2CAN069109, Application for Amend to License NPF-6,changing Tech Spec Figure 3.4-2 Re RCS Temp Limitations for 0 to 10 Yrs of Full Power Operation,Per Rev 2 to Reg Guide 1.99.C-E Rept A-MPS-ER-002 Re Pressure/Temp Limits Encl

Application for Amend to License NPF-6,changing Tech Spec Figure 3.4-2 Re RCS Temp Limitations for 0 to 10 Yrs of Full Power Operation,Per Rev 2 to Reg Guide 1.99.C-E Rept A-MPS-ER-002 Re Pressure/Temp Limits Encl

ML20081K975
Person / Time
Site:	Arkansas Nuclear
Issue date:	06/18/1991
From:	Carns N ENTERGY OPERATIONS, INC.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20081K976	List: 2CAN069109, Application for Amend to License NPF-6,changing Tech Spec Figure 3.4-2 Re RCS Temp Limitations for 0 to 10 Yrs of Full Power Operation,Per Rev 2 to Reg Guide 1.99.C-E Rept A-MPS-ER-002 Re Pressure/Temp Limits Encl ML20081K979 ML20081K982
References
RTR-REGGD-01.099, RTR-REGGD-1.099 2CAN069109, 2CAN69109, NUDOCS 9107020250
Download: ML20081K975 (11)

Text

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Eniergy tai.< u oe.,. u n..i c.

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T ei f>0964 -M88 Neil S. " Buss" Larns  !

Vu Preusem Orates ANO Juno 18, 1991 2CAN069109 U. S. Nuclent Regulatory Commission Document Control Desk Mail Station PI-137 Washington, DC 20555

SUBJECT:

Arkansas Nucient One - Unit 2 Docket No. 50-358 License No. NPF-6 s Proposed Chnnge ta the Technica1 Specification Pressure / Temperature himits Gentlemon:

Figurn 3.4-2 of the Arkansas Nuclent on9, Unit 2 (ANO-2) Technical Specifications, entitled "Renct or Coolant System Temperat urn himitations for o o 10 Years of Full Pownr Operation", expires af ter 8.24 of fect ivn full power years (EfPY) of operation (10 years of 2900 MW,r peintion nt, nn With ANO-2 60% capacit.y fact or; actual thermal rating of only 2815 approaching the expirnt ion of the current pressure /tempernMC[u)r. e curves, Entergy Operations is proposing chnnges to this figurn, Technical Specification 3/4.4.9, and the associated Bases to reflect operational limitations through 21 EFPY.

These proposed changes are not due to Generic hetter 88-11. "NRC Position on Rndintion Embrittlement of Reactor Vessel Materials and its impact on Pinnt Opnrations". This generic Intter noted that the methods of Envision 2 to Regulatory Guide 1.99 should he used to predict thn effect of neutron rndintion on rnactor vessel material. Entergy Operations has previously determined that t.ho use of tho Revision 2 methodology did not have a major impact on the current pressure /temperaturn limits (lettet DCAN018917, dated Janunry 27, 1989). TN changes current.ly proposed worn, however, developed using th, prd icti sn met hods of Regulatory Guide 1.99, Revision 2.

Tho effects of thn proposed pressure / temperature limits for low temperature overpressu e protect ion (hTOP) and for the Prersurized Thermal Shock analysis '. ore considered. A discussion of these nnnlyses is att ached.

The proposed changes hnve bean evaluated in accordance with 10CFR50.91(n)( 1), us ing t he cri t erin in 10CFR50.92(c) and it has been detnrmined that this r erlue s t involves no significnnt hazards considerations. The bosis for this determinn' ion is includnd in thn 9107020250 910618 PDR ADOCK 05000368 P PDR h}

l

e 'a U. S. NRC June 18, 1991 Page 2 enclosed submittal. AllB Combustion Engineering report number A-MI'S-ER-002 is also attached which provides the required 10CFR50, Appendix G analysis utilized for development of the pressure / temperature limits.

