ML20212G638
| ML20212G638 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 02/25/1999 |
| From: | ENTERGY OPERATIONS, INC. |
| To: | |
| Shared Package | |
| ML20212G627 | List: |
| References | |
| NUDOCS 9909300063 | |
| Download: ML20212G638 (400) | |
Text
{{#Wiki_filter:,. , ARKANSAS NUCLEAR ONE-UNIT 2 I DOCKETNO. 50-368 LICENSE NO. NPF-6 1 10CFR50.59 REPORT FOR 1999 This report contains a brief description of changes in procedures and in the facility as described in the Safety Analysis Report (SAR), tests and experiments conducted which were not described in the SAR, and other changes to the SAR for which a safety analysis was conducted. The report also contains the safety evaluation for each change. This report is applicable for the period from April 11,1998 through February 25,1999. The safety evaluations included in this report were performed in accordance with 10CFR50.59 and determined that none of the changes involved an unreviewed safety question. i i I i 9909300063 990923 PDR ADOCK 05000368 R PDR
i Initiatian Dec. Descrintion 1 CALC 91E011601 Revision to Net Positive Suction Head Available to the High Pressure Safety Insection and Containment Spray Pumps ; 2 CALC %E002',01 Steam Generator Tube Plugging Limits for Reactor Coolant System Flow Variations 3 CALC 973976E201 60PA Equivalent Replacement 4 CALC 97R201802 ANO-2 Cycle 14 Reload Analysis Report 5 CALC 97R201803 Changes to the COLR to Support Cycle 14 Operation 6 CR l-96-0199 Change to Penetration Protection Wording 7 CR 2-94 0374 Emergency FeedwaterFlow Reduction 8 CR2 95-0434 Air Maintenaw Device Air Supply C-ion 9 CR 2-96-0078 Fuel Pool Cooling System Design Basis Discussion ) 10 CR 2-98-0168 Removal of Fire Bamer Penetration Seal Details i 11 CR 2 99 0073 Breaker Size Correction for 2D2224 12 DCP 946012D201 Containment Vent Header / Waste Gas System Modifications , 13 DCP 962006D202 Contamment Penetration Building Electncal Penetration Upgrade 14 DCP %3230D201 Replacement of Original Main Condenser Tube Bundles With Shop Fabricated Titanium Bundles 15 DCP 903242D201 Battery Charger Replacement 16 DCP %3254D201 In-Mast Fuel Sipping Modifications 17 DCP %3523D202 Emergency Feedwater System Valve Actuator Replacement l 18 DCP 963559D301 C ,=Pr and T{*-7 = Rooms Power and AC Upgrade 19 DCP 973950D201 PW-t of the NaOH Addition System with Three Trisodium Phosphate Raakets 20 DCP 975015D201 Reduction of the Steam Generator Low Pressure Trip Setpoint as Approved in Amendment 189 to the Technical Specifications 21 DCP 975015D201 Evaluation of the Pfinwee of Rapid Depressurization of the Steam Generators on level Instrumentation l l I
i laitiatina Dac. Descrintion 22 ~ DCP 975015D201 Incorporation of Reactor Coolant System and Main Steam Isolation System Technical Specificauon Changes 23 DRN 9840641 Plant Heaung System Valve Alignment 24 DRN 9843021 Chemical Addition and Pandensate Punp Recirrmlantan Valve Positions 2$ DRN 9843022 Nitrogen Addition System Valve Positions at Full Power 26 . DRN 9f,43023 Main Steam Valve Positions at Full Power
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27 ER 9.i3555E204 Removal of the Reactor Building Equipenent Hatch Missile Shield Blocks 28 ER 974550E201 Corression of Quality Assurance Summary Imel Q-List 29 ER 980302E108 High Presame Iq)ection Full Flow Test Gage Inmilatian 30 ER 980579E201 Setpoint Revision for the Circulatmg Water Pump T , --e Recorder 31 ER 980898E201 Use of the Alternate AC Generitor as a Peaking Unit 32 ER 981035E201 Palaentian of Emergency Light Battenes 2EL-71 and 2EL-72 33 ER 980706E101 Main Tutune Quarterly Valve Stroke Test Deferral 34 ER 991335E201 mapime=nant of the Plant Hosting Boaler Burner 35 LCP 958078L201 Sodium Hydroxide Tank Imel Transmitter Repl=essnent ! 36 LCP M2018L201 Addition of Spectacle Flanges to Plant Heatmg System 37 LCP M320$L201 Removal of Emergency Diesel GeLarator Exhaust Rain Hoods ' 38 LCP 974379L201 Resolution of Vent / Drain Line Code C ,4== Issues 39 LDCR Turbane Building Exhaust Flow Rate Correcuan ! 40 LDCR P ade% Liquid Release with Less than Two Circulating Water Pumps in Operation , i 41- LDCR Removal of a Vent Valve in the Liqued Rad Waste and Boron ; Management Systems l 42 IDCR Post Racirculanon Actuasum SystoniImv Presome Safety Iqjection Pump P,' --+ for Operanon 43 LDCR Reduced Circulaung Water Pump Flow 44 LDCR Clan 6cahon of Feedwater Control System Description
i Initiatina Dec. Description 45 LDCR Revision to Peak Cladding Temperature and Maximum Cladding
. Oxidation for the Limiting Large Break IDCA 46 LDCR Transfer Informatian from QAMO to SAR '47 NC 980366N201 Removal of the Pacidag Imakoff Lines from Main Feedwater Isolation Valves CV-1024-1 and 2CV-1074-1 48 PC 958007P201 Stator IAak Monitonag System Installation 49 PC 95-8033 I:ytsezi.g AnalyzerRF at 50 PC 962029P201 Reptar=nant of Main Feedwater Recirwlation Valve Operators and Controllers 51 PC %3203P201 Installatian of Additional Spent Fuel Poollevel Indicators 52 PC 963219P201 Alternate Power Supply for 2XL43 and 43LA 53 PC 96-8016 Medincations to the Pneumatic Control Circuit of the Service Water Squeeze Valve 54 PC 973636P201 Plant Heatsag System Chemical Addition Pot Installation 55 - PC 973673P201 DiNorentsal Pressure Isthentian for Service Water Basket Strainers 56 PC 973744P201 Revision of the Method ofI= ' ;- '=2 Test Verification of the Shunt Trip Coil in the Reactor Trip Switchgear Circuit 57 PC 973786N201 Replacement of the High Pressure and Iow Pressure Tube Bundles h. the Moisture Separator Pahanters 58 PC 973905N201 Removal of the Main Turbine khacir Circuit 59 PC 973932P202 Relacaenan of ANO-2 Outage Control Center 60 PC 973%7P301 Alternate Reliable Power for VSF-9 61 PC 974062P201 Reactor Coolant System Ietdown Backpressure Valve Signal Drivers 62 PC 974326P201 External Limit Switch Removal for 2CV-5859-2 and 2CV-5852-2 63 PC 974343P201 Addition of Flange Pair to the One Inch Diameter Piping on Each Side of E-MyFeedwaterSuctaonReliefValve 2PSV0706.
64 PC 974346P201 ('antainment Spray Rehef Valve Flange Addition 65 PC 974369P201 Motor Operated Valve Limit Switch Configuration Change L
f jalllstlgg]gt. Descrintion 66 PC 974820P201 Spent Fuel Pool High Temperature Alann Setpoint Change 67 PC 974899P201 Service Water Pipe Pg'
-^
68 PC 974904P201 Main Chill Water Compresmon/Frpanelan Tank Vent Valve Addition 69 PC 974991N202 Steam Generator Blowdown Filtration System Inmallation 70 . PC 975054P201 Reactor Coolant System Refueling 14 vel hbing Modification 71 PC 975109P201 Stator Water Cochng Conductivity Detector Replama*=* and PS Snubber Inspitatian 72 PC 980184P201 Valve Modificanian in the Steam Generator and Feedwater Chemical Feed System 73 PC 980243P201 Reactor Coolant System Isolation Valve 2PS-162 Rapl===*=# 74 PC 980628P201 Parla-aaat of a Drain Line Manifold on the B E-gf eedwater F Line 75 PC 980704N201 Installation of a Partial Flow Condensate Filter in Senes with the Existing Startup and Blowdown Dominerahzers 76 PC 98088301 Installaelan of Diesel Fuel Storage Vault Security Door and Full Access Hardware 77 PEAR 95 0082 "B" Feedwater Sodium Analyzer Valve Configuration 78 PROC 1000.043 Incorporation of Revision 4 to the EPRI PWR Sw./ y Water Cl=nistry Gnutelinen 79 PROC 1000.043 Padar*ian in Required Frequency of Steam Generator Silica Analysis 80 PROC 1000.152 Fire Pump Suction Requirements 81 PROC :005.002 New Fuel Handling and Shipping Design Imads 82 PROC 1402.133 Operation of the Spent Fuel Crane, L-3 83 - PROC 2102.002 Alignment of Reactor t'aal=nt Pump's Controlled Bleed Off Lines 84 PROC 2102.004 Increased T-hot to C , -- - for Steam Generator Tube Plugging 85 PROC 2104.008 - CirculaCag Water Pumps and Motor Valve ConAguration 86 PROC 2104.018 Removal of Valve 2BM-1035 Downstream of the Boric Acid Concentrator Rupture Disc
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87 PROC 2104.019 Clean Resia Transfer Valve Lineup Change ;
i Initiatina Doc. Description 68 PROC 2104.032 Incorporation ofInformation Obtained from Remntly Performed Valve Lineups 89 PROC 2104.036 Alternate Flow Path for Fuel Oil to the Emergency Diesel Generator Day Tank 90 PROC 2104.036 Installation of a Bypass Valve Around the E. , y Diesel Generator Fuel Oil DayTankInlet Valve 91 PROC 2106.014 Alignment of Domestic Water System Valve 2DW-174 92 PROC 2106.015 Revision to the Normal Position of Condensate 'Ausfer Valve 2CT-21 93 PROC 2106.024 Startup & Blowdown Demineralizer System Valve Position Change (2BD-1003 & 2BD 1004) 94 PROC 2106.024 Startup Blowdown Demineralizer System Valye Position Change (2BD-45A & 2BD-45B) 95 PROC 2304.165 Calibration Values and Instructional Steps Required to Perform Calibration for Newly Installed Radioactive Radwaue System Instrumentation 96 PROC 2305.057 Guidance to Detennine the Leak Rate Associated with Certain ESF Minimum Recimulation isolation MOVs 97 PROC 2311.002 Design Service Water Flow Requirement for 2VUC5B 98 PROC 2311.002 Radioactive Liquid Release with Less than Two Circulating Water Pumps in Operation l 1 99 PROC 2401.131 Installation of Temporary FME Carr While PSVs Are Removed 100 PROC 2402.214 Raising RCS Levels During N9000 RCP Shaft Seal Cartridge Removal 101 PROC 2628.012 Operation of Copper Corrosion Inhibitor System at Cooling Tom.r 102 TAP 97-2 010 Temporary Pressunzer Level Setpoint 103 TAP 97-2-018 Temporary Alteration to the Cooling Tower Acid Pump Suction 104 TAP 97-2 023 Installation of Temporary Bypass Circuit to Control Room Ammeter 105 TAP 97-2 024 2CV-1051 Instrument Air Root Valve Bypass 106 TAP 97-2 027 Use of Unit 2 Corulensate Storage Water for Cleaning of Unit l's T-12 Tanks
E i Initiatian Doc. Descrintion 107 TAP 98-2 001 Temporary Isolation of Dameanic Water Supply to 2P3A/B Gland Cooling 108 TAP 98-2402 Use of Unit 2 Condan==** Storage Water as a Cleaning / Wash Water Supply for Cleaning the Unit 1 T-12 Tanks 109 TAP 98 2 003 Startup Channel #1 Single Fission Chamber Configuration implesmentatinn 110 TAP 98-2-004 Blind Flange for Turbine Gland Scaling System Relief Valve 111 TAP 98-2405. 2CV-1074-1 Packing Laak Off Line Isolation 112 TAP 98-2-011 Power Source to a Temporary Power Panel I4cated Beside each Moisture Separator Reheater 113 TS BASES 3/4 3.1 Clarincation of the Plant Protection System Matnx logic Surveillance Requirements Related to Tri annual Channel Functional Testing 114 WP 2409.583 Instrucuons for Back Flushing the Normal Pressurizer Spray Lines 115 WP 2409.588 Instrucuons for Freezing the inlet to 2PSV-4878 for Removal, Repair, and peinsennaeum 116 WP 2409.597 Collection of Data on the Condeaanse and Steam Generator Blowdown for Selection ofFilterMedia i l j l 1 l 1 l
I ARKANSAS NUCLEAR ONE FORM TITLE: PaDe 1 FORM NO, REV. 10CFR50.59 DETERMINATION 1000.131 A 3 PC-1 This Document contains 3 Pages. Document No. 91 E-0116-01 Rev1 Change No. 2 Title Minimum Containment water level during recirculation to provide adequate NPSH. Brief description of proposed change: This calculation does not reflect any physical changes to the plant. The NPSH available to the HPSI and CS pumps as determined by the calculation is changed by revision 2 due to a new determination of minimum su water level and by consideration of the effect of insulation on the sump screen. Will the proposed Activity: j 1. Require a change to the Operating License including: I 1 Technical Specifications (excluding the bases)? YesO NoS Operating License? YesO No@ Confirmatory Orders? YesO No@ 2. Result in information in the following SAR documents (including drawings and text) being (a) no longer true or accurate, or (b) violate a requirement stated in the document: SAR (multi-volume set for each unit)? Yes@ NoO Core Operating Limits Report? YesO No@ Fire Hazards Analysis? YesO No@ Bases of the Technical Specifications? ! YesO No@ Technical Requirements Manual? YesO No@ NRC Safety Evaluation Reports? Yes@ nod
- 3. Involve a test or experiment not described in the SAR?
(See Attachment 2 for guidance) YesO nod
- 4. Result in a potential impact to the environment? (Complete Environmental Impact Determination of this form.)
YesO No@ 5. Result in the need for a Radiological Safety Evaluation per section 6.1.57 YesO No@
- 6. Result in any potential impact to the equipment or facilities utilized for Ventilated Storage Cask activities per Section 6.1.67 YesO No@
- 7. Involve a change under 10CFR50.54 for the following SAR documents per Section 6.1.77 QAMO?
YesO No@ E-Plan? YesO No@
ARKANSAS NUCLEAR ONE Page 2 FORM TITLE: FORM NO. REV. ) 10CFR50.59 DETERMINATION 1000.131 A 3 PC-1 Document No. 91-E-0116-01 RevlChange No. 2 Basis for Determination (Questions 1,2, & 3): Calculation 91-E-0116-01 has been revised. Revision 2 changes the NPSH margin which is mentioned in section 7.3. The calculation revision is based on a minimum containment sump water level at elevation 342.17', a total flow of 4025 gpm per train, a density of 59.347 lb/ft , and the NPSHA for the HPSI pumps is approximately 18 ft, requiring a change to section 6.3.2.14. In addition table 6.2-18 will have to be changed to indicate the minimum NPSHA for the CS purnps of 17.95 ft. This calculation does not reflect any change to the operation or function of any component or system. No tech spec is affected. Section 6.3.4 of Supplement 2 to the original SER is now untrue (the margin between required and available NPSH is no longer 1.7 feet, it is now v., proximately 0.6 feet). O Proposed change does not require 10CFR50.59 Evaluation per Attachment 1, item # (if checked, note appropriate item #, send LDCR to Licensing). Search Scope: f List sections reviewed in the Licensing Basis Documents specified in questions 1,2 and 3. If a keyword search was done on LRS, "all" may be entered under "Section" with the keyword (s) used in parentheses. Controlled hard copies of the documents shall be reviewed (LRS is not verified and searches only text, not figures or drawings). Attach and distribute a completed LDCR per Section 6.1.2 if LBD changes are required. Document Section LRS; All. keywords NPSH. net positive suction head. nosh available. noshr. nosh reouired. HPSI. hioh pressure safety iniection. CSS. containment sorav system. sum.,_p. containment sumo. reactor buildino sumo MANUAL SECTIONS: Unit 2 SAR sections 7.3. 6.2. 6.3 Unit 2 T.S. sections 3/4.5.2. 3/4.5.3. 3/4.6.2 FIGURES: E_~ John Richardson 12/17/97 Ce ied Reviewers Signature Printed Name Date Reviewers certification expiration date: 5/31/98 Assistance provided by: Printed Name Scope of Assistance Date Sesrch Scope Review Acceptability (NA, if performed by Technical Reviewer per 1000.006) eO. Nr ae/ 0 Jerry W. Howell 1/6/98 Certg.egers Signature Printed Name Date
l 1 ARKANSAS Wei. EAR Cf'i Page 3 FORM TITLE: FORM NO. REV. 10CFR50.59 DETERMINATIUN 1000.131 A 3
\
ENVIRONMENTAL IMPACT DETERMINATION (UNIT 1 and UNIT 2) Document No. 91 E-0116-01 Rev1 Change No. 2 Complete the following Determination. If the answer to any item below is "Yes", an Environmental Evalu required. See Section 6.1.4 for additional guidance. Will the Activity being evaluated: Yes No O B Disturb land that is beyond that initially disturbed during construction (i.e., new construction of buildings creation or removal of ponds, or other terrestrialimpact)? See Unit 2 SAR Figu 2.5-17. This applies only to areas outside the protected area. O B Increase thermal discharges to lake or atmosphere? O 2 increase concentration of chemicals to cooling lake or atmosphere through discharge canal or tower? O B increase quantity of chemicals to cooling lake or atmosphere through discharge canal or tower? O E Modify the design or operation of cooling tower which will change drift characteristics? O B install any new transmission lines leading offsite? O @ Change the design or operation of the intake er discharge structures? O E Discharges any chemicals new or different from that previously discharged? O S Potentially cause a spill or unevaluated discharge which may effect neighboring soils, surface water or ground water? l j O E involve burying or placement of any solid wastes in the site area which may effect runoff, surface water or ground water? O M Involve incineration or disposal of any potentially hazardous materials on the ANO site? O @ Result in a change to nonradiological effluents or licensed reactor power level? O E Potentially change the type or increase the amount of non-radiological air emissions from the ANO site.
_ ARKANSAS NUCLEAR ONE Pace 1 FORM UTLE: FORM NO. REV. 10CFR50.59 SAFETY EVALUATION 1000.131B 3 This Document contains 1 Page. Document No. 91-E-0116-01 Rev1 Change No. 2 10CFR50.59 Eval. No. FF Al@CM ; (Assigned by PSC) Title Minimum Containment water level during recirculation to provide adequate NPSH. A WRITTEN RESPONSE PROVIDING THE BASIS FOR THE ANSWER TO EACH QUESTION MUST BE ATTACHED. EACH QUESTION MUST BE ANSWERED SEPARATELY. A SIMPLE STATEMENT OF CONCI.USION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES GUIDANCE FOR RESPONSE. If the answer to any question on this form is "Yes," then an unreviewed safety question is involved. If the answer to all questions is "No,' then the proposed change does not involve an unreviewed safety question.
- 1. Will the probability of an accident previously evaluated in the SAR be increased? YesO No @
- 2. Will the consequences of an accident previously evaluated in the SAR be increased? Yes O No @
- 3. Will the probability of a malfunction of equipment important to safety be increased?
YesO No @
- 4. Will the consequences of a malfunction of equipment important to safety be increased?
Yes O No @
- 5. Will the possibility of an accident of a different type than any previously evaluated in the SAR be created? YesO No @
- 6. Will the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated i
in the SAR be created? Yes O No @
- 7. Will the margin of safety as defined in the bases for any technical specification be reduced? Yes O No @
ef_ . 4Cer'tified' Reviewer's Signature sohn R,cha,eson 12,17,97 Printed Name Date Reviewer's certification expiration date: 5/31/98 Assistance provided by:
)
Printed Name Scope of Assistance Date P PSC review by: U xS- Date: I 7.~) 81V
ARKANSAS NUCLEAR ONE Pace i FORM TITLE: FORM NO. REV.
.10CFR60.89 REVIEW CONTINUATION PAGE 1000.131C 3 l
. Document No. 91-E-0116-01 RevdChange No. _2 10CFR50.59 Review Continuation Pace i
1. Will the probability of an accident previously evaluated in the SAR be increased? No. No modification is represented by this revision. Corrections were made to the NPSH calculation which revealed a decreased NPSH available to the building spray and HPSI pumps. There are no chang to modes of operation, performance characteristics or requirements, or operating procedures. 2. Will the consequences of an accident previously evaluated in the SAR be increased? No. This revision will not change the way in which the HPSI and CS systems respond under accident conditions (or any other conditions). The NPSHA was determined to be adequate before, and it is still adequate. 3. Will the probability of a malfunction of equipment important to safety be increased?
. No. This calculation revision does not represent any physical change to the plant. The Margin between required and available NPSH is revealed to be less than previously thought, however, there is still a positive margin even under conditions which would exist with the most conservative assumptions for level, flow and sump b:ockage.
4. Will the consequences of a malfuntion of equipment important to safety be increased? No. The decreased NPSH margin revealed by this calculation will not change the operating characteristics of the pump or system. The HPSI and CS systems will respond in the same way as before, except that there is less room for modifications or procedural changes which would increase flow resistance or result in lower sump levels.
' 8.
Will the possibility of an accident of a different type than any previously evaluated in the SAR be created?
^ No. The HPSI and CS systems will be operated the same as before and the system response is not changed. While the NPSH available has been reduced, there is still adequate margin for NPSH and the difference in pump discharge pressure (<1 psi ) is not significant enough to make any difference in the characteristics of the system or in its interaction with other systems.
- 8. )
Will the possibility of a malfunction of equipment important to safety of a different type thta any previously evaluated in the SAR be created? ; t No. The calculation assumes minimum sump water level and pump runout flow. Even under these conditions, there is adequate NPSH available to provide the proper pump suction conditions. Pump . performance is not affected by a larger NPSH margin. Provided that the margin is positive, the pumps can be expected to perform as designed.
- 7. Will the margin of safety as defined in the bases for any technical specification be reduced?
No. The technical specifications require that the HPSI and CS systems be " operable". No margin of safety is defined which will be impaired by reduced NPSH margin, as long as MPSHA exceeds NPSHR.
ARKANSAS NUCLEAR ONE FORM TITLE: Page 1 FORM NO. REV. 10CFR50.59 DETERMINATION 1000.131A 3 PC-1 This Document contains 4 Pages. Document No. __ CRN 99-007 to 96-E-0025-01 Rev1 Change No. O Title SG Tube Plugging Limits for RCS Flow Variations Brief description of proposed change: This change extends total tube plugging and tube plugging asymmetry limits for RCS core in out to 30% and 1000 tubes respectively, to match the limitations of other analysis input assump Will the proposed Activity: 1. Require a change to the Operating License including: Technical Specifications (excluding the bases)? YesO No@ Operating License? YesO No@ Confirmatory Orders? YesO No@ 2. Result in information in the following SAR documents (including drawings and text) being (a) no longer true or accurate, or (b) violate a requirement stated in the document: SAR (multi-volume set for each unit)? Yes@ nod Core Operating Limits Report? YesO No@ Fire Hazards Analysis? YesO No@ Bases of the Technical Specifications? YesO No@ Technical Requirements Manual? YesO No@ NRC Safety Evaluation Reports? YesO No@
- 3. Involve a test or experiment not described in the SAR?
(See Attachment 2 for guidance) YesO No@
- 4. Result in a potentialimpact to the environment? (Complete Environmental Impact Determination of this form.)
YesO No@ 5. Result in the need for a Radiological Safety Evaluation per section 6.1.57 YesO No@
- 6. Result in any potential impact to the equipment or facilities utilized for Ventilated Storage Cask activities per Section 6.1.6?
YesO No@
- 7. Involve a change under 10CFR50.54 for the following SAR documents per Section 6.1.7?
QAMO? YesO No@ E-Plan? YesO No@
ARKANSAS NUCLEAR ONE FORM TITLE: . Page 2 FORM NO. REV. 10CFR50.59 DETERMINATION 1000.131 A 3 PC-1,2 Document No. CRN 99-007 to 96-E-0025-01 Rev1 Change No. _0 Basis for Determination (Questions 1,2, & 3): See Attached O Proposed change does not require 10CFR50.59 Evaluation per Attachment 1, item # (if checked, note appropriate item #, send LDCR to Licensing). Search Scope: List sections reviewed in the Licensing Basis Documents specified in questions 1,2 and 3. If s performed on LRS, the LRS search index should be entered under *Section" with the se parentheses. Controlled hard copies of the documents shall be reviewed (LRS is not ve text, not figures or drawings). Attach and distribute a completed LDCR per Section 6.1 required. Document Section LRS: All (" asym
- w/25 plug *")
MANUAL SECTIONS: SAR Section 4.4 (all subsections) FIGURES: _ / Bryan Dalber Cert!fiedAevie'wers Signature Printed Name l M* I7 Date Reviewers certification expiration date: _3-18 2000 Assistance provided by: Printed Name Jacque Lingenfelter Scope of Assistance Date LRS and SAR search. draft determination 1/19/99 Search Scope Review Acceptability (NA, if performed by Technical Reviewer per 1000.006)
% :e) u). fod Certified Reviewers Signature to20-11 Printed Name Date 4
I J FORM TITLE: ARKANSAS NUCLEAR ONE Page 3 10CFR50.59 DETERMINATION FORM No. REV. 1000.131A 3 ENVIRONMENTAL IMPACT DETERMINATION (UNIT 1 and UNIT 2) Document No. _ CRN 99-007 to 96-E-0025-01 RevlChange No. O required, See Section 6.1.4 for additional guidance. Complete the uation is follo Will the Activity being evaluated: I Y.af No O 2 buildings, creation or removal of ponds, or oth 2.5-17. This applies only to areas outside the protected area. O E increase thermal discharges to lake or atmosphere? ! O O Increase tower? concentration of chemicals to cooling lake or atmosphere thro O @ increase tower? quantity of chemicals to cooling lake or atmosphere through disc O S Modify the design or operation of cooling tower which will change drift charac ' O E instati any new transmission lines leading oftsite? O E . 1 Change the design or operation of the intake or discharge structures? l O E Discharges any chemicals new or different from that previously discharged? O E Potentially water or groundcause water? a spill or unevaluated discharge which may effect neigh O 2 ! involve surface burying water orwater? or ground piacement of any solid wastes in the site area which mI 0 B Involve incineration or disposal of any potentia!!y hazardous materials on the l O E Result in a change to nonradiological effluents or licensed reactor power level? O E Potentially ANO site. change the type or increase the amount of non-radiological air em
ARKANSAS NUCLEAR ONE F41RM TITLE: Page 4 FORM NO. REV. l 10CFR50.59 REVIEW CONTINUATION PA'2E 1000.131C 3 I Document No. CRN 99-007 to 96-E-0025-01 RevJChange No. 0 10CFR50.59 Review Continuation Paae Backaround The extensive plugging of ANO-2 SG tubes led to a comprehensive review of the bases. This effort resulted in Technical Specification Amendments 189 and 190 whic MSIS trip setpoint and reduced the required RCS flow to 90% of its original value the I.DCR supporting these TS amendments (LDCR Tracking Number 2-4.4-1 and 10CFR Number FFN-98-097) included several changes that were not included in the TS ame One such change added limits to steam generator tube plugging based on analyses the core inlet. While most aspects of the safety analyses restrictive, especially for 3 pump operating conditions. These limitations were based on flow variations performed by ABS-CE using the SYSFLOW code, and documented in 0025-01. addresses reactor vessei flow distribution but had not previo These tube plugging limits have been revised by CRN 99-007 to 96-E-0025-01, based o obtained from the steam generator replacement effort. The new information provid how tube plugging and plugging asymmetry affects core inlet flow variations, such th can be predicted with substantially higher tube plugging limits. The allowable core inlet flow va themselves are unchanged, thus maintaining the assumptions used in safety analyses code. The new limits now conform with the 30% and 1000 tube asymmetry limits us , proposed SAR change modifies the previous change to reflect these new limits. Basis for Determination , The specific limitations on tube plugging with respect to core inlet flow distribution licensing basis documentation until the 30% tube plugging teanalysis efforts leading to Amendments 189 and 190 were completed. The SAR changes supporting those TS amend plugging limitations to SAR section 4.4.2.7.1 that were more restrictive than those applie section is the only licensing documentation that addresses reactor vessel flow distribution in of detail. Consequently, these limits were not used anywhere else and this change af document. l l
FORM UTLE: ARKANSAS NUCLEAR ONE Pago 1 FORM NO. REV. 10CFR50.f.9 SAFETY EVALUATION 1000.131B 3 PC.2 This Document contains 3 Pages. Document No. _ CRN 99-007 to 96-E-0025-01 RevlChange No. 0 10CFR50.59 Eval. No. FEN 46.J. Title _ SG Tube Plugging Umits for RCS Flow Variations (Assigned by PSC) ATTACHED. EACH QUESTION MUST BE AN CONCLUSION IS NOT SUFFICIENT ATTACHMENT 2 PROVIDES to all questions is "No," then the proposed change does . 1. Will the probability of an accident previously evaluated in the SAR be increased? YesO No @ 2. Will the consequences of an accident previously evaluated in the SAR be increased? YesO No @ 3. Willthe probability of a malfunction of equipment important to safety be ! increased? YesO No @ { 4. Will the consequences of a malfunction of equipment important to { safety be increased? YesO No @ 5. Willthe possibility of an accident of a different type than any previously YesO evaluated in the SAR be created? No @ 6. Will the possibility of a malfunction of equipment important to safety of a different type than any previous'y evaluated in the SAR be created? YesO No @
- 7. Will the margin of safety as defined in the basis for any technical specification be reduced? YesO No @
_ /
- Bryan Dalber WrtifiedTieviewer's Signature Printed Name
)- 10-77 Date Reviewer's certification expiration date:
3-18-2000 l Assistance provided by: Printed Name Jacque Lingenfelter Scope of Assistance Draft Evaluation Date 1/19/99 PSC review by: 3 '. c h-- Date: 0 b h .S
FORM TITLE: ARKANSAS NUCLEAR oNE Pact 2
' FORM NO. REV.
10CFR$0.89 REVIEW CC'NTINUATION PA2E 1000.131C 3
. Document No. CRN 99-007 to 96-E-00*!5-01 Rev/ Change No. 0 IQQUR50.59 Review Continuation Paae A summary of the changes being covered by this evaluation are delineated in the to the determination for more information with respect to the changes being mad information.
1. Will the probability of an accident previously evaluated in the SAR be increased? No to core inlet flow variations listed in the SAR. There are , These tube plugging limits, which are now consistent wit aspects of the plant design and licensing bases, are in no way related to any acciden
.the modification of these limits will have no impact on the probability of an accident p SAR.
2. Will the consequences of an accident previously evaluated in the SAR No be increase The proposed changes increase the limits on tube plugging and tube pluggin thus maintaining the assumptions used in the safety an , substructures, design changes physical alterations, or operating procedure , this change. proposed changes. The consequence,s of accidents previously evaluated in the SAR 3.
' Will the probability of a malfunction of equipment important to safety No be increased?
The proposed changes increase the limits on tube plugging and tube pluggin thus maintaining the assumptions used in the safety an , substructures, design changes, physical alterations, or operating procedure c , this change. These tube plugging limits are now consistent with the tube plugging l aspects of the plant design and licensing bases, wNch have been separately evaluated and documented. increased.' Consequently, the probability of a malfunction of equipment importa 4. Will be the consequences increased? of a malfunction of equipment important to safety No The proposed changes increase the limits on tube plugging and tube plugging ' thus maintaining the assumptions used in the safety ana , substructures, design changes, physical alterations, or operating procedure ch , this change. . The changes support an increase in tube plugging limits to be consiste plugging limits that have been evaluated for accident analysis impact in separate 10 The acceptable results of the accident analyses, which consider all appropriate single fa .
~malfunction cr different operating conditions for equipment importa of equipment important to safety will not be increased.
