCA, will be added to the listing of approved methodologies listed in Technical Specification 6.9.5.1. The addition of this methodology to TS 6.9.5.1 is evaluated in a subsequent discussion.

In addition, the " Orifice Size" column, of TS Table 3.7-5, " Steam Line Safety Valves", which specifies a design requirement for the MSSVs, is deleted from the table in accordance with the standardized TS format and content requirements identified by NUREG-1432, Standard TechnicalSpecyications Combustion Engineering Plants, Revision 1, dated April,1995. The minimum orifice size is identified in the ANO-2 SAR and is appropriately controlled under the SAR revision process.

Attachment to

. 2CAN069804 Page 5 of 9 l

l ' Immets Due to Revision of PSV Lift Settina To';rance The PSVs consist of two totally enclosed, back pressure compensated, spring-loaded pressure relief vr.lves located on top of the pressurizer. Design and operational characteristics for these valves are given in SAR Table 5.5-10.

The PSVs must pass sufficient pressurizer steam to limit the RCS pressure to less than 110 percent of design pressure following a complete loss of turbine-generator load with no simultaneous reactor trip. The reactor is assumed to trip on a high pressurizer pressure signal.

In determining the maximum steam flow through the PSVs, the MSSVs are assumed to be operational. Conservative values for all system parameters, delay times, and core moderator coefficient were assumed during analysis for peak RCS pressure.

Accident events and AOOs for the PSV lift setting tolerance revision were evaluated. Based on the review cf SAR, the feedwater line break is the limiting accident event for the determination of the peak RCS pressure and peak steam generator secondary pressure. This event was assessed for Cycle 13 operation. The event assessment included up to 30% of steam generator tubes being plugged and a 10% reduction in RCS flow. This assessment was performed with the NRC approved, Combustion Engineering Nuclear Transient Simulation (CENTS), analytical methodology.

Sensitivity studies were performed on the RCS flow rate and the amount of steam generator I tube plugging. The results of these analyses indicated that 0% tube plugging resulted in the maximum analyzed RCS pressure. RCS flow showed no effect on the peak pressure so a nominal Bow of 120.4 x 10' lb/hr was assumed Based on these sensitivity analyses results, 0% tube plugging and nominal RCS flow were assumed in the feedwater line break accident  !

analysis.

The RCS flow sensitivity studies determined that the peak RCS pressure would be 2730.1 psia and peak secondary pressure would be 1163.8 psia. Both the RCS and secondary peak pressures are within the acceptance criteria of 110% of design pressure (2750 psia and 1210 psia, respectively). As these results are within the acceptance criteria, there are no increased ,

risks in the analytical assumptions related to radiological dose or accident consequences The limiting peak RCS pressure event for AOOs is the LOCV event. This event was assessed 4 for Cycle 13 operation, including the proposed increase in lift setting tolerance. The assessment, which also included up to 30% steam generator tube plugging and a 10%

reduction in the RCS flow, was performed using the CENTS methodology.

The AOO assessment also includes a sensitivity study. This study determined that the peak RCS pressure was based on maximum RCS flow and no plugged steam generator tubes. The effect of tube plugging and RCS flow had a very small impact on the analysis results.

i Attachment to

. 2CAN069804 Page 6 of 9 The results of the AOO analysis indicate that the peak RCS pressure is 2683 psia. The RCS l

' pressure is within the acceptance criteria of 110% of design pressure (2750 psia). As these l results are within the acceptance criteria, there are no increases in radiological dose or i accident consequences.

Revision of the Maximum Allowable Linear Power Level - High Tria Setooint With Three Inocerable Steam Line Safety Valves l l

The revision of the linear power level trip setpoint, TS Table 3.7-1, is necessary to utilize the increased MSSV lift setting tolerances.

1 Consistent with the latest revision of NUREG 1432, an LOCV event analysis can also be used /

l to define the linear power setpoints in TS Table 3.7-1. This method is different from the

! methodology and results currently defined in the TS bases The current values are:

Maximum Allowable Linear Power Level - 1 Maximum Number ofInoperable MSSVs on High Trio Setooint i Any Operating Steam Generator (% ofRated THERMAL POWER)

I 1 91.0 2 67.7 I 3 45.0 l

l ANO has chosen to use the results from the current method and the LOCV analysis. In the LOCV analysis, trip setpoints of 103%, 74%, and 36%, for 1, 2, and 3 inoperable MSSVs, respectively, were determined using CENTS for the LOCV events at these power ratings. TS Table 3.7-1 has been revised to reflect the most conservative trip setpoint that is determined from either the current methodology in the bases, or from the LOCV analysis. The trip setpoint values for one or two inoperable MSSVs, which were derived from the formula referenced in the TS Bases, remain unchanged while the trip setpoint for three inoperable MSSVs has been revised to 36% of rated thermal power consistent with the results from the LOCV analysis.

The bases for TS 3/4.7.1 was also revised. Text changes reflect that an increased allowable setpoint tolerance of

  • 3% is no longer " conservative" with respect to ASME Section XI requirements, but is " consistent" with those requirements. The design steam flow values (steam flow specified as lbs/hr and as an equivalent %) calculated from the 102% rated thermal power limit are deleted from the bases. These design steam flow values are identified I in the SAR. Furthermore, the use of the LOCV event to d(me the linear power setpoint is included as an additional methodology permissible for determination of trip setpoints. The l most conservative setpoint resulting from the results of the two methodologies is selected.

L The specific performance requirements for the valves in the text of the bases, and the addition i ot' the LOCV methodology for determining setpoints, are allowed to be deleted by the l standardized format ofNUREG-1432.