The circumstances of this request are not exigent or emergency, llowever, AND-2 will reach the current operational limits set forth in the Technical Specifications on November 18, 1991, assuming continued 100% power operation. Theref ore, the changes should be processed accordingly.

We request that the effective date for this change be November 18, 1991.

Very truly yours,

<> atum, f

l--> ,:L fg., c NSC:sgw Attachments cc: Mr. Robert Martin U. S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive, Suite 1000 Arlington, TX 76011 NRC Senior Resident inspector Arkansas Nuclear One - ANO-16 2 Number 1, Nuclear Plant Road Russellville, AR 72801 Mr. Thomas W. Alexion NRR Project Manager, Region IV/ANO-1 U. S. Nuclear Regulatory Commission NRR Mall Stop 11-D-23 One White Flint North 11555 Rockville Pike Rockville, Maryland 20852 Ms. Sheri R. Peterson NRR Proj mt Manager, Region IV/ANO-2 U. S. Nuclear Regulatory Commission NRR Mail Stop 11-D-23 One White Flint North 11555 Rockville Pike Rockville, Mary 1nnd 20852 Ms. Greta Dieus, Director Division of Radiation Control and Emergency Management Arkansas Department of ilealth 4815 West Markham Street Little Rock, AR 72201

e s STATE OF ARKANSAS )

) SS COUNTY OF LOGAN )

0ATil

1. J. W. Yelverton, being duly sworn, subscribe to and say that 1 6m General Mannger, Plant Operations ANO for Entergy Operations, Inc.; that I have full authorit3 to execute this oath; that I have read the locument numbered 2CAN069109 and know the contents thereof; and that to the best of my knowledge, information and belief, the statements in it are true.

0 .10 ?/6 ,-e

/, ' . W. N41verton SUBSCRIBED AND SFORN TO before me, a Notary Public in and for the County and State above named, this uO

/rfd day of ,_ /

1991.

AMA/ 1 %ftj$ 07% )_

ff N'otary Public My Commission Expires:

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e e.

PROPCSED TECilNICAL SPECIF] CATION CHANGE AND RESPECTIVE SAFETY ANALYSIS IN Ti;E liATTER OF AMENDING LICENSE NO. NPF-6 ENTERGY OPERATIONS INC.

ARKANSAS NUCLEAR ONE, UNIT 2 DOC!'ET NO. 50-368 i

0 t PROPOSED CHANGE Figure 3.4-2 of the Arkansas Nuclear One, Unit 2 (ANO-2) Technical Specifications (TS), entitled "Rcoctor Coolant System Temperature himitations for 0 to 10 Years of Full Power Operation", expires af t er 8.24 ef fective full power years (FFPY) of operation (10 years 2900 MW 7 of operation at an 80%

capacity factor; actual thermal rating of only 2815 MWT ). With ANO-2 approaching 8 EFPY, Entergy Operations is proposing changes to this Figure. TS 3/4_4.9, and the associated Bases to reflect operational limitations through 21 EFPY.

BACKGROUND Appendix A, General Denign Criteria for Nuclear Power Plants, of 10CFR50 establishes the minimum design requirements for nuclear power plants. Criterion 14 dictates that the reactor coolant system (RCS) pressure boundary be designed, fabricated, crected and tested to ensure a low probaallity of abnormal leakage, of rapidly propagating failure and of gross rupture. In additicn, Criterion 31 provides general guidance in the prevention of fractures of the RCS pressure boundary while 10CFR50, Appendix G provides the fracture toughness requirements.

Both regulations dictate that the pressure boundary be designed with sufficient margin to withstand any condition during normal operation including anticipated operational occurrences and system inservice leak and hydrostatic tests.

As a result, all components of the ANO-2 RCS are designed to withstand the effects of cyclo loads due to system temperature and pressure changes. The integrity of the RCS pressure boundary; however, is directly linked to the operation of tne unit. TS 3/4.4.9, " Pressure /Temperat.ure Limits", specifies the RCS operat ional limitations necessary to guaranten that the pressure boundary components continue to meet the design requirements specified in the regulations.