ARKANSAS NUCLEAR ONE FORM TITLE: Psot 3 FORM NO. REV. 10CFR50.59 REVIEW CONTINUATION PACE 1000,131C 3 5. Will the possibility of an accident of a different type than any previously evaluated in the SAR be created? No The proposed changes increase the limits on tube plugging and tube plugging a to core inlet flow variations listed in the SAR but do not change the allowable flow variat thus maintaining the assumptions used in the safety analyses. The changes support a plugging limits to be consistent with other tube plugging limits that have been evaluated which consider all appropriate single failures (malfunctio demonstrate that these plugging limits produce no new or different operating cond this change; therefore no new accidents are created a more limiting. 6. Will the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR be created? No There are no new systems, components, substructures, physical design changes, nor operating procedure changes being proposed by this change. The proposed changes inc limits on tube plugging and tube plugging asymmetry with respect to core inlet flow varia the SAR but do not change the allowable flow variations themselves, thus maintai used in the safety analyses. The changes support an increase in tube plugging limits to b with other tube plugging limits that have been evaluated for accident analysis impact in i 10CFR50.59 Reviews. As there are no physical changes to the plant, the possibility of a equipment created. important to safety of a different type than any previously evaluated in the SAR 7. Will the margin of safety as defined in the bases for any technical specification be reduced? No The proposed changes increase the limits on tube plugging and tube plugging asy to core inlet fiow variations listed in the SAR but do not change the allowable flow var thus maintaining the assumptions used in the safety analyses. The changes support an in plugging limits to be consistent with other tube plugging limits that have been evaluated fo analysis impact in separate 10CFR50.59 Reviews. There are no new systems, componen substructures, this change. design changes, physical alterations, or operating procedure change The limits on tube plugging and tube plugging asymmetry with respect to core inlet flow varia not mentioned in any technical specification or technical specification bases. The margi defined by the bases for the technical specifications are unaffected by these changes. 1 l l j
ARKANSAS NUCLEAR ONE FORM TITLE. FORM NO. REV. 10CFR50.88 DETERMINATION 1906.131A 2 PC s.3 Page 1of Document No. ER 973976E201 Rev1 Change No. 1 Title GOPA Eaulvalent Ree'ee= ment Willthe proposed Activity:
- 1. Require a change to the Operating License induding:
Technical Specifications (excluding the bases)? YesO NoS Operating License? YesO NoS Confinnatory Orders? YesO NoE
- 2. Result in information in the following SAR documents (including drawings and text) being (a) no longer true or accurate, or (b) violate a requirement stated in the document:
Core Operating Limits Report YesO NoS SAR (multi-volume set for each unit)? YesE NoO QAMO?* YesO NoS E-Plan?* YesO NoS FHA YesO NoS Bases of the Technical Specifications? YesO NoS NRC Safety Evaluation Reports? YesO NoS
- 3. Involve a test or experiment not described in the SAR?
YesO NoS
- 4. Result in a potential impact to the environment? (Complete Environmental impact Checklist of this form.) YesO NoS
- 5. Result in the need for a Radiological Safety Evaluation per section 6.2.4.A? YesO No@ l
- 6. Result in any potential impact to the equipment or facilities utilized for Ventilated Storage Cask activities per Section 6.2.4.87 YesO No@
Basis for Determination: ER 973976E201 replace" GOPA alone with it's internal fused discennect switches (60PA-1. 2. & 3) with
^"
- r'vi-.:. r -- --- ; he==d. fumad d!seennect switches (2824. 5. & 8) and a 36" Isas 8" x 4" 4,;.;.;;i.
The criminal: ' .. a; and it's eauivalent ree!ecament provides falced iso!=-:-a to 2CV-1205. 8 & 9 inside the cooline tower. SAR Floure 8.3-83. semen D. dee!& an :::vstiori view of the sene: arrance.. a; adiacent to the cooline tower irie!ad!na SOPA. Since the ree!ers.en; eau;s.T.ent consists of four seoarate housinos (one for =ech of the d!seennects and one for the wirewavt the existina SAR fleure is no lonaer accurate and will be chanced to reflect the new canel arraneament. Changes to these documents requ!re an evaluation in accordance with 10CFR50.54. See Section 6.2.1.B. P. 7
AfuUWsAs NUCLEAR ONE FORM TITLE. FORM NO REV. 10CRIAS.89 DET5tMMATioN 1000,131A 2 PC.2.3 Page 2 of Document No. 60 PA Eaulvalent Ree!scement Rev1 Change No. A
References:
List sections reviewed in the Licensing Basis Documents, specified in questions 1,2 and 3. If a keyword search was done on LRS, "all" may be entered under "Section" with the keyword (s) used in parentheses. Controlled hard copies of the documents shall be rev; awed as computer based searches such as LRS are not controlled and search text only, not figures or drawings. Attach a completed LDCR if LBD changes are required. Document Section 80.59 Unit 2 &l1 60PA. 2CV-1205. 2CV-1208. 2CV-1209. Coolina Tower and De-lce. De lc* Steven L. Smith 4/17/97 CerHfied Reviewers Signature Printed Name Date Reviewer's certification expiration date: 3/8/99 Assistance provided by: Printed Name Scope of Assistance Date 1 R8
f i ARKANSAS NUCLEAR ONE f0RM TITLE. FORM NO. REV. 10CPR88.sl DETERMINATION 1eet.131A 2 PC.2.3 Pagelof ENVIRONMENTAL IMPACT CHECKLIST (UNIT 1 and UNIT 2) Document No. ER973976E201 RevlChange No. 1 Complete the following checklist. If the answer to any checklist item is "Yes", an Environmental Evaluation is required. See Section 6.2.1.E for additional guidance. Willthe Activity being evaluated: ! Yes N_o O E Disturb land that is beyond that initially disturbed during construction (i.e., new construction of buildings, creation or rernovel of ponds, or other terrestrialimpact)? See Unit 2 SAR Figure 2.5-17. This applies only to areas outside the protected area. O E increase thermal discharges to lake or atmosphere? O E increase concentration of chemicais to cooiing iake or atmosphere through discharge canai or tower? O E increase quantity of chemicais to cooling take or atmosphere through discharge canal or tower? I O E Modify the design or operation of cooling tower which will change drift characteristics? O E instati any new transmission lines leading offsite? O E Change the design or operation of the intake or discharge structures? O E Discharges any chemicais new or different from that previously discharged? O E Potentially cause a spill or unevaluated discharge which may effect neighboring soils, surface water or ground water? O E involve burying or piecement of any solid wastes in the site area which may effect runoff, surface water or ground water? O E involve incineration ordisposal of any potentially hazardous materials on the ANO site? O E Result in a change to nonradiological effluents or licensed reactor power level? O E Potentially change the type or increase the amount of non-radiological air emissions from the ANO site.
)
1
s ARKANSAS NUCLEAR ONE FORM TITLE. FORM NO. REV. 10CFR50.59 EVALUATION 13es.mg 3 Pageiof 10CFR50.59 Eval. No FFW-97-054 (Assigned by PSC) Document No. ER 973976E201 Rev/ Change No. 1 Title GOPA Eauivalent Re=8ece neset l A WRITTEN RESPONSE PROVIDING THE BASIS FOR THE ANSWER TO EACH QUESTION MUST BE ATTACHED. EACH QUESTION MUST BE ANSWERED SEPARATELY. A SIMPLE STATEMENT OF CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES GUIDANCE FOR RESPONSE. If the answer to any question on this form is "Yes." then an unreviewed safety question is involved. If the answer to all questions is "No." then the proposed change does not involve an unreviewed safety question.
- 1. Will the probability of an accident previous!y evaluated in the SAR be increased?
Yes O No @ The roolacement of SOPA with eauivalent. separatelv hee ==d. d!?cann=ct= and a w .e.;;; will not increase the orahah!!"v of an sc-:! den; previous!v evaluated in the ??R b=cEcee it will not desi.de the orie:ne ' ==ls5 h =le of the coe!!na tower csei=issa or lEd!cstion. 8: ace the eau lv-:;..; i '::__.. C of SOPA fuE-^!sse exactiv as the original and ca:V the essensen 5; of the ves; sus d!Ecssa.;;e and cer ;6 w... ways and conde"s will be chessed. the .. .=_ . .... will have no r
.;.;t on the sie t"tw of any ecE!d:M sa.;susiv evaluated in the ???..
- 2. Will the consequences of an accident previously evaluated in the SAR be increased?
Yes No S The rentacament of SOPA with =ae8 valent. 7 ser ^ 's her::1 d;;-:saa-d and a w ia ;; will not
;acr --
the cer _ : == of an EEE;-i .; si;;; sus!v ;;;;aated in the !?R because it will not cesseG. the orioinal d=e!en hae!? of the caa!ine ;c;;.i sse =isc-5 or lEd!- E^'Es. 8: ace the eaulv;;.M i~s'- : i- % of SOPA fun-3;sse exactiv as the orlainal and only the .....-__.._... of the various G;essi.6ect. and ceitsin w;ie ;se and conduits will be chanced. the sissceed ect Atv will have no
.G.ct on the cer.: .:-c:a;ee of any sce! dant Lievisusiv evaluated in the SAR.
- 3. Will the probability of a malfunction of equipment important to safety be increased?
Yes O No @ The ses!Ec:n_: of SOPA with eau lv;:.M. separately houeed. dis- sseects and a w;re.sev will not increase the si-i-as;;!!i of a malfunden of eau l i.ea :.c:-:-r.eni to safety t+ =e== lt will not
":_.0. the si e:. I d=e;-:-n basis of the coe!ine tower operetion or indication. S;Gce the eaulvalent see;es...
of SOPA fuEct-c-se exactiv as the orlainal and only the erie 6Geinedt of the various gisconnects and cgite;n w....;.e; and conde"= will be chana-d. the siesc:;- ec;;4:v will have no effect on the LiebELility of a malfunction of eeu;i,.aent linecitent to safety.
- 4. Will the consequences of a malfunction of equipment important to safety be increased?
Yes O No @ The teolacement of 60PA with eaulv;;.at. essei;^.:. V hoessd. disccr.a=c'= and a w;i.;; V will not in..-___ the coneemes;=+5 of a malfunction of eauisin.nt important to safety ber?e== it will not d+-iece the ci;s;nel dee;ss basis of the coolino tower operation or indicstion. Since the eauiv;;;al replacement of SOPA functly; exactiv as the orioinal and only the arranoement of the various
- f. l O
ARKANSAS NUCLEAR ONE FORM TITLE. FORM NO. REV. 1eCFReeAs aVALUATIO* 100s.1313 3 disconnects and certain wireways and ennduits will be chanced. the proposed activity will have no effect on the ce.x_:::r.ces of a malfunction of =aalament important to safety.
- 5. Will the possibility of an accident of a different type than any previously evaluated in the SAR be created? Yes O No E The see!Ecm..; of 60PA with aae!v;;; .t. ME+__le housed. d;Ecs .. se*= and a wireway will not create the possibility of an accident of a different tvoo than any previeeshr evaluated in the SAR because it will not dearade the orlainal desian h==i= of the cooline tower oper=*ien or indie=*len.
Since the eauivalent roolacement of 60PA functions e==e'iv as the oriainal and only the arranaement of the various disconnects and certain wireways and conduits will be chanced. the proposed activity will have no effect on the nessibility of an see! dent of a different tvos than oftviousiv evaluated in the SAR.
- 6. Will the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR be created? Yes O No E The replacement of 60PA with eaulvalent. aE+__:4 hoe ==d. disconnae*= and a wireway will not
^
create the possibility of a malfunction of souloment important to safety of a different tvos than any previousiv evaluated in the SAR because it will not dearade the orlainal desian h==i= of the coolina tower operation or indicatio i. Since the eaulvalent replacement of 60PA functions a-=e+!v as the orlainal and only the arranaement of the various disconnects and certain wireways and conduits will be chanced, the proposed activity will have no effect on the possibility of a malfunction of eauipment important to safety of a different tvos than any previously eve!9a*=d in the f.*R.
- 7. Will the margin of safety as defined in the bases for any technical specification be reduced? Yes O No E The replacement of 60PA with coulvaient. separately heuea*8. disconnects and a wireway will not reduce the maroin of safety as defined in the bases for s'iv technical speciflCation because it will not dearade the orlainal desian haele of the coolina tnwer oper=*ien or indie=* ion. Since ttg ecuivalent replacement of SOPA functions exactiv as the oriainal and only the arranaement of the various disconnects and certain wirewevs and conduits will be chanced. the oraaa==** activity will have no effect on the maroin of safety as defined in the be==s for any technical saacWication.
s Steven L Smith 4/18/97 Certlfied Reviewers SigTiature Printed Name Date Reviewers certification expiration date: 3/5/99 Assistance provided by: Printed Name Scope of Assistance Date PSC review by: Dh/ M N Date: [IM f II l
aruvutena nuva.can une FORM TITLE: -
. FORM NO. REV 10CFRSS.60 DETERMINATION 1000.131A 3 .*C 1 Page 1 of.23 ' Document No. 97-R 2018-02 RevdChange No. 9 Title ANO-2 Cycle 14 Reload Analysis Rooort Brief description of proposed change:
Implement reauired channes to the LBDs to sunoort the Cycle 14 operation documented in the Re!end An:!veis Ressit bed"lenal desc.r;--1;en is alven in the followino o&GGE.1 ' Will the proposed Activity:
- 1. Require a change to the Operating License including:
- Technical Specifications (excluding the bases)?
YesO No@ Operating License? YesO No@ Confirmatory Orders? YesO No@ 2. Result in information in the following SAR documents (including drawings and text) being
. (a) no longer true or accurate, or (b) violate a requirement stated in the document:
SAR (multi-volume set for each unit)? Yes@ NoO g Core Operating Limits Report Yes@ NoO Fire Hazards Analysis? YesO No@ Bases of the Technical Specifications? YesO No@ Technical Requirements Manual? - YesO No@ NRC Safety Evaluation Reports? YesO No@
- 3. Involve a test or experiment not described in the SAR? .
(See Attachment 2 for guidance) YesO NoS
- 4. Result in a potential impact to the environment? (Complete the Environmentalimpact Determination of this form.)
YesO No@
- 5. Result in the need for a Radiological Safety Evaluation per section 6.1.5?
YesO No@
- 6. Result in any potentialimpact to the equipment or facilities utilized for Ventilated Storage Cask activities per Section 6.1.67 YesO No@
- 7. Involve a change under 10CFR50.54 for the following SAR documents per Section 6.1.7:
QAMO?' YesO No@ E-Plan? YesO No@ I J
ARKANSAS NUCLEAR ONE FORM TITLE: FORM NO. REV. 10CFRS0JS DETERMINATION 1000.131A 3 PC.1,8 L
\
Page 2 of.22 Document No. 97-R-2018-02 Rev1 Change No. 9 Basis for ti"_ r..lre'-:-r (na_e=*!ons 1. 2 & 31: See attached discussion. O Proposed change does not require 10 CFR 50.59 Evaluation per Attachment 1, item # ,(if checked, note appropriate item #, send LDCR to Licensing). Search Scope: List sections reviewed in tha Licensing Basis Documents specified in Question 1,2 and 3. If a search was I performed on LRS, the LRS search indet sitould be entered under "Section* with the search statement (s)I used i parentheses. Controlled hard copies o' the documents shall be reviewed (LRS is not verified and searches only text, not figures or drawings). Attach and distribute a completed LDCR per Section 8.* ' If LSD changes are required. l Document Section LRS: l 80.69 - ANO-2 ALL (See attached for keywords) MANUAL SECTIONS: s See attached See attached FIGURES: : See attached See attached - Me - -N E "rdd John T. Sankoorikal 12/9/98 Celtified Reviewers signature Printed Name Date Reviewers certification expiration date: 8/19/2000 Assistance provided by: Printed Name Scope of Assistance Thomas L Lotz Date Core Design - Neutronics 12/9/98 Larry Hu Core Design - Mechanical Lori Ann Potts 12/9/98 Thermal Hydraulics 12/9/98 Robert W. Clark ' Safety Analysis, CEA Ejection, Boron Dilution Larry D. Young 12/9/98 Safety Analysis, MSLB 12/9/98 Tim J. Rush Safety Analysis, Seited Rotor Jacque Lingenfelter 12/9/98 ECCS 12/9/98 Search Scope Review Acceptability (NA, if performed by Technical Review per 1000.006)
'I M, NA Daniel W. Fouts 12/9/98 Certified Reviewers Signature Printed Name Date
FORM MLE: 10cFR60.89 DETERMINATION g , ENVIRONMENTAL IMPACT DETERMINATION (UNIT 1 and UNIT 2) Document No. 97-R-2018-02 RevlChange No. 0 ) Complete the following Determination. If the answer to any checklist item is "Yes", an Environmental Evaluation is required. See Section 6.1.4 for additional guidance. Will the Activity being evaluated: Yes g O @ Disturb land that is beyond that initially disturbed during construction (i.e., new construction of buildings, creation or removal of ponds, or other terrestrialimpact)? See Unit 2 SAR Figure 2.5-17. This applies only to areas outside the protected area. O E Increase thermal discharges to lake or atmosphere? O E Increase concentration of chemicals to cooling lake or atmosphere through discharge cana' or tower? O S Increase quantity of chemicals to cooling lake or atmosphere through discharge canal or tower? O @ Modify the design or operation of cooling tower which will change drift characteristics? O E install any new transmission lines leading offsite? O E Change the design or operation of the intake or discharge structures? O E Discharges any chemicals new or different from that previously discharged? O E Potentially cause a spill or unevaluated discharge which may effect neighboring soils, surface water or ground water? O @ involve burying or placernent of any solid wastes in the site area which may effect runoff, surface water or ground water? O E Involve incineration or disposal of any potentially hazardous materials on the ANO site? O 2 Result in a change to nonradiological effluents or licensed reactor power level? O E Potentially change the type or increase the amount of non-radiological air emissions from the ANO site.
ARKANSAS NUCLEAR ONE FORM TITLE: FORM No. REV. 10CFRS0A0 REMW CONTINUATION PAGE 1000.131C 3 Document No. 97-R-2018-02, Rev 0, h = tion 4/23 10CFR50.59 Review Continuation Paae Document Section LRS: 50.59, ANO-2 ALL (reload *, Batch *, Cycle *, fuel, fuel w/10 mechanical, grid, plenum, max
- w/10 burnup, poison, gadolin*, gd2o3, gd7o*,
power w/10 peaking, critical boron, CEA, ejection, withdrawal, boron, dilution, MTC, " moderator temperature coefficient", COLR, COLSS, LHR, PLHR, " linear heat rate", PDIL, " seized rotor", seized, seizure, " fuel failure", " thermal hydraulic", " thermal hydraulic?", thermal-hydraulic, shim, shims, DNBR, MSLB, SLB, Steam Line Break. "4.5-1", "4.3-29") MANUAL SECTIONS: ANO-2 TS and Bases 2.1.1, 3/4.2, 3.2.1/4.2.1, 3.2.4/4.2.4, 3.2.5/4.2.5, 3/4.9.12, 6.9.5.1. ANO-2 SAR 15.1, 15.1.4, 15.1.20, 1.5.2, 4.1, 4.2, 4.2.1.2.3, 4.3, 4.3.2,4.3.3.1, 4.4, 4.4.1, 4.4.2, 4.4.3, 4.5, 4.6, 4.7, 15.1.5.2.1.2, 15.1.5.2.2.2, 15.1.5.2.3, 15.1.5.3.2, 15.1.14.1, 15.1.20, and associated Tables. ANO-2 COLR ALL (Additional manual search information is attached to the end of this Determ FIGURES: I ANO-2 SAR All Chapter 3, 4, 5, 7, 9, and 15 Figures i l SUPPLEMENTAL INFORMATION The ANO-2 Cycle 14 Reload Analysis Report describes tlie Cycle 14 core and provides an evaluation of the design and performance of the core during Cycle 14. The imnact of the new core design on the licensing analysis that qualifies the ANO-2 core is also evaluated. The ANO-2 licensing Basis Documents (LBDs) were reviewed with respect to the new core design and analyses changes to determine the revisions required to the LBDs. The '
' changes that affect the LBDs are discussed below.
l.
AemAN6AS NUCLEAR ONE
, . FORM No. REtf.
10CFRSS.se REVIEW CONTINUATION PAGE 1000.131C 3 Document No. 97-R-201s-02, Rev 0. DG " tion 5/23 ANO-2 CYCLE 14 RELOAD CORE MECHANICAL AND NEUTRONIC DESIGN CHANGES Cycle 14 Core Neutronics and Fuel Management The Cycle 14 core consists of 80 fresh Batch S assemblies and 97 previously irradiated assemblies from fuel Batches M (1 assembly), P (20 assemblies), and R (76 assemblies) which are utilized in a "very low leakage" fuel management scheme. The Cycle 14 core has been configured to provide longer cycle energy capability than the previous cycle. The Cycle 14 Reload Analysis Report supports operation to 557 Effective Full Power Days (EFPD) at the Cycle 13 long endpoint (548 EFPD) and to 572 EFPD at the Cycle 13 short endpoint (519 EFPD). Cycle 13 was licensed to 532 EFPD at the Cycle 12 long endpoint and to 558 EFPD at the Cycle 12 short endpoint. The Cycle 13 endpoint is predicted to be in the above licensed window (about 535 EFPD) based on the planned shutdown date and coastdown scenario. The Cycle 14 analysis window supports burnup beyond expected full power capability at 0 ppm critical boron concentration, which is 546 EFPD at the short Cycle 13 endpoint and 531 EFPD at the long endpoint. Batch S fuel employs gadolinia as an integral burnable poison. The Batch S integral poison pellets contain 6.0 weight percent 2 Ud 03 and 94.0 weight percent UO . The uranium in 2 the Batch S integral poison rods is enriched to 2.50 weight percent Uranium 235. The Cycle 13 (Batch R) loaded assemblies utilized the same integral poison pellets design, but with a 2.30 weight percent Uranium 235 carrier enrichment. Both the Batch R and Batch S reload designs are consistent with and meet the requirements for Gadolinia-Urania core design methods (Reference 9). All fuel assemblies loaded prior to Cycle 13 have utilized Al2O3 -B4C pins as neutron absorbers. Integral burnable poison permits the loading of more fuel material per bundle into the reactor core, which results in a lower core average linear power density. Gadolinia-poisoned cores are characterized by higher intra-asse power peaking and more pronounced core average axial power peaks Gadolinia has been widely utilized in both U.S. NRC licensed BWR and PWR cores. The ABB-Combustion Engineering neutronics methods are U.S. NRC licensed for gadolinia-poisoned core (Reference 9).' 1 The non-gadolinia fuel pins in the Cycle 14 fresh fuel are ofinitial enrichments 4.50 and i '~ 4.00 weight percent Uranium-235. The Cycle 13 Batch R non-gadolinia fuel pins are of initial enrichments 4.40 and 3.90 weight percent uranium-235. The increased enrichments employed in Cycle 14 are permitted by ANO-2 Technical Specification 3.9.12, which was implemented by Amendment 178 to the ANO-2 operating license (Reference 10). 4-Power peaking in the Cycle 14 core design is similar to the previous cycle. The power distribution behavior versus burnup of the Cycle 14 Batch S new fuel assemblies is similar to the Cycle 13 Batch R new fuel assemblies. The Cycle 14 BOC peak Fxy is 1.55 vith the
}
k ARKANSAS NUCLEAR ONE FORM TITLE: FORM No. REE i 10CFR40.49 REVIEW COhmNuATION PAGE 1000.131C 3 Document No. 97.R-2018-02, Rev 0, Deternunation 6/23 maximum Cycle 14 Fxy of1.59 occurring at MOC which is similar to the Cycle 13 BOC peak Fxy of 1.54 with the maximum Cycle 13 Fxy of1.58 occurring at MOC. The Cycle 14 maximum BOC assembly relative power density is 1.33 is similar to the Cycle 13 maximum BOC assembly relative power density of 1.32. The Cycle 14 Axial Shape Index (AS1)is shown to be bounded by the Cycle 13 ASI results. Hence, the Cycle 14 axial power shapes will be less challenging to the core operating limits than the Cycle 13 core operation. The neutronic parameters important to safety were calculated for Cycle 14 and any differences from those of Cycle 13 were noted and the appropriate reference cycle analyse were also noted. All applicable coefficier,ts, kinetic parameters, power peaking factors, rod burnups, rod wonhs, and boron worths were generated using U.S. NRC approved codes and methods appropriate for the ANO-2 Cycle 14 core. Extensive analyses of the reactivity performance of the ANO-2 Cycle 14 core were performed. The physics data, which are supplied as input to the Cycle 14 safety r,nd fuel
. performance analyses, were generated using U.S. NRC approved methodology appropriate for ANO-2 Cycle 14. The neutronics input to the safety analyses includes rod worths, . power peaking factors, kinetics parameters, pin census data, soluble boron concentrations and worths, and reactivity coefficients. All Cycle 14 neutronic and fuel management parameters have been evaluated and found to be either bounded by the existing analysis of record (AOR) or a Cycle 14 specific analysis performed for the Batch S reload assemblies and the Cycle 14 core design. The input parameters for Steam Line Break (SLB) and Control Element Assembly (CEA) ejection events were not determined to be clearly bounding. The SLB event was reevaluated with bounding input parameters for Cycle 14.
The final Cycle 14 SLB results were determined to be bounded by the Analysis ofRecord (AOR) from Cycle 13. The CEA ejection event required reevaluation due to a more positive Cycle 14 moderator temperature coefficient (MTC). The Cycle 14 MTC was used to define a new Core Operating Limits Report (COLR) upper MTC limit (Reference 11). The final Cycle 14 CEA ejection results were determined to be bounded by the Analysis of Record (AOR) from Cycle 13. The Cycle 14 Core Operating Limits Report (Reference 11) section of the ANO-2 Technical Specifications requires a minimum shutdown margin of 5.0%Ap for Modes 1-5, which has not changed from the required shutdown margin in Cycle 13. The shutdown margin, assuming the most reactive control rod stuck out, was calculated for Cycle 14 and was found to exceed the required COLR limits. Tae power dependent control rod insertion limits (PDIL) have been established, in part, to ensure'that adequate shutdown margin exists throughout the cycle. The PDIL portion of the COLR has been updated for Cycle 14. The only change is the removal cf the prohibition on simultaneous insenion ofBanks 6 and P above 80% power except during a downpower.
AsuuumAS NUCLkAM ONt: FORM TITLE: . FORM No. REV. 10CFRSS.88 REVIEW CoNTihuaTioN PAGE 1000.131C 3 M-t No. 97-R-2018 02, Rev 0, Detenahation 7/23 Approved methodologies were utilized to verify that reactivity coefficients and shutdown margin meet the Cycle 14 COLR requirements in modes 1 and 2. Note that the reactivi balance update provides the means to satisfy the surveillance requirements on shutdown margin. Data covering the full range of operating conditions is generated for the reacti balance update. The COLR upper moderator temperature coefficient (MTC) for Cycle 14 is slightly more positive than the limit in the Cycle 13 COLR. Calculations ofMTC throughout Cycle 14 at various power levels demonstrate compliance of the Cycle 14 core design with the new Cycle 14 MTC COLR requirements (Reference 11). Additionally, the applicability of the - current licensing AORs has been assessed for ANO-2 Cycle 14 and have shown that they are still applicable and bound'mg. The Cycle 14 projected maximum integrated fuel rod burnup for Cycle 14 is 58,726 mwd /MTU, including an uncertainty of 750 mwd /MTU. This value is less than the 60,000 mwd /MTU licensed maximum fuel rod expos >2re limit imposed by the Operating Limit condition (See ABB-CE topical report, Reference 7, which is NRC approved). The impact of Cycle 14 fuel on the fuel storage criticality calculations was assessed. The current fresh and spent fuel criticality analyses were verified to be applicable to Batch S fresh fuel and to all fuel previously loaded at ANO-2. These analyses verify that compliance with Tech Spec 3.9.12 can be maintained and that, e The new fuel storage configuration is sufficient to maintain a k.gless than 0.98 with _ optimum moderation and 0.95 under normal conditions, and The calculated k.a for the spent fuel racks is less than 0.95 under all conditions. e The calculated k.a for the upender and the containment temporary storage racks is less than 0.95 under all conditions. Consequently, the Cycle 14 fresh fuel can be safely stored without restriction in the new fuel storage rack, the containment temporary storage rack, and the upender. In accordance with Tech Spec 3.9.12, storage ofBatch S fresh fuel in Region 1 of the spent fuel rack requires cross-hatch fuel spacing; in Region 2 of the spent fuel rack, two-of-four spacing in'a checkerboard configuration is required for storage offresh Batch S fuel. The effects of the minor design changes to the Batch S new fuel assembly, which are detailed in the next section, do not affect the applicability of the criticality analyses. The as-built variations in enrichment and stack height density are considered by the evaluation performed in Reference 5.
6 ARKANSAS NUCLEAR ONE FORM TITLE: l FORM No. RElf. 10CFRSS.s0 REVEW CohmNuATioN PAGE 1000.131c 3 Document No. 97-R-2018 02,Rev 0 D J=Gon 8/23 Reload Fuel Assembly Design Changes With the exceptions discussed below, the mechanical design of the Cycle 13 Batch R new fuel and Cycle 14 Batch S new fuel bundle assemblies is identical. The mechanical bases have not changed since the initial core loading fuel design. Two minor mechat%al design changes are implemented for the Batch S grid cages. These changes are made mainly to improve fuel vendor's component manufacturing process in order to 1) reduce the frequency ofcomponent rework, yielding a more consistent prod and 2) permit use of standardized components for all ABB 16x16 fuel assembly designs These design changes are listed below. Modification of the design for thejoint between the guide tubes and lower end fitting The materials, thread size, bolt head, lower end fitting counterbore and through-hole, and locking disk are unchanged. The bolt shaft length has been reduced by 0.425", an the length of the female threads in the guide tube plug has been reduced by 0.406". The overall length of the guide tube assembly has been retained by increasing the tube length to compensate for the shorter plug. The change to the assembly geometry d result in a very small downward shift of the dashpot and cooling hole elevations in the outer guide tubes. Modiscation of the upper end fitting outer and center post heads. The combination of a thick chrome plating requirement and the geometric features of the Batch R-type post heads causes a wide range of thicknesses to be deposited on the post heads) plating. As a result of the heavy localized chrome plating, grinding and re-plating of high percentage ofposts was iequired to meet design requirements. Therefore, to reduce this rework, in Datch S the chrome pieting thickness is now 0.0005" to 0.0025" thick instead of the 0.002 to 0.0045 thickness range for the Batch R, thus returning
- the same range that was speci6ed in Batches A through M. To further improve the plating process, the sharp corners on the corners have been blended.
The diameter of the centerbore at the base of the center post has been increased by 0.0 to reduce the hand grinding in this area after the post is welded to the flow plate. This change will also ease the remote fitup between the upper end fitting and the center g tube should reconstitution become necessary. These changes have been reviewed and approved by Entergy for implementation st with the Batch S new fuel (Reference 12). The review concluded that the ov assembly mechanical, thermal-hydraulic, and nuclear performances have not been altered by any of these design changes. These design improvements do not require any modifications to existing design criteria or methodologies and the ANO-2 Technical Specifications are not impacted. I ___ -----_-------J
l N WCLEAR ONE FORM TITLEr
. FORM No. REV.
10CFRSS.89 REVIEW CotmNuaTIoN PAGE 1000.131C 3 Docanent N197-R-2018-02, Rev 0.DG= d on 9/23 i The thermal perfonnance of composite fuel rods that envelope the fuel rods of the fuel batches present in ANO-2 Cycle 14 have been evaluated using the FATES 3B version the C-E evaluation model, and the licensed C-E methodology for the core conta gadolinia burnable absorber. The analysis was performed using a power history tha enveloped the power and burnup levels representative of the peak fuel rod for each batch
- each burnup interval, from begimJng of cycle to end ofcycle burnups. The burnup analyzed is in excess of that expected at the end of this cycle.
As was performed for Cycle 13, additional fuel performance analyses were performed fo Cycle 14 to show that the gad rods are bounded by the Urania rods with respect to rod internal pressure, fuel centerline temperature, and power-to-melt criteria. Predicted maximum rod internal pressures are less than the nominal operating system pressure o 2200 psia. The cold intemal rod pressure was calculated to remain below the NRC
. Regulatory Guide 1.25 value of 1200 psig (Reference 13).
