ML060130639
| ML060130639 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 01/13/2006 |
| From: | Holland D G Plant Licensing Branch III-2 |
| To: | Forbes J S Entergy Operations |
| Holland D G, NRR/DLPM, 415-1436 | |
| Shared Package | |
| ML060130643 | List: |
| References | |
| TAC MC9437 | |
| Download: ML060130639 (16) | |
Text
January 13, 2006Mr. Jeffrey S. ForbesSite Vice President Arkansas Nuclear One Entergy Operations, Inc.
1448 S. R. 333 Russellville, AR 72802
SUBJECT:
ARKANSAS NUCLEAR ONE, UNIT 1 (ANO-1) - ISSUANCE OF AMENDMENTRE: REVISION OF THE ALLOWABLE VALUE FOR EMERGENCY FEEDWATER INITIATION AND CONTROL FUNCTION (TAC NO. MC 9437)
Dear Mr. Forbes:
The Commission has issued the enclosed Amendment No. 227 to Renewed Facility OperatingLicense No. DPR-51 for ANO-1. The amendment consists of changes to the Technical Specifications (TSs) in response to your letter dated January 3, 2006, as supplemented by letters dated January 6 and 10, 2006. This amendment revises TS 3.3.11, "Emergency Feedwater [EFW] Initiation and ControlSystem Instrumentation." Entergy Operations, Inc. (Entergy) requests a revised TS allowable value of 9.34 inches and a limiting trip setpoint value of 10.42 inches in order to achieve andmaintain 100 percent power. An actuation time delay of 10.4 seconds is also provided tominimize the possibility of inadvertant actuations during anticipated transients such as mainfeedwater transients, main turbine trips, etc. Operation at 100 percent power with the currentallowable value involves an increased risk of spurious EFW initiation. Entergy requested this amendment as an emergency license amendment to allow for full, rated power operation (100 percent) power at ANO-1 while preventing unnecessary plant transients.A copy of the Safety Evaluation is also enclosed. The Notice of Issuance will be included in theCommission's biweekly Federal Register notice.Sincerely,/RA/Drew Holland, Project ManagerPlant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor RegulationDocket No. 50-313
Enclosures:
- 1. Amendment No. 227 to DPR-512. Safety Evaluationcc w/encls:See next page
ML060130639NRR-058OFFICELPL4/PMLPL4/LAEICB/BCSPWB/BCOGC NloLPL4/BCNAMEDHollandLFeizollahiAHoweJNakoskiMWoodsDTeraoDATE1/13/061/13/061/13/061/13/061/13/061/13/06 ENTERGY OPERATIONS, INC.DOCKET NO. 50-313ARKANSAS NUCLEAR ONE, UNIT 1AMENDMENT TO RENEWED FACILITY OPERATING LICENSEAmendment No. 227Renewed License No. DPR-511.The Nuclear Regulatory Commission (the Commission) has found that:A.The application for amendment by Entergy Operations, Inc. (the licensee) datedJanuary 3, 2006, as supplemented by letters dated January 6 and 10, 2006, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I;B.The facility will operate in conformity with the application, the provisions of theAct, and the rules and regulations of the Commission;C.There is reasonable assurance (i) that the activities authorized by thisamendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with theCommission's regulations;D.The issuance of this license amendment will not be inimical to the common defense and security or to the health and safety of the public; andE.The issuance of this amendment is in accordance with 10 CFR Part 51 of theCommission's regulations and all applicable requirements have been satisfied. 2.Accordingly, the license is amended by changes to the Technical Specifications asindicated in the attachment to this license amendment, and Paragraph 2.c.(2) of Renewed Facility Operating License No. DPR-15 is hereby amended to read as follows: (2)Technical SpecificationsThe Technical Specifications contained in Appendix A, as revised throughAmendment No. 227, are hereby incorporated in the license. EOI shall operatethe facility in accordance with the Technical Specifications.3.The license amendment is effective as of its date of issuance and shall be implementedwithin 7 days from the date of issuance.FOR THE NUCLEAR REGULATORY COMMISSION/RA/David Terao, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical SpecificationsDate of Issuance: January 13, 2006 ATTACHMENT TO LICENSE AMENDMENT NO. 227RENEWED FACILITY OPERATING LICENSE NO. DPR-51DOCKET NO. 50-313Replace the following pages of the Appendix A Technical Specifications with the attachedrevised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change. Remove PageInsert Page3.3.11-23.3.11-23.3.11-33.3.11-3