The fracture toughness of the reactor vessel materials; however, changes due to neutron irradiation throughout the service life and is required to be monitored.

TS 4.4.9.1.2 requires that vessel specimens be removed periodically and examined to determine the magnitude of any change. The examination results are then utilized to update the pressure / temperature limitations to account for the aging of the vessel.

To date only one specimen has been examined. This specimen was the 97-degree surveillance capsule. Battelle Columbus Laboratories issued a report on their examination, testing and evaluation of the specimen on May 1, 1984.

A corresponding summary report was submitted t o the NRC (2CAN028503 dated February 8, 1985) in accordance with 10CFR50 Appendix II. Based upon the data from the first surveillance capsuln analysis, as described in the summary report, the ANO-2 upper shelf energy is not predicted to drop below the 50 f t-lb minimum set by 10CFR50 Appendix G during its operational life.

DISCUSSION The 21 EFPY limits for normal operation (both heatup and cooldown) and inservice hydrostatic tests were developed in accordance with the requirements of 10CFR50, Appendix G and the prediction methods of Regulatory Guide 1.99, Revision 2.

Attachment I contains ABB-Combustion Engineering report number A-MPS-EP-002 detailing the development of the pressure /temperatare limits.

. i The Adjusted Reference Temperature (ART) values for the controlling beltline material, based on Regulatory Guide 1.99, Revision 2, Position 1.1, were i calculated. Position 1.1 dictates the calculational method for determining the ART values when two or more credible surveillance data sets are not available.

Currently, Entergy Operations has analyzed one surveillance specimen for ANO-2 in accordance with TS Table 4.4-5. The results of this examination, along with the relationship for fluence as a function of reactor vessel depth as given in Regulatory Guide 1.99, Revision 2, were utilized by Entergy Operations to derive the fluence which is an input to the ART values. The fluence value used in the pressure / temperature limits development was verified to be conservative by comparing it to fluences derived from ANO-2 cavity dosimetry measurements.

The pressure / temperature limits cited in report A-MPS-ER-002 (Attachment 1) include corrections for elevation differences, flow induced pressure drops and instrumentation uncertainties. The instrumentation uncertainty values used in the report were nominal values (85 psi snd 30 degrees F). Entergy Operations has calculated plant specific values of 85 psi and 12 degrees F. The proposed limi+s which utilize bounding instrument uncartainty values of 85 psi and 20 degrees F are uniformly 10 degrees F less than that used in the report and are bounded by actual instrument uncertainty values.

In the development of the new pressure /temperatare limits, different heatup and cooldown scenarios were considered in order to optimize future operating flexibility and margin. The heatup limits are based on a constant rate over the operating range of temperatures.

Two scenarios are considered for cooldowns. Curves were developed that modeled ramped cooldowns and stepped temperature decreases. The ramped scenario considered operation when a constant cooldown rate is maintained while the stepped case assumes an instantaneous decrecsc in temperature and a subsequent thirty minute hold period. The most limiting of the two cooldown cases is presented in the proposed changes.

The proposed limits corresponding to a heatup rate of 50 degrees F/hr. , 60 degrees F/hr., 70 degrees F/hr., or 80 degrees F/hr. for all temperatures are delineated in the proposed TS. The limits corresponding to tuo cooldown rates of 100 degrees F/hr. for RCS temperatures above 220 degrees F; 60 degrees F/hr.

In the 220 degrees F to 140 degrees F range; and 25 degrees F/hr. below 140 degrees F are also delineated in the proposed changes. The cooldown limits allow for an instantaneous drop in temperature of 50 degrees F, 30 degrees F and 12.5 degrees F in the 100 degrees F/hr., 60 degrees F/hr., and 25 degrees F/hr.

ranges respectively, in any half hour period. These rates and limits will provide the flexibility necessary to operate ANO-2 effectively and safely through 21 EFPY.