The metallurgical requirements of the fuel cladding and the fuel assembly components f the Batch S fuel are the same as those used in Cycle 13. Since there will be no chang core operating environment and the duty cycle, the chemical or metallurgical performance of the Batch S fuel will be similar to the Batch R fuel. To date the Batch R fue assemblies have performed satisfactorily. It should be noted that the overall envelopes of the fuel rod assembly, the poison rod assembly and the fuel bundle assembly all remain unchanged from the previous reloads. The
' fuel assembly mass is also unchanged, which is a bounding value based on as-built of fuel components and assumes all 236 rod locations are occupied by UO2 fuel rods (Reference 1). The HID-IL Zircaloy spacer grids are the same for Batch S new fuel as was used for Batch R. Adequate shoulder gap is predicted for all of the Batches offuel in Cyc
- 14. The CEA guide tube assembly is also unchanged, with the exception of the very small downward shift of the dashpot and cooling hole elevations in the outer guide tubes
. mentioned above.
During the fabrication ofBatch S fuel assemblies, a previously unrecognized quality issue was discovered. Some of the laser welded spacer grids were found to have welds contaminated with stainless steel constituent elements. ABB has since concluded, with Entergy's concurrence (Reference 14) that 1) this does not constitute a 10CFR21 reportable condition for previously delivered ANO-2 fuel,2) this does not present r. safety concern for the Batch S fuel and current fuel in core to satisfy their safety-related function
- during seismic /LOCA accident conditions, and 3) the level of stainless steel contamination on these grid welds will not affect the ability of the Batch S fuel to satisfy its functional requirements, and that the presence ofstainless steel constituents will not cause any increase in fuel failure probability during normal operation.
t
ARKANSAS NUCLEAR ONE FORM TRLE: FORM No. REV. 10CFRSO.50 REVIEW CONTINUATION PAGE 1000.131C 3 Document No. 97-R-201842, Rev 0, &ml1.etion 10/23 l As an extra layer of defense for some of these cages, Entergy will impose restrictions on t
-their core residence time and core location to further minimize the probability of fuel fretting failure due to the very minimal incremental risk introduced by the stainless steel contamination. These restrictions do not impose any penalties wiiatsoever on cycle ener fuel cycle cost, operational flexibility, operational or safety margins.
This restriction takes advantage of the projected fuel shuffling and fuel management j schedule by placing those assemblies having the most contaminated seam welds into sub- { batches S2 and S3 fuel that are scheduled to be in the core for no more than two cycles will not be loaded on the core periphery, This will minimize the probability of fuel frettingj by limiting the core residence / exposure and avoiding core peripheral locations that were known to be more susceptible offuel rod fretting due to local high cross flow conditions. { l Rosemount Nuclear Instruments 10CFR21 Notification Rosemount Nuclear Instruments issued a notification under 10CFR21 regarding the Model 1154 Series H Alphaline Nuclear Pressure Transmitters on October 14,1998. The notification was to inform licensees that certain data contained in the qualification and test repons for the subject transmitter, which were used to specify the radiation performance identified in relevant product data sheets and manuals, were incorrectly tabulated and reported. This results in a small change to the transmitter's 30-minute radiation accuracy specification. In response to this notification, an evaluation was performed on the impact of the increased radiation effect. It was determined that eight (8) subject transmitters were in sersice at ANO. The tag numbers and functions of the transmitters are listed as follows: 2PT-4600-1 DSS Pressurizer Pressure 2PT-4600-2 DSS Pressurizer Pressure 2PT-4600-3 DSS Pressurizer Pressure 2PT-4600-4 DSS Pressurizer Pressure 2PT-4601-1 RCS Pressurizer Pressure 2PT-4601-2 1 RCS Pressurizer Pressure l 2PT-4601-3 RCS Pressurizer Pressure 2PT-4601-4 RCS Pressurizer Pressure Based on the evaluetions that were performed in support of this notification, the current calculations at ANO bound the increase in the radiation effect given in the Pan 21 notification, and the notification has no impact on the ANO calculations. (Reference CR-2-98-0393)
-- o,.
FORM TITLE:
, FORM No, GEV.
10CFR80.59 REVIEW CONTINUATION PAGE 1000.131C 3 Document No. 97 R-201s 02, Rev 0, Nd= dan . 11/23 T/H Screening Methodology 10CFR21 Notification Thermal hydraulic safety analysis is performed for a selected set oflimiting assemblie determine the CETOP/ TORC benchmarking overpower penalties in thermal hydrau analyses to determine inputs to specific transient events, as needed. To detennine the selected set oflimiting assemblies, the thermal hydraulics analyst ha traditionally first identiSed a subset oflow inlet flow assemblies that are likely candida for the most limiting assembly (s). The physics analyst subsequently used Ws subset of assemblies to survey for core power and local assembly power distribution cata for each of the subset members the flattest pin power distribution (i.e. correspond lowest pin-to-node). The thermal hydraulics analyst used these power distributions for th subset oflow inlet flow assemblies to perform further scoping thermal hydraulic evaluations to determine the most limiting assembly (s). The TORC code was used to calculate the inlet flow distributions. The C is benchmarked against the TORC model to ensure its conservatism over specified operating ranges. If CETOP-D is not conservative relative to TORC, a penalty factor m be applied to CETOP-D so that CETOP-D will calculate a conservative DNBR at a given condition over the specified operatmg ranges. ' The TORC model is based on a limiting assembly in which minimum DNBR will occur. A potentiallimiting assembly is selected in accordance with the following screening criteria: Low flow to the assembly and its four neighbors e Assembly maximum pin peak 2 90% of the maximum pin peak in the core e Flat pin power distribution (maximum pin peak / node average peak closest to 1.0 ABB Combustion Engineering identified a deficiency in the current screening metho (listed above) for determining the limiting assembly for detailed PWR thermal hydraulic analysis. The de6ciency is that the screening methodology based on low inlet flow m consistently identify the most limiting assembly in the presence of very flat core and assembly power distributions (lower pin-to-node ratio). The deficiency in the current screening methodology was concluded to constitute a defect as defined by 10CFR21. The ofBeial notification to the NRC was made on August 13,1998. (Reference CR-2-09-0 The thermal hydraulic analysis methods themselves (utilizing ABB-CE computer codes TORC and CETOP-D), when applied to the limiting assembly, yield valid results and are not at issue. The screening methodology was revised to correct the noted deficiency. The revised screening methodology was used in the analyses ofthe ANO-2 Cycle 14 reload core. The evaluation ofimpacts on the thermal-hydraulics analyses is presented below.
ARKANaAS NUCLEAR ONE FORM TITLE: FORM No. REV. 10CFR60,58 REVEWCONTINUATioN PAGE 1000.131C 3 P-:- ment No. 97-R 2018-02, key 0, h - - tion 12/23 Reactor Vessel Fluence A review ofthe impact of the Cycle 14 reload core on the fast neutron fluence to the 1 reactor vessel was performed using the methodology outlined in 96-R-2030-02, " Revised Reactor Vessel Fluence Determination". The results of this review indi values at the % vessel wall tl.ickness and the % vessel wall thickne the values listed in the bases of the ANO-2 Technical Specifications for the pressure / temperature limits. The h* justed reference temperatures at these locations are also less than the values listed Lt the bases ofthe ANO-2 Technical Specifications. Therefore the Cycle 14 reload core does not negativelyimpact the bases of the pressure / temperature limits in the Technical Specifications. Thermal Hydraulic Analysis The thern'al hydraulic analysis is described in Section 4.0 of the RAR. The thermal hydraulic characteristics of the Cycle 14 core necessitating re-evaluation are: The core average heat flux for Cycle 14 is lower due to the fact that Cycle 14 utilize 272 B.C shims, while Cycle 13 used 1292 B 4C shims. , e The limiting Pin-to-node factors with 1% uncertainty included for Cycle .4 are smaller than the Cycle 13 limiting values. Results of the re-analyds show that the following primary design bases criteria continu be met for the Cycle 14 core at ANO-2. There is at least a 95% probability at a 95% confidence level that the limiting fuel ro in the core does not experience departure from nucleate boiling (DNB) during n operation or anticipated operational occurrences; and, e The worst case hydraulic loads do not exceed the holddown capability of the fuel assemble (gravity plus holddown spring capability) during normal operation. The re-analysis generated revised 3-pump TORC data for use in the seized rotor fuel failure analysis. This analysis is described in Section 5.3 of the RAR. Transients Design Basis Events (DBEs) currently analyzed in the ANO-2 SAR was evaluated with respect to four criteria: Offsite Dose, Reactor Coolant System Pressure, Fuel Perform
. = - w .u.
FORM TITI.E: . FORM No. REV. 10CFR$0.89 REVIEW CONTINUATION PAGE 1900.131c 3 Document No. 97-R-2018-02, Rev 0 Determination 13/23 and Loss of Shutdown Margin. All events were reanalyzed or reevaluated to assure that they :neet their respective critedon at a reactor thermal power rating of 2815 MWt. It was dete. nined that Boron Dilution, CEA Ejection, Main Steam Line Break and Seized Rotor events need reanalysis. Boron Dilutiqm The Boron Dilution event was reanalyzed for Cycle 14 for the purpose of producing bounding critical boron concentration (CBC) / inverse boron worth (IBW) limit lines for use in the Reload Process Improvement (RPI) process in the current and future cycles. This is one portion ofthe ANO-2 Cycle 14 Reload Report. For Cycle 14, a different approach was taken than in the past. The dilution equations were rearranged to solve for relationships between IBW and CBC. That is, for a given dilution time constant, suberiticality at the time of alarm and the time from alarm to criticality, various values of CBC were input to the equations and the corresponding IBWs were calculated. The resulting CBC/IBW limit lines are used to determine the ace ntability of a cycle's core design by verifying for each mode and alarm configuration that the cycle specific CBC and IBW values fall within the acceptable region. These limits are based on the Standard Review Plan's criteria for boron dilution. The Cycle 14 specific CBC and IBW values have been compared to the mode and alarm dependent CBC/IBW limit lines. The comparison demonstrates acceptable results in all
-Cases.
An evaluation will be required to incorporate changes to the LBDs. ' CEA Eiection: The CEA Ejection analysis was performed for Cycle 14 for the following reasons: To perform cycle burnup dependent analyses for the purpose of setting less restrictive F, limits for the event.
- l To address more positive (or less negative) upper MTC limits in the COLR MTC limit curve at 20% and 50% power. l e
To perform generic transient analysis parametric studies for the purpose ofproducing ) i bounding core and RCS data for use in the RPI process in Cycle 14 and future cycles. ! The results of the analysis shows a slightly more restrictive F, limits for the reanalyzed points. The reason for this is slightly more positive MTC limits than what was used in ' previous analyses. These limits are valid for any future cycle for ANO-2 unless there is a change in the COLR MTC. l l
ARKANSAS NUCLEAR oNE FORM TITLE: FORM No. REY. 10CFR$0.59 REVIEW CONTINUATION PAGE 1000.131C 3 D-t No. 97-R-201s-02, Rev 0, Determmation
. 14/23 1
The cycle specific results confirm these limits are met for Cycle 14. Since these limits are based on protecting the energy deposition limits and will not result in exceeding the Pea Linear Heat Rate SAFDL, no fuel failure is predicted to occur due to this event. An evaluation will be required to incorporate changes to the LBDs. MSLB: The Main Steam Line Break (MSLB) analysis was assessed for Cycle 14. The methodology generally follows the Cycle 13 assessment with changes made to more realistically model the plant configuration and reflect Cycle 14 physics changes. The results confirm that there is no calculated fuel failure and that the licensing basis remains bounding for Cycle 14. An evaluation will be required to incorporate changes to the LBDs. Seized Rotor Event: The Seized Rotor event was reanalyzed for Cycle 14 to evaluate the impact of the revised TORC model and the Cycle 14 specific pin census. Consistent with the previous two cycles, the current methodology includes determining a POL with CETOP and conservatively reducing the mass flow rate to the 3-pump asymptotic flow while maintaining 100% heat flux. Since the 3-pump asymptotic flow is used, no transient response is necessary and the only physics data required is the pin census. The calculated fuel failure for Cycle 14 is less than the value determined in the previous analysis of record (Cycle 3). The offsite dose remains well within 10CFR100 limits. There are no changes to the LBDs related to this event and therefore an evaluation will not be i required. ECCS Analysis The Cycle 14 ECCS Performance analysis demonstrated that the results of the Reference Cycle analyses for LBLOCA and SBLOCA apply to Cycle 14. The limiting break, i.e., the break that results in the highest peak cladding temperature, from the Reference Cycl analyses is the 0.6 DEG/PD (Double Ended Guillotine / Pump Discharge) break. Conformance to the ECCS acceptance criteria for the limiting break is summarized below. The results are for a PLHGR of13,5 kW/ft and a power level of 2900 MWt (rated core power of 2815 MWt plus a 3% power measurement uncertainty). 4
~ .~.o~,. w_e m unc FORM TITLE:
. FORM No.. REV. 10CFRse.88 REVIEW CONTINUATION PAGE 1900.131C 3 Document No. 97-R-2018-02, Rev 0. Detennmation 15/23 Parameter Criterion Value Peak Cladding Temperature, 'F s2200 2158 Maximum Cladding Oxidation, % sl7 7.2 Maximum Core-Wide Cladding Oxidation, % s1 <0.99 Maintain Coolable Geometry Yes Yes There are no changes to the LBDs resulting from the ECCS performance analysis an therefore, an evaluation is not required. Reference to Figures 4.5-1 and 4.3-29 in the SAR. SAR section 4.5 references the Figure 4.5-1 in two places and section 4.3 references Figure 4.3-29 in four places. Cycle 13 reload 50.59 (96-E-0018-01, Rev. 0) deleted this figure since the information was included in the Core Operating Limits Report (COLR SAR needs to be corrected by replacing the reference to Figure 4.5-1 with the reference to the COLR MTC curve and the reference to 4.3-29 with the COLR PDIL figure. This c hange is beyond the scope of what is discussed in the operating license and does not involve a test or experiment. Changes to the sections 4.3 and 4.5 of the SAR are r but an evaluation is not required per 1000.131 Attachment E.1, " Correction ofMistakes made in incorporation ofrequested changes (the correct information has previously received a 10CFR50.59 Evaluation, which is still considered sufficient justification)". Th change will be included in the Chapter 4 LDCR prepared for the Cycle 14 reload. BASES FOR DETERMINATION
- 1. Require a change to the Operating License:
1(a) Require a change to the ANO-2 Technical Specifications? No The Cycle 14 RAR describes and addresses the design, accident analyses, and performa of the ANO-2 Cycle 14 core. The Cycle 14 core design and reload analysis results fully complies with the criteria discussed in the Technical Specifications. All the input assumptions and methods are consistent with or conservative with respcct to the current ANO-2 Technical Specifications. All cycle-specific limits for operation of the Cycle 14 core are located in the Cycle 14 COLR (98-R-2012-01, Revision 0). The remaining Technical Specification Safety Limits, Limiting Safety Settings, and Limiting Conditions of Operations (LCOs) governing the operation of the Cycle 13 core are also bounding for the { Cycle 14 core. A
ARKANSAa NUCLEAR ONE FORM TITLE: FORM No. RElf. 10CFRSS.SS REVIEW CONTINUATION PAGE 1000.131C 3 Document No. 97-R-2018-02, Rev 0. Dets - - = don 16/23 Although not specified in the Reload Report, a Technical Specification Change Req (TSCR) was submitted to the NRC and is required to be approved before the start
- 14. 2CAN059801 requested to change the MSSV setpoint uncertainty to 3% from 1% and the associated Linear Power Level High trip setpoint for three inoperable safety valve any steam generator to be changed from 45% to 36%. This was necessary due to the more positive MTC value for Cycle 14 at the 45% power level and the desirability to use a mor realistic valve setpoint uncertainty value. This change also involves a revision to the relevant TS bases section.
Other than the TSCR described above, no additional ANO-2 Technical Specification changes are required for the operation of the Cycle 14 core. The PSC concurrence with this 10CFR50.59 review is contingent upon approval ofthe ISCR by the NRC. 1(b) Require a change to the ANO-2 Operating License? No The results of the reload analyses presented in the Cycle 14 RAR fall within the requirements for operating the ANO-2 Cycle 14 core as referenced or described in the ANO-2 Operating License. Therefore, no changes to the ANO-2 Operating License are required to support the operation of the Cycle 14 core. 1(c) Require a change to the ANO-2 Confirmatory Orders? No The results of the reload analysis are within the requirements for operating the ANO-2 Cycle 14 core as referenced or described in the ANO-2 Confirmatory Orders and the l 1 specific results of the analysis are beyond the scope of the Confirmatory Orders. T no changes to the ANO-2 Confirmatory Orders are required to support the operation of t Cycle 14 core.
- 2. Require a change to SAR documents:
2(a&b) Result in information in the ANO-2 SAR or COLR being no longer true or accurate or violate a requirement stated in the document? Yes The Cycle 14 RAR describes and addresses the design, accident analyses, and perfo of the ANO-2 Cycle 14 core. The results of the reload analysis are within the requir for operating the ANO-2 Cycle 14 core as referenced or described in the ANO-2 SAR except for some cycle specific information in Chapters 4,15, and the COLR. ; i Chapter 4 of the SAR describes the fuel, reactor internals, reactivity control systems, nuclear and thermal and hydraulic design and the testing and verification of the ANO-2 reactor. Chapter 15 describes the various safety analyses. Information presented in the Cycle 14 RAR impacts the information presented in these chapters. Therefore, a 1 10CFR50.59 evaluation is required. LDCRs have been prepared for these changes.
\
a ARKANSAS NUCLEAR ONE FORM TITLE:
. FORM No. REV.
10CFR$0.69 REVIEW CommNUATioN PAGE 1900.131C 3 Document No. 97-R-2018-02, Rev 0 Detcir.di.ation 17/23 Boron Dilution is described in Section 15.1.4 of the ANO-2 SAR. With the different approach than what has been utilized in the past, the SAR needs to be updated. Therefore, a 10CFR50.59 evaluation is required. A LDCR has been prepared for the changes associated with this event. The boron dilution analyses assumed a shutdown margin of 5% for Modes 1 through 5. This is the current limit listed in the Cycle 14 COLR. Therefore, no changes to the COLR are required for this event. CEA Ejection is described in Section 15.1.20 of the ANO-2 SAR. This section needs to be updated to reflect the more limiting physics data and results. Therefore, a 10CFR50.59 evaluation is required. A LDCR has been prepared for the changes associated with this event. The CEA Ejection analysis, in part sets the PDILs that are listed in the ANO-2 Cycle 14 COLR. The PDILs are not being revised due to this analysis. Therefore no changes to the ANO-2 Cycle 14 COLR ars required to support this evaluation. The revised core physics input data, maximum post-trip fission power, minimum DNBR, and maximum post-trip reactivity values for Hot Full Power (HFP) need to be updated to reflect the Cycle 14 specific values in the SAR Section 15.1.14.1 on MSLB. The remaining SAR documents will not be affected by the proposed changes. The COLR is considered to be part of the SAR and reflects the specific operationallimits for the Cycle 14 core. In accordance with Technical Specification 6.9.5, the COLR is to be revised for each reload as a minimum ifchanges are needed. The Cycle 14 COLR has been prepared and reviewed. A separate 10CFR50.59 review of the COLR has been prepared. Change to the MTC Limit curve in the COLR is discussed in this 50.59; all the other changes to the COLR are addressed in the COLR 10CFR50.59 review. 7he PSC concurrence with this 10CFR50.59 review is contingent upon approval of the COLR 10CFR50.59 review. 2(c) Result in information in the Fire Hazards Analysis being no longer true or accurate or violate a requirement stated in the document? No The specific results of the reload analyses are beyond the scope of the FHA. Therefore, no changes to the FHA are required to support the operation of the Cycle 14 core. 2(d). Result in information in the Bases of the Technical Specifications being no ! longer true or accurate or violate a requirement stated in the document? No
.' Die resuhs of the reload analysis are within the requirements for operating the Cycle 14 core as referenced or described in the bases of the ANO-2 Technical specifications and does not result in invalidating any information presented in it. Therefore, no changes are required to the ANO-2 Technical Specification bases. l i
ARKANSAS NUCLEAR ONE FORM TITLE: FORM No. REV. 10CFRse.se REVIEW CONTINUATION PAGE 1000.131C 3 Document No. 97-R-2018-02, Rev 0, h = tion 18/23 Changes to the bases associated with the Technical Specification Change Request to th NRC described earlier are not addressed here. 2(e) i Result in information in the Technical Requirements Manual bemg no j longer true or accurate or violate a requirement stated in the document? No The results of the reload analysis are beyond the scope of the Technical Requirements Manual and do not result in invalidating any information presented in the TRM. There no changes to the TRM are required for this event. 2(f) Result in information in the ANO-2 SERs being no longer true or accurate or violate a requirement stated in the document? No The results of the reload analysis are within the requirements for operating the ANO-2 Cycle 14 core as referenced or described in the NRC SERs and does not result in invalidating any information presented in the ANO-2 NRC SERs.
- 3. Involve a test or experiment not described in the SAR?
No The results of the reload analysis are consistent with the current licensing basis documents The proposed changes do not involve any test or experiment that is not described in the SAR
- 4. Result in a potentialimpact to the environment?
No Page 3 of this 10CFR50.59 Determination verifies that there are no potential environmental impacts as a result ofoperating the plant with the Cycle 14 core. 5. Result in the need for a Radiological Safety Evaluation? No Operating the plant with the Cycle 14 core does not involve the processing of radioactive material outside the Auxiliary Building, Reactor Building, or the Low Level Radwaste Storage Building. In addition, a new pathway outside of the monitored ventilation or drainage pathways is not created with the operation of the Cycle 14 core. Thus, there is need for a radiological safety evaluation. 6. Result in any potentialimpact to the equipment and facilities utilized for VSC Ctivities? No The loading and operation of the Cycle 14 core does not require any spent fuel Ventilated Storage Cask or related activities. Therefore, no change to the equipment and facilities used for the VSC activities is required.
ARKANaAa NUCLEAR ONE FORM TITLE:
. FORM No. REV.
i 18CFR$0.58 REVIEW CONTINUATION PAGE 1000.131C 3
. Docwnent No. 97-R-2018-02, Rev 0, r1 tion 19/23 7(a) Involve a change under 10CFR50.54 for the QAMO? No The Cycle 14 RAR describes and addresses the design, accident analyses, and performance of the ANO-2 Cycle 14 core. The specific results of the analyses are the scope of the QAMO. Therefore, no changes to the QAMO are required to supp operation of the Cycle 14 core.
7(b). Involve a change under 10CFR50.54 for the E-Plan? No The Cycle 14 RAR describes and addresses the design, accident analyses, and performance of the ANO-2 Cycle 14 core. The specific results of the analyses are bey the scope of the E-Plan. Therefore, no changes to the E-Plan are required to support operation ofthe Cycle 14 core. ' REFERENCES
- 1. Arkansas Nuclear One Unit 2 Cycle 14 Reload Analysis Report, AN-FE-0354 Rev.00, 10/16/98, File 224-03, ANO No. 97-R-2018-01, Rev. O.
- 2. Vendor Document Review:' Arkansas Nuclear One Unit 2 Cycle 14 Reload Analysis Report, AN-FE-0354 Rev.00, File QR-026-27 '
- 3. ANO-2 Final Safety Analysis Report (as of11/98 through Amendment 14).
- 4. ANO-2 Technical Specifications (as of11/98 through Amendment 192).
- 5. CP: ANO-2 Cycle 14 (Batch S) Criticality Assessment, T. L. Lotz, NEAD-SR-98/048.R0,10/98, File QR-204-37. (CP - Calculational Package).
- 6. ABB/ Combustion Engineering Proprietary Analyses for ANO-2 Cycle 14:
A-AN-FE-0205 Revision 0, ANO-2 Cycle 14 Design ROCS / CORD Models and Depletions, ) 5/4/98.- A-AN-FE-0205 Revision 1, ANO-2 Cycle 14 Design ROCS / CORD Models and Depletions, 7/2/98. A-AN-FE-0213 Revision 0, ANO-2 Cycle 14 PAC Amermarnant, 9/15/98. A-AN-FE-0206, Revision 0, ANO-2 Cycle 14 Physics Input to Thermal Hydraulics Analysis, 5/11/98. i i
' A-AN-FE-4206, Revision 1, ANO-2 Cycle 14 Physics Input to Thennal Hydraulics Analysis, 7/2/98.-
A-AN-FE-0207 Revision 0, ANO-2 CEAC CEA Deviation Uncenannty Analysis for CEA Withdrawal,5/14/98. A.AN-FE-208, Revision 0, ANO-2 CEA Ejection RPI Parametrics Analysis, 9/21/98, ' A AN-FE-232, Revision 0, ANO-2 Cycle 14 Slow-Trip CEA Ejection Analysis for 50%,20%, and 0% Powers,10/9/98. ' A-AN-FE-0211 Revision 0, ANO-2 Cycle 14 Physics Data for the CEA Deviation within Deadhand Analysis,7/21/98.
ARKANSAS NUCLEAR ONE FORM TITLE: FORM No. REif. 10CFR50.88 REVEW CONTINUATION PAGE 1000.131C 3 Document No. 97-R 2018-02, Rev 0, D:r Hon 20/23 A-AN-FE-0210 Revision 0, ANO-2 Cycle 14 Generation of Additional MTC Data and +0.308 Axial Shape,7/7/98. A-AN-FE-0227 Revision 0, ANO-2 Cycle 14 Uncontrolled Boron Dilution Incident,10/12/98. A-AN-FE-0220 8/31/98. Revision 0lIhermal Hydraulics Checklist Assessment for ANO-2 Cycle 14, M-AK2-FMDE-98-402, Revision 0, Transmittal ofMechanical Design Section of the Comprehensive Chackhet for ANO-2 Cycle 14,8/31/98.
'AN-FE-340, Revision 0, Evaluation of the Fuel Performance Section of the Comprehensive Check List and Ammanament for Arkansas Nuclear One - Unit 2 Cycle 14, 7/13/98.
A-AN-FE-0175, 10/12/98. Revision 1, ANO-2 ECCS Performance Analysis Comprehensive Check, A-AN-FE-0209, Revision 0, ANJ-2 Cycle 14 Fuel Performance Analysis,8/31/98 A-AN-FE-0161, Rev 0, ANO-2 EPAC Equivalence Analysis,10/1/98. A-AN-FE-0214, Rev 0,1,7, ANO-2 Cycle 14 Thermal Hydraulic Analysis,11/5/98. A.AN-FE-0215, Rev 0, DO-2 Cycle 14 Post-Trip SLB Transient Analysis, 9/1/98. A.AN-PE4217. 10/19/98. Rev 0, ANO-2 RPI Pai.ii. sinc of CEA Withdrawal Within Deadband Ana A-AN-FE-0225, Rev 0, ANO-2 Cycle 14 ECCS Performance Analysis, 11/12/98. A-AN-FE 0226, Rev 0, ANO-2 Cycle 14 Addendum to Physics input to Thermal Hydraulics, 8/19/98. A-AN-FE 0228, Rev 0, ANO 2 Cycle 14 Seized Rotor Fuel Failure Analysis,10/1/98. A-AN-FE 0229, Rev 0, ANO-2 Reload Process Improvement - Comprehensive Checklist for N LOCA Transient Analysis,10/16/98. A-AN-FE 0230, Rev 0, ANO-2 Cycle 14 COLSS Margin Analysis,10/9/98. AN-FE-215. Rev 1,2, ANO-2 EPAC and APAC,10/1/98. AN-FE-0218, Rev 3, Release of Revision 03 of the ANO-2 PAC Methodology, 7/2/98. AN-FE4333, Rev 0, ANO-2 Cycle 14 Draft Reload Analysis Report, 8/25/98. AN-FE-0353, CW* 10/16/98. Rev 0, ANO 2 Cycle 14 - Overall Assessment ofReload Analysis Compre AN FE-0354, Rev 0, ANO-2 Cycle 14 Final Reload Analysis Report,10/16/98. AN-FE 0361, Rev 0, ANO-2 Cycle 14 Core Operating Limits Report Changes,11/9/98. ST-98-542, Rev 0, Revised Screening Methodology for Determining the Limiting Ass Thermal HydraW Analysis,9/30/98.
- 7. CEN-386-P-A, Verification of the Acceptability of a 1-Pin Burnup Limit of 60 MWD /kgU for Combustion Engineering 16x16 PWR Fuel, August 1992.
- 8. CEN-372-P-A, Fuel Rod Maximum Allowable Gas Pressure, Combustion Enginecring, Inc., May 1990.
- 9. CENPD-275-P-A, C-E Methodology for Core Designs Containing Gadolinia-Urania Burnable Absorbers, May 1988.
- 10. " Issuance ofAmendment No.178 to Facility Operating License No. NPF Arkansas Nuclear One, Unit 2 (TAC No. M96478)", NRC letter from Kombiz Salchi (NRC) to C. R. Hutchinson (EOI), Dated: 1/14/97, Docket No. 50-368.
a _-_______-__
wavuu m nucu. axone FORM TITLE
. FORM No. REV.
10CFR80.59 REVIEW CONTINUATION PAGE 1000.131C 3 Document No. 97.R-2018-02, Rev 0. Detenmnation 21/23
- 11. Letter, ANO-2 Cycle 14 Core Operating Limits Report (COLR) Changes, 2. E.
Karoutas (ABB) to J. H. Willoughby (EOI), ABB/CE Number: A-98-030, EOI Number: AN-FE-0361 Revision 0, November 9,1998, File 224-03.
- 12. Letter, Fuel Mechanical Design Improvements for Waterford 3 and ANO-2, R. M.
Wilkins (EOI) to Z. E. Karoutas (ABB), CEXO-97/00557, December 10,1997, Files 320-70,224-70
- 13. CP: MAXTMUM COLD INTERNAL PRESSURE OFA PWR FUEL ROD, D.L. Smith, NEAD-SR-95/037.R0,10/98, File QR-178-15. (CP - Calculational Package).
- 14. Vendor Document Review: Application ofNew Criteria andDCRfor Grid Weld impurity, ABB-CENO Letter A-98-029 and attachments, October 24,1998, Files QR-026-27,224-70, and 320-70.
i k t
ARKANSAS NUCLEAR ONE FORM TITLE: FORM NO. REV. itCFR50.59 REVIEW CONTI MIATION PAGE 1000.131C 3 . l Document No. 97-R 2018-02, Rev 0, Detenmnation l 22/23 _ Additional Manual search infor.eation ) Document Section ANO-2 Tech Spec ALL (boron, dilution, Ejection) t 4.1.1.1.1; 4.1.1.2; 3/4.1.1.3; 4.1.1.4.2; 3.1.2.1; 3.1.2.2; 3.1.2.3; 3.1.2.6; 3/4.1.2.7; 3/4.1.2.8; 3/4.4.1.2; 3/4.4.1.3; 3/4.5.1; 3/4.5.4; j 3/4.9.1; 3/4.9.2; 3.9.8.1; 3/4.9.12, -Ejection- 6.9.5.1 { ANO 2 OL ALL(Same keywords as above) ANO-2 Confirmatory Orders ALL (Same keywords as above)
"Ortier for Modification of License Conceming Environmental Qualification of Safety-related Electrical Equipment" ANO-2 SAR ALL (Same keywords as above) I 1.2.2.1.1; 1.2.2.3.1; 1.2.2.7; table 1.3-1; table 1.7-1; table 1.7-2;2.2.2.1; 2.2.2.4; 2.3.2.4; 2.3.4.1; 2.3.4.2; 2.3.4.3; 2.3.4.4; 2.3.5.1; 2.3.5.2; 3.1.2; 3.1.3; table 3.2-2; table 32-3; table 3.2-5; table 3.2-6; table 3.3-1; 3.5; table 3.5-1; 3.6.4.3.3.6; 3.6.4.4.3; table 3.6-27; 3.8.4.1.1; 3.8.5.1.3.1; 3.8.5.1.3.3; table 3.8-7;4.1; 4.2.1.1.5; 4.2.1.2.2; 4.2,1.2.4.10; 4.2.3; 4.2.32.2; 4.2.3.3.2.2; 4.2.3.3.2.6; table 4.2-1; 4.3.1.6; 4.3.1.9; 4.3.2.1; 4.3.2.3.2; 4.3.2.3.3; 4.3.2.3.4; 4.3.2.3.5; 4.3.2.4; 4.3.2.5; 4.3.2.5.2; 4.3.2.5.3; 4.3.2.5.4; 4.3.2.6; 4.3.3.2.4; 4.3.4; table 4.3-1; table 4.3-2; 4.5.2.1; 5.1; 5.2.1.5; 6.2.2.2.1; 6.2.3.2.1; 6.2.3.3.1.3;6.3.2.2.2;6.3.2.10;6.3.2.13;6.3.2.21;6.3.3.13.3; 6.3.3.15; table 6.3-1; 7.1.1.2; 7.2.2.1.1; 7.2.2.2.2: 7.4.1.3; !