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATIONRELATED TO AMENDMENT NO. 227 TORENEWED FACILITY OPERATING LICENSE NO. DPR-51ENTERGY OPERATIONS, INC.ARKANSAS NUCLEAR ONE, UNIT 1DOCKET NO. 50-31
31.0INTRODUCTION
By application dated January 3, 2006 (Agencywide Documents Access and ManagementSystem (ADAMS) Accession No. ML060050448), Entergy Operations, Inc. (Entergy/the licensee) requested an emergency Technical Specification (TS) amendment for (Arkansas Nuclear One, Unit 1 (ANO-1). By letters dated January 6 (ADAMS Accession No. ML060110060) and January 10, 2006 (ADAMS Accession No. ML060120152), Entergy supplemented its January 3, 2006, license amendment request with additional information requested by the Nuclear Regulatory Commission (NRC) staff. The licensee indicates that the change is needed because phenomena not related to once-through steam generator (OTSG) level tend to reduce the measured value of the level during normal operation by an amount greater than anticipated. The request would lower the OTSGlevel - low allowable value of Limiting Condition for Operation (LCO) 3.3.11, "Emergency Feedwater Initiation and Control (EFIC) System Instrumentation," from 11.1 inches to9.34 inches with the limiting trip setpoint 10.42 inches. In addition, an actuation time delay of10.4 seconds is also proposed. The licensee indicated the lower allowable value is intendedto alleviate its concern that operating at 100 percent power with the current allowable valueinvolves an increased risk of spurious emergency feedwater (EFW) initiation, and the actuation time delay is intended to minimize the possibility of inadvertent actuations during anticipatedtransients such as main feedwater (MFW) transients and main turbine trip.
2.0REGULATORY EVALUATION
In promulgating Section 50.36 of Title 10 of the Code of Federal Regulations (10 CFR), theNRC established its regulatory requirements related to the content of TSs. In doing so, theNRC emphasized those matters related to the prevention of accidents and mitigation ofconsequences of such accidents. As recorded in the Statements of Consideration accompanying the rule, "Technical Specifications for Facility Licenses; Safety Analysis Reports"(33 FR 18612, December 17, 1968), the NRC noted that licensees are expected to incorporateinto their plant TSs those items that are directly related to maintaining the integrity of thephysical barriers designed to contain radioactivity. Pursuant to 10 CFR 50.36, TSs are required to include items in five specific categories related to station operation. Specifically, those categories include: (1) safety limits, limiting safety system settings (LSSSs), and limiting controlsettings; (2) LCOs; (3) surveillance requirements (SRs); (4) design features; and (5) administrative controls. However, the rule does not specify the particular requirements to be included in a plant's TSs.Nuclear power plants are licensed to operate at a specified core thermal power. The instrumentmeasurement uncertainty should be considered to avoid exceeding the power level assumed in the design-basis transient and accident analysis. The safety-related instrument trip setpointsare calculated to ensure that sufficient allowance exists between the trip setpoint and the safetylimit to account for instrument uncertainties. The Commission's regulatory requirements related to this review can be found in 10 CFR 50.36(c)(1)(ii)(A) which requires that, where a limiting safety system setting (LSSS) is specified for a variable on which a safety limit has been placed,the setting must be so chosen that automatic protective action will correct the abnormalsituation before they exceed a safety limit. LSSSs are for automatic protective devices related to variables having significant safety functions. Setpoints found to exceed TS limits are considered a malfunction of an automatic safety system. Such an occurrence could challengethe integrity of the reactor core, reactor coolant pressure boundary (RCPB), containment, and associated safety systems. Regulatory Guide (RG) 1.105, Revision 3, "Setpoints for Safety-Related Instrumentation," is used to evaluate the conformance of setpoints with 10 CFR 50.36.