The maximum pressve for shutdown cooling (SDC) operation previously shown on the pressure / temperature limitation curves has been removed since this is provided as an operatinp limit and is not required for pressure / temperature limit protection.

The Bases have been modified to reflect the new operational limits and approach.

The previous Bases Figure B3/4.4-1 is being deleted to avoid conflict with the guidance given in Regulatory Guide 1.99, Revision 2.

OTHER RELATED ASSESSMENTS Low Temperature Overpressure Protection

-The Low Temperature Overpressure Protection (LTOP) system provides backup protection to that provided by the pressure / temperature limits. The LTOP setpoint and enable temperature are typically derived from the pressure / temperature limits in accordance with Standard Review Plan 5.2.2.

Therefore, a change in the pressure / temperature limits necessitates a review of-the LTOP setpoint and enable temperature. Based on the review, the new LTOP relief valve design setpoint is 450 psig and the new enable towperature design setpoint is 200 degrees F.

L Entergy Operations is submitting under separate cover, an ANO-2 LTOP related proposed TS in response to Generic Letter 90-06. In that submittal, a discussion of the methodolegy used to determine the new setpoint and enable temperature is provided.

Pressuri ed Thermal Shock The effect on the Pressurized Thermal Shock (PTS) analysis must be reconsidered as part of the development of a pressure / temperature limits submittal. The PTS Rule (10CFR50.61) requires that the assessment of the RT PT Temperature) be updated whenever changes in core loadings,S (R ference surveillance measurements or other information indicates a significant change in the .

projected values. In addition, in letter 2CNA078703, dated July 20, 1987, the NRC requested a reasnessment of the PTS reference temperature be submitted to the NRC when any pressure / temperature submittals are made. Lastly, the PTS Rule (10CFR50.61) was revised, effective June 14, 1991, which also requires a reassessment of the PTS reference temperature to be aubmitted. In compliance with the above, the PTS analysis has been re-evaluated and adjusted to reflect the assumptions made in the development of the new pressure / temperature curves, consistent with the June 14, 1991, revision to the PTS Rule.

The proposed pressure / temperature curves are based on-somewhat different

.information than the original PTS evaluation. The original PTS assessment used a different- reactor vessel- surface fluence value than is being-utilized for the 21 EFPY pressure / temperature limitations. In addition, some of the chemical

-compositions and material properties differ from those submitted in the original PTS evaluation (letter 2CAN018605, dated January 22, 1986). Attachment 2

documents the basis for the changes in the chemical compositions and initial RT . These changes are a result of a more accurate determination of the ANO-2 makkIialconditionsbaseduponoriginalmaterialtestreports.

The results of the first surveillance capsule analysis, as described in the previously submitted summary report, indicate a fast neutron fluence rate of 1.4 x 10'^nyt/EFPY for the inside surface of the ANO-2 reactor vessel. The surveillance capsule was removed from the reactor following Cycle 2, which L represented 1.69 EFPY of operation. In Cycle 6 ANO-2 incorporated a low leakage core design for fuel economics that had an added benefit of reducing the vessel fluence rate. Therefore, the predicted fluence rate of 1.4 x 10 *nyt/EFPY 2

is conservative with respect to current operating practice at ANO-2.

A bounding, conservative reactor vessel surface fluence rate of 1.78 x i 10'*nyt/EFPY has been assumed for the proposed 21 EFPY pressure / temperature limitations. In the previous ANO-2 PTS submittal a similar bounding, conservative fluence rate of 1,73 x 10 *nyt/EFPY was used. This fluence was 2