7.4.1.3.1; 7.4.1.3.4; 7.4.1.5.2; 7.7.1.1.1; 7.7.1.1.6; 7.7.1.5; Table 7.5-1; table 7.5-3; 9.1.2.3; 9.1.4.3.2; 9.1.4.3.3; Table 9.1-2; i 9.3.2.1; 9.3.4.1.1; 9.3.4.1.2; 9.3.4.2.1; 9.3.4.2.2; 9.3.4.2.3; l 9.3.4.2.4;9.3.4.2.5;9.3.4.3.4;9.3.4.3.5;9.3.4.3.10;9.3.4.4.1; { Table 9.3-4; Table 9.3-6; table 9.3-15; table 9.3-21; table 9.3-22; i 10.4.5.2; 10.4.5.5; table 10.3-2; 11.1.1; 11.1.3; table 11.1-1; I table 11.1-10; 11.2.1; 11.2.2.1; 11.2.2.2; 11.2.3; 11.2.4.1; 11.2.4.1.8; 11.2.7; 11.2.8; 11.2.9; Table 11.2-1; table 11.2-2; table 11.2-6; table 11.2-7; table 11.2-14; 11.3.1; 11.3.8; 11.4.2.1.5;11.6.6;12.1.2.1;14.1.4;14.1.4.2;14.1.4.3.1; table 14.1-1; Table 14.1-3; table 14.1-4; 15.1.0.5.2; 15.1.0.6.1; Table 15.1.0-5; 15.1.1.1; 15.1.4; table 15.1.4-1; table 15.1.10-4; 15.1.13.4.1; table 15.1.13-1; 15.1.14.1.4.1; 15.1.14.1.4.3; table 15.1.14-28; table 15.1.14-29; 15.1.14-30; table 15.1.14-31; table 15.14-33; table 15.1.14-34; table 15.1.14-35; table 15.1.14-36; 15.1.18.2.1; 15.1.23.1; 15.1.32-Ejection-3.1.3; 3.5.2.2.2.2; 3.5.2.2.2.3; 3.5.2.2.2.4; 3.5.2.3.2; 4.2.1.1.1.1; 4.2.3.3.2.5; 4.2.3.3.2.7; Table 4.3-3; 7.2.2.1.2; 7.5.1.4.1; Table 7.3-2; Table 14.1-4; 15.1.20; Table 15.1.0-1; Tables 15.1.20-1 through 15.1.20-17
-was nuum um:
FORM TITLE: FORM NO. REV. 14CFRSO.SS REVIEW CONTNJATION PAGE 1000.131C 3 Document No. 97-R-2018-02, Rev 0 Deter-mation 23/23 ANO-2 COLR ALL (Same keywords as above)
-Ejedion-V FHA ALL (Eame keywords as above) 5.5.2; 5.6.2; 5.8.1; 6.2.5; Fire Areas K and 2020-JJ Bases ALL (Same keywords as above) 3/4.1.1.1; 3/4.1.1.2; 3/4.1.1.3; 3/4.1.1.4; 3/4.1.2; 3/4.4.1; 3/4.5.1; 3/4.5.4; 3/4.9.1; 3/4.9.8; 3/4.9.12-Ejedion-3/4/1/3 TRM ALL (Same keywords as above)
NRC SER ALL (Same keywords as above) NS.E 1.6; 4.1; 4.2.4; 4.3.2; 6.3.3; 7.7; 9.4.2; Table 15.1; 15.3; 15.3.1; 15.4.6-Ejedion-15.2; Table 15.1; 15.4; 15.4.1; Table 15.6 E.E8 2;6;24;32;43;58;81;82;86; 95;104;111;126;152;178;189; 190; 192-Ejedion-24; 37; 83; 111; 138; 157; 159; 190 VSC ALL (Same keywords as above) Table 1.2-2; 2.1; 6.1; 11.2.8.2; 12.2.2; 12.2.2.5 VSC SER ALL (Same keywords as above) 7.3; 11.3; 14.1.2; 14.2.6; Table 14-2 QAMO ALL (Same keywords as above) Eplan ALL (Same keywords as above) Table D-3; table D-4; 'l' 2.2.5 and 2.2.6 i
i ARKANSAS Nucl. EAR oNE FORM TITLE: FORM NO. REV. 10CFR80.50 EVALUATION 1000.131a 3 PC 2 Page1 of 19
. 10CFR50.59 Eval. No. b N -
(Assigned by PSC) Document No. 97-R-2018-02 RevlChange No. 9 Title ANO-2 Cycle 14 Re!aed Analysis Rescit , A WRITTEN RESPONSE PROVIDING THE BASIS FOR THE ANSWER TO EAC ATTACHED. EACH QUESTION MUST BE ANSWERED SEPARATELY. A SIMPLE CONCLUSION IS f40T SUFFICIENT. ATTACHMENT 2 PROVIDES GUIDANCE FO If the answer to any question on this form is "Yes," then an unreviewed safety question is involved.i
' to all questions is "No," then the proposed change does not involve an unreviewed safety question.~
1. { Will the probability of an accident previously evaluated in the SAR be ) I increased? Yes O No @ See attached discuss!en. 2. Will the consequences of an accident previously evaluated in the SAR be increased? Yes O No @ See attached discussion.
' 3.
Will the probability of a malfunction of equipment important to safety be increased? Yes O No @ See attached discuss!sn. 4. Will the consequences of a malfunction of equipment important to safety be increased? I Yes O No @ See attached discussion.
. 5.
Will the possibility of an accident of a different type than any previously evaluated in the SAR be created? YesO No @ See attached discuss!en. 6.
. Will the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR be created?
Yes O No @ See attached disces!en.
- 7. Will the margin of safety as defined in the basis for any technical specification be reduced?
Yes O No @ See attached discussion.
FORM TITLE: anruwsman Nuc'.t:AH ONE FORM NO. REV. 10CFR50.89 EVALUATION 1000.131B 3 Page 2 of_19 Document No. 97-R-2018-02 Rev1 Change No. O ML,w elserdaI Csdified Reviewers Signature John T. Sankoorikal 12/9/98 Printed Name Date Reviewers certification expiration date: 8/19/2000-Assistance provided by: Printed Name Scope of Assistance Date Thomas L Lotz Core Design -Neutronics Larry Hu 12/9/98 Core Design - Mechanical Lori Ann Potts 12/9/98 Thermal Hydraulics Robert W. Clark 12/9/98 Safety Analysis, CEA Ejection, Boron Dilution 12/9/98 Larry D. Young Safety Analysis, MSLB Tim J. Rush 12/9/98 Safety Analysis, Seized Rotor Jacque Lingenfelter 12/9/98 ECCS 12/9/98 PSC review by: W Date: O M D
ARKANSAS NUCLEAR ONE FORM TITLE: FORM No. REV. 14CFRSc.89 REVEW CONTINUATION PAGE 14ee.131c 3 Document No. 97-R-2018-02, Rev 0, Evaluation 3/19 10CFR50.59 Review Continuation Paae A summary of the changes being incorporated by this evaluation has been made in the determination section. The information in the detennination section delinea this evaluation. Please refer to the determination for a discussion of the changes and background with respect to these changes. 1. Will the probability of an accident previously evaluated in the SAR be increased? No Cycle 14 Core Neutronics and Fuel Management The neutronic fuel design changes (e.g., slightly longer cycle length, higher initial enrichment, larger batch size, etc) were properly incorporated in the neutronics models. Appropriate methodologies were used which have been approved by the U.S. NRC for application to ANO-2 reloads. The core loading pattern and Gadolinia integral poison rod were configured to maintain acceptable power peaking margin throughout Cycle 14 while meeting the Cycle 14 energy requirements. The resulting small differences between C 13 and Cycle 14 neutronics parameters are consistent with expected cycle-to-cycle variations. All Cycle 14 neutronic and fuel management parameters have been evaluated and found to be either bounded by the existing analysis ofrecord (AOR) or a Cycle 14 specific analysis performed for the Batch S reload assemblies and the Cycle 14 co For example, new bounding COLR upper moderator temperature coefficient (MTC) limi were utilized in the Cycle 14 input to the safety analyses. The power dependent insertion limits have been established to ensure adequate shutdown margin during normal conditions. A conservative evaluation of the shutdown margin, assuming the maximum worth rod is stuck in the full out position, demonstrates that the Cycle 14 shutdown margin meets the required acceptance criteria. The moderator temperature coefficient for Cycle 14 operation has been demonstrated to fully comply with the Cycle 14 COLR input to the safety analyses. Criticality analyses have been performed on the fresh, spent, and temporary storage and on the upender to ensure that the established criticality criteria are met at ANO-2. The current fuel stotage criticality analysis (See Amendment No.178 to Facility Operating License No. NPF-6) is applicable to the fresh Batch S fuel, as well as to all previous loaded fuel at ANO-2. Storage ofBatch S fresh fuel in Region 1 of the spent fuel rack requires cross-hatch fuel spacing. Storage ofBatch S fresh fuelin Region 2 of the sp fuel rack requires two-of-four spacing in a checkerboard configuration.
m,w oe cu m ura: 7 FORM TITLE:
. FORM No. REV.
10CFRSS.50 REVEW CoNTINUAT1oN PAGE 1000.131C 3 j
. Document No. 97.R-2018-02, Rev 0. Evaluation 4/19 The fuel rod internal pressure remains below the nominal operating system pressure fo projected Cycle 14 maximum burnup.
The Cycle 14 core configuration does not require any changes to the plant equipment initiators to accidents previously evaluated in the LBD are not affected, and the p of an accident is not increased due to the gadolinia poison, the increased feed enrichm or the longer fuel cycle length. Reload Assembly Design Changes Cycle 14 core reload differs slightly from Cycle 13 core reload due to changes in fuel management scheme and minor modifications in Batch S fuel design compared to Batch R and Batch P fuel design. The following provides an assessment of the impact of the changes in Batch S fuel design on the probability ofoccurrence of accidents documented in the ANO-2 FSAR. Evaluations ofthe impact of the design changes implemented for the Batch S reload fuel assemblies have shown that none of the fuel assembly mechanical structural design criteria [FSAR 4.2.1] for the normal operating and upset conditions, emergency conditions, and faulted conditions are violated. The fuel performance ofthe fuel designs in Cycle 14 has been evaluated using U.S. NRC approved codes (FATES 3B) and all design criteria were confirmed to be met. The maximum fuel rod burnup projected for ANO-2 Cycle 14 is 58,726 mwd /MTU, including a 750 mwd /MTU uncertainty, and is less than the 60,000 mwd /MTU licensed limit. The Cycle 14 burnup will be well within the industry experience base. The use of these design change features in Batch S assemblies has not significantly altere the enrichment and burnable poison loading scheme ofpellets and rods in a fuel assemblI nor the low-leakage loading scheme offuel assemblies in the Cycle 14 core. ABB/CE has successfully provided full batch application of these design features to other Combustion Engineering plants. There are no required changes in any vendor's quality control procedures, quality surveillance programs, or fabrication processes to ensure correct ' loading of fuel and burnable poison in assemblies and in the core. Therefore, the ' probability of erroneous loading of fuel pellets or fuel pins of different enrichment in a fuel assembly or erroneous placement or orientation offuel assemblies in the core [FSAR 15.1.15] due to these design change features in Batch S assemblies is not increased. The probability of fuel failure due to mechanical or flow induced vibration and fretting with the spacer grids [FSAR 4.2.1.3.6] will not be increased. The Batch S fuel assembly has the J
ARKANSAa NUCLEAR ONE FORM TITLE: FORM No. REV. 10CFRS8.s0 REVIEW CONTINUATION PAGE 1000.131C 3 Document No. 97-R-2018-02, Rev 0. Evaluation 5/19 same structural cage as the previous reload. Its fuel rods and poison rods have the same external dimensions, materials, clad thickness, and approximate mass as the Cycle 13 rod The probability of a fuel hand!ing accident [FSAR 15.1.23] will not be increased. These assemblies have the same snuctural cage as that previously used at ANO-2 and will be capable ofwithstanding the expected handling loads. These assemblies will continue to be compatible with the fuel handling equipment. The manner ofhandling the new fuel assemblies will be unchanged. The envelope of the new fuel is no different than that of the past. The mass of these new assemblies remains unchanged compared to the previous batch. Hence, the probability of a fuel handling accident is not increased. The probability of CEA misoperation [FSAR 15.1.3] is not increased The dimensions and positions of the CEA guide tube assemblies are unchanged compared to the assemblies used in the previous cycles. Also, any dimensional changes due to irradiation, such as assembly bow, will not be altered since no changes in the guide tubes material have occurred. No changes to the plant equipment or any significant changes to operating procedures are required for Cycle 14 due to the reload assembly design. No impact to any accident initiator occurs due to the Cycle 14 fuel. Therefore, the probability of an accident j previously evaluated in the FSAR will not be increased due to the Cycle 14 core loadin changes. j i Thermal Hydraulic Analysis
- The probability of an accident evaluated in the SAR will not increase due to the change in .
the thermal hydraulic design of the core for Cycle 14. The lower core average heat flux and smaller limiting pin-to-node factors have been evaluated to show that the current DNB )
' limit applies for Cycle 14 Transients 1 Boron Dilution:
In previous boron dilution analyses, the cycle specific CBC and IBW values were determined as part of the reload process. From these given values, the time from the alarm until the loss of shutdown margin was then explicitly calculated for each mode and plant configuration. These values were then compared to the values previously reported to the Staff and the Standard Review Plan (SRP) (for Modes 3 through 6) acceptance criteria. In Cycle 14, the same equations were used but the process was worked in reverse. Starting with the SRP acceptance criteria, a one minute margin was added to it (e.g., if the i
wwem mucuAM ONE l FORM TITLE: . [ FORM No. REV.
,10CFR50.59 REVWW CONTWuATioN PAGE I 1000.131C 3 Document Ns. 97-R-2018-02, Rev 0. Evaluation 6/19 I
SRP acceptance criteria is 15 minutes, then the analysis acceptance criteria is 16 Using these criteria and the relationship between the CBC and IBW values, the limits for the CBC and IBW for each mode and plant configuration were developed. The cycl specific CBC and IBW values were then compared to the established limits. For Cy all the data points fall within the acceptable region of all the curves. Bounding values for the worths of the shutdown banks and shutdown margin were assumed. The Uncontrolled Boron Dilution event was reanalyzed for Cycle 14 for the purpose o producing bounding CBC/IBW limit lines for use in the RPI process in the current and
. future cycles. The cycle-specific CBC and IBW values have been compared to the mode dependent CBC/IBW limit lines, which ensure that the acceptance criteria will be met for this event. The initiation of an inadvertent dilution requires a series ofmisoperations. The .
physics parameters; however do not effect the initiation of this event. Therefore, the I likelihood ofa boron dilution event is unchanged. . CEA Ejection: The CEA Ejection event was reanalyzed for Cycle 14 due to key physics input parame not being bounded by the Analysis ofRecord. The initiation of the event is the comple circumferential rupture of the CEDM housing or the CEDM nozzle. This analysis did no require any changes to the plant equipment or plant operations. The manner in which the
. unit is operated is not altered by these evaluations. The initiators to any of the accidents previously evaluated in the ANO-2 SAR related to this event are not affected and the probability of this or any other accident is not increased.
MSLB: The Cycle 14 operating conditions for the main steam line break evaluation remains unchanged such that accident initiators remain unaffected. Therefore, the probability of a main steam line break accident previously evaluated in the SAR will not be increased for Cycle 14 operation.
'I . In conclusion, based on the above discussion of changes due to the Cycle 14 reload core and the associated analyses results, the probability of an accident previously evaluated in the ANO-2 SAR will not be increased. !
i
ARKANSAS NUCLEAR ONE FORM TITLE: FORM No. REV. 10CFRSS.50 REVIEW CONTINUATION PAGE 1000.131c 3 Document No. 97-R 2018 02, Rev 0, Evaluation 7/19 2. Will the consequences of an accident previously evaluated in the SAR be
. increased?
No Cycle 14 Core Neutronics and Fuel Management All Cycle 14 neutronic and fuel management parameters have been evaluated and fo be either bounded by the existing analysis ofrecord (AOR) or a Cycle 14 spe performed for the Batch S reload assemblies and the Cycle 14 core design.
\
The power dependent insertion limits have been established to ensure adequate shu margin during normal conditions. A criterion has been established to ensure that the core is capable of providing adequate shutdown margin under both normal and accident conditions with the single most reactive control rod fully withdrawn. The available scram worth under such conditions was shown to remain above the requirements. Criteria have been established for MTC behavior as a function of core power level to ensure that the consequences of various accidents are acceptable. The Cycle 14 fuel characteristics are consistent with the assumptions used in the spent fuel pool ra containment temporary storage rack, new fuel vault, and fuel handling equipm analyses. Since the peak rod exposure, shutdown margin, MTC, criticality criteria, an other physics input requirements are met by the Cycle 14 core configuration, the consequences of accidents previously evaluated in the LBD will not be increased for the Cycle 14 fuel. The current fuel storage criticality AOR have been confirmed to be applicable to the Bi ! S reload fuel assembly design. The current storage and handling criteria and requ will preclude criticality during normal and postulated events, so the consequences a affected. Reload Assembly Design Changes The mechanical design of the Batch S and Batch R reload fuel bundle assemblies are virtually identical. The mechanical design bases have not changed since the orig; design. The thermal performance of composite fuel rods that envelop the fuel rods of the fuel batches present in ANO-2 Cycle 14 has been evaluated. The analysis was pe a power history that envelopes the power and burnup levels representative of the pe rod for each batch at each burnup interval, from beginning of cycle to end of cyc The burnup range analyzed is in excess of that expected at the end of this cycle.
FORM TITLE: FORM NO. REV. 10cFR60.59 REVIEW CONTINUATION PAGE 1000.131c 3 Document No. 97-R-201842, Rev 0 Evaluation 8/19 The fuel performance of the fuel designs at Cycle 14 burnups hcs been evaluated usin U.S. NRC approved codes (FATE 23B), and all design criteria were confirmed to be met. The maximum fuel rod burnup projected for ANO-2 Cycle 12 is 58,726 mwd /MTU, including a 750 mwd /MTU uncertainty, and is less than the 60,000 mwd /hfrU licensed limit. The Cycle 14 burnup will be well within the industry experience base. No chang clad barrier performance will occur. Adddonal fuel performance analyses were performed to show that the gadolinia rods are bowde( by the urania rods with respect to rod internal pressure, fuel centerline temper. mre, and power-to-melt criteria. The fuel rod internal pressure remains below the nominal operating system pressure for the projected Cycle 14 maximum burnup. The cold internal rod pressure was calculated to remain below the NRC Regulatory Guide 1.25 value of 1200 psig. These Batch S fuel assemblies have the same envelope, materials, dimensions, structural cage, and virtually the same mass as that previously used at ANO-2. Hence, the number offuel pins that will fail during a fuel handling accident will not be more than the present analyzed pin failures. Therefore, consequences of a dropped bundle accident [FSAR 15.1.23] are also not increased. Adequate shoulder gap is predicted for all of the batches offuel in Cycle 14. The chemical and metallurgical performance of the Batch S fuel will be similar to the Batch P and Batch R fuel. As such, no change will occur in the radiological release rate / duration, no new release mechanisms can be postula no impact will occur to any radiation release barriers. Therefore, the consequences o accident previously evaluated in the LBD will not be increased because of the use of Batch S fuel assemblies. Thermal Hydraulic Analysis i The consequences of an accident previously evaluated in the SAR will not increase due to I the changes in the thermal hydraulic design of the core for Cycle 14. The lower core I average heat flux and smaller limiting pin-to-node factors have been evaluated to show that the current DNB limit applies for Cycle 14. The cycle specific Thermal hydraulic analy did generate revised 3-pump TORC data for use in the seized rotor fuel failure analys The results of this analysis show that the calculated fuel failure as a result of this accident is less than the value determined in the previous analysis ofrecord. There will not be an change to the radiological release rate or duration, no new release mechanisms can be l postulated, and there will not be any impact to the radiation release barriers. Thus, the consequences of this accident are actually decreased. 1 r
FORM TITLE: FORM No. REV. I s 1scPRes.se REVIEW CONTINUATION PAGE 1000.131C ! 3
- Docwnent No. 97-R-2018 02, Rev 0. Evaluation 9/19 Transients Boron Dilution:
i
! The Cycle 14 nec:ific CBC and IBW values were within the acceptance region of the limits that were detmed. Based on this, the boron dilution analyses for Cycle 14 does not challenge the results or criteria previously accepted by the NRC. As such, there is no change to the radiological release rate / duration, no new release mechanism can be postulated and no impact will occur to any radiation release barrier. Therefore, the consequences of a boron dilution event will not be increased
' CEA Ejection:
The proposed changes to ANO-2 SAR are to describe the input parameters and results of the reevaluation ofthe CEA Ejection event. This analysis did not require any revisions to either plant operations or plant equipment. As such, no changes will occur to any radiological release rate / duration, no new release mechanisms can be postulated and no impact will occur to any radiation release barriers. In addition, the analysis determined that there is no fuel failure due to this event.- Therefore, the consequences ofany accident previously evaluated in the ANO-2 SAR will not be increased. MSLB: The Cycle 14 accident conditions for the main steam line break remain unchanged in regard to offsite dose consequences. The current Technical Specification limits on primary and secondary activity remain unchanged and in coabination with no physical plant changes will ensure no increase in the offsite dose consequences for a main steam line break. Therefore, the consequences of an accident previously evaluated in the SAR will not be increased. In conclusion, based on the above discussion ofchanges due to the Cycle 14 reload core and the associated analyses results, the dose consequences of any accidents previously evaluated in the ANO-2 SAR will not be increased
- 3. .
.Will the probability of a malfunction of equipment important to safety be
-increased?
No Cycle 14 Core Neutronics and Fuel Management All equipment important to safety will function in the same manner with the Cycle 14 reload core as with the previous reload configuration. No characteristic of the Cycle 14 core, which is different from previous reload cores, increases the probability ofa f i-
m,ww.a. cuwoce vose FORM TITLa: FORM No. REV. 10CFRse.89 RaVEW CONTINUATION PAGE 1000,131c 3 Document Ns. 97-R-2018 02. Rev 0. Evaluation 10/19 malfunction of equipment important to safety. Therefore, the probability of an occurrence of a malfunction of equipment important to safety is not increased due to the Cycle 14 core reload. Reload Assembly Design Changes The Batch S assemblies are materially, dimensionally, and structurally the same as previous fuel designs used during Cycle 13. The minor changes in the grid cage assembly design are considered in the Cycle 14 reload analyses. The very small downward shift of the dashpot and cooling hole elevations have a minimal impact on the CEA operation. The stiffness of the guide tube assembly and the assembly bow are not impacted. Therefore, no change in the CEA insertion performance is affected. In addition, the cold internal pressure of the Cycle 14 fuel rods will continue to be limited to below 1200 psig. No changes in the assumptions concerning equipment availability or failure modes are made. Therefore, the mechanical changes do not increase the probability of a malfunction of any equipment important to safety. Thermal Hydraulic Analysis The probability of a malfunction of equipment imponant to safety will not increase due to the changes in the thermal hydraulic design of the core for Cycle 14. The lower core average heat flux and smaller limiting pin-to-node factors have been evaluated to show that the current DNB limit applies for Cycle 14. The Cycle 14 core is similar to past cores and does not alter plant operation such that equipinent imponant to safety will be affected. Transients Boron Dilution: For the Boron Dilution event, the changes to equipment imponant to safety involves the i 1 us2 of CEA Banks A and B for cocked rod protection under certain conditions. The use of Banks A and B for cocked rod protection requires Banks A and B to be fully withdrawn from the core. The withdrawal and maintenance of these CEAs in this configuration are within the design and licensing bases for the unit and does not impact the function of the supporting equipment. Therefore, there is no increase in the probability of occurrence of a malfunction ofequipment important to safety.
ARKANSAS NucLaAR oNE FORM TITLE: FORM No. REV. 10CPRst.s0 REVEM CONT 38uATIoN PAGE { 1000.131C 3 Document M. 97-R-2018-02, Rev 0. Evaluation
/
11/19 CEA Ejection: The purpose of the proposed changes is to implement the bounding ejection analysis assumptions and results. This analysis does not require any changes to the plant equipmen or operations and none of the accident initiators are impacted by the proposed changes. As such, the function and the duty of the equipment imponant to safety are not altered. Therefore, the probability of a malfunction of equipment imponant to safety is not increased due to this analysis. MSLB: The Cycle 14 operating conditions for the main steam line break evaluation remain unchanged such that the licensing basis limits for the specified acceptable fuel design limits on maximum linear heat generation rate or minimum depanure from nucleate boiling ratio (DNBR) remain bounding. Therefore, the probability of a malfunction of equipment important to safety will not be increased 7 iIn conclusion, based on the above discussion of changes due to the Cycle 14 reload core and the associated analyses results, the probability of a malfunction of equipment imponan to safety will not be increased. 4. Will the consequences of a malfunction of equipment important to safety be increased? No Cycle 14 Core Neutronics and Fuel Management The core loading pattern has been configured to maintain acceptable radial peaking throughout Cycle 14. The core configuration, including increased feed enrichment and gadolinia poison in the fresh fuel, has been appropriately incorporated into the Cycle 14 neutronic model. All Cycle 14 neutronic and fuel management parameters have been
' evaluated and found to be either bounded by the existing analysis ofrecord (AOR) or a Cycle 14 specific analysis performed for the Batch 5 reload assemblies and the Cycle 14
. core design to show acceptability. For example, new bounding COLR upper moderator temperature coeHicient (MTC) was utilized in the Cycle 14 input to the safety analyses.
Total and stuck rods wonh calculations performed for Cycle 14 demonstrate that the available shutdown margin exceeds the required criterion. These evaluations do not require any changes to the assumptions concerning equipment availability or failure modes. The suberiticality margins for all fuel used through Cycle 13 have been confirmed for the containment tenporary storage racks, the spent fuel pool, and the fuel handling equipme L
g NN HAM ONE FORM TITLE: FORM NO. Rett 10CPR$0.s0 REVEW CoNTBluaTioN PAGE 1000.181C 3 Dwument No. 97-R-2018@, Rev 0. Evaluation
> 12/19 The Batch S fuel meets the fuel storage acceptance criteria for the new fuel vau fuel pool, the containment temporary storage racks, and the fuel handling equipment.
No changes in the assumptions concerning equipment availability or failure modes were made, and no new equipment or operational changes are required. Therefore, the consequences of a malfunction of any equipment important to safety will not be increased with the gadolinia poison, the increased feed enrichment, or the longer fuel cycle len Reload Assembly Design Changes
, The minor changes in the grid cage assembly arti considered in the Cycle 14 reload analyses. The Batch S assemblies are materially, dimensionally, and structurally the sam as previous fuel designs used during Cycle 13. In addition, the cold internal pressure of the Cycle 14 fuel rods will continue to be limited to below 1200 psig. As such, no change wi occur in the radiological release rate /duratio' n , no new release mechanisms can be postulated, and no impact will occur to any radiation release barriers.
{ The fuel performance of both the Guardian and non-Guardian fuel designs at the Cycle 1l burnups has been evaluated using U.S. NRC approved codes (FATES 3B), and all desi
~
criteria were confirmed to be met. The maximum fuel rod burnup projected for ANO-2 I Cycle 14 is 58,726 mwd /MTU, including a 750 mwd /MTU uncertainty, and is less than
' the 60,000 mwd /MTU licensed limit. The Cycle 14 bumup will be well within the iindustry experience base. No change in clad barrier performance will occur.
No changes in the assumptions concerning equipment availability or failure modes are made. Therefore, the mechanical changes do not increase the consequences of a malfunction of any equipment important to safety. Thermal Hydraulic Analysis The consequences of a malfunction of equipment important to safety will not increase due to the changes in the thermal hydraulic design of the core for Cycle 14. The lower core average heat flux and smaller limiting pin-to-node factors have been evaluated to show that the current DNB limit applies for Cycle 14. The Cycle 14 core is similar to past cores and does not alter plant operation such that operation or effectiveness of equipment important
' to safety will be affected. l i
i
ARKANSAS NUCLEAR ONE FORM TITLE: FORM No. REV. 10CPR80.50 REVEW CONTINUATION PAGE 1000.131c 3 e Document No. 97-R-2018-02, Rev 0, Evaluation 13/19 i Transients - Boron Dilution: The withdrawal ofCEAs for cocked rod protection during shutdown conditions does not affect the overall function of the system and its capability to trip when needed. The function and duty of the equipment imponant to safety is not altered by this event. As such, the initiators to this event are not effected by the reanalysis and the consequence malfunction of equipment imponant to safety are not increased. CEA Ejection: The analysis of the CEA ejection event does not require any changes to the assumpt concerning equipment availability or failure modes. No equipment or operational chan
- is caused by this analysis. The function and duty of the equipment important not altered. Thus the initiators to these events are not affected by this analysis and the consequence of a malfunction of equipment important to safety is not increased.
MSLB: The Cycle 14 operating conditions for the main steam line break remains unchanged in regard to offsite dose consequences. The current Technical Specification limits on p and secondary activity remain unchanged and in combination with no physical plant changes will ensure no increase in the offsite dose consequences for a main steam line break. Therefore, the consequences of a malfunction ofequipment important to sa not be increased. In conclusion, based on the above discussion of changes due to the Cycle 14 reload core and the associated analyses results, the consequences of a malfunction of any equipme important to safety will not be increased. 5.- Will the possibility of an accident of a different type than any previously evaluated in the LBD be created? No Cycle 14 Core Neutronics and Fuel Management There are no new system interactions or connections associated with the ANO-2 Cycle i core reload. Therefore, operation of the ANO-2 Cycle 14 core will not initiate an accident of a different type than previously evaluated in the FSAR.
semarrnas anA4.a:AN ONti FORM TITLE: . FORM No. REV. 10CFRSS.Ss REVIEW COlmNuaTioN PAGE 1900.131C 3 Document No. 97-R 2018 02, Rev 0, Evaluation 14/19 All Cycle 14 neutronic and fuel management parameters have been evaluated and found to be either bounded by the existing analysis ofrecord (AOR) or a Cycle 14 specific an performed for the Batch S reload assemblies and the Cycle 14 core design. For exam new bounding COLR upper moderator temperature coefficient (MTC) was utilized in the Cycle 14 input to the safety analyses. Total and stuck rods worth calculations performed j for Cycle 14 demonstrate that the available shutdown margin exceeds the required criterion. There is no new equipment associated with the use ofBatch S fuel. The gadolinia poison the increased feed enrichment, and the longer cycle length will not alter the way in which the plant operates. No changes in the failure modes of the equipment important to safe were assumed in the Cycle 14 analyses. No initiator to any of the accidents was impacted. Therefore, the possibility of an accident of a different type than any previously evaluated will not be created due to the Cycle 14 core reload. Reload Assembly Design Changes . The fuel performance of the Guardian and non-Guardian fuel designs at the Cycle 14 burnups has been evaluated using U.S. NRC approved codes (FATES 3B), and all des criteria were confirmed to be met. The maximum fuel rod burnup projected for ANO-2 Cycle 14 is 58,726 mwd /MTU, including a 750 mwd /MTU uncertainty, and is less than the 60,000 mwd /MTU licensed limit. The Cycle 14 bumup will be well within the industry experience base. No change in clad barrier performance will occur. The FATES 3B fuel performance analysis has demonstrated that no change will occur in the radiological release rate / duration, no new release mechanisms can be postulated, and no impact will occur to any radiation release barriers. The changes will not require new equipment or alter the way in which the plant operates. No changes in the failure modes of the equipment important to safety were assumed in these analyses. No initiators to any of the accidents are impacted. Therefore, the changes will not result in an accident of a different type than previously evaluated. Thermal Hydraulic Analysis The possibility of an accident of a different type than any previously evaluated in the SAR is not created by the changes in the thermal hydraulic design of the core for Cycle 14. The Cycle 14 core is similar to past cores and does not alter the operation or the required configuration of the plant. As such, no new initiators will be created due to the slight thermal hydraulic changes for Cycle 14.
i [! FORM TITLED ARKANSAS NUCt.aAR oNE i FORM No. RElf. 10CPRse.s0 REVEWCoNTRIUATIoN PAGE ' 1ste.131C 3 Document No. 97-R-2018-02, Rev 0. Evaluation 15/19 f ,
, Transients - !