3.0TECHNICAL EVALUATION
3.1Description of the Proposed ChangesIn its application, the licensee proposed the following changes to TS 3.3.11, "EFICSystem Instrumentation:"1.Add two footnotes to Surveillance Requirements (SRs) 3.
3.11.2 and 3.3.11.3 onperforming the channel functional test and channel calibration, respectively.2.Revise the allowable value for the EFW initiation-SG level - low function in Table3.3.11-1, "Emergency Feedwater Initiation and Control System Instrumentation."3.Add two footnotes (c) and (d) to Table 3.3.11-1 to the allowable value for theEFW initiation-SG level - low function.The NRC staff's evaluation of these TS changes is given below:3.2Thermal Hydraulics Evaluation The EFW system is used to protect the core against the c onsequences of overheatingconditions upon a loss of MFW or a loss of the reactor coolant system (RCS) forced circulation. TS Table 3.3.11-1 specifies the EFIC system instrumentation for four plant conditions thattrigger the EFW system initiation as: loss of MFW pumps, Steam Generator level - low, OTSGpressure - low, and reactor coolant pumps (RCP) status (loss of all four RCPs). In addition, an Emergency Safeguard Actuation System actuation or a Diverse Reactor Overpressure Protection System/Anticipated Transient Without Scram Mitigation System Actuation Circuitry actuation will also initiate the EFW system. The licensee proposed to reduce the OTSG level -low instrumentation allowable value from 11.1 inches to 9.34 inches, to introduce other setpoint- related constraints and include a limiting setpoint of 10.42 inches and an actuation time delay of10.4 seconds, while the other actuation settings remain intact. Although many transients andaccidents analyzed in the ANO-1 Safety Analysis Report (SAR), Chapter 14, such as loss of alternating current power, loss of reactor coolant flow, and steam line break, require the EFW system for overpressure mitigation, these events result in plant conditions other than low OTSGlevel that actuate the EFW. The loss of feedwater (LOFW) transient is the only event thatresults in the actuation of the EFW due to low OTSG level, and is, therefore, the only transientaffected by a change to the OTSG level - low actuation setpoint. Since the OTSG level - low actuation setpoint is credited in the mitigation of the LOFW event,the licensee performed an analysis of the LOFW event to demonstrate that acceptance criteriaassociated with the LOFW are met with the revised OTSG level - low setpoint value and timedelay. This analysis is performed with the assumpti on that the EFW system will be actuatedwhen the OTSG level falls below 6 inches above the lower tube sheet plus a time delay of 10.4 seconds. This safety analysis actuation setpoint of 6 inches corresponds to the reduced allowable value of 9.34 inches and limiting as-left setpoint value of 10.42 inches (the evaluation of the instrument settings is addressed later in this section).The analysis of the LOFW event is performed with the RELAP5/MOD2-B&W [Babcock andWilcox] code and the methodology described in the NRC-approved topical report BAW-10193P-A, "RELAP5/MOD2-B&W for Safety Analysis of B&W-Designed Pressurized Water Reactors," dated January 2000. The analysis follows the guidance outlined in Appendix A of BAW-10193P-A, including the use of the "large detail" RELAP5/MOD2 plant model, and conservative initial and boundary conditions for the heatup transients. The large detail model is used for those transients such as LOFW in which the performance of the OTSG and/or secondary plant dominates the transients. Tables 1 and 2 in Attachment 1 to Entergy's January 3, 2006, letter list the input assumptions of initial and boundary conditions. These initial and boundary conditions are consistent with the guidance of BAW-10193P-A regarding the incorporation of uncertainties associated with initial power, reactor coolant pressure, temperature and flow, pressurizer level, and OTSG mass inventory, and boundary conditions regarding reactor trip setpoints, core decay heat, safety valve setpoints, reactivity coefficients, and EFW actuation setpoint and time delay, and the EFW flow rate. The analysis assumes 102 percent of rated thermal power plus the addition of reactor coolant pump heat, lower SG mass inventory at 62 percent operating range, high pressurizer level, pressurizer safety valve lift setpoint of 2500 pounds per square inch-gage (psig) plus 3 percent tolerance, moderator temperature coefficient (MTC) of