based upon fluence estimates made in EPR1 report NP-4238, " Neutron Energy Spectra in the Core and Cavity of the AND-2 PWR" For conservatism this value was used since it was higher than that determined by Battelle Columbus baboratories. It should be noted that the values of fluence stated for the 1/4 vessel thickness location of interest ( for pressure / temperature limits) is less in both the Bette11e report (8.45 x 10 'nyt/EFPY or 2.70 x 10nyt at 32EFPY) and the EPRI report (9.46 x 10nvt/EFPY or 3.03 x 10nyt at 32EFPY) than in the ANO-2 Safety Analysis Report and basis for the current pressu:e-temperature limits (3.47 x 10'*nyt at 32EFPY). In order to maintain consisten:y with the proposed 21 EFPY pressure / temperature limits, a reactor vessel surface fluence rate of 1.78 x 10'*nyt/EFPY has been used to update the PTS evaluation, except for the Upper to Intermediate Shell Girth Weld which is based on 5.13 x 10nyt/EFPY. This value is greater than that used in the previous PTS evaluation by the same proportion as the other materials being evaluated (i.e.

1.78/1.73).

The Intermediate Shell Plate C-8009-1 (heat number C8161-3) continues to be the limiting material with a RT at license expiration (currently estimated as29.56EFPY)of126degreesF,wbhisa47degreesFreductionfromthe previous calculation. This dif ference is primarily due to the use of an initial RT ~

EI""* * *"* E** " ##" ""E" RuhDT

e. The calculated PTS reference temperature provides significant margin to the PTS screening criteria of 270 degrees F. ables 1 and 2 document the new fluence and RT values, requested by the new PTS Rule, along with the chemical compositionanb,nitialRT ND'r changes.

As previously noted, the PTS update is required by the recent revision to 10CFR50.61 and was also requested by the NRC as part of each pressure / temperature limitation submittals. The adjustments to the PTS analysis are presented to provide consistency between the assumptions in the pressure / temperature IJmitation development and the PTS evaluation.

SUMMARY

Entergy Operations is proposing to amend the pressure / temperature limits through 21 EFPY for ANO-2. These limits will be used during normal heatup, normal cooldown, inservice inak and hydrostatic tests, and core operation as required by 10CFR50, Appendix G. The limits have invoked additioaal operational restraints that must be maintained for future operation, but allow enough flexibility to ef fectively and sa fely manage the plant.

Protection against nonductile failure la ensured by maintaining the KCS pressure below and to the right of the limits of the pressure / temperature limit curves.

In the unlikely event that the administrative controls, systems and procedures for normal operation fail to prevent. violation of these limits, the LTOP system provides a backup means of protection against nonductile failure of RCS components. The ANO-2 LTOP setpoint has also been reassessed, with the results being a revised LTOP setpoint of 450 psig and an enable temperature of 200 degrees F.

~

APTSovaluationfisdocumented'inTables1.and2. This evaluation follows the methods established in the recent revision'to the PTS-Rule. The fluence rate

- and material properties' utilized in the pressure / temperature limitations development are' also used :in this assessment to provide consistency. The final RT values are decreased from the values report ed- originally .and remain well

. be $o.g.3w the PTS screening criteria, BASIS FOR PR0p0 SED NO SIGNIFICANT llAZARDS CONSIDERATION DETERMINATION

'Entergy Operations has concluded that the proposed changes to TS 3/4.4.9, its associated Bases, and Figure 3.4-2 do not involve a significant hazards consideration because the operation of Arkansas Nuclear One, Unit 2 in accordance with these changes would not:

(1) Involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed changes will not increase the probability or consequences of any accident previously evaluat'ed since the proposed changes revise the pressure / temperature limits in accordance with 10CFR50, Appendix G utilizing the latest NRC guidelines in Regulatory _ Guide 1.99, Revision 2 relative to-estimating neutron irradiation damage to the reactor vessel.

The proposed changes also maintain the conservative limits with respect to the I. TOP system.

(2) . Create the possibility of a new or different kind of accident from any previously evaluated.

The proposed changes will not create the possibility of a new or dif ferent kind of accident from _ any previously analyzed since they do not introduce-new systems, failure modes, or other_ plant perturbations. .The proposed changes revise the pressure / temperature limits in accordance with 10CFR50, Appendix G utilizing the latest NRC guidelines in Regulatory Guide 1.99, Revislom 2 relative to estimating neutron irradiation damage to the reactor vessel.