Boron Dilution: CEA Banks A and B must be pulled under certain conditions to satisfy the boron dilu event criteria. The withdrawal of the CEAs during shutdown conditions does not new equipment or alter the way the plant operates. The withdrawal of the CEAs does n create an additional failure mode than what has already been analyzed. No initiator to thi
. event or any of the other accidents are impacted. Therefore, the possibility of an accident of a different type than any previously evaluated in the ANO-2 SAR will not be create CSA Ejection:
The analysis of the CEA Ejection event does not require new equipment or which the plant is operated. This analysis did not assume an additional failure mode what has already been analyzed. No initiator to any other accidents is impacted. ; Therefore, the possibility of an accident of a different type than any previously eval the ANO-2 SAR will not be created. 1
-MSLB:
The Cycle 14 operating conditions for the main steam line break and the physica
~
conditions remain unchanged. This ensures that the circumstances considered by) analyses remain bounding. Therefore, the possibility of an accident of a different any previously evaluated in the SAR is not created. In conclusion, based on the above discussion of changes due to the Cycle 14 reload c and the associated analyses results, the possibility of an accident of a different type any previously evaluated in the ANO-2 SAR will not be created.
- 6. _
Will the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR be created? No Cycle 14 Core Neutronics and Fuel Management l Installation of a reload core cannot cause the malfunction ofequipment important in a different manner than that already evaluated in the FSAR. Equipment important safety will function in the same manner with the reload core as with the previous core. Th changes in core characteristics do not change any parameter that would affect the functio of equipment important to safety. There are no new methods of failure associate of the changes previously identified for Cycle 14, i
p N muutnag oug FORM TITLE: FORM No, REV. 10cFR$0.89 REVWW coNrNduaTION PAGE 1000.131C 3
, . Document No. 97-R-201 s-02, Rev 0. Evaluation i
16/19 i Reload Assembly Design Changes There is no new equipment associated with the use ofBatch S fuel. No new systems or substructures are involved. The changes will not alter the way in which the plant op No changes in the failure modes of the equipment imponant to safety were assumed in the Cycle 14 analyses. Therefore, the possibility of a malfunction of equipment imponant to safety of a different type than any previcusly evaluated will not be created. Thermal Hydraulic Analysis The possibility of a malfunction of equipment imponant to safety of a different type than any previously evaluated in the SAR is not created by the changes in the thermal hydraulic design of the core for Cycle 14. The lower core average heat flux and smaller limiting ' to-node factors have been evaluated to show that the core remains cooled during abnormal or anticipated occurrences. The Cycle 14 core does not alter the operation or the requir configuration of the plant. Since no new initiator has been created, consequences to existing analyzed abnonnal or anticipated occurrences have not worsened, or no new equipment is required for mitigation of an event, the possibility of a malfunction of ' equipment important to safety of a different type than previously evaluated is not created. Transients' ! Boron Dilution: ( CEA Banks A and B are utilized for cocked rod protection during cenain shutdown ' conditions. The boran dilution event does not require new equipment or alter the way in which the plant operates. No changes in the failure modes of the equipment imponant to safety were assumed in the analyses. No initiators to any of the accidents are impacted. L Therefore, the possibility of a malfunction ofequipment important to safety of a different type than any previously evaluated in the ANO-2 SAR will not be created. CEA Ejection:
- i The CEA Ejection analysis does not require new equipment or alter the way in which the plant operates. No changes in the failure modes of the equipment importart to safety were !
assumed in the analysis. No initiators of any of the accidents are impacted. Therefore the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the ANO-2 SAR will not be created. i 4 l-
ARKANaAa NUCLEAR oNE FORM TITLE: FORM No. REV. 19CFRSS.s0 REVWWcoNTINuaTIoN PAGE 1000.131C 3 Document No. 97-R-201s 02, Rev 0, Evaluation i 17/19 MSLB: l The Cycle 14 operating conditions for the main steam line break and the physical plant conditions remain unchanged. This ensures that the malfunctions considered by previous analyses remain boundmg Therefore, the possibility of a malfunction ofequipment important to safety of a different type than any previously evaluated in the SAR is not created. In conclusion, based on the above discussion of changes due to the Cycle 14 reload core and the associated analyses results, the possibility ofmalfunction ofequipment important to safety of a different type than any previously evaluated in the ANO-2 SAR will not be created. 7. Will the magia of safety as defined in the basis for any technical specification be reduced? No Cycle 14 Core Neutronics and Fuel Management The Batch S fresh fuel assemblies in Cycle 14 are loaded such that cycle energy potential maximized and local peaking is minimi~d. The gadolinia-poisoned rods, the increased fresh fuel feed enrichment, and the increased fuel cycle length change have been modeled in the neutronics calculations, which are performed using U.S. NRC licensed methodology j appropriate for ANO-2. The changes in the core Cycle 14 core configuration have been ' found to introduce small differences in the neutronics parameters that are consistent with expected cycle-to-cycle variations. The calculated moderator temperature coefficients are consistent with the COLR MTC requirements throughout Cycle 14 operation. The pover dependent insertion limits have been established to ensure adequate shutdown margin _ during normal conditions, and the shutdown margin exce'eds the COLR acceptance criteria of 5.0%Ap with the maximum worth rod stuck out. Criticality analyses have demonstrated that the Batch S fuel can safely be stored in the
. spent fuel pool racks, the containment temporary storage racks, the fuel handling equipment, and the new fuel vault. Fuel from previous cycles can also be safely stored.
The spent fuel racks, the containment temporary storage reck, and the fuel carrier maintain a k-effective less than 0.95 under all conditions. The new fuel vault maintains a k-effective less than 0.95 under normal conditions and less than 0.98 under optimum moderation conditions. Therefore, the margin to safcty as defined in the bases for the criticality technical specification will not be reduced for Cycle 14 operation with the introduction ofBatch S fuel.
ARKANSA NuCLEARoNE FORM TITLE: .
. FORM No.
~ REV.
10CPRSS.89 REVEW CONTINUATION PAGE 1000.131C 3 D-==t No. 97-R 2018 02, Rev 0, Evaluation 18/19 Reload Assembly Design Changes The fuel performance of the fuel designs at higher Cycle 14 burnups has been evaluated using U.S. NRC approved codes (FATES 3B), and all design criteria were confirmed to be met. The maximum dadding plastic strain will remain below 1.0% within the anticipated fuel assembly burnup, and fuel melt will continue not to occur. Therefore, the margin safety will not be reduced due to the Batch S reload assembly design changes. Thermal Hydraulic Analysis The margin of safety as defined in the bases for any technical specification will not be reduced as a result of the change in the thermal hydraulic design of the core for Cycle 14 The Cycle 14 core, with lower core average heat flux and smaller limiting pin-to-node factors, has been evaluated consistent with the margins as prescribed in the bases of the i Technical Specifications to show that the current DNB limit applies for Cycle 14. Transients Boron Dilution: No acceptance criteria were found in the Bases of the ANO-2 Technical Specifications which is explicitly associated with the Boron Dilution event. Therefore, the margin of safety as defined in the bases for any ANO-2 Technical Specification will not be reduced. CEA' Ejection: The CEA ejection analysis has been performed consistent with the Technical Specification requirements, with NRC approved methods as listed in the COLR reference list and in a manner consistent with the approval. The Transient Insertion Limits and the Shutdown Insenion Limits " ensure the potential effects of a CEA ejection accident are limited to acceptable limits." Since these limits are not revised for Cycle 14 and conservative inputs were used in the analysis, the margin of i safety as defined in the bases of ANO-2 Technical Specification 3/4.1.3 is not reduced. Therefore, the margin of safety as specified in the basis for any ANO-2 Technical Speci6 cation will not be reduced. MSLB: The reactor fuel linear heat rate and DNBR will be maintained within the previously established acceptable limits and thus maintain the margin ofsafety. Therefore, the margin
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antwvana nuvocan unc FORM TITLE: FORM NO. REV. 10CFR50.59 DETERMINATION 1000.131A 3 Pc 1 Page 1 of.11 Document No. 97-R-2018-03 Rev./ Change No. 9 Title Core Operatina Limits Report for Cvele 14 Brief description of proposed chtrige: Implement reauired chanoes to the COLR to suDoort the Cvele 14 operation. (additional description is alven in the followina paces.) Will the proposed Activity:
- 1. Require a change to the Operating License including:
Technical Specifications (excluding the bases)? YesO No@ Operating License? YesO No@ Confirmatory Orders? YesO No@ 2. Result in information in the following SAR documents (including drawings and text) being (a) no longer true or accurate, or (b) violate a requirement stated in the document: SAR (multi-volume set foreach unit)? . YesO NoS Core Operating Limits Report Yes@ nod Fire Hazards Analysis? YesO No@ Bases of the Technical Specifications? YesO No@ Technical Requirements Manual? YesO No@ l NRC Safety Evaluation Reports? YesO No@
- 3. Involve a test or experiment not described in the SAR?
(See Attachment 2 for guidance) YesO No@
- 4. Result in a potential impact to the environment? (Complete the EnvironmentalImpact Detennination of this form.)
YesO No@ S. Result in the need for e Radiological Safety Evaluation l per section 6.1.57 i YesO No@
- 6. Result in any potential impact to the equipment or facilities utilized for Ventilated Storage Cask activities per Section 6.1.67 YesO No@
- 7. Involve a change under 10CFR50.54 for the following SAR documents per Section 6.1.7:
QAMO? YesO No@ j i E-Plan? YesO No@ !
ARKANSAS NUCLEAR ONE FORM TITLE: FORM NO. REV. 10CFR60.89 DETERMINATION . 1000.131A 3 Pc 1.2 Page 2 of n Document No. 97-R 2018-03 Rev> Change No. A Basis for Determination (Questions 1. 2 & 31: I See attached dmo ===!on. I 0 Proposed change does not require 10 CFR 50.59 Evaluation per Attachment 1, item.#(if checked, l j note appropriate item #, send LDCR to Licensing). l Search Scope: List sections reviewed in the Licensing Basis Documents sp;cified in Question 1,2 and 3. If a search was performed on LRS, the LRS search index should be entered under "Section" with the search statement ( parentheses. Controlled hard copies of the documents shall be reviewed (LRS is not verified and search text, not figures or drawings). Attach and distribute a completed LDCR per Section 6.1.2 If LBD changes required. Document Section LRS: 50.59 - ANO-2 ALL (See attached for keywords) MANUAL SECTIONS: See attached See attached FIGURES: See attached See attached
* ^ ' '
John T. Sankoorikal 12/22/98 CMtified Reviewers Signature Printed Name Date Reviewers certification expiration date: 8/19/2000 Assistance provided by: 1 Printed Name Scope of Assistance Date Robert W. Clark Generalinput 12/22/98 Todd Erskine Generalinput 12/22/98 Jacque Lingenfelter Linear Heat Rate 12/22/98
)
Search Scope Review Acceptability (NA, if performed by Technical Review per 1000.006) , M Daniel W. Fouts 12/z2/f/ Certified Reviewefs Signature Printed Name Date l
AMAANbAh NUCLhAR ONti FORM TULE: FORM NO. REU. 10CFR50.89 DETERMINATION 1000.131A 3 Page a of.11 ENVIRONMENTAL IMPACT DETERMINATION (UNIT 1 and UNIT 2) Document No. 97-R-2018-03 RevlChange No. g Complete the following Determination. If the answer to any checklist item is "Yes", an Environmental Evaluatio is required. See Section 6.1.4 for additional guidance. Will the Activity being evaluated: Yes No O E- Disturb land that is beyond that initially disturbed during construction (i.e., new construction of buildings, creation or removal of ponds, or other terrestrial impact)? See Unit 2 SAR Figure 2.5-17. This applies only to areas outside the protected area. O E increase thermal discharges to lake or atmosphere? O B Increase concentration of chemicals to cooling lake or atmosphere through discharge canal or
. tower?
O E increase quantity of chemicals to cooling lake or atmosphere through discharge canal or tower? O 3 Modify the design or operation of cooling tower which will change drift characteristics? O S instati any new transmission lines leading offsite? O 9 Change the design or operation of the intake or discharge structures? O 2 Discharges any chemicals new or different from that previously discharged? O B Potentially cause a spill or unevaluated discharDe which may effect neighboring soils, surface water or ground water? O E involve burying or placement of any solid wastes in the site area which may effect runoff, surface water orground water? O E Involve incineration or disposal of any potentially hazardous materials on the ANO site? O E Result in a change to nonradiological effluents or licensed reactor power level? O @ Potentially change the type or increase the amount of non-radiological air emissions from the ANO site. { I
L ~~ ARKANSAS NUCLEAR ONE FORM TITLE: FORM No. REV. 10CFR40.89 REVIEW CONTINUATION PAGE 1000.131C 3 Document No. 97-R-2018-03, Rev 0. D=Tuhation 4/11 10CFR50.59 Review Continuation Paae Document Section LRS: 50.59, ANO-2 ALL (LHR, " linear heat rate", "13.3", "13.5", " energy redistribution factors, PDIL, Power dependent insertion limit, MTC, Moderator Temperature Coefficient, COLSS, Core operating limits supervisory system, DNBR, ASI, Axial Shape Index, CEAC, OOS, Out ofservice, SALARM, CWP, CEA Withdrawal Prohibit, CEA misoperation, CEA drop, CEA Ejection, deviation, deadband) MANUAL SECTIONS: ANO-2 TS 2.1.1, 3/4.2.1, 3/4.1.1.4, 3/4.1.3.1, 3/4.1.3.6, 3/4.2.4 ANO-2 SAR 4.1, 4.3.2.6, 4.5.2.3, 4.5.3.5, 15.1 ANO-2 COLR ALL NRC SER 190 ANO-2 TS Bases 3/4.2.1, 3/4.1.1.4, 3/4.1.3, 3/4.2.4 FIGURES: ANO-2 COLR ALL Figures SUPPLEMENTAL INFORMATION ANO-2 Technical Specification 6.9.5 requires that the core operating limits be established and documented in the Core Operating Limits Report (COLR) for each reload cycle. The following is a summary of the changes from the Cycle 13 COLR for Cycle 14. The Linear Heat Rate with COLSS out of service has been revised from 513.3 k to 513.5 kW/ft.
j
' avuwww.w nuws.com vmc FORM TITLE:
! FORM No. REV. l 10CFRSS.50 REVEW continuation PAGE 1900.131c 3 Document No. 97-R 2018-03, Rev 0, Dewmination 5/11 l L e The positive COLR Moderator Temperature Coefficient Limit line has been raised to accommodate more positive MTC values in Cycle 14 (Figure 1). ! e Banks 6 'and P Power Dependent Insertion Limits versus Thermal Power differs from the Cycle 13 PDIL in that it removes the prohibition on simultaneous insertion of
. Banks 6 and P above 80% power except during a down power (Figure 3).
. DNBR Margin Operating Limit Based on Core Protection Calculators for COLSS Out of Service, neither CEAC Operable was revised to make it consistent with the non-LOCA Transient Analysis requirements (Figure 5).
General editorial changes. The methodologies section in Cycle 13 COLR was deleted, since this information is an exact copy of Technical Specification 6.9.5.1. Deletion of this information from the COLR does not result in a change to the overall information present in the LBDs. The proposed change does not require 10CFR 50.59 Evaluation per Attachment 1, item C, " Incorporation ofinformation submitted to and approved by the Commission", of procedure 1000.131. The blank page that contained Figure 4 in Cycle 13 was deleted. This and the above change resulted in editorial changes of i , renumbering sections, figures, and page numbers. These changes do not require 10CFR l 50.59 Evaluation per Attachment 1, item A, " Editorial changes in text, tables or figures", of procedure 1000.131. Linear Heat Rate For Cycle 14, the limit on Linear Heat Rate with COLSS out of senice has been increased from s 13.3 kW/ft to s 13.5 kW/ft. This change reverse the 0.2 kW/ft reduction made for Cycle 13 to compensate for an error found in the large break LOCA energy redistribution factar (ERF). The error in the ERF was identified by ABB-CE in a 10CFR21 report to the NRC in ' August of 1997. In order to suppon Cycle 13 operation, a preliminary evaluation of the effects of the error demonstrated that a 0.2 kW/ft reduction of the linear heat rate limit would offset the error. The COLR was therefore revised to reflect this reduction. Subsequent evaluations of the error determined that cycle specific conservatisms in the ANO-2 Cycle 13 core design more than offset the effects of the error, without reliance on the reduced linear heat rate limit. Specifically, the pin-to-box ratio (a parameter which directly affects the ERF) for Cycle 13 was much higher than the minimum pin-to-box ratio used in the LOCA analyses. The additional margin in this ratio more than offset the erTor in the ERF.
ARKANaAS NUCLEAR ONE FORM TITLE: FORM No. REV. 10CFRSO.59 REVIEW CONTINUATION PAGE 1000.131C 3 Document No. 97-R-201 s o3, Rev 0, Detzs=uion 6/11 ABB-CE has determined a new minimum pin-to-box ratio (ABB-CE Letter ST-97-0551 that compensates for the ERF error. This new ratio has been established as a limit for the evaluation of the ANO-2 Cycle 14 core design. The Cycle 14 pin-to-box ratio is greater than the minimum value. This adequately compensates for the ERF error, such that the linear heat rate limit may be restored to the 13.5 kW/ft value used in the LOCA analyse MIC The Cycle 14 core design resulted in a slightly more positive MTC at 50% power than the target value of 0, which was the case for Cycle 13 To account for this the COLR upper moderator temperature coefficient (MTC) limit for Cycle 14 was selected to be slightly more positive than the limit in the Cycle 13 COLR. The most negative MTC remains the same as the previous cycle value. All analyses for Cycle 14 was done at the appropriate Cycle 14 COLR MTC limits. The results of all Cycle 14 ANO-2 licensing AORs were found to be bounding and acceptable, which verifies the acceptability of the COLR MTC. Cycle 14 RAR 50.59 provides more inforination about the COLR MTC. PDIL The change from the Cycle 13 is that the prohibition of simultaneous insertion ofBanks 6 and P above 80% power except during a down power is removed. The allowed PDIL limits in the COLR define the initial condition for the CEA misoperation analysis. Two events, the CEA Withdrawal Within Deadband analysis and the Slow Trip CEA Ejection analysis, were mainly used to verify the acceptability of the new PDIL in the COLR in Cycle 14. In the CEA withdrawal analysis the magnitude of the deadband was reduced by reducing the CEA deviation within subgroup alarm setpoint given by the CEAC (SALARM) from 5" to 4". The impact of this change would be a reduction in the margin to the alarm and also provides a CEA Withdrawal Prohibit for certain CEDMCS control modes. The uncenainty calculation for CEA deviation analysis (Reference 4), reduces the uncertainty in position detcrmination from 3.955" to 2.32 ". This is done by reducing the RSPT resistance and deadband errors and also by using more realistic but conservative values (current values are too conservative) for the drift and calibration tolerance for the RSPT Power supply. More restrictive values for the RSPT power supply allowable drift (i20 mV) and calibration telerance (il5 mV) is proposed in the analysis. This calibration tolerance is already incorporated into the CPC Triannual Channel Functional Test procedures. Changing SALARM and the CEA position uncertainty is a more restrictive change. This change is
' incorporated in the above analysis. An operator action of re-aligning the CEAs within 15 minutes of the deviation alarm is credited in the analysis Procedures will be modified to ensure that this will be the case.
9
p ~ ~ . = .. FORM TITLE: FORM No. REV. 10CFRSS.80 REVRW CoNTWuATIoN PAGE it00.131C 3 Document No. 97-R-201s.03, Rev 0, Detenmnation 7/11 The CEA withdrawal within the deadband safety analysis with the unrestricted PDIL for Cycle 14 showed that there is sufficient initial thermal margin, or the Required Over Power
- Margin, for not resulting in fuel pins violating DNBR SAFDL. The slow trip ejection analysis showed that for ejection from PDIL limits the core powers and coolant -
temperatures are such that the peak linear heat rate SAFDL is not violated. The Cycle 14 RAR 50.59 addresses the adequacy of the shutdown margin requirement and the PDIL. The footnote to this figure is modified with a clarification that the Full Out Position for CEAs implies at or above the programmed insertion limit (Amendment 24). The proposed change does not require 10CFR 50.59 Evaluation per Attachment 1, item C,
" Incorporation ofinformation submitted to and approved by the Commission", of procedure 1000.131.
I DNBR.Marain COLSS / CEACS OOS. Finure 5. The revision to this figure for Cycle 14 is more restrictive than the figure for Cycle 13. This was necessary because of an inconsistency between the margin reserved in the figure for Cycle 13 and the margin required by CEA drop event analysis. The limit line also has > additional conservatism built into it for making it bounding and cycle independent. Cycle speciSc analysis was used to justify the curves in previous cycles. The Cycle 14 core operating limits were determined using the analytical methods listed in ANO-2 Technical Specification 6.9.5.1. All the Cycle 14 core operating limits have been determined so that all applicable limits of the safety analyses for Cycle 14 are met. The changes to the Core Operating Limits for LHR, MTC, and PDIL are consistent with the Cycle 14 groundrules. The change to Core Operating Limit for DNBR Margin - COLSS out of service and neither CEAC operable - is required as a result of the Cycle 14 1
. reload analyses. l The Cycle 14 specific Reload Analysis Report (RAR) addresses the design and !
performance of the Cycle 14 reload core. The core design included the determination of l the limits defined in the COLR. The RAR was reviewed and a separate 50.59 has been
~
prepared for it. Duis determinationfor the Cycle 14 COLR assumes that the 10CFR50.59 reviewfor the Cycle 14 RAR willdemonstrate that there are no unreviewedsafety \ questionsassociatedwith the RAR. ( l
ARKANSAS NUClaAR ONE FORM TITLE: FORM No. REV. 10CFR80.80 RIMEW CONTINUATION PAGE 1000.131C 3 Document No. 97-R-2018-03, Rev 0 Drk.etion S/11 BASES FOR DETERMINATION
- 1. Require a change to the Operating License:
1(a) Require a change to the ANO-2 Technical Specifications? No All cycle-specific limits for operation of the Cycle 14 core are located in the COLR. The remaining Technical Specification Safety Limits, Limiting Safety Settings, and Limiting Conditions of Operation (LCOs) governing the operation of the core are bounding. Therefore, no changes to the ANO-2 Technical Speci6 cation are required. 1(b) Require a change to the ANO-2 Operating License? No { All cycle-specific operating limits for the ANO-2 Cycle 14 core are within the requirements for operating the ANO-2 core as referenced or described in the ANO-2 Operating License. The speci6c values for these limits are beyond the scope of the Operating License. Therefore, no changes to the ANO-2 Operating License are required. 1(c) Require a change to the ANO-2 Confirmatory Orders? No All cycle-specific operating limits for the ANO-2 Cycle 14 core are within the requirements for operating the ANO-2 core as referenced or described in the ANO-2 Confirmatory Orders. The speci6c values for these limits are beyond the scope of the Confirmatory i Orders. Therefore, no changes to the ANO-2 Confirmatory Orders are required. i i
- 2. Require a change to SAR documents:
2(a&b) Result in information in the ANO-2 SAR or COLR being no longer true or accurate or violate a requirement stated in the document? Yes The Cycle 14 COLR reflects the cycle-specific operating limits for the ANO-2 Cycle 14 i core. The changes to the COLR described above will ensure that the Cycle 14 core is operated in a manner consistent with the assumptions used in the analyses described in the SAR. A 10CFR50.59 evaluation is required to incorporate these changes to the COLR, except for the ones excluded by the execptions described above. Changes to the SAR sections due to the actual analyses ofthe core for Cycle 14 will be addressed in the 50.59 i for the RAR.
mum.m., num . unc FORM TITLE:
. FORM No. REV.
itcFR50.59 RmnEWCOhmNuaTioN PAGE 1000,131c 3 Document N3. 97-R-2018-03, Rev 0. W .- '-
- 9/11 2(c) Result in information in the Fire Bazards Analysis being no longer true or accurate or violate a requirement stated in the document?
No The Cycle 14 COLR addresses the cycle-specific operating limits for the ANO-2 Cycle 14 core. The specific values for limits are beyond the scope of the FHA Therefore, no changes to the FHA are required to support the operation of the Cycle 14 core. i 2(d) Result in information in the Bases of the Technical Specifications being no 1 longer true or accurate or violate a requirement stated in the document? No i The Cycle 14 COLR addresses the cycle-specific operating limits for the ANO-2 Cycle 14 core. These limits are within the requirements for operating the ANO-2 core as referenced or described in the Bases ofthe ANO-2 Technical Specifications. These limits do not result in invalidating any infonnation presented in the Bases of the ANO-2 Technical Specifications. Therefore, no changes to the Bases of the ANO-2 Technical Specifications are required to suppon the operation of the Cycle 14 core. 2(e) Result in information in the Technical Requirements Manual being no longer true or accurate or violate a requirement stated in the document? No The Cycle 14 COLR addresses the cycle-speci6c operating limits for the ANO-2 Cycle 14 core. These limits are beyond the scope of the Technical Requirements Manual and do not result in invalidating any information presented in the TRM. Therefore, no changes to the TRM are required. 2(f) Result in information in the ANO-2 SERs being no longer true or accurate or violate a requirement stated in the document? No The Cycle l4' COLR addresses the cycle-specific operating limits for the ANO-2 Cycle 14 core. These limits are within the requirements for operating the ANO-2 core as referenced or described in the ANO-2 NRC Safety Evaluation Reports. These limits do not result in invalidating any information presented in the ANO-2 NRC Safety Evaluation Reports. Therefore, no changes to the ANO-2 NRC Safety Evaluation Reports are required to l suppon the operation ofthe Cycle 14 core. .i
i ARKANSAS NUCLEAR ONE FORM TITLE: FORM No. REV. SecFRso.s3 REVIEW CONTINUATION PAGE 1000.131C 3 Document No. 97-R-2018-03. Rev 0, Lttermmation 10/11
- 3. Involve a test or experiment not described in the SAR? No All cycle-specific operating limits for the ANO-2 Cycle 14 core are consistent with the ,
current licensing basis documents. The proposed changes do not constitute a test, do not I invalidate any previously identified test, nor do they postulate a condition that would invalidate a previously performed test.
- 4. Result in a potentialimpact to the environment? No Page 3 of this 10CFR50.59 Determination, Environmental Impacts Checklist, verifies that there are no potential environmental impacts as a result of operating the plant with the Cycle 14 core.
5. Result in the need for a Radiological Safety Evaluation? No Operating the plant with the Cycle 14 core does not involve the processing of radioactive material outside the Auxiliary Building, Reactor Building, or the Low Level Radwaste Storage Building. In addition, a new pathway outside of the monitored ventilation or drainage pathways is not created with the operation of the Cycle 14 core. Thus, there is no need for a radiological safety evaluation. 6. Result in any potential impact to the equipment and facilities utilized for VSC activities? No The Cycle 14 COLR addresses the cycle-specific operating limits for the ANO-2 Cycle 14 core. The loading and operation of the Cycle 14 core does not require any spent fuel Ventilated Storage Cask or related activities. Therefore, no change to the equipment and facilities used for the VSC activities is required. 7(a) Involve a change under 10CFR50.54 for the QAMO? No The Cycle 14 COLR addresses the cycle-specific operating limits for the ANO-2 Cycle 14 core. The specific values for limits are beyond the scope of the QAMO. Therefore, no changes to the QAMO are required to support the operation of the Cycle 14 core. 7(b) Involve a change under 10CFR50.54 for tl.e E-Plan? No The Cycle 14 COLR addresses the cycle-specific operating limits for the ANO-2 Cycle 14 core. The specific values for limits are beyond the scope of the E-Plan. Therefore, no changes to the E-Plan are required to support the operation of the Cycle 14 core.
m, u -~ , ~ m_- m,w. FORM TITLE: FORM No. REV. 10CFR80J9 REVIEW CONTINUATION PAGE 1000.131C 3 Document No. 97-R-2018-03, Rev 0. Detenmnation 11/21 REFERENCES
- 1. Arkansas Nuclear One Unit 2 Cycle 14 Reload Analysis Report, AN-FE-0354 Rev.00, 10/16/98, File 224-03, 97-R-2018-02, Rev 0.
- 2. ANO-2 Final Ssfety Analysis Report (as of11/98 through Amendment 14).
- 3. ANO-2 Technical Specifications (as of11/98 through Amendment 192).
- 4. ABB/ Combustion Engineering Proprietary Analyses for ANO-2 Cycle 14:
A-AN-FE-0207 Revision 0, ANO-2 CEAC CEA Deviation Uncertainty Analysis for CEA Withdrawal,5/14/98. A-AN-FE-208, Revision 0, ANO-2 CEA Ejection RPI Parametrics Analysis,9/21/98. A AN-FE-232, Revision 0, ANO-2 Cycle 14 Slow-Trip CEA Ejection Analysis for 50%,20%, and 0% Powers,10/9/98. A-AN-FE-0211 Revision 0, ANO-2 Cycle 14 Physics Data for the CEA Deviation within Deadhand Analysis,7/21/98. A-AN-FE-0210 Revision 0, ANO-2 Cycle 14 Generation of Additional MTC Data and +0.308 Axial Shape,7n/98. A-AN-FE-0217, Rev 0, ANO-2 RPI Parametric of CEA Withdrawal Within Deadband Analysis, 10/19/98. AN-FE-0333, Rev 0, ANO-2 Cycle 14 DraR Reload Analysis Report,8/25/98. AN-FE 0354, Rev 0, ANO-2 Cycle 14 Final Reload Analysis Report,10/16/98. AN-FE 0361, Rev 0, ANO-2 Cycle 14 Core Operating Limits Repon Changes,11/9/98.
- 5. "ANO-2 Cycle 14 Fuel Management Information and Groundrules," 97-R-2018-01, Revision 2, September 8,1998.
- 6. " Arkansas Nuclear One Unit 2 Core Operating Limits Report," 97-R-2018-03, Rev 0, 12/10/98.
A
ARKANSAS NUCLEAR ONE FORM TITLE: FORM NO. CEV. 10cFR60JS EVALUATs3N 1000.1318 3 PC 8 Page1 of) 10CFR50.59 Eval. No. FFA)- Qc s .0 oa (Assigned by PSC) Document No. 97-R-2018-03 Rev1 Change No. 9 Title Core operatina Limits Report for Cvele 14 A WRITTEN RESPONSE PROVIDlNG THE BASIS FOR THE ANSWER TO EACH QUEST ATTACHED. EACH QUESTION MUST BE ANSWERED SEPARATELY. A SIMPLE STATEM CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES GUIDANCE FOR RESPON If the answer to any question on this form is "Yes,"then an unreviewed safety question is involved. If the answer to all questions is "No," then the proposed change does not involve an unreviewed safety question.-
- 1. Will the probability of an accident previously evaluated in the SAR be increased?
. Yes O No @
See attached discussion.
- 2. Will the consequences of an accident previously evaluated in the SAR be increased?
Yes O No @ g_ee attached discussion.
- 3. Will the probability of a malfunction of equipment important to safety be increased?
Yes O No B See attached discussion.