. Attachment to 2CAN069804 Page 7 of 9 l Addition OfThe Latest NRC Approved SBLOCA For CE Designed Plants to TS 6.9.5.1 Addition of CENPD-137, Supplement 2-P-A, is necessary to utilize increased MSSV lift l setting tolerances. Technical Specification 6.9.5.1 provides a specific list of the " analytical methods used to determine the core operating limits addressed by the individual Technical Specifications".

Each reload-related accident analysis addressed in the SAR is considered in the reload report for each fuel cycle. The reload report documents and evaluates cycle-specific parameters in

order to ensure that thermal performance during hypothetical transients is acceptable. For l each core reload, the margins of safety for fuel system design, nuclear design, and thermal-hydraulic design are addressed in a reload report. The applicable operational limits and setpoints (listed in the Core Operating Limits Report) for a particular fuel cycle are determined to be within allowable design limits and TS requirements for safe operation of the reactor. Each reload cycle is evaluated under the provisions of 10CFR50.59, Changes, tests, experiments.

l SBLOCA analysis results are utilized as inputs to the moderator temperature coefficient analysis, linear heat rate analysis, azimuthal power tilt analysis, and axial shape index core operating limits analysis. In order to adopt improved methodologies, it has become desirable to utilize CENPD-137, Supplement 2-P-A, for the determination of the limits previously listed. The Staff approved the use of this methodology in the safety evaluation dated December 16,1997.

l DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION An evaluation of the proposed change has been performed in accordance with 10CFR50.91(a)(1) regarding no significant hazards considerations using the standards in 10CFR50.92(c). A discussion of these standards as they relate to this amendment request  ;

follows:

Criterion 1 - Does Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated.

This change allows for a larger

  • 3% tolerance versus +1%, -3% as-found lift setting tolerance. The proposed change does not involve any change to the physical characteristics of the main steam safety valves (MSSVs) and pressurizer safety valves (PSVs), and will have no impact on the as-left settings. During testing, the MSSVs and PSVs will continue to adjusted to 1% of the Technical Specification (TS) lift setting.

The impact on the Safety Analysis Report (SAR) analyses when the as-found lift setting

. tolerances are increased has been evaluated and the effects upon the impacted events have been found to be within acceptable limits, providing the allowable linear power level with L_._ _ _ _ _ _ _ _ _ _ __ ._ . . . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _

Attachment to

, 2CAN069804 Page 8 of 9 three inoperable MSSVs is revised from 45% to 36%, and that the latest NRC approved C-E

' small break loss of coolant analysis (LOCA) evaluation model, CENPD-137, Supplement 2-P-A, is included as a methodology for determination of operating parameters identi6ed within the core operating limits report (COLR). With these concurrent changes, plant systems required for safe operation and shutdown will continue to be available to fulfill their safety function as described in the SAR. Steam production in excess of relief capacity is precluded by the physical design of the plant and operation of the reactor protection system. Revision of the MSSV as-found lift setting tole ance from +1%,-3% to

  • 3% does not alter safety analyses conclusions.

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

Criterion 2 - Does Not Create the Possibility of a New or Different Kind of Accident from any Previously Evaluated.

This change does not create any new plant configuration or operational mode. This proposal to increase the MSSV and PSV as-found lift setting tolerance does not modify equipment or change the manner in which the MSSVs and PSVs will be operated. ASME design requirements for maintaining system operating pressure limits below the maximum design pressure of 1210 psia for plant secondary systems, and 2750 psia for the reactor coolant system (RCS) are not impacted. The reduction in allowable linear power level when three MSSVs are inoperable assures plant operation within current analysis assumptions. The addition of topical report CENPD-137, Supplement 2-P-A, as a reference to develop the COLR is bounded by assumptions within the existing safety analysis. The cycle specific COLR analyses will continue to be perforrd utilizing NRC approved methodologies. The TS changes do not require any new equiprtpat be included in the design basis, and current equipment will continue to be operated in a manner consistent with its design.

Therefore, this change does g create the possibility of a new or different kind of accident from any previously evaluated.

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

The upper tolerance limit for design pressure is not affected by this change. During the most severe anticipated operational transient, the Secondary System pressure and RCS pressure wi!!

not exceed 110% of design pressure. The MSSV and PSV lift settings will continue to b e set within 1% of the TS lift setting during surveillance testing.

l The decrease in the peak cladding temperature of the reactor fuel, due to a change in the methodology for analysis, does not significantly impact previous analytical results. The l

l current and previous analytical methodologies are approved by the Staff.

i The impact of the proposed changes on the ANO-2 SAR analyses have been evaluated. The evaluation demonstrates that the results of the impacted events remained within the acceptable t

Attachment to 2CAN069804 ,

Page 9 cf 9 limits providing the maximum linear power level percentage for three inoperable MSSVs is

' reduced. This reduction in maximum allowable linear power level assures that adequate steam relief capacity will be available to prevent overpressurizing the secondary steam system during the most severe anticipated operational transient.

Addition of topical report CENPD-137, Supplement 2-P-A, will not reduce the existing TS operability and surveillance requirements. The cycle specific COLR limits for future reloads will continue to be developed based on NRC-approved methodologies. The ANO-2 TSs will continue to require that the core be operated within these limits.

The cumulative impact of all of the proposed changes and the results of the impacted events have been found to be within acceptable limits. The system capabilities to mitigate and/or prevent accidents will be the same as they were prior to these changes.

Therefore, this change does n91 involve a significant reduction in the margin of safety.

Therefore, based upon the reasoning presented above and the previous discussion of the amendment request, Entergy Operations has determined that the requested change does n91 involve a significant hazards consideration.

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