(3). Involve a significant- reduction in the margin of safety.

The proposed changes will not involve-a significant reduction in the margin of safety since. equal or more stringent pressure / temperature limitation requirements for reactor operation will be applied. The proposed changes were derived in accordance with~ approved NRC methodology which was developed to assure the RCS pressure boundary is designed with sufficient margin to withstand any condition during normal operation including anticipated operational occurrences and system inservice leak and hydrostatic tests.

These requirements were revised in accordance with.10CFR50, Appendix G utilizing the' latest NRC guidance in Regulatory Guide 1.99, Revision 2 relative to estimating neutron irradiation damage to the reactor vessel.

The .LTOP system limits were also reanalyzed for the proposed changes.

Therefore, based on the evaluation discussed above, Entergy Operations has concluded that the proposed changes do not involve a significant hazards consideration.

l._ - _ _ - - _ _. - _. . _ . _ . _ _ _ _ . . _ _ _ _ _ _ , , -

TABLE 1 10CFR50.61 PTS EVALUATION OF ARKANSAS NUCLEAR ONE - UNIT 2 I

Chemical Constants for Inside Calculations, F Surface Fluence, nyt Calculated RT g Composition, w/o RT PTS CurrentCondit$ns Current Conditions Reactor Vessel Identification Margin (8 EFPY) (8 EFPY)

Beltline Region Location Number Copper Nickel Initial RT 'DT I

66 1.42x10 ' 61

.05 .18 -56 Intermediate Shell 2-203- A, B , C Longitudinal Weld I

66 1.42x10 ' 61 3-203-A, B, C .05 .18 -56 Lower Shell Longitudinal Weld I

56 1.42x10 ' 85 9-203 .05 .08 -10 Intermediate to Lower Shell Girth Weld I

66 0.41x10 40 8-203 .23 .18 -56 Upper to Intermediate Shell Girth Weld 1

34 1.42x10 ' 100 C-8009-1 .12 .63 -26 Intermediate Shell 19 90

.08 .59 0 34 1.42x10 C-8009-2 34 1.42x10 ' 90

.60 0 C-8009-3 .08 10 34 1.42x10 ' 102 C-8010-1 .08 .59 12 Lower Shell Plates l

34 1.42x10 ' 54 C-8010-2 .07 .66 -28 lo 34 1.42x10 ' 52 C-8010-3 .07 .65 -30 i

TABLE 2 10CFR50.61 PTS EVALUATION OF ARKANSAS NUCLEAR ONE - UNIT 2 Chemical Constants for Inside alculations, T Surface Fluence, nyt Calculated RT FTS Composition, w/o RT PTS License Expire License Expire Reactor Vessel Identification Margin (29.6 EFPY) (29.6 EFPY)

Beltline Region Location Number Copper Nickel Initial RT@T I9 76 2-203-A, B, C .05 .18 -56 66 5.26x10 Intermediate Shell Longitudinal Weld 19 76 3-203-A, B, C .05 .18 -56 66 5.26x10 Lower Shell Longitudinal Weld 19 96

.05 .08 -10 56 5.26x10 Intermediate to Lower 9-203 Shell Girth Weld IO 68

.18 -56 66 1.52x10 Upper to Intermediate 8-203 .23 Shell Girth Weld 19 126

.12 .63 -26 34 5.26x10 Intermediate Shell C-8009-1 19 106

.59 0 34 5.26x10 C-8009-2 .08 1 106 I

.08 .60 0 34 5.26x10 '

C-800"-3 I

34 5.26x10 118 C-8010-1 .08 .59 12 l

Lower Shell Plates 10 68

.07 .66 -26 34 5.26x10 '

C-8010-2

.07 .65 -30 34 5.26x10I ' 66 C-8010-3

_ _ _ -