- 4. Will the consequences of a malfunction of equipment important to safety !
be increased? Yes O No @ l See attached discussion. \ { 5. Willthe possibility of an accident of a different type than any previously evaluated in the SAR be created? Yes O No @ See attached discussion.
- 6. Will the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR be created?
Yes O No @ See attached discussion.
- 7. Will the margin of safety as defined in the basis for any technical specification be reduced?
Yes O No @ See attached discussios
a.- - ~..wo = vac 10CFR50.59 EVALUATION 1000.1318 3 Page2of5 Document No. 97-R-2018-03 Rev./ Change No. R
#9 John T. Sankoorikal 12/22/98 Certified Reviewers Signature Printed Name Date Reviewers certification expiration date: 8/19/2000 Assistance provided by:
Printed Name Scope of Assistance Robert W. Clark Date Generalinput Todd Erskine 12/22/98 Generalinput Jacque Lingenfelter 12/22/98 Linear Heat Rate 12/22/93 1 PSC review by. 4 vh I Date: ( ,1 93 {
FORM TITLE: FORM No. REV. 19cPR50.50 RIMEW CONTINUATION PAGE 1000.131C 3 Document No. 97-R-2018 03, Rev 0, Evaluation 3/5 10CFR50.59 Review Continuation Paoe A summary of the changes being incorporated by this evaluation has been made in the determination section. The information in the determination section delineates the nee ' this evaluation. Please refer to the determination for a discussion of the changes and background with respect to these changes.
- 1. Will the probability of an accident previously evaluated in the SAR be increased?
No The changes described above will ensure that the unit is operated during Cycle 14 in a manner that is consistent with the assumptions used in the safety analyses for Cycle 14. The appropriate actions required if these limits are violated are in the Technical Specifications and are not being changed. The changes to the COLR affects only the operational limits and has no impact on the initiating events of any accident previously evaluated in the SAR. Therefore, the probability of an accident previously evaluated in the SAR will not be increased.
- 2. Will the consequences of an accident previously evaluated in the SAR be increased?
No l The changes described above will ensure that the unit is operated during Cycle 14 in a manner that is consistent with the assumptions used in the safety analyses for Cycle 14. The appropriate actions required if these limits are violated are in the Technical Specifications and are not being changed. Increasing the linear heat rate limit will not result in a change to
. the evaluated consequences of the accident. The analyses for the reload has been performed with NRC approved methodology to ensure that SAFDLs will not be violated and the' dose consequences are bounded by the results in the licensing basis analyses.
Therefore, the consequences of an accident previously evaluated in the SAR will not be increased. L
,-.~..=.v.-c FORM TITLE:. .
FORM No. REV.
- 14CFR80.58 REMEW CONTINUATION PAGE 1000.131C 3 Document NA 97-R 201s43, Rev 0, Evaluation 4/5 3.
. Will the probability of a malfunction of equipment important to safety be increased?
No The changes to the COLR affect only the operational limits and ensure that the core is operated consistent with the assumptions used in the analyses. The changes described above do not involve any changes in equipment. These changes will alter the manner in which the unit _is operated; however, the function and duty of the equipment important safety is not altered. These changes do not affect the initiators to any event defined in the SAR. Therefore, the probability of a malfunction of equipment important to safety will not be increased. 4. Will the consequences of a malfunction of equipment important to safety be increased? No The changes described above do not require any changes to the assumptions concerning equipment availability or failure modes. These changes do not involve any changes in equipment. In addition, these changes do not impact negatively the overall function or du of the equipment important to safety. These changes will not in result in a change to the evaluated consequences of the accidents, which also included consideration of all relevant equipment malfunctions. Therefore, the consequences of malfunction of equipment important to safety will not be increased. 5. Will the possibility of an accident of a different type than any previously evaluated in the LBD be created? No The changes described above will ensure that the unit is operated during Cycle 14 in a manner that is consistent with the assumptions used in the Cycle 14 safety analyses. These changes do not create an additional failure mode than what has already been analyzed. No initiators to any of the accidents are impacted by this modification. No new operating conditions or plant configurations are created that could lead to an accident of a different type than any previously evaluated in the SAR. Therefore, the possibility of an accident of a different type than any previously evaluated in the SAR will not be created.
j NUCLEAR ONE FORM TITLE: FORM NO. REV. 10CFRSS.SS REVWW CONTINUATION PAGE 1000,131C 3 P--r _t No. 97-R-2018-03, Rev 0, Evaluation 5/5 6.- Will the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR be created? No No changes in the failure modes of the equipment important to safety are assumed in the changes described above, No initiators to any of the accidents are impacted. No new operating conditions or plant configurations are created that could lead to a malfunction of equipment of a different type than any previously evaluated in the SAR. Therefore, the possibility ofmalfunction of equipment important to safety of a different type than previously evaluated in the SAR will not be created. 7. Will the margin of safety as defined in the basis for any technical specification be reduced? No The changes described above will ensure that the unit is operated during Cycle 14 in a manner that is consistent with the conservative assumptions used in the Cycle 14 safety analyses. The analyses were performed consistent with the requirements of Technical Specifications and COLR limits and demonstrate that acceptance limits approved by the NRC are not exceeded. The changes described above do not modify the limits or the
. margin to the limits for these parameters as described in the bases of the Technical Specification. The results ofthe reload analyses that utilized the COLR limits were found to be acceptable with respect to the margin described in the bases for these specifications.
The COLR limits assure that SAFDLs are not violated for AOOs and the effects of accidents are within acceptable limits. Therefore, the margin of safety as defined in the basis for any ANO-2 technical specification will not be reduced. i l l l l i i i
l ARKANSAS NUCLEAR ONE Page 1 FORM TITLE: FORM NO. REV. 10CFR50.59 DETERMINATION 1000.131A 3 PC-1 l This Document contains 8 Pages. Document No. CR-2 97-0199 CA#3 Rev> Change No. O Titie CHANGE TO PENETRATION PROTECTION WORDING IN THE UNIT 2 SAR Brief description of proposed change: See attached. Will the proposed Activity:
- 1. Require a change to the Operating License including:
Technical Specifications (excluding the bases)? YesO No@ Operating License? YesO No@ Confirmatory Orders? YesO No@ 2. Result in information in the following SAR documents (including drawings and text) being (a) no longer true or accurate, or (b) violate a requirement stated in the document: SAR (multi-volume set for each unit)? Yes@ NoO Core Operating Limits Report? YesO No@ Fire Hazards Analysis? YesO No@ Bases of the Technical Specifications? YesO No@ Technical Requirements Manual? YesO No@ NRC Safety Evaluation Reports? YesO No@
- 3. Involve a test or experiment not described in the SAR7 (See Attachment 2 for guidance) YesO No@
- 4. Result in a potentialimpact to the environment? (Complete Environmental impact Determination of this form.)
YesO No@
- 5. Result in the need for a Radiological Safety Evaluation per section 6.1.57 YesO No@
- 6. Result in any potential impact to the equipment or facilities utilized for Ventilated Storage Cask activities per Section 6.1.67 YesO No@
- 7. Involve a change under 10CFR50.54 for the following SAR documents per Section 6.1.77 QAMO?
YesO No@ E-Plan? YesO No@
t ARKANSAS NUCLEAR ONE FORM TITLE: PaDe 2 FORM N3. REV. 10CFR60.69 DETERMINATION 1000.131 A 3 PC-1
\
Document No. CR-2-97-0199 CA#3 Rev1 Change No. 0 Basis for Determination (Questions 1,2, & 3): See attached. O Proposed change does not require 10CFR50.59 Evaluation per Attachment 1, item # (if checked, note appropriate item #, send LDCR to Licensing). Search Scope: List sections reviewed in the Licensing Basis Documents specified in questions 1,2 and 3. If a keyword search was done on LRS, all" may be entered under"Section" with the keyword (s) used in parentheses. Controlled hard copies of the documents shall be reviewed (LRS is not verified and searches only text, not figures or draw Attach and distribute a completed LDCR per Section 6.1.2 if LBD changes are required. Document S9ction LRS: All (maanetic. Denetration w/10 orotection, thermal-maanetic. thermal maanetic. thermal-maa. thermal maa) MANUAL SECTIONS: Unit 2 SAR Sections 8.3.1.1.13 and 8.3.1.2. FIGURES: w oL DA Ce fied Reviewers Signature 7 & Printed l n 0.16Name tJnhe D' ate Reviewers certification expiration date: 04-2.3 '/7 Assistance provided by: Printed Name Scope of Assistance Brad Risner LRS Search Date Monte Morris LRS Search Search Sco Review c ptability (NA, if performed by Technical Reviewer per 1000.006) 0;0 A 94vid A. Lo//wcw C;rtified Reviewers Signature 2/1U9% Printed Name Date
. _ - _ _ _ _ _ _ - - - _ ~ - - - -
ARKANSAS NUCLEAR ONE Page 3 FORM TITLE: FORM NO. REV.
- 10CFR50.59 DETERMINATION 1000.131 A 3 ENVIRONMENTAL IMPACT DETERMINATION (UNIT 1 and UNIT 2)
Document No. CR 2-97-0199 CA#3 Rev1 Change No. O Complete the following Determination. If the answer to any item below is "Yes", an Environmental Evaluation is required. See Section 6.1.4 for additional guidance. Will the Activity being evaluated: Yes No O @ Disturb land that is beyond that initially disturbed during construction (i.e., new construction of buildings, creation or removal of ponds, or other terrestrial impact)? See Unit 2 SAR Figure 2.5-17. This applies only to areas outside the protected area. O E increase thermal discharges to lake or atmosphere? O B increase concentration of chemicals to cooling lake or atmosphere through discharge canal or tower? O E increase quantity of chemicals to cooling lake or atmosphere through discharge canal or tower? O E Modify the design or operation of cooling tower which will change drift characteristics? O E instati any new transmission lines leading offsite? O @ Change the design or operation of the intake or discharge structures? O E Discharges any chemicals new or different from that previously discharged? O @ Potentially cause a spill or unevaluated discharge which may effect neighboring soils, surface water or ground water? ' O B Involve burying or placement of any solid wastes in the site area which may effect runoff, surface water or ground water? O E involve incineration or disposal of any potentially hazardous materials on the ANO site? O E Result in a change to nonradiological effluents or licensed reactor power level? O O Potentially change the type or increase the amount of non-radiological air emissions from the ANO site. 1
CR 2-07-0199 CA#3, Rev. o ARKANSAS NUCLEAR ONE Page 4 FORM TITLE: FORM NO. REV. 10CFR50.59 SAFETY EVALUATION 1000.131B 3 Document No. CR 2-97-0199 CA#3 Rev1 Change No. 0 10CFR50.59 Eval. No. P M-9f 010 (Assigned by PSC) Title Chance to Penetration Protection Wording in the Unit 2 SAR A WRITTEN RESPONSE PROVIDING THE BASIS FOR THE ANSWER TO EACH QUESTION MUST BE ATTACHED. EACH QUESTION MUST BE ANOWERED SEPARATELY. A SIMPLE STATEMENT OF CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES GUIDANCE FOR RESPONSE. If the answer to any question on this form is "Yes," then an unreviewed safety question is involved. If the answer to all questions is "No," then the proposed change does not involve an unreviewed safety question.
- 1. Will the probability of an accident previously evaluated in the SAR be increased? YesO No @
- 2. Will the consequences of an accident previously evaluated in the SAR be increased?
YesO No @
- 3. Will the probability of a malfunction of equipment important to safety be increased?
YesO No @
- 4. Will the consequences of a malfunction of equipment important to safety be increased?
YesO No @
- 5. Will the possibility of an accident of a different type than any previously evaluated in the SAR be created?
YesO No @
- 6. Will the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR be created?
Yes O No @
- 7. Will the margin of safety as defined in the bases for any technical specification be reduced?
YesO No @ 0A -
$ b lt m n O u N3 2- /d '} d y Certified Reviewers Signature Printed Name '
D6te R: viewers certification expiration date: 0 t- 2. 3 ') 'l Assistance provided by: Printed Name Scope of Assistance AA o s, i e. Annyvn1 Date Lp. + So v 5 C . 5 */ w, A e t,f 'L A s.Je a f h PSC review by: 4 J - Date: L t/
)
CR 2 07.o199 CAe3, Rev. O ARKANSAS NUCLEAR ONE Paw 5 - FORM TITLE: FORM NO. REV.
~ 10CFR50.89 REVIEW CONTINUATION PACE 1000.131C 3 Document No. CR 2 97-0199 CA#3 Rev1 Change No. 0 10CFR50.59 Review ContinWien Pane Discussion CR-2-97-0199 was written to document that Procedure 2307.008 (Unit 2 Containment Penetration Protective Device Testing) did not comply with the applicable calculations (94-E-0018-01 Rev. 2,85-E-0118.01 Rev. 2, 84-E-0103-47 Rev. 6, and 84-E-0103-01 Rev. 7). The conflicting information involved 480V breaker sizes, types (thermal-magnetic or magnetic only), and setpoints. There were four discrepancies which required a field walkdown to determine the actual plant configuration. As documented by Action item 01 of CR-2-97-0199, Design Engineering evaluated the as-found configuration of these four breakers and determined that they were operable. Although all four breakers met the required protection for the containment penetrations, two of the breakers did not conform to the specific wording found in Section 8.3.1.2 of the Unit 2 SAR. The SAR states that backup protection for containment penetration circuits is provided by a thermal-magnetic breaker in series with j
- the primary MCC breaker. Note that the primary MCC breakeris normally a magnetic-only breaker. Design Engineering's evaluation concluded that backup protection could be provided by a magnetic-only breaker (vs. a thermal-magnetic breaker) as long as the primary and backup breakers in series provide the necessary protection. The electrical penetrations are required to maintain their integrity as an isolation barrier for the containment building. This integrity could be challenged by high currents of long duration. Therefore, the primary and backup circuit breakers are required to limit the current's magnitude and duration and thus maintain the integrity of the penetration. Part of the reply to Action item 03 of CR 2-97-0199 is to revise the Unit 2 SAR to state that in general one of the two series circuit breakers is a thermal magnetic breaker, but that whether a thermal-magnetic or magnetic oaly breaker is used, at least one of the breakers provides reasonable overload protection and both provide shoit cin: ult protection. This reply reflects the intent presented in ANO's response to the NRC by correspondence dated August 20,1993, in that correspondence, ANO stated that for the series breakers which protect electrical penetrations, at least one of the two breakers provides reasonable overload protection and both provide maximum short circuit protection. Therefore, it is Design Engineering's position that with proper application of the breakers and their settings, the requirrd protection will be provided regardless of the combination of breaker types used and therefore it is not necessary that the backup breaker be a thermal-magnetic breaker.' in conclusion, this SAR revision does not alter the degree of protection for the electrical penetrations. Basis for Determination
- 1. Technical Specifie** ions The Tech Specs were reviewed to detumine if the proposed change to the SAR would require changes to these documents. Although the penetration protective devices are discussed in Tech. Spec. 3.8.2.5, the type of overcurrent device being used (thermal magnetic or magnetic only) is not stated in the Tech. Spec.
No other references to the electrical penetration protective devices were found. Therefore, the Tech. Specs do not require any changes. Operatino License The Operating License was reviewed to determine if the proposed change to the SAR would require changes to these documents. No reference to the type of overcurrent device being used for electrical
. penetration protection was found in these documents.
. Confirmatory Orders The Confirmatory orders associated with Unit 2 were reviewed to determino if the proposed change to the SAR would require changes to these documents. No reference to the type of overcurrent device being l
l
i j CR 2-07-0190 cAf3, Rev. o ARKANSAS NUCLEAR ONE PaDe 6 FORM TITLE: FORM NO. REV. 10CFR50.89 REVIEW CONTINUATION PAGE 1000.131C 3 used for electrical penetration protection was found in these documents. Therefore, no revision to the CO's is required.
- 2. Core Operatino Limits Report The changes made to the SAR do not affect the com Operating Limits Report and thus no changes are required. )
SAR Unit 2 S.C mction 8.3.1.2, item C under Regulatory Guide 1.6.3 regarding electrical penetrations will be revised to kate, "in general, one of the two breakers used for electrical penetration protection of circuits supplied from 480 V MCC's will be a thermal-magnetic type. However, whether a thermal-magnetic or magnetic only breakeris utilized, one of the two breakers will provide reasonable overload protection and both will provide maximum short circuit protection." No other references to the type of overcurrent protection used in these circuits were found in the SAR. FHA The Fire Hazards Analysis does not describe the type of overcurrent devices used for electrical penetrations. Therefore, no changes are required to the FHA. Tech Spec Bases The Tech Spec Bases for Tech Spec. section 3/4.8 does state that the molded case protective devices utilize magnetic or thermal-magnetic overcurrent elements, but does not specify that one of the series breakers must be a thermal-magnetic type. No other reference to penetration protective devices could be found in the Tech Spec Bases, thus, no change is required. S_gg These documents were reviewed to determine if the proposed change to the SAR would require any additional changes to these documents. No reference was found in the SER regarding the type of protective device used for electrical penetration protection.
- 3. Involve a test or experiment not described in the SAR7 The proposed change is a text revision only and does not involve a design change or a new test or experiment or change the requirements of a test or experiment described in the SAR.
- 7. QAMO The QA Manual does not describe the type of overcurrent devices used for electrical penetrations.
Therefore, no changes are required to the QA Manual. Ef!M! The E-Plan does not describe the type of overcurrent devices used for electrical penetrations. Therefore no changes are required to the E-Plan.
CR 2 97-0109 CA#3. Rev. o ARKANSAS NUCLEAR ONE Page 7 FORM TITLE: FORM NO. REV. 10CFR80.89 REVIEW CONTINUATION PACE 1000.131C 3
~ ~
Basis for Evaluation Question 1 Will the probability of an accident previously evaluated in the SAR be increased? Discussion Neither the breakers discussed in this Unit 2 SAR change nor the electrical penetration assemblies that they protect are credited with initiating any of the accidents evaluated in the SAR. This SAR change will not create cny new conditions that would increase the likelihood of an initiating event which is evaluated in the SAR. Question 2 Will the consequences of an accident previously evaluated in the SAR be increased? Discussion The breakers discussed in this Unit 2 SAR change provide overload and short circuit protection for the electrical penetrations. Without proper protection, it is possible that the penetration can be damaged by high current which could cause a breach in containment. This SAR change does not involve a physical change in the plant. Neither does it involve a change in the degree of protection applied to the electrical penetrations since proper application of the breakers and their settings will provide the required protection regardiess of the combination of breaker types. This change merely removes the
' restriction that the backup breaker in the penetration protection scheme be a certain type of breaker, namely, a thermal-magnetic breaker. It is the proper application of these breakers and their ::ettinos, regardless of the type of breaker used, that provides the required protection for the penetrations and thus maintains their integrity.
Thus, the consequences of a DBA along with a containment penentration cable fault is not increased with this SAR change. Therefore, since the penetrations will continue to be protected from high current, they will not fait due to a cable fault and there will be no increase in dose to the public. Quest son 3 o Will the probability of a malfunction of equipment important to safety be increased? Discussion This SAR change does not affect the degree of protection provided by the breakers for the electrical penetrations. The required degree of protection is maintained by properly sizing the breakers and selecting their setpoints and not by merely selecting a particular type of breaker. Therefore, the probability of a breach in containment due to i the failure of an electrical penetration caused by high current in not increased. Question 4 Will the consequences of a malfunction of equipment important to safety be increased? Discussion I } 1-The consequences of a malfunction of equipment important to safety will not be affected by this SAR change since the degree of protection provided for the electrical penetrations is not changed and no other equipment important to safety is affected.
CR 2 07 o100 CAs3. Rev. o ARKANSAS NUCLEAR OE PaDe B FORM TITLE: - FORM NO. REV. 10CFR$0.89 REVIEW CONTINUATION PAGE 1000.131C 3 Question 5 Will the possibility of an accident of a different type than any previously evaluated in the SAR be created? Discussion This SAR change removes the requirement that the backup breaker in the electrical penetration protection scheme be a thermal-magnetic breaker. Protection is still provided for the penetrations by property applying the breakers sizes and settings. No new accidents will be created by this changes. Question 6 Will the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR be created? Discussion There are no new types of malfunctions of equipment important to safety created by this SAR change. This is because there is no new equipment being installed and the method of selecting and applying the breakers which protect the electrical penetrations is not being changed. Question 7 , Will the margin of safety as defined in the bases for any technical specification be reduced? Discussion While the bases of the Technical Specifications state that molded case protective devices for the electrical penetrations utilize magnetic or thermal-magnetic overcurrent elements, they do not contain a margin of sa forthe protection of the electrical penetrations. 1
AmouesAs NUCUUWt ONE F,0.sM TITLE: seCrneo.ee per;mMcNATION FORM NO. REV. 1000.131A { 3 PC.1 l Page1 of,3 Document No. LDCR SAR SECTION 10.4.9 Rev/ Change No. g
' Title ANO-2 EFW FLOW REDUCTION SAR CHANGES Brief description of proposed change:
Update the SAR to reflect the EFW desion flow reouirements and j incorporate Amendmer,t 188 results. Will the proposed Activity:
- 1. Require a change to the Operating License including:
Technical Specifications (excluding the bases)? YesO No@ Operating License? YesO No@ i Confirmatory Orders? ' YesO No@ 2. Result in information in the following SAR documents (including drawings and text) being (a) no longer true or accurate, or (b) violate a requirement stated in the document: SAR (multi-volume set for each unit)? . Yes@ nod Core Opeiating Limits Report YesO No@ Fire Hazards Analysis? YesO No@ Bases of the Technical Specifications? YesO No@ Technical Requirements Manual? YesO No@ NRC Safety Evaluation Reports? YesO No@
- 3. Involve a test or experiment not described in the SAR? l (See Attachment 2 forguidance) YesO NoE !
i
- 4. Result in a potential impact to the environment? (Complete the Environmentalimpact Determination of this form.)
YesO No@
- 5. Result in the need for a Radiological Safety Evaluation per section 6.1.57 YesO No@ ;
- 6. - Result in any potential impact to the equipment or facilities utilized for Ventilated Storage Cask activities per Section 6.1.67 YesO No@
- 7. Involve a change under 10CFR50.54 for the following SAR documents per Section 6.1.7:
QAMO? YesO No@ E-Plan? YesO No@ , Basis for Determination (Questions 1. 2 & 31: See Attached O Proposed change does not require 10 CFR 50.59 Evaluation per Attachment 1, item # . (if checked, note appropriate item #, send LDCR to Licensing).
ARKANSAS NUCLEAR ONE FORM TITLE: FORM NO. RElf. 10CFRSS.80 DETERMINATION 1000.131A 3 PC 1 Page 2 of.9 Document No. LDCR SAR SECTION 10.4.9 RevjChange No. ! Search Scope: List sections reviewed in the Licensing Basis Documents specified in Question 1,2 and 3. If a keyword s was done on LRS, "all" may be entered under "Section" with the keyword (s) used in parentheses. Controlle copies of the documents shall be reviewed (LRS is not verified and searches only text, not figures o Attach and distribute a completed LDCR per Section 6.1.2 if LBD changes are required. Document LRS; lllection All (Emergency Feed
- water, EFW, EFWS, EFS, 485, 575, 560)
MANUAL SECTIONS: SAR (Sections 10.4.9, Table 10.4-10) FIGURES: None Bryan Daiber 6-29 98 celtifietfReviewers Signature Printed Name Date ! Reviewers certification expiration date:_ 3-18-2003 Assistance provided by: Printed Name Scope of Assistance Jacove Linoenfelter Date d Drafted packaoe and assissted in SAR searches. 6-29 98 Search Scope Review Acceptability (NA, if performed by Technical Review per 1000.006) b k.42' L;d J. F.wk 1-I-98 Certified Reviewers Signature Printed Name Date
ARMANSAS NUCLEAR ONE FORM TITLE: PORM NO. REtf. 10CPRSS.88 DETERMINATION 1000.131A 3 PC-1 Page a of,2 j ENVIRONMENTAL IMPACT DETERMINATION (UNIT 1 and UNIT 2) I Document No. LDCR SAR SECTION 10.4.9 ' Rev/ Change No. 1
' Complete the following Determination. If the answer to any checklist item is "Yes", an Environmental E is required. See Se Aion 6.1.4 for additional guidance. ..Will the Activity behg evaluated:
Yes p O- 0 Disturb land that is beyond that initia!!y disturbed during construction (i.e., new construction of buildings, creation or removal of ponds, or other terrestrial impact)? See Unit 2 SAR Figure 2.5-17. This applies only to areas outside the protected area. O E Increase thermal discharges to lake or atmosphere? O E increase concentration of chemicals to cooling lake or atmosphere through discharge canal or tower? O _E increase quantity of chemicals to cooling lake or atmosphere through discharge canal or tower? O E Modify the design or operation of cooling tower which will change drift characteristics? O E Install any new transmission lines leading offsite? O E- Change the design or operation of the intake or discharge structures? O E Discharges any chemicals new or different from that previously discharged? O E Potentially cause a spill or unevaluated discharge which may effect neighboring soils, surface water or ground water? O E Involve burying or placement of any solid wastes in the site area which may effect runoff, surface water or ground water? O E Involve incineration or disposal of any potentially hazardous materials or Je ANO site? O E Result in a change to nonradiological effluents or licensed reactor power level? O E Potentially change the type or increase the amount of non-radiological air emissions from the ANO site. 1
FORM TITLE: ARKANSA"! NUCLEAR ONE 10CFR00.49 RRVIEW CONTINUATN3N PAGE FORM NO. REV. 1000.131C 3 Page e of,3 Document No. LDCR SAR SECTION 10.4.9 Rev> Change No. g 10CFR50.59 Review Continuation Paae Backeround The expected rate of EFW flow to the steam generators as a function of steam genera
- slightly since initial plant operation. His reduced flow is due to the combined e performance, system configuration, modeling of the system and the application of meas new set testing. of design basis muumum flow requirements (head curve) has been develo
' He new set ofdesign basis nummum flow requirements has been incorporated in seve analyses. The two accident acalyses that are most significantly affected by the reduced of ncrmalintofeedwater incorporated (LOFW) the SAR by nparate LDCRsewnt main feedwater line break (M{
and theReview. and 10CFR50.59 The Technical Specifications for the Emergency Feedwater System has been modified b to remove specific flow requirements. Testmg will be performed in accordance with the provisio Specification 4.0.5 and the in service testmg program EFW pump performance test accep developed from piping loss analyses to deternune the pump perfonnance needed in the te u.c design basis mmimum flow vs. SG pressure assumptions used in the accident analy will be incorporated into the operating / test procedures by a separate 10CFR50.59 Review. The SAR changes proposed in the *=Ad LDCR modify discussions of EFW flow requirem with the revised LOFW and MFLB accident analyses. He proposed changes expand the Se of the design basis capacity requirements, and clarify the distinction between design requirements and pump rated flow data. Discussion - 1 l Of all the accident analyses, the loss of normal feedwater and main feedwater line break; 1 to the mimmum assumed capacity of the emergency feedwater pumps. Other analyses m flow, but the impact of EFW flow on the other analyses is either insignificant or the cl are boimdad by the LOFW or MFLB events. As previously stated, the reduced flow assumpt MFLB events are being evaluated separately De minimal impact of the flow reduction ) , will be discussed later in this evaluation. The SAR changes proposed by the attached LDCR fo clarification of the EFW flow requirements discussed outside the accident analyses. ! i As stated in SAR Section 10.4.9.3, "The design basis requirement for the emergen that function the as pump a ==ad must be capable of delivering sufficient emergency feedwater to the ste y heat sink for nonnat shutdowns. He pump must also provide sufficient feedwater in combination with pressurized sprays or the safety valves to preclude over pressurization of the l line break accidents." he qualitative criteria for the feedwater line break accident can be ea i quantitative limits for that event. The results of the MFLB analysis must demonstrate that R remain less than 110% of design, or 2750 psig. However, quantitative EFW flow criteria for normal shutdowns can not be derived easil i requirement for preserving the ==ade y heat sink function. The current SAR states that E
a.u . am. ce me FORM TITLE: FORM NO. REV. 10CFRSO.80 REVIEW CONTINUATION PAGE 1900,131C 3 Page 5 of 9 Document No. LDCR SAR SECTION 10.4.9 Rev/ Change No. 2 to maintain steam generator level when either generator is removing decay heat equivalent to 3.5% vailue was given as both 3.5% and 2.95% in the FSAR and modified to 3.5% in Amendment 6 to the SAR. T basis for the original value may be found in a response to NRC questions from an extensive post TMI r EFW System (2CANO18024). In the response to the NRC, the requirements for EFW flow were evaluated for several postulated transie accidents. The response identified plant cooldown and the main feedwater line break as the design bas sizing the EFW pumps. Qualitative criteria, effectively the same as the SAR criteria discussed above, w presented for the MFLB event and the plant cooldown. These criteria were translated into quantitative crite the sizing of the pumps. These criteria included a flow sufEcient to remove decay heat equivalent to 2.95% power (500 gpm) at a steam generator pressure of 1220 psia. With the addition of 75 gpm to accommodate recirculation flow, this resulted in a pump rated flow of 575 gpm at a head of 2800 feet (1220 psia), wh design point on the pump head curves. Note that line losses between the pump and the steam _ genera considered in the discussion, nor were the loads from RCP heat or water and metal sensible heat addressed. Additionally, for the plant cooldown, the response presented a graph of EFW flow and steam genera function of time after shutdewn for a " representative" cooldown. This graph shows an EFW flow oI to the steam generator beguunng at about 350 seconds followmg a trip. Steam generator pressure is gi psia at this time, indicating nommal control by the steam dump and bypass control system. Notes indicate the assumption of four RCP operation. A flow of 615 gpm at a temperature of 100*F to the steam generators at a pressure of 1000 psia will remove decay heat equivalent to approximately 2.95% full power and 18A MWt pump heat. A flow of 615 gpm to the steam generators at 1000 psia is consistent with the manufacturer's pump head curve assuming line losses of about 100 psi and nominal condensate storage tank conditions. Mo:t impartantly, the response provided a clear distinction bmo the uses of the pump rated flow values, and th design basis muumum flow assumed in the accident analyses. The response identified a conservative value of 485 gpm, as the minimum flow rate assumption specifically for the accident analyses. (This value subsequently the Technical Specification flow requirement.) The response demonstrates that the evaluation of EFW p capacity to support normal cooldown was based on rated values rather than the nummum flows used in the acciden analyses. The basis for the Amendment 6 SAR change from 2.95% to 3.5% full power equivalent decay heat is no clear. Using pump rated data, 3.5% full power decay heat could be removed, with or without RCPs in oper assuming steam generator pressures at or below the nominal post shutdown values. Regardless of the basis, it clear that the value of 3.5% (and the original value of 2.95%) full power provides only a subjective demonstration the acceptability of the nominal pump rated flow. No basis was provided to establish 3.5% full power as a quantitative limit for the preservation of the waadary heat sink function. The SAR values for time from shutdown to reach decay heat levels of 3.5% (2.95%) and 2.4%, 200 (350) seconds and 10 minutes respectively, were not ' linked to any operator action or event sequence. These tunes only apply to the time required to reach the indicated decay heat levels; they do not address the other factors affectmg prunary to secondary heat removal, inclu heat, prunary and semadary water and metal sensible heat, and secondary pressure conditions. 'Ihese values were provided only as part of the subjective demonstration of the =dv=q of the pump rated flow. The use of pump rated flow values in demonstratmg adequate flow for normal cooldown is appropriate. For the evaluation of normal operating conditions, the application of conservative safety analysis type assumptions is not necessary given the assumed availability of most non-safety-grade equipment. However, a plant cooldown using safety grade equipment may follow a number of postulated transients and accidents. It may, therefore, be useful to
PORM TITLa: Ansoussas NuctaAR ONE FORM NO. REV. 10CPRSS.80 RaVIEW CONTINUATION PAes 1900.131C 3 Page 1 of_t Document No. 1.DCR SAR SECTION 10.4.9 Rev/ Change No. O L...si L- that EFW capacity is sufficient to support a plant cooldown by secondary heat removal f
- transients and accidents in which the =caad y system remains intact.