+1.3 x 10-5 k/k/oF. A failure of the motor-driven EFW pump isassumed, and the remaining EFW flow is conservatively assumed to be 500 gpm, versus the design flow of 700 gpm, equally split between the two OTSGs. The time delay for the EFW delivery to the OTSG is assumed to be 80 seconds to account for the EFW actuation delay time of 10.4 seconds pump start and time to reach full speed conditions and valve opening. No loss of offsite power is assumed consistent with the ANO-1 licensing basis.Figures 5 through 12 in Attachment 1 to Entergy's January 3, 2006 letter, depict the analysisresults, and Table 3 summarizes the LOFW event sequence. With the loss of MFW flow that ramps down at 7 seconds, resulting in boil-off of OTSG inventory and heat-up and pressurization of the primary side, the reactor trips at about 17 seconds due to high RCS pressure. After the reactor trip, the reactor power is reduced to the decay heat and the RCP heat. The RCS pressure continues to increase, resulting in the safety valve lifting in about 20 seconds. The OTSG secondary inventory continues to boil off, and at 83 seconds, the OTSG level reaches the EFW actuation setpoint of 6 inches. Assuming an 80 seconds delaytime for the EFW to reach the OTSGs, EFW starts to deliver 250 gallons per minute (gpm) to each OTSG at 183 seconds. This flow is initially not sufficient to remove the primary side heat until approximately 10 minutes when the decay heat has decreased sufficiently. Subsequently,the EFW is sufficient to remove the heat and pressure begins to decrease.
For the transient event, the peak RCS pressure is calculated to be 2704 pounds per square inch-absolute (psia), below the 100 percent design pressure of 2500 psig (i.e., 2750 psig). The analysis did not explicitly calculate the departure from nucleate boiling ratio (DNBR) for theLOFW transient. This is because the core thermal conditions during the transient do not approach the limiting conditions for departure from nucleate boiling (DNB). Prior to the reactortrip, the RCS temperature and pressure increase. At the time of reactor trip, the core inlet temperature has increased by less than 3 oF while the RCS pressure has increased more than200 pounds per square inch (psi), and the core power increases slightly due to a slightly positive MTC. However, the RCPs continue to maintain forced circulation flow through the core and keep the core in subcooled condition. The DNBR would decrease from the initial value but would still be above the DNBR limit. After the reactor trip at 17 seconds, the reactor powerdrops very quickly resulting in an increase in the DNBR. Therefore, the specified acceptable fuel design limit is met and there would be no fuel failure. Therefore, the safety analysis EFW OTSG level - low actuation setpoint of 6 inch and 10.4 seconds actuation time delay areacceptable.3.3Initial Control Transient Entergy was asked to confirm that the increased OTSG level control error upon initiation ofEFW would not adversely affect the operation of the system. The increased initial control errorwould be due to the increased difference between the (reduced) initiation setpoint and the (unchanged) control setpoint. Specifically, the licensee was asked to show that the controlovershoot magnitude and period, or the additional time required to approach the control setpoint, if the system is underdamped, does not compromise the conclusion that acceptablesystem behavior is maintained. The license replied that the OTSGs are expected to be emptyby the time EFW arrives, and that all of the EFW introduced into them would also be evaporated until after the RCS pressure and temperature had already begun to decrease.