De ability of the muumum EFW flow capacity to assure maintenance of steam generator lev can be Aw.si.L.kJ by the analysis of the loss of feedwater event. With respect to maintaining sec removal capability, this event conservatively bounds the conditions ofnormal operations and other accidents which the seco.J.ty system remains intact. The analysis of the loss of feedwater event, demonstrates t generator level is maintained by one EFW pump following the loss of all main feedwater. This analysis steam generator pressure control by the SDBCS at 900 psia, continued operation of the reactor coolant the failure of the motor driven EFW pump. (A parametric study was performed which concluded that lowe generator inventones occurred for lower steady state steam generator pressures.) He design basis minimum from the turbine driven EFW pump (which has a slower response time than the motor driven pump g availability of offsite power), combined with the nummum steam generator inventory available at the start of t transient, is sufficient to assure that steam generator level and pnmary to secondary heat transfer are mainta throughout the transient. Dese results demonstrate that the EFW pump capacity is sufficient to assure ma of steam generator level and primary cooling during accident conditions which bound the conditions of normal operations. De proposed changes to SAR Section 10.4.9.3 clarify the original basis for evaluating the adequacy pump capacity for normal cooldowns The applicability of the pump rated data for this evaluation is c A quantitative summary of the limiting point (highest decay heat) of the graphical evaluation in the response (2CAN018024) has been added in place of the ambiguous discussion of decay heat levels versus tim reference to Figwe 6.3-1 A for decay heat values versus time have been replaced by the use of ANS/ ANS Although this =si+hlogy is less conservative than that used for Figure 6.3-1 A, it is an industry sta beat curve. Additionally, this is the v.eiodology noted and approved in the CENTS methodology whic in the MFLB and LOFW analyses. He paragraphs discussing the response to accidents have been rearranged to clarify the analysis of the fee line break as the limiting accident analysis. Words have been added to clarify the significance of the the analysis of the main steam line break. Since the mimmum pump capacity is not critically important to MSLB, the sentence describing decay heat levels followmg the event has been removed. One minor clari6 cation has been made to SAR Section 10.4.9.2.1. The words "At rated flow from the second sentence of the fifth paragraph. Each pump is capable of providing sufficient water at well below the pump rated flow. No other SAR changes are required to support the reduction in EFW pump muumum flow. He flow data in Table 10.4-10 is the pump rated flow which remams nachmaged. It is important to note that the chan muumum EFW flow have no signi6 cant impact on the duration of a normal plant cooldown using EFW. The EFW flow avadable at the reduced steam generator pressures required for cooldown, is sufficient to support c the maximum allowable rates. Consequently, the discussions ofcooldown durations in various sections of the SAR are not affected by the EFW minimum flow changes. In the safety analyses presented in Chapter 15, only the loss of feedwater and feedwater line break analy the design basis mmimum EFW flow. Changes to these analyses are addressed in separate 10CFR50.59 Review Other Chapter 15 safety analyses mention the response of the emergency feedwater system. However, the ac~ptable results of these analyses are not danaadant on the muumum EFW flow. These analyses include:
ARMANSAs NUCLEAR OfE FORM TITLE: FORM NO. RElf. 10CFR$0.89 REVIEW CONTINUATION PAGE 1960.is1C 3 Page Z of.9 Document No. LDCR SAR SECTION 10.4.9 Rev/ Change No. g, Uncontrolled CEA Withdrawal from Critical Con 6tian_= The inadvertent CEA withdrawal analyses are short term (relative to the long tenn cooling requi EFW) pnmary system power and pressure transient analyses. The automatic initiation of emer feedwater to assure secondary heat removal, is mentioned in the SAR presentation, but the EFW flow not considered in these analyses, ne heat loads generated by this event would not be significantly from nonnal cooldown and would be haand~i by the loss of feedwater event. Loss of All Normal and Preferred AC Power to the Statian Auxiliaries This analysis evaluates the pnmary and secondary system response to the simultaneous loss o loss of feedwater flow and loss of reactor coolant flow. He duration of this analysis includes the actuatio of emergency feedwater, and muumum EFW pump flow is assumed in this analysis (but not spe i mentioned in the SAR). However, the reduction in reactor coolant flow reduces the demand on EF capacity by reducing the rate ofprimary to secondary heat transfer and elirnia=*ing the heat added RCPs. He loss of feedwater flow analysis is essentially the same transient except that the RCPs continue to operate. With respect to EFW pump capacity, the loss of feedwater event bounds the loss of AC power event. Excess Heat Removal Due to Secondary System Malfunction One of the postulated malfunctions in this collection ofevents, is the startup of an EFW pump. As an overcooling event, the maxunum flow values are assumed. Changes to the muumum flow requirements ha no impact on this event. Maior Secondary System Pine Breaks With or Without a Concurrent Loss of AC Power (Main Steam Line Breaks) ne main steam line break is an overcooling event which is currently analyzed with maximum EFW flows. Long term cooling by the EFW system is important to this event, but the demands on the EFW pump capacity are lower than for the feedwater line break event. Changes to the muumum flow do not impact t event. Steam Generator Tube Ruoture With or Without a Concurrent Loss of AC Power For the steam generator tube rupture with a loss of offsite power, the EFW system assures amad='y h i removal during the plant cooldown. The heat loads from this event would be bounded by those from the loss of feedwater event. The changes to the mimmum EFW flow rate do not impact this event. He small break loss of coolant analyses also consider the muumum EFW flow. He minimum flow is mo Sled to reduce heat removal by the amadary system. However, the impact of EFW flow is very small and acceptable analysis results are not 4 =-%t on the assumed flow values. De SAR does not address the specific assumptions for EFW flow, so no changes are required.
ARKANaAa NUCLEAR ONE FORM TITLE: FORM NO. REV. 10CFRSS.80 REVEW CONTINUATION PAGE ' 10o0.131C 3 Page1of,1
. Document No. LDCR SAR SECTION 10.4.9 Rev1 Change No. g Basis For C;;;.
- M=E: .
.1) Require a change to the Ope..uug License . Technical Speedication 3/4.7.1.2 was modified by License A==A Et 188 to remove specific flo He EFW pump capacity is not addressed by the Operstmg License or any Confirmatory Orders. No furt changes to these documents are required to support the proposed SAR changes.
2)- Result in information in the following SAR h=ats (including drawings and text) being (a) true or accurate, or (b) violate a requirement stated in the document: De discussions of EFW pump capacity in SAR Section 10.4.9 require clarification of the d
. basis minimum flow and pump design flow values. The proposed changes to Section 10.4.9 are changes made to the scrident analyses. The pump design flow values are not changed so data in not affected. De test criteria of the original test program, in SAR Table 14.1-1 are historic data and are affected. No other sections of the SAR contain information on EFW pump capacity.
This effort is focused on SAR Section 10.4.9 and has no impact on the ~COLR, or FHA. None o addresses EFW pump capacity. The Bases ofTechnical Specification 3/4.7.1.2 was modified by A=4=^ 188 to remove specific flow requirements. EFW pump capacity is noted in the original NRC Safety Evaluation Report. Section 10.5 of the SE design point rated value of 575 spm in a descriptive discussion of the system, but did not functional criterion. De SER for Amendment 51 evaluated a new Technical Specification on E SER for A* 136 evaluated a Technical Specification change to the required steam pressure f turbine driven pump. Both of these SERs mentioned se design basis muumum ficw value of 485 g
'188 to the Technical Specifications supersedes these SERs.
- 3) Involve a test or experiment not described in the SAR?
No tests or Ewimets are proposed or affected by this change. Rese changes only relate to clarificati Section 10.4.9. t
- 4) : Result in a y**ial impact to the environment?
No plant modifications or conditions are being proposed by this change. Rese changes are re discussions of EFW pump capacity in SAR Section 10.4.9. No physical modifications to the implemented by this change.
- 5) Result in the need for a Radiological Safety Evaluation per section 6.1.57 No Safety activity involving Evaluation the processing of radioactive material is being proposed by this change is not needed 6)
Result in any pa* a'i=1 impact to the equipment or facilities utilized for Ventilated Storage Cask per Section 6.1.67 1 1 ___--__A
ARKANSAS NUCLEAR ONE FORM TITLE: FORM NO. REV. 10CPR40.59 REVIEW CONTINUATION PAGE 1900.131C 3 Page 9 of 9 Document No. LDCR SAR SECTION 10.4.9 RevlChange No. g There are no proposed activities by this change which would involve any aspect of the VSC. This propose is limited to the SAR Section 10.4.9. 7) Involve a change under 10CFR50.54 for the following SAR documents per S.:ction 6.1.7: a) QAMO E-Plan? ~
\
l There are no proposed activities by this change which would affect the QAMO or E-Plan. This propos limited to SAR Section 10.4.9 relating to the EFW flow requirements. I 1 l I I I I i l i
ARKANSAS NUCLEAR ONE FORM TITLE: FORM NO. REE
. Socrase.se EVALUATION 1000.1313 3 Page1 off 10CFR50.59 Eval. No. NO M 'OS $
(Assigned by PSC) Document No. LDCR SAR SECTION 10.4.9 Rev/ Change No. ! Title ANO-2 EFW FLOW REDUCTION SAR CHANGES A WRITTEN RESPONSE PROVIDING THE BASIS FOR THE ANSWER TO EACH Q ATTACHED. EACH QUESTION MUST BE ANSWERED SEPARATELY. A SIMPLE STA CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES GUIDANCE FOR RE If the answer to any question on this form is "Yes," then an unreviewed safety question is involved. If to all questions is "No,'" then the proposed change does not involve an unreviewed safety question.
- 1. Will the probability of an accident previously evaluated in the SAR be increased?
Yes O No @
- 2. Will the consequences of an accident previously evaluated in the SAR be increased?
Yes O No @ 3. Will the probability of a malfunction of equipment important to safety be increased? Yes O No @ 4. Will the consequences of a malfunction of equipment important to safety be increased? Yes O No @ 5. Will the possibility of an accident of a different type than any previously evaluated in the SAR be created? Yes O No @ 6. Will the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR be created? Yes O No @ 7. Will the margin of safety as defined in the bases for any technical specification be reduced? Yes O No @ l
anmans nuvLean une FORM TITLE: FORM NO. REV. 10CFR80.59 EVALUATION 1000.131B 3 Page 2 of,4 e Bryan Dalber Certified Rsviewers Signature 6-29 98 Printed Name Date Reviewers certification expiration date: 3-18-2000 Assistance provided by: Printed Name Spope of Assistance Date Jacaue 1.inoonfeNer ReMeatsino Basis. 6-29-98 PSC review by:
/c - -
/I _-
Date: 7 8 /, i . 7 1
ARKaNaAS NUCLEAR ONE FORM TITLE: FORM NO. REV. 10CFRSS.89 REMW CONTINUATION PAGE 1000.131C 3 Page 3 off Document No. LDCR SAR SECTION 10.4.9- RevlChange No. 0 10CFR50.59 Review Continuation Pane A summary of the changes being covered by this evaluation are delineated in the determmation. Re detennination for more information with respect to the changes being made and the background in 1. Will the probability of an accident previously evaluated in the SAR be increased? No The proposed changes clarify the SAR presentation of EFW pump capacity requirements. There ar systems, components, Kbstructures, design changes, physical alterations, or operating proced proposed by this change. The capacity requirements of the EFW pumps are in no way related t precursor. evaluated in theHe clarification SAR. of these requirements will have no impact on the probability of 2. Will the consequences of an accident previously evaluated in the SAR be increased? No
%e proposed changes clarify the SAR presentation of EFW pump capacity requirements. Here are systems, components, substructures, design changes, physical alterations, or operating procedu proposed by this change. He changes support a reduction of the design basis mmimum flow requ been evaluated for accident analysis impact in separate 10CFR50.59 Reviews. The evalua to support normal cooldown is not part of the accident analyses and has no defmable consequences. N the proposed changes clanfy the EFW capacity requirements for cooldown and demonstrate that the r design basis mimmum flow does not alter the conclusion the EFW system can effectively support norm The change in decay host models quoted for EFW flow verification is a conversion to a newer approv maeadalogy and consistent with the accident analysis methodology. His method is still conserv to the actual decay heat expected following reactor trip. Additionally, the timmg for the decay heat values provided only as part of a subjective aw-%. tion of the EFW pump adequacy. Therefore, the conse accidents previously evaluated in the SAR are unchanged by the proposed changes.
3. Will the probability of a malfunction of equipment important to safety be increased? No The proposed changes clarify the SAR presentation of EFW pump capacity requirements. Here are no new systems, v=i=-4, substructures, design changes, physical alterations, or operating procedure changes bei proposed by this change. He changes support a reduction of the design basis mmimum flow requirements been evaluated for accident analysis impact in separate 10CFR50.59 Resiews. The acceptable results of accident analyses Lawn ie that the reduction in the mimmum flow produces no new or different ope conditions for equipment important to safety. Similarly, the proposed changes demonstrate that the reductio pump mimmum flow produces no new or different operating conditions for equipment important to saf support of normal cooldowns. Consequently, the probability of a malfunction of equipment importan not be increased 4. Will the consequences of a malfunction of equipment important to safety be increased? No De proposed changes clarify the SAR presentation of EFW pump capacity requirements. There are no new systems, ca..gs..s, substructures, design changes, physical alterations, or operating procedure changes b proposed by this change ne changes support a reduction of the design basis minimum flow requirements th been evaluated for accident analysis impact in separate 10CFR50.59 Reviews. The acceptable results of th b
.-- -== wuctaAR ONE FORM TITLE.
FORM NO. REtf. 10CPRSS.80 REVIEW CONTINUATION PAGE 100s.131C 3 PaSe f of 4 Document No. LDCR SAR SECTION 10.4.9 Rev/ Change No. 1 accident analyses, which consider all appropriate single failures (malfunctions of equipment demonstrate that the reduction in the muumum flow produces no new or different operating important to safety. Similarly, the proposed changes demonstrate that the reduction of the pum produces no new or different operatmg conditions for equipment important to safety in the support cooldowns. Consequently, the probability of a malfunction of equipment important to safety will not be 5. Will the possibility of an accident of a different type than any previously evaluated in the SAR be created 7 No No new initiators or accidents are caused by this change. He proposed changes clarify the SAR p EFW pump capacity requirements. The changes support a reduction of the design basis muu that have been evaluated for accident analysis impact in separate 10CFR50.59 Reviews. He acce the revised accident analyseo, which consider all appropriate single failures (malfunctions of eq safety), h,eionnie that the reduction in the mmimum flow produces no rew or different operatmg cond plant modi 6 cations, new components, physical alterations, nor operatmg conditions are being change; therefore no new accidents are created and no currently non-limiting events are beram 6. Will the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR be created? No There are no new systems, components, substmetures, physical design changes, physical alt procedure changes being proposed by this change. He proposed changes clarify the SAR presentatio pump capacity requirements. As there are no physical changes to the plant, the possibility of a malfunction o equipment important to safety of a different type than any previously evaluated in the SAR will not be created 7. Will the margin of safety as defined in the bases for any technical specification be reduced? No De proposed changes clarify the SAR presentation of EFW pump capacity requirements. There are no new systems, %-==ts, substructures, design changes, physical alterations, or operatmg procedure changes bein proposed by this change. De changes support a reduction of the design basis mmimum flow requiremen been evaluated in separate 10CFR50.59 Reviews. EFW flow requirements were removed from the bases ofTechnical Speci6 cation 3/4.7.1.2 by Amendme EFW pump capacity requirements are not mentioned in any other technical specification or technica bases. De margin of safety as denned by the bases for the technical speci6 cations are unaffe' I 4 l
, PC-963090P2o1 Rev.o ARKANSAS NUCLEAR ONE FORM TITLE: Peoe 1 FORM NO. REV. 10CFR50.59 DETERMINATION 1000.131A 2 This Document contains 3 Pages. Document No. ER963090P201 Rev1 Change No. O Title Sparing 2UAV 3207 Air Maintenance Device Air Supply Connection Will the proposed Activity:
- 1. Require a change to the Operating License including:
Technical Specifications (excluding the bases)? YesO No@ Operating License? YesO NoS Confirmatory Orders? ! YesO NoS i 2. Result in information in the following SAR documents (including drawings and text) being (a) no longer true or accurate, or (b) violate a requirement stated in the document: SAR (multi-volume set for each unit)? Yes@ NoO QAMO?* YesO No@ E-Plan? - YesO No@ FHA i YesO NoS Bases of the Technical Specifications? . YesO No@ NRC Safety Evaluation Reports? YesO No@
- 3. Involve a test or experiment not described in the SAR?
YesO No@
- 4. Result in a potentialimpact to the environment? (Complete Environmental impact Checklist of this form.)
YesO No@
- 5. Result in the need for a Radiological Safety Evaluation per section 6.2.4.a?
YesO No@
- 6. Result in the need for a 10CFR72.48 Review per section 6.2.4.b?
YesO No@ Basis for Determination: The air supply to 2UAV-3207 air maintenance device is being spared in place at 2SA-206 since supervision is no needed per NFPA 13. The control room annunication for panel 2C22-F8 (Supervisory air press low) will also be spared. This will be reflected on drawings that appear in the SAR. Changes to these documents require an evaluation in accordance with 10CFR50.54. See Section 6.2.1.B.
rewn ~' rscv. v Pc-a63090P201, Rev. o ARKANSAS NUCLEAR ONE Page 2 FORM TITLE: FORM NO. REV. 10CFR50.59 DETERMINATION 1000.131 A 2 Document No. ER 963090P201 Rev/ Change No. O
References:
Ust sections reviewed in the Licensing Basis Documents, specified in questions 1,2 and 3. If a keyword search was done on LRS, "all* may be entered under "Section" with the keyword (s) used in parentheses. Controlled hard copies of the documents shall be reviewed as computer based searches such as LRS are not controlled and search text only, not figures or drawings. Attach a I completed LDCR if LBD changes are required. Document Section Tech Spec, Operating Ucense &
*All", " Air Maintenance Device",
- Containment Suppression System',
Confiramatory Orders "2UAV 3207' SAR, QAMO, E-Plan, FHA, Tech Spec *All", Same as above Bases & NRC SE l Nw D .bMe Thomas D. Robinson 6/26/97 Certified Reviewers Signature Printed Name Date Reviewers certification expiration date: 5/5/99 Assistance provided by: Printed Name Scope of Assistance Date 1
l PC-963090P201, Rev. O ARKANSAS NUCLEAR ONE Page 3 FORM TITLE: FORM NO. REV. 10CFR50.59 DETERMINATION 1000.131A 2 ENVIRONMENTAL IMPACT CHECKLIST (UNIT 1 and UNIT 2) Document No. ER 983090P201 Rev1 Change No. O Complete the following checklist. If the answer to any checklist item is "Yes", an Envimnmental Evaluation is required. See Section 6.2.1.E for additional guidance. ; Will the Activity being evaluated: I X.tk Ut 1 O E Disturb land that is beyond that initially disturt>ed during construction (i.e., new construction of buildings, creation or removal of ponds, or other terrestrial impact)? See Unit 2 SAR Figure 2.5-17.' This applies only to areas outside the protected area. 4 O E increase thermal discharges to take or atmosphere? O E increase concentration of chemicais to cooling lake or atmosphere through discharge canal or l' tower? O E increase quantity of chemicais to cooling lake or atmosphere through discharge canal or towct? -- O E Modify the design or operation of cooling tower which will change drift characteristics? O E install aniy new transmission lines leading offsite? O E Change the design or operation of the intake or discharge structures? O E oischarges any chemicais new or different from that previously discharged? O @ Potentially cause a spill or unevaluated discharge which may effect neighboring soils, surface water orground water? O E involve burying or placement of any solid wastes in the site area which may effect runoff, surface water or ground water? O E involve incineration or disposal of any potentially hazardous materials on the ANO site? O E Result in a change to nonradiological effluents or licensed reactor power level? O E Potentially change the type or increase the amount of non-radiological air emissions from the ANO site. l
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- ARKANSAS NUCLEAR ONE FORM TITLE.
FORM NO. N EVALMA M 1000.1315 REV. i s l l Page i of f j 10CFR50.59 Eval.No. FFN OrM35 (Assigned by PSC) Document No. ER 963090P201 RevlChange No. A j Title Soarina 2UAV Air Maintenance Device Air SessN Conasctlen i A WRITTEN RESPONSE PROVIDING THE BASIS FOR THE ANSWER TO EACH ATTACHED. EACH QUESTION MUST BE ANSWERED SEPARATELY. A SIMPLE S CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES GUIDANCE FOR I If the answer to any question on this form is "Yes," then an unreviewed safety question is involved. to all questions is "No," then the proposed change does not involve an unreviewed safety question. 1
- 1. Will the probability of an accident previously evaluated in the SAR be increased?
Yes O No @
- 2. Will the consequences of an accident previously evaluated in the SAR be
- Increased?
Yes O No @ 3. Will the probability of a malfunction of equipment important to safety be increased? Yes O No @ 4. Will the consequences of a malfunction of equipment important to safety be increased? Yes O No @ 5. Will the possibility of an accident of a different type than any previously evaluated in the SAR be created? Yes O No @
- 6. Will the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR be created?
Yes O No @
- 7. Will the margin of safety as defined in the bases for any technical specification be reduced?
Yes O No @ l l l
ARKANSAS NUCt. EAR ONE FORM TITLE: FORM NO. REV. 10CFR80.88 EVMM 1000.1315 2 Nws b.Sh> Thomas D. Robinson 6/26/97 Certified Reviewers Signature Printed Name Date Reviewers certification expiration date: 3/5/99 Assistance provided by: 1 Printed Name
- Sco ofAssistance Date PSC review by:
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- ARKANSAS NUCLEAR ONE FORM TITLE.
FORM NO. REV. 10CFRse.s0 REVEWCoNTINUATioN PAGE 1ees.131C 2 Page i of 3 Document No. ER 963090P201 Rev> Change No. 1 10CFR50.59 Review Continuation Paae Backoround: The containment cable spreading area suppression systems (one system per cable spreadi equipped with an air maintenance device that is designed to be used to notify the control room in the ev ' damaged head. However, the system is isolated (ref. procedure 1015.016F, " Shift tumover Modes 5 & 6) except in modes 5 & 6. In the event of a fire, the cable spreading area detection system the control room of fire conditions who would visually verify the fire before tripping the suppression needed. Since access to containment is very limited in modes 14, it is highly unlikely that a sprinkler hea be damaged during those modes of operation. Since the sprinkler systems are in an area that accessible, it is highly unlikely that a sprinkler head could be damaged in modes 5 & 6. NFPA 13 requ maintenance device on sprinkler systems with more than 20 heads in areas that are easily accessible o , area that is not in a normal traffic area. If a sprinkler head were damaged, air would escape and cause an a in a remote occupiedytation,(contpl room). This mod will also spare the control room annunciatio 2C22 (Annunciation 2K22, ' Supervisory Air Pressure Low"). i Since it is highly unlikely that a head could be damaged in any operating mode and the system isolated and the NFPA code does not require it, the air supply is being disconnected. Evaluation:
)
1. Will the probability of an accident previously evaluated in the SAR be increased? There are no accidents involving fires in the containment building that have Mtc malyzed in the SAR fire is not a licensing bases accident. Thus, there are no frequency class hges as a result of this modification and no change in frequency class nor significant move in one frequency class. Thus, the probability of an accident previously evaluated in the SAR will not be increased. 2. Will the consequences of an accident previously evaluated in the SAR be increased? The purpose of the Containment suppression system is to protect the cable spreading areas and water to the containment hose reels. The system is normally isolated (2UAV 3200 closed) except in modes 5 & 6 and when a fire in Containment is verified. By disconnecting the air supply to the air maintenance device the ability of the cable spreading area sprinkler system will not be hinder nor prevented from operating nor will the containment isolation valve be prevented from operating in the event of a containment isolation signal. Thus, the offsite dose as a result of an accident previously evaluated wi
- ARMANsAs NUCLEAR OfE 1 FORM TirLE.
FORM NO. REV. 1eCPRst.s0 REMEW CONTDauATION PAGE tees.131C 2 not be increased beyond the licensed limit. Therefore, the consequences of an accident previously evaluated in the SAR will not be increased.
- 3. Will the probability of a malfunction of equipment important to safety be increased?
The purpose of the Containment fire water suppression system is to provida fire water to the Containment cable spreading areas and hose reels. The system is isolated except in modes 5 & 6 but may be opened in any mode in the event of a fire. The containment cable spreading area suppression system is not considered equipment important to safety nor affects any equipment to safety nor would prevent safely shutting down the unit in the event of a fire . By disconnecting the air supply to the air maintenance device the operation of the suppression system valve will not be impaired since it may still be closed to isolate Containment. As a re'sult, the probability of the failure of any equipment important to safety to perfo specific safety function desc 1 bed in the SAR will not be increased.
- 4. Will the consequences of a malfunction of equipment important to safety be increased?
i
- The fire water suppression system for the Containment cable spreading areas and hose reels are not considered equipment that is importa' n t tot safe'y. The operation of the containment suppression system will not be hindered nor containment isolation. Thus, the offsite does will not be increased due to the malfunction of the supssion system nor will this modification affect any other equipment that may result in an increase in the offsite dose limits.
5. Will the possibility of an accident of a different type than previously evaluated in the SAR be created? A fire in containment is not an accident that has been evaluated in the SAR nor are there any other accidents that could be affected by this modification. The removal of the air maintenance device will have no affect on the operation of the sprinkler valve nor will it affect the operation of the fire water containment isolation valve. Thus, the possibility of an accident of a different type than previously evaluated in the SAR will not be created. 6. Will the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR be created? A fire in containment is not an accident that has been previously evaluated in the SAR thus there should be no other acciderst conditions considered or evaluated. There should be no other initiators nor fai considered that could potentially create a malfunction of any type. Thus, ditconnecting the air supply to the air maintenance device will not create the possibility of a malfunction of equipment important to safety of a different type than previously evaluated in the SAR.
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, ARKANSAS NUCLEAR ONE FORM NO REV' 10CFR60J8 REVEW CONTWUATION PAGE 1000,131C 2 l
- 7. Will the margin to safety as defined in the Bases of any technical specifications be reduced?
The containment suppression system is not addressed in the Bases of any technical specifications. l m* 4 @ ,
mm.me euuwm una FORM TITLE: FORM NO. REV. 10cFR60.55 DETERMINATION 1000,131A 3 Pc.1 Document No. CR-2-96-0078-18 Rev1 Change No. O Title ANO-2 SAR LDCR - Fuel Pool System Brief description of proposed change: Modifies SAR text and tables containino information related to the Fuel Pool System in order to make that information accurately reflect the existino Diant confiouration. This chance corrects and clarifies the fuel pool coolina system desion basis discussion. Existino details that described assumptions used to calculate fuel pool heat load were not entirely correct. Many of these details have been removed since they reside and are controlled in the covemino enoineerino calculations. Some have been replaced with a more oeneral statement such as " maximum theoretical heat load". Clarified the sionificance of the 120*F and 150'F 0001 temperatures discussed in this section. Clarified the basis for the oriainal selection of 85'F as the desion SW temperature. Noted that administrative controls are in place to minimize the potential of exceedino a pool temperature of 150*F. Corrected the code of construction listed for the FP heat exchanoer. 2E-
- 27. in Table 9.1-3.
Corrected information in Table 9.2-1 related to heat loeds and reouired SW flow ra associated with the FP heat exchancer and the main chiller condensers. Will the proposed Activity: - 1. Require a change to the Operating License including: Technical Specifications (excluding the bases)? YesO NoS Operating License? YesO No@ Confirmatory Orders? YesO No@ 2. Result in information in the following SAR documents (including drawings and text) being (a) no longer true or accurate, or (b) violate a requirement stated in the document: SAR (multi-volume set for each unit)? Yes@ nod Core Operating Limits Report YesO No@ Fire Hazards Analysis? YesO No@ Basas of the Technical Specifications? YesO No@ Technical Requirements Manual? YesO No@ NRC Safety Evaluation Reports? YesO No@
- 3. Involve a test or experiment not described in the SAR?
(See Attachment 2 for guidance) YesO No@ 4. Result in a potentialimpact to the environment? (Complete the Environmentalimpact Determination of this form.) YesO No@ 5. Result in the need for a Radiological Safety Evaluation per section 6.1.57 ! YesO No@ l 6. Result in any potential impact to the equipment or facilities utilized for Ventilated Storage Cask activities per Section 6.1.6? . YesO No@
ARKANSAS NUCLEAR ONE FORM TITLE: FORM NO. REV. 10cFR80.59 DETERMINATION 1000.131A 3 PC.1 6. Result in any potentialimpact to the equipment or facilities utilized for Ventilated Storage Cask activities per Section 6.1.6? YesO No@ 7. Involve a change under 10CFR50.54 for the following SAR documents per Section 6.1.7: QAMO? YesO No@ E-Plan? YesO No@ Basis for Determination fQuestions 1. 2 & 31: The chances are on a level of detail not discussed in the Ooeratino License. The activity is in i the 2SAR in section 9.1.3. table 9.1-3. and fioure 9.2-1. The activity is limited to a SAR chance constitute a test or experiment. O Proposed change does not require 10 CFR 50.59 Evaluation per Attachment 1, item #
. (if checked, note appropriate item #, send LDCR to Licensing).
Search Scope: List sections reviewed in the Licensing Basis Documents specified in Question 1,2 and 3. If a was done on LRS, 'all" may be entered under "Section" with the keyword (s) used in parentheses. Attach and distribute a completed LDCR per Section 6.1.2 if LBD Document _Section LRS: 50.59 - Unit 2 All (" spent fuel' or " fuel pool" or 2e27 or 2P40; heat and pool) MANUAL SECTIONS: 2SAR 9.1.3, Tables 9.1-3, 9.1-6, & 9.2-1 FIGURES: 2 b - Edward Paul Blackard Certified Reviewers Signature 4/23/98 Printed Name Date Reviewers certification expiration date:_ 11-25-98 Ass ~ stance provided by: Printed Name Scope of Assistance Date SearsflTeep'e
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cceptability (NA, if performed by Technical Review per 1000.006) l w ' 791H/),uf C '. ': Certified Reviewers Signature 9c. Printed Name f Date
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MWhAS NUCLAAM ONE FORM TITLE:
. FORM NO. REV.
10cFR50.50 DETERMINATION 1000.131A 3 PC 1 Page - of ENVIRONMENTAL IMPACT DETERMINATION (UNIT 1 and UNIT 2) i Document No. CR-2-96-0078-18 RevjChange No. 2 Complete the following Determination. If the answer to any checklist item is "Yes", an Environmenta is required. See Section 6.1.4 for additional guidance. Will the Activity being evaluated: Yes No O @ Disturb land that is beyond that initially disturbed during construction (i.e., new construction o buildings, creation or removal of ponds, or other terrestrialimpact)? Sce Unit 2 SAR Fig 2.5-17. This applies only to areas outside the protected area. O @ increase thermal discharges to lake or atmosphere? O @ Increase concentration of chemicals to cooling lake or atmosphere through discharge cana tower? O E increase quantity of chemicals to cooling take or atmosphere through discharge canal or tower? O @ Modify the design or operation of cooling tower which will change drift characteristics? l 0 @ instali any new transmission lines leading offsite? O E Change the design or operation of the intake or discharge structures? O @ Discharges any chemicals new or different from that previously discharged? O @ Potentially cause a sp!!! or unevaluated discharge which may effect neighboring soils, surface water or ground water? l O @ Involve burying or placement of any solid wastes in the site area which may effect runoff, surface water orground water? O @ involve incineration or disposal of any potentially hazardous materials on the ANO site? l O @ Result in a change to nonradiological effluents or licensed reactor power level? l 0 E Potentially change the type orincrease the amount of non-radiological air emissions from the ANO site. i
ARKANSAS NUCLEAR ONE Page 3 FORM TITLE: FORM NO. REV. 10CFR50.59 SAFETY EVALUATION 1000.131B 3 Document No. CR-2-96-007818 RevlChange No. 0 10CFR50.59 Eval. No. FFMi-fM,J Title ANO 2 SAR LDCR - Fuel Pool System (Assigned by PSC) A WRITTEN RESPONSE PROVIDING THE BASIS FOR THE ANSWER TO EACH QU ATTACHED. EACH QUESTION MUST BE ANSWERED SEPARATELY. A SIMPLE STA CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES GUIDANCE FOR RE If the answer to any question on this form is "Yes," then an unreviewed safety question is involved. If the an to all questions is "No,* then the proposed change does not involve an unreviewed safety question. 1. Will the probability of an accident previously evaluated in the SAR be increased? YesO No @ The proposed activity does not adversely affect systems, structures, or components whose failure or degradation is considered to be an initiator of an accident previously evaluated in the SAR. The is limited to corrections and clarifications of SAR text which does not introduce new or modify accident initiators. The change makes SAR text more accurately reflect and represent the existi configuration. 2. Will the consequences of an accident previously evaluated in the SAR be increased? YesO No @ This change to SAR text does not affect fission product barriers or introduce new pathways for o release of radioactive material, nor does it create new or aggravate existing onsite dose conseq that might restrict access to vital areas or otherwise impede mitigating actions.
- 3. Will the probability of a malfunction of equipment important to safety be increased? i YesO No @ '
This activity does not represent a change to the existing design basis of the affected system, nor d introduce new or different interactions with other structures, systems, and components that would increase the probability of a malfunction of equipment important to safety. 4. Will the consequences of a malfunction of equipment important to safety be increased? I Yes O No @ The activity does not complicate or worsen the consequences of malfunctions of existing equipm important to safety, nor does it introduce new equipment whose failure would create new dose consequences. The fundamental design functions and existing interactions are unaffected by the chan 5. Will the possibility of an accident of a different type than any previously evaluated in the SAR be created? YesO No @ The change is limited to corrections and clarifications of SAR text which does not introduce existing accident initiators. The change therefore does not introduce the possibility of an accident of a different type. ; I I l
ARKANSAS NUCLEAR ONE FORM TITLE: PaQe 3 FORM NO. REV. 10CFR50.59 SAFETY EVALUATION 1000.131B 3 6. Will the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR be created? YesO No @ The change makes SAR text more accurately reflect and represent the existing plant configurat ; physical modifications to equipment important to safety are involved. Neither are new or different ' equipment or system interactions created. The possibility of a different type of malfunction of essent equipment other than that previously evaluated is therefore not created. 7. Will the margin of safety as defined in the bases for any technical specification be reduced? l YesO No @ The subject of the text affected by this change is not discussed in the Technical Specifications or thei bases. The value for ric .dnal water level assumed in the thermal / hydraulic analysis (sec. 9.1 incorrectly listed as 23 feet and is corrected by this change to 23.5 feet. Although Technical Sp 3.9.9 enforces a minimum level of 23 feet during movement of assemblies, the basis for this levelis ti to the assumptions used in the fuel handling accident analysis - not pool thermal / hydraulics. Edward Paul Blackard 4/23/98 Certified Reviewers Signature Printed Name Date Reviewers certification expiration date: 11/25/98 Assistance provided by: Printed Name Scope of Assistance Date
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CR.2-96 0168 ARKANSAS NuCLEMt ONE Page 1 FcRM u n.E: FORM NO. REV. 10CFR60.89 DETERMINATION 1000.131 A 3 PC 1 This Document contains 3 Pages. Document No. CR-2-98-0188 Rev/ Change No. O Title REMOVAL OF FIRE BARRIER PENETRATION SEAL DETAILS FROM SAR Brief description of proposed change: Figures 8.3-87,8.3-88 and 8.3-95 are subsets of Figure 8.3 95A. All of these figures illustrate penetration se details that are shown on A-2600 series of drawings. Two of these figures (88 and 95) are referenced in Section 8.3.1.4.6.1 along with Figures 8.3-89 and 8.3-90. This section of the SAR refers to these figures as provi details for 3-hour rated penetration seals for blockouts/ sleeves below electrical panels. The reference of 8.3-89 and 8.3-90 as rated details is in error, since only A-2600 provides rated fire barrier penetration details. The reference of 8.3-88 and 8.3 95 is in error since they are associated with a piping penetration. The intent of the information provided by the figures is historical in nature and beyond the level of detail necessary. As a result of the above, it is recommended that the details be deleted from the SAR along with the statement in Section 8.3.1.4.6.l. Section 6.4.5 of the FHA currently contains a statement listing A-2600 as the drawing that contains all fire rated penetration seals. Willthe proposed Activity:
- 1. Require a change to the Operating License including:
Technical Specifications (excluding the bases)? - YesO No@ Operating License? YesO No@ Confirmatory Orders? YesO No@ 2. Result in information in the following SAR documents (including drawings and text) being (a) no longer true or accurate, or (b) violate a requirement stated in the document: SAR (multi-volume set for each unit)? Yes@ NoO Core Operating Limits Report? YesO No@ Fire Hazards Analysis? YesO No@ l Bases of the Technical Specifications? ' YesO No@ Technical Requirements Manual? YesO No@ NRC Safety Evaluation Reports? YesO No@
- 3. Involve a test or experiment not described in the SAR?
(See Attachment 2 forguidance) YesO No@ - 4. f Result in a potential impact to the environment? (Complete Environmental impact Determination of this fonn.) YesO No@
- 5. Result in the need for a Radiological Safety Evaluation per section 6.1.5?
YesO No@
- 6. Result in any potentialimpact to the equipment or facilities utilized for Ventilated Storage Cask activities per Section 6.1.67 YesO No@
- 7. Involve a change under 10CFR50.54 for the following SAR documents per Section 6.1.7? {
I QAMO? YesO No@ E Plan? YesO No@ { i a
ARKAN-As NUCLEAR ONE FORM TITLE: - PeGe 2 FORM NO. REV. 10CFR80.89 DETERMINATION 1000.131A 3 PC-1,2 Document No. CR-2-98-0168 RevJChange No. O Basis for Determination (Questions 1,2, & 3):
' Specific penetration seal details are below the level of detail required for inc the exception of the U2 SAR, nn other SAR related documents contain illustrations of these fir The FHAwith associated does contain this changed. a discussion of penetration seats but no changes are required O PiW change does not require 10CFR50.59 Evaluation per Attachment 1, item #
(if checked, note appropriate item #, send LDCR to Licensing). Search Scope: List sections reviewed in the Licensing Basis Documents specified in questions 1,2 and 3. If sear performed on LRS, the LRS sean:h index should be entered under "Section" with the search parentheses. Controlled harti copies of the documents shall be reviewed (LRS is not verifi text, not figures or drawings). Attach and distribute a completed LDCR per Section 6.1.2 If required.
, Document intGugD LRS:
50.59 Unit 2 - ("8.3-87") ("8.3-88") F8.3-95") ("8.3-95A") 50.59 CG,T T,00 - ("A.__wtnn-) ( ,agag g__;gg ,! ) i MANUAL SECTIONS: SAR - Section 8.3.1.4.8.l. Tahle 1.7-2 FIG f ES: SAR - Floures B 3-87. 88. 89. 90. 95 and 95A h A Woody Walker
' Certified R - rs Signature 12/10/98 Printed Name Date Reviews s certification expiration date: 5/21/99 j
Assistance provided by: Printed Name Scope of Assistance Date Search Scope Review Acceptability (NA, if performed by Technical Reviewer per 1000.006)
%wbb-Certified Reviewers Signature "Tdos 'C2-4 U a so 4 Irdst/16 Printed Name Dnth l
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E i ARKANSAS NUCLEAR ONE
' FORM TITLE: Paos 3 FORM NO, REV. I 10CFR50.89 DETERMINATION 1000.131A 3 ENVIRONMENTAL IMPACT DETERMINATION (UNIT 1 and UNIT 2) l Document No. CR-2-98-0188 Rev/ Change No. O Complete the following Determination. If the answer to any item below is Yes", an Environmental Evalur'u7 required. See Section 6.1.4 foradditionalguidance.
Willthe Acilvity being evaluated: Y_en He O S Disturb land that is beyond that initially disturbed during construction (i.e., new construction of buildings, creation or removal of ponds, or other terrestrialimpact)? See Unit 2 SAR Figure 1 I 2.5-17. This applies only to areas outside the protected area. I O @ increase thermal discharges to lake or atmosphere? O @ Increase concentration of chemicals to cooling lake or atmosphere through discharge canal or tower? ' O S Increase quantity of chemicals to cooling lake or atmosphere through discharge canal or tower? O E Modify the design or operation of cooling tower which will change drift characteristics? O 2 install any new transmission lines leading offsite? O E Change the design or operation of the intake or discharge structures? O @ Discharges any chemicals new or different from that previously discharged? O @ Potentially cause a spill or unevaluated discharge which may effect neighboring soils, surface water or ground water? O E involve burying er place'nent of any solid wastes in the site area which may effect runoff, surface water or giand water? O E involve incineration or disposal of any potentially hazardous materials on the ANO site? O @ Result in a change to nonradiologica emuents or licensed reactor power level? I O E Potentially change the type orincrease the amount of non-radiological air emissions from the ANO site. l l ( e I
CR 2-96-o168 I ARKANSAS NUCLEAR ONE Page 1 FORM TITLE: FORM NO. REV. 10CFR50.69 8AFETY EVALUATION 1000.131B 3 PC-2 This Document contains 1 Page. Document No. CR-2 98-0168 Rev/ Change No. 0 10CFR50.59 Eval. No. [FC-99-Odd Title (Assigned by PSC) Removal of Fire Barrier Penetration Seal details from U2 SAR A WRITTEN RESPONSE PROVIDING THE BASIS FOR THE ANSWER TO EACH QUE ATTACHED. EACH QUESTION MUST BE ANSWERED SEPARATELY. A SIMPLE STATE CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES GUIDANCE FOR RESPO i if the answer to any question on this form is "Yes," then an unreviewed safety question is involved. If the a to all questions is 'No," then the proposed change does not involve an unreviewed safety question. I
- 1. Will the probability of an accident previously evaluated in the SAR be increased? YesO No @
- 2. Willthe consequences of an accident previously evaluated in the SAR be increased? Yes O No @
3. Will the probabilliy of a malfunction of equipment important to safety be YesO No @ tocreased?
- 4. Will the consequences of a malfunction of equipment important to safety beincreased? YesO No @
5. Will the possibility of an accident of a different type than any previously YesO No @ eJaluated in the SAR be created? 6. Will the possibility of a malfunction of equipment important to safety of a YesO different type than any previously evaluated in the SAR be created? No @
- 7. Will the margin of safety as defined in the basis for any technical specification be ?
YesO No @ Woody Walker Certif Reviewer's Signature 12/10/98 Printed Name Date Reviewe s certification expiration date: 5/21/99 Assistance provided by: Printed Name Scope of Assistance Date PSC review by: & \ Date: h kSS I
~ cR-2 m otes AmuvsAs NuCLaAR ONE Page 1 FORM TITLE: FORM No. REV. 10CFR50.89 REVIEW CONTINUATION PACE 1000.131C 3 Document No. CR-2-g8-1068 Rev/ Change No. _ 0 10CFR50.59 Review Continn*Wn Pace Response to Questions from 1000.131B
- 1. Fire is not a design basis accident and therefore has not been evaluated as such in the SAR. Th Hazards Analysis contains a descririlon of the measures taken to prevent, suppress and mi removal of select penetration detal.Mrom the documentation included in the SAR will not invalidate evaluation included in the FHA. Therefore, the probability of an accident occurring will not be in 2.
Removal of specific fire barrier penetration seal c"~ ::.s from the documentation listed in the effect on the off-site dose release rates. The function of the fire banter penetration sea 3. Fire banter penetration seals are nM classified as equipment important to safety. The removal details from the SAR documentation will not change tfie function of the penetration seals. The detail specified on drawing A 2600 provide 3-hourfire rated assemblies. 4. Removal of specific fire barrier penetration seal details from the documentation listed in the SA effect on the off-site dose release rates. The function of the fire banter penetration seals 5. Removal of the fire banter penetration seal details from the documentation listed in the SA the function of the seals.10CFR50 Appendix R requires penetration seals to have a 3-hour ra accident scenarios can be created by this documentation change. 6. This documentation change performs no physical changes to plant systems, structures and c interface between the fire seals and any components is unchanged as the basic function of seals. Therefore, this change will not affect the probability of any equipment malfunction.
- 7. Fire barrier penetration seals have no impact on any margin of safety identified in the Bases of Specificat!ons. The specified documentation change does not alter the function of penetration sea it reduce the required rating of such seals.
ARKANSAS NUCLEAR ONE Page 1
. FORM TITLE:
FORM NO. REV. 10CFR50.59 DETERMINATION 1000.131 A 3 PC-1 This Document contains 3 Pages. Document No. CR-ANO-2-1999-0073 Rev/ Change No. O Title Correct Breaker Size for 2D2224 on E-2022 sh.1 Brict description of proposed change: Drawing E-2022 sh.1 will be correct to reflect breaker 2D2224 as a 20 amp breaker. Per CR ANO-2-1999 0073 l action item 01, breaker 202224 should be a 20 amp breaker as reflected in calculation 86-E-0020-02 and { drawing E 2432 sh. 3. E 2022 sh.1 is also contained in the Unit 2 SAR as figure 8.318.
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Willthe proposed Activity:
- 1. Require a change to the Operating License including:
Technical Specifications (excluding the bases)? YesO No@ Operating Ucense? YesO NoS Confirmatory Orders? YesO No@ 2. Result in information in the following SAR documents (including drawings and text) being (a) no longer true or accurate, or (b) violate a requirement stated in the document: SAR (multi-volume set for each unit)? Yes@ NoO Core Operating Limits Report? YesO No@ Fire Hazards Analysis? - YesO No2 Bases of the Technical Specifications? YesO No@ Technical Requirements Manual? YesO No@ NRC Safety Evaluation Reports? YesO No@ t
- 3. Involve a test or experiment not described in the SAR? I YesO No@
(See Attachment 2 for guidance)
- 4. Result in a potential impact to the environment? (Complete Environmental 1 Impact Determination of this form.)
YesO No@
- 5. Result in the need for a Radiological Safety Evaluation per section 6.1.57 YesO No@
- 6. Result in any potential impact to the equipment or facilities utilized for Ventilated Storage Cask activities per Section 6.1.67 YesO No@
- 7. Involve a change under 10CFR50.54 for the following SAR documents per Section 6.1.77 QAMO?
YesO No@ l E-Plan? YesO No@ )
)
ARKANSAS NUCLEAR ONE Page 2 FORM TITLE: FORM NO. REV. 10CFR80.89 DETERMINATION 1000.131 A 3 PC-1,2 Document No. CR-ANO-2-1999-0073 Rev/ Change No. 00 Basis for Determination (Questions 1,2,86 3): See attached. O Proposed change does not require 10CFR50.59 Evaluation per Attachment 1, item # (If checked, note appropriate Rem #, send LDCR to Licensing). Search Scope: List sections reviewed in the Licensing Basis Documents specified in questions 1,2 and 3. If search was peiformed on LRS, the LRS search index should be entered under "Section" with the search statement (s1 j parentheses. Controlled hard copies of the documents shall be reviewed (LRS is not verified and sear - text, not figures or drawings). Attach and distribute a completed LDCR per Section 6.1.2 if LBD changes required. Document Section LRS: 50.59-Unit 2 (2d22, distribution w/10 panel, panel w/10 breaker,2d2224) MANUAL SECTIONS: Unit 2 SAR 8.1.2, 8.1.4, 8.3.1.2, 8.3.2, Table (s) 8.3-2, 8.3-48, 8.3-11 Unit 2 TS 3.8.2.4, 3.8.2.3 Unit 2 TS Bases 3 / 4.8, Table 3.8-1 FIGURES: Unit 2 SAR 8.3-6,8.3-16,8.3-18
/ Robert Buser 1/27/99 Certifiiid Reviewp's Signature '
Printed Name Date Reviewers certification expiration date: 04/02/99 Assistance provided by: Printed Name Scope of Assistance Date Search Scope Review Acceptability (NA, if perfonned by Technical Reviewer per 1000.006) Reviewers Signature S.titoson SM Printed Name
//Why Date l
ARKANSAS NUCLEAR ONE Pag) 1 FORM TITLE: FORM NO. REV. 10CFR50.59 REVIEW CONTINUATION PACE 1000.131C 3 Page 3 of 3 Document No. CR-ANO-2-1999-0073 Rev1 Change No. 00 10CFR50.59 Review Continuation Paae Determination Continuation Per CR-ANO-2-1999-0073 action item 01, drawing E-2022 sh.1 will be correct to reflect breaker 2D2224 as a 20 amp breaker. The correct breaker size was verified from calculation 86-E-0020-02 and drawing E-2432 sh. 3. The breaker manufacture and model/ type remain unchanged only the rating is changed. 1. This breaker size change does not require a change to the Operating License, since the scope of the changes is below the level of detail of these documents. 2. ANO drawing E-2022 sh.1 will be revised to reflect 2D2224 as a 20 amp breaker since this drawing is also contained in the Unit 2 SAR as figure 8.3-18 it will also be revised. An evaluation per 10CFR50.59 is attached. The changes include minor revisions to figure 8.318 to reflect revised breaker ratings listed above. 3. This change does not involve any tests or experiments not described in the SAR. This drawing change will not require any unusual operating conditions or startup tests.
- 4. This breaker size change will not result in any adverse impacts to the environment as documented in the attached EnvironmentalImpact Checklist.
5. This change will not require a Radiological Safety Evaluation (RSE) since it does not involve processing any radioactive material outside of the Auxiliary Building, Reactor Building, or Low Level Radwaste Building or create a new pathway for an unmonitored release. 6. This breaker size change does not involve any impact to the Ventilated Storage Cask, including any loading equipment or facliities, monitoring activities, load path / crane changes, associated analysis or spent fuel poolimpacts. 7. This breaker size change will not affect the E-plan or the QAMO, since the scope of the changes is below the level of detail of these documents. I l
ARKANSAS NUCLEAR ONE Pagt 3 FcRM TITLE: f FORM NO. REV. i 10CFR50.59 DETERMINATION 1000.131 A 3 { I ENVIRONMENTAL IMPACT DETERMINATION (UNIT 1 and UNIT 2) Document No. CR-ANO-2-1999-0073 Rev> Change No. 00 { Complete the following Determination. If the answer to any item below is "Yes", an Environmental Evaluation is required. See Section 6.1.4 for additional guidance. Will the Activity being evaluated: xn ug 1
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O E Disturb land that is beyond that initially disturbed during construction (i.e., new construction of buildings, creation or removal of ponds., or other terrestrial impact)? See Unit 2 SAR Figure
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2.5-17. This applies only to areas outside the protected area. O 2 Increase thermal discharges to lake or atmosphere? ! O 2 increase concentration of chemicals to cooling lake or atmosphere through discharge canal or tower? O E increase quantity of chemicals to cooling lake or atmosphere through discharge canal or tower? O E Modify the design or operation of cooling tower which will change drift characteristics? O E install any new transmission lines leading offsite? O @ Change the design or operation of the intake or discharge structures? O B Discharges any chemicals new or different from that previously discharged? O E Potentially cause a spill or unevaluated discharge which may effect neighboring soils, surface water orground water? O E involve burying or placement of any solid wastes in the site area which may effect runoff, surface water orground water? O 2 involve incineration or disposal of any potentially hazardous materials on the ANO site? l 0 2 Result in a change to nonradiological effluents or licensed reactor power level? O @ Potentially change the type or increase the amount of non-radiological air emissions from the ANO site. i t i
ARKANSAS NUCLEAR ONE Paga 1 FORM DTLE: FORM NO. REV. ' 10CFR50.59 SAFETY EVALUATION 1000.131B 3 PC 2 This Document contains 1 Page. Document No. CR-ANO-2-1999-0073 Rev1 Change No. 00
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100FR50.59 Eval. No. @A) 99 -Ol'l j (Assigned by PSC) Title Con ect Breaker Size for 2D2224 on E 2022 sh.1 l
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I A WRITTEN RESPONSE PROVIDING THE BASIS FOR THE ANSWER TO EACH QUEST ATTACHED. EACH QUESTION MUST BE ANSWERED SEPARATELY. A SIMPLE STATEMEN { CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES GUIDANCE FOR RESPO ' If the answer to any question on this form is "Yes," then an unreviewed safety question is involved. If the answer to all questions is 'No," then the proposed change does not involve an unreviewed safety question.
- 1. Willthe probability of an accident previously evaluated in the SAR be increased?
YesO No @
- 2. Willthe consequences of an accident previously evaluated in the SAR be increased?
YesO No @
- 3. Willthe probability of a malfunction of equipment important to safety be increased?
YesO No @ c Willthe consequences of a malfunction of equipment important to safety be increased? YesO No @
- 5. Willthe possibility of an accident of a different type than any previously evaluated in the SAR be created?
YesO No @
- 6. Will the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated i i
in the SAR be created? d YesO No @
- 7. Will the margin of safety as defined in the basis for any technical specification be reduced?
YesO No @ M/ Robert Buser 1/27/99
- Certified'R iewe?T$lgnature i Printed Name Date Reviewer's certification expiration date: 04/02/99 Assistance provided by:
Printed Name Scope of Assistance Date PSC review by: Date: ibMS
ARXANSAS NUCLEAR UNE FARM TITLE: Pagi 1 FORM NO. REV. 10CFR50.59 REVIEW CONTINUATION PA?E 1000.131C 3 Page 2 of 3 Document No. CR ANO-21999-0073 Rev/ Change No. _ 00 Evaluation: 10CFR50.59 Review Continuation Pace Scope Per CR-ANO-2-1999-0073 action item 01, drawing E 2022 sh.1 will be to reflect breaker breaker. The correct breaker size was verified from calculation 86-E-0020-02 and drawing E-2432 breaker manufacture and model/ type remain unchanged only the rating is changed.
- 1. I Will the probability of an accident previously evaluated in the SAR be increased?
( This breaker size change has been reviewed against all of the accidents analyzed in the Unit! DC system is discussed in section 15.1.31 of the Unit 2 SAR. The breaker size change wil of breaker will remain unchanged only the trip rating will chang on 2022. 2. Will the consequences of an accident previously evaluated in the SAR be increased? This breaker size change will not alterthe offsite dose consequent;es of any accident previous SAR. This change will not create any new pathways for release of radioactive material. This chan dose to the public from any previously analyzed event. Changing the trip rating of 2D2224 from amps on codes applicable figureand 8.3-18 will meet design criteria for cable protection as noted in other cales, drawings, an standards. 3. Willthe probability of a malfunction of equipment important to safety be increased? This change involves a change to documentation only (i.e. U2 SAR figure 8.3-18). The correct size br already installed as 2D2224. The brea'ner currently installed was procured as original plant equipme application and meets all applicable design requirements such as seismic and electricalisolation change will correct U2 SAR figure 8.3-18 such that it reflects the proper breaker size for this application accordance with existing calculations, drawings, and applicable codes and standards. Design features an requirements have been analyzed to ensure this equipment cannot increase the probability of a malfunc equipment important to safety. 4. Will the consequences of a malfunction of equipment important to safety be increased? This breaker size change will not affect the offsite dose consequences due to malfunctions of equ to safety. This breaker size change does not change or prevent actions assumed to occurin response to a malfunction of equipment important to safety nor does it alter any assumptions used in evaluating tM consequences of equipment failures. No equipment classified as important to safety will be relocat - This change cannot increase the consequences of failure of equipment important to safety. S. Will the possibility of an accident of a different type than previously evaluated in the SAR be created? The accident types in chapter 15 of the Unit 2 SAR were reviewed. No new accidents could be postulated d this breaker size change. The 20 amp breaker will have the same system function as the existing breakers drawing is being changed to reflect the proper size breakerin accordance with other ANO drawings, cales. applicable codes and standards. Therefore, the possibility of an accident of a different type than previous evaluated in the Unit 2 SAR will not be created.
ARKANSAS NUCLEAR ONE Page 2 FORM TITLE: FORM NO. REV. 10CFR60,59 REVIEW CONTINUATION PAGE 1000.131C 3 Page 3 of 3 t I Document No. CR-ANO-2-1999-0073 Rev/ Change No. 00 l 10CFR50.59 Review Continuation Pace
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6. Will the possibility of a malfunction of equipment important to safety of a different type than that previousl evaluated in the SAR be created? The breaker size change willimprove overcurrent protection for panel 2D22 and the cable fed by 2D2224. This change will not cause any other affects on plant systems. The 20 amp breaker is sufficient to serve loads connected to 2D2224 per calc 86-E-0020-02. The 20 amp breaker currently installed was supplied as original plant equipment and therefore, meets requirements for installation in 2022. No now failure modes are introduced to any system by this change. All of these design features have been analyzed to ensure that this breaker size change cannot create the possibility of a malfunction of equipment important to safety of a different type than previously evaluated in the Unit 2 SAR.
- 7. Will the margin to safety as defined in the basis of any technical specification be reduced?
Unit 2 technical specification basis section 3 / 4.8 addresses operability of the A.C. and D.C. ecuer systems and associated distribution systems. No limits or margins are defined in the Unit 2 technical specifications basis for the equipment affected by this change. Therefore, this change will not reduce the margin of safety defined in the technical specification basis. l l l
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FORM TITLE: FORM NO. REv. 10CFR$0.80 DETERMedATION 1000.131A 2 Pc 2.3 Page1 of [4 Document No. 946012D201 DCPR 3 RevjChange No. 0 Title Containment Vent Header / Waste Gas System Modifications PAGE ~ REV.C Will the proposed Activity: . 1. Require a change to the Operating License including: Technical Specifications (excluding the bases)? YesO No@ Operating License? YesO No@ Confirmatory Orders? YesO .No@ 2. Result in information in the following SAR documents (inc;uding drawings and text) being (a) no longer true or accurate, or (b) violate a requirement. stated in the document: Core Operatikg Limits Report YesO No@ SAR (multi-volume set for each unit)? Yes@ nod QAMO?* YesO No@ E-Plan?" YesO No@ FHA YesO No@ Bases of the Technical Specifications? YesO No@ NRC Safety Evaluation Reports? YesO No@ 3. Involve a test or experiment not described in the SAR7 YesO No@ 4. Result in a potential impact to the environment? (Complete Environmental impact Checklist of this form.) YesO No@ 5. Result in the need for a Radiological Safety Evaluation per section 6.2.4.A? YesO No@ 6. Result in any potential impact to the equipment or facilities utilized for Ventilated Storage Cask activities per Section 6.2.4.87 YesO No@ Basis for Determination: see continuation oaoe Changes to these documents require an evaluation in accordance with 10CFR50.54. See Section 6.2.1.B.
FORM TITLE: M NUCLEAR ONE ) FORM NO. REU. 10CFR80.89 DETERMINATION 1000.131A 2 PC 2.3 I i Page[ of ( Document No. 946012D201 DCPR 3 Rev> Change No. 0
References:
Ust sections reviewed in the Licensing Basis Documents, specified in qu in parentheses. Controlled hard copies of the do ' searches such as LRS are not controlled and seamh text only, not figures completed LDCR if LSD changes are required. Document 1 Section PAGE REY. O Unit 2 SAR Figure 9.3 2 M-2237 Sh.1 Figure 9.3 2 M-2237 Sh. 3 Figure 11.3-1 M-2215 Sh.1 Figure 5.1-3 M-2230 Sh. 2 Figure 11.2-1 M-2213 Sh. 6 Figure 9.2-6 M-2234 Sh. 2 Figure 9.3-1 M-2218 Sh. 3 Figure 3.2-5 M-2239 Sh.1 Figure 8.3-65 E-2866 Sh.1 Figure 8.3 66 E-2867 Sh.1 Figure 8.3-68 E-2873 Sh.1 Figure 8.3-71 E-2876 Sh.1 Figure 8.3-72 E-2877 Sh.1 Table 11.2-14, Table 6.2-26, Table 9.3-3, Table 9.3-1 Table 15.1-3 Table 5.2-2 Unit 2 Tech Specs, Unit 2,SAR, Operating License, ' All ( 2SV-4632, 2CV-2400-2, containment vent header, pressu Confirmatory Orders, QAMO, va,cuum degasifier,2PT-2210) Unit 2 SERs, E-Plan 0-/
/\ J %9 A mm% , _
Stephen J. Lynn Certifsid Rfviewe Signqure 3/29/97 Printed Name Date Reviewer's certification expiration date: 5/26/97 Assistance provided by: I Printed Name- i Scope of Assistance
' Date I
1 I
FORiA MLE: FORM NO. REV. 10CFR80.88 DETERMINAT1oN 1000.131A 2 PC 2.3 Page3 of b ENVIRONMENTAL IMPACT CHECKLIST (UNIT 1 and UNIT 2) Document No. 946012D201 DCPR 3 PAGE / M REY. O Rev3 Change No. O_ Complete the following checklist. If the answer to any checklist item is "Yes", an Environm required. See Section 6.2.1.E for additional guidance. Will the Activity being evaluated: Yes No O @ Disturb land that is beyond that initially disturbed during construction (i.e., ne , buildings, creation or removal of ponds, or other terrestrial impact)? See U! 2.5-17. This applies only to areas outside the protected area. O E increase thermal discharges to lake or atmosphere? O e increase concentration of chemicals to cooling lake or atmosphere tnrough dis tower? O O increase tower? quantity of chemicals to cooling lake or atmosphere through discharge O O Modify the design or operation of cooling tower which will change drift characteristi O a instati any new transmission lines leading offsite? O @ Change the design or operation of the intake or discharge structures? O O Discharges any chemicais new or different from tnat previously discharged? O E Potentially water or groundcause water?a spill or unevaluated discharge which may effect neighborin O E Involve burying or placement of any solid wastes in the site area which may surface water or ground water? O E Involve incineration or disposal of any potentially hazardous materials on the ANO s O @ Result in a change to nonradiological effluents or licensed reactor power level? O B Potentially ANO site. change the type or increase the amount of non-radiological air emis
FORM TITLE: ANKANsAs NUCLEAR ONE . FORM NO. REV. secnue.as nevewcommNUATIONPAGE 180s.131c 2 Pagej of_fg Document No. 946012D201 DCPR 3 Rev1 Change No. 2 10CFR50.59 Review Continuation Pane PAGE M REV.( Description of Chanoe DCPR 3 is the 2R12 outage related portion of this design change. DCPR a on i for the hydrogen oxygen analyzer and the new waste gas compressors m er for and the vacuum degasifier tank affected some SAR Tables and Figureseand ow and the addressed in this 50.59 Review also. This 50.59 Review is a stand alon a revision to the original 50.59 Review. DCPR #3 This DCPR will replace the existing pressurizer steam space solenoid ra e gate valve 2CV-4832. Leakage past 2SV-4832 causes contamination of o presently isolated by manual isolation valves located in the containment sample is therefore not available. The improved leak tightness of the new assist with degasification of the RCS prior to an outage by directing t 2C116 sample panel connections. This portion of the DCPR is "Q'. This DCPR will modify the vent path for the quench tank and RDT so th new waste gas compressors for storage in the waste gas decay tanks. Th through containment that was originally used for a quench tank liquid . sample a penetration 2P37 by containment isolation valves 2SV-5878-1 and 2SV-5871-2. c was The originally used to collect waste gases from the quench tank and the RDT cormsion and leaks. Portions of the CVH located outside of containment containment isolation valve 2CV-2400 2. Other safety related componen valve operator, environmentally qualified position switches 2ZS 2400-2A/B , r supply SOV 2SV 2400-2, and manual vent and test connection valves. Th will be capped just outside the containment wall. This portion of the DCPR is "Q'. Other portions of the Unit 2 Gaseous Waste System will be removed by this DCPR
. These components include the vacuum degasifier pumps, motors, separator, recirculation pump, and asso s and piping.
The vacuum degasifier tank and emuent pumps will remain as theymoved . will be components of the vacuum degasifier system have been unused for a c , corrosion, and unreliability of components. The new waste gas compres suction from the vacuum degasifier tank which will now be operated at , tank liquid sample pump vendor skid will be removed. A liquid sample from the qu required nor 4
FORM TITLE: FORM NO. REU. 10CFR50.80 REVIEW COhmNUATION PAGE 4 1000.131C 2 DCP 948012D201 DCPR #3 PAGE /O// REV. O needed. Existing temperature and level indication alerts Operations to any