Once the water level begins to recover and level control becomes important, the heat load from the reactor core will already have begun to diminish. Therefore, the greatest heat load to theOTSG will occur when the water level is zero. The OTSG does not depend upon a fixed EFWinventory to accomplish its safety function, and so fluctuations in water level due to level control considerations are not important to the successful operation of the OTSG under EFWconditions. The licensee also pointed out that under normal operation the EFIC level error signal is artificially biased, so that the integral function in the Proportional-Plus-Integralcontrollers are already saturated when a need for EFW occurs. As a result, the control response is already saturated, and the specific water level in existence at the time of controlinitiation has little influence over the initial control response. Therefore, the original concern expressed above is resolved. 3.4Setpoint-Related TS ConsiderationsA.Total Loop UncertaintyThe Total Loop Uncertainty calculation was provided in the licensee's January 6,2006, supplement. This calculation accounts for process-induced errors, inherent instrument errors, environmental effects, errors in test and measurement equipment, drift over the interval between setpoint measurements, and other effects. The calculation also derives the other quantities applied in the TS as discussed below.B.Limiting SetpointThe licensee indicates that the limiting setpoint (10.42 inches) is computed byadding the total loop uncertainty to the OTSG level assumed in the safety analysis for level-based initiation of EFW, with the addition of a small arbitrary safety margin.C.As-Found Setpoint Evaluation1)Deviation Limit The proposed TSs require that the as-found setpoint be within the acceptabletolerance band about the previous as-left setting. The associated bases establish the tolerance band as +/-1.08 inches. The 1.08 inch limit is derived in the uncertainty calculation as the combination of instrument reference accuracy, calibration tolerance (M&TE uncertainty), and limiting anticipated drift over the interval between tests.2)Allowable Value The licensee indicates that the allowable value is computed by means ofISA67.04, "Setpoints for Nuclear Safety-Related Instrumentation," Part 2,Method 3. The proposed TS require that the channel be declared inoperable if the as-found setpoint is nonconservative relative to the allowable value.3)As-Left Requirement The proposed TS requires that the setpoint be reset to a value not lessconservative than the established limiting setpoint.The licensee has proposed various footnotes to address the setpoint limits and the associatedcalculational methodology associated with SRs 3.3.11.2 and 3.3.11.3 and with the EFWinitiation-SG level - low function in Table 3.3.11-1. The NRC staff concludes that the setpointlimits and the methodology which has been used by the licensee to determine those limits willprovide reasonable assurance that the setpoint-related assumptions of the safety analysis willbe met. 3.5ConclusionsThe NRC staff has reviewed Entergy's proposed TS changes concerning allowable values andlimiting setpoints, and the notes proposed to be added to the TS relating to testing of instrumentation associated with the EFIC initiation function. The addition of the notes and otherTS changes provides reasonable assurance that the plant will operate in accordance with thedesign and licensing basis, and also that the operability of the instrumentation is ensured. TheNRC staff has evaluated the proposed TS values with respect to the thermal-hydraulicperformance of the OTSGs, the RCS, and the reactor fuel for the LOFW event. The LOFW isthe only licensing basis event potentially impacted by the proposed TS values. The evaluationby the NRC staff of RCS and OSTG performance using the new setpoint and time delay havebeen determined to be acceptable. In addition, core thermal limits are met. Therefore, the establishment of the new level setpoint limits and time delay has been evaluated and is acceptable to the NRC staff based on our review of the licensee's submittals and the addition ofthe TS notes.Based on (1) the proposed EFW OTSG level - low actuation setpoint of 6 inches and10.4 seconds actuation time delay is acceptable to assure no fuel failure (the proposed actuation setpoint limit of 10.42 inches with 10.4 seconds time delay provides reasonable assurance that the analytical limit of 6 inches will not be exceeded), (2) the original concern hasbeen resolved, and (3) the setpoint methodology assures that the LSSS shall be met, the NRCstaff concludes that the proposed TS changes meet 10 CFR 50.36. Based on this, the NRCstaff concludes that the proposed amendment is acceptable.4.0STATEMENT OF EMERGENCY CIRCUMSTANCES10 CFR 50.91(a)(5) states:Where the Commission finds that an emergency situation exists, in thatfailure to act in a timely way would result in derating or shutdown of a nuclear power plant, or in prevention of either resumption of operation or of increase in power output up to the plant's licensed power level, it may issue a license amendment involving no significant hazards consideration without prior notice and opportunity for a hearing or for public comment. In such a situation, the Commission will not publish a notice of proposeddetermination on no significant hazards consideration, but will publish anotice of issuance under § 2.106 of this chapter, providing for opportunity for a hearing and for public comment after issuance. The Commission expects its licensees to apply for license amendments in timely fashion. It will decline to dispense with notice and comment on the determination ofno significant hazards consideration if it determines that the licensee hasabused the emergency provision by failing to make timely application forthe amendment and thus itself creating the emergency. Whenever an emergency situation exists, a licensee requesting an amendment must explain why this emergency situation occurred and why it could not avoid this situation, and the Commission will assess the licensee's reasons forfailing to file an application sufficiently in advance of that event. The original OTSG orifice plates contained 3/4" gaps and were only adjustable between fullyclosed (0 Percent open) and 25 percent open. Operating at greater than 25 percent openresulted in excessive OTSG (EOTSG) instability. In an effort to provide more adjustability of the EOTSG orifice plates while maintaining flow stability, the design value of the radial gap between the plate outside diameter and shell inside diameter was reduced to 5/8". It wasalso desired to set the EOTSG orifice plates such that the operating range was similar tothat of the original OTSG. The OTSG operating ranges (unfouled) are on the order of 55 percent to 60 percent. The minimum required operating range is 50 percent, and the maximum inventory from flooding of the aspirator ports is 80 percent. Based on reducing the potential of being less than 50 percent or greater than 80 percent, a target operatingrange of 65 percent was selected. The 65 percent target value also provides calculated stability ratios that are improved from those of the original OTSG.The RELAP5 thermal hydraulic computer code was used to model both the original OTSGsand EOTSGs using plant startup and operating level data. Data taken at 98 percent powerduring startup after outage 1R19 agreed well with these predictions. However, RELAP5 predictions of the OTSG EFIC low level response are not as easily predictable. This is due to difficulties in computing localized void distributions in unheated (non-boiling) regions (i.e.,EFIC instrument span) rather than large heated regions that occur over the startup andoperating ranges.As a result, EFIC low level predictions for the EOTSG were performed on a comparativebasis. OTSG EFIC level data was adjusted by the difference in EOTSG and orifice plate resistance. This adjustment assumed design dimensions, for the orifice plates and shell dimensions, and also assumed that the OTSG orifice plate was fully closed. Based on a relatively small difference in orifice plate pressure losses, only about a 10-inch difference in level was predicted at 100 percent power versus the approximately 50 inches experienced.
Therefore, the reduced operating margin for the EFIC low level trip was not able to be readily foreseen. Because the instrument inaccuracy was unexpected, the licensee could not have anticipated the need for a license amendment that would allow for a 30-daycomment period. Therefore, the licensee concluded that an emergency TS change is appropriate to avoidcontinued derate of ANO-1 and to allow for an expeditious resumption of power to 100 percent. Based on the above, Entergy believes the conditions for an emergency TS change are met. Additionally, the proposed amendment involves no significant hazards as the licensee provided.On the basis of the above discussion, the NRC staff has determined that, in accordance with10 CFR 50.91(a)(5), emergency circumstances exist and that the licensee used its best efforts to make a timely application and could not avoid the emergency situation.
5.0FINAL NO SIGNIFICANT HAZARDS CONSIDERATION
DETERMINATIONThe Commission's regulations in 10 CFR 50.92(c) state that the Commission may make afinal determination that a license amendment involves no significant hazards consideration ifoperation of the facility in accordance with the amendment would not: (1)Involve a significant increase in the probability or consequences of an accidentpreviously evaluated; or(2)Create the possibility of a new or different kind of accident from any previouslyevaluated; or(3)Involve a significant reduction in a margin of safety.
The following analysis was provided by the licensee in its January 3, 2006, letter.1.Does the proposed change involve a significant increase in the probability orconsequences of an accident previously evaluated?Response: No.
The allowable value for actuation of the EFIC system is not an accidentinitiator and, therefore, cannot increase the probability of an accident. TheEFIC and associated Emergency Feedwater (EFW) systems are componentscredited to mitigate the consequences of an accident. However, the small reduction in the SG Level - Low allowable value in conjunction with the addition of an actuation time delay still affords ample volume in the SGs[steam generators] to remove decay heat in a timely manner from the Reactor Coolant System (RCS) following a design[-]basis accident described in the ANO-1 Safety Analysis Report (SAR). The revised allowable values for both the SG Level- Low and the delay time will continue to enable the EFWsystem to maintain plant parameters within SAR limits for the previously evaluated accidents. The analysis results do not impact the dose consequences of any accidents previously analyzed.Therefore, the proposed change does not involve a significant increase in theprobability or consequences of an accident previously evaluated.2.Does the proposed change create the possibility of a new or different kind ofaccident from any accident previously evaluated?Response: No.
The proposed change to the TSs does not require any physical alteration tothe plant or alter plant design, other than the slight reduction in SG Level -Low allowable value and the addition of an actuation time delay, as associated with EFIC actuation. The proposed change does not present a significant adverse impact on the EFIC function or EFW systems as credited in any safety analyses for the prevention or mitigation of any accident. The proposed change is associated with mitigating systems and the changecannot, in itself, initiate an accident of any type.Therefore, the proposed change does not create the possibility of a new ordifferent kind of accident from any previously evaluated. 3.Does the proposed change involve a significant reduction in a margin ofsafety?Response: No.
The proposed change does not significantly impact a margin of safety for anyaccident previously evaluated. Based on the revised safety analysis, the proposed change in EFIC low level initiation and the addition of an EFW actuation delay time will still assure adequate margin for EFW actuationunder a Loss of Feedwater event, but will minimize inadvertent EFWactuation due to other plant transients. Therefore, the proposed change does not involve a significant reduction in amargin of safety.The NRC staff has reviewed the licensee's analysis and, based on this review, hasconcluded that operation of the facility in accordance with the proposed amendmentsatisfies the three standards of 10 CFR 50.92(c). Therefore, the NRC staff determines thatthe proposed amendment involves no significant hazards consideration.
6.0STATE CONSULTATION
In accordance with the Commission's regulations, the Arkansas State official was notified ofthe proposed issuance of the amendment. The State official had no comments.
7.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to installation or use of a facilitycomponent located within the restricted area as defined in 10 CFR Part 20. The NRC staffhas determined that the amendment involves no significant increase in the amounts, and nosignificant change in the types, of any effluents that may be released offsite, and that thereis no significant increase in individual or cumulative occupational radiation exposure. The Commission has made a final no significant hazards finding with respect to this amendment.
Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
8.0CONCLUSION
The Commission has concluded, based on the considerations discussed above, that: (1) the amendment does not (a) involve a significant increase in the probability orconsequences of an accident previously evaluated, or (b) create the possibility of a new ordifferent kind of accident from any previously evaluated, or (c) involve a significant reduction in a margin of safety and, therefore, the amendment does not involve a significant hazardsconsideration; (2) there is reasonable assurance that the health and safety of the public willnot be endangered by operation in the proposed manner, (3) such activities will be conducted in compliance with the Commission's regulations, and (4) the issuance of theamendment will not be inimical to the common defense and security or to the health andsafety of the public.Principal Contributors: Yi-Hsiung Hsii P. Rebstock
Date: January 13, 2006 Arkansas Nuclear One cc:
Senior Vice President & Chief Operating Officer Entergy Operations, Inc.
P. O. Box 31995 Jackson, MS 39286-1995Director, Division of Radiation Control and Emergency Management Arkansas Department of Health 4815 West Markham Street, Slot 30 Little Rock, AR 72205-3867Winston & Strawn1700 K Street, N.W.
Washington, DC 20006-3817Senior Resident InspectorU.S. Nuclear Regulatory Commission
P. O. Box 310 London, AR 72847Regional Administrator, Region IV U.S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-8064County Judge of Pope County Pope County Courthouse Russellville, AR 72801Vice President, Operations SupportEntergy Operations, Inc.
P. O. Box 31995 Jackson, MS 39286-1995Wise, Carter, Child & Caraway P. O. Box 651 Jackson, MS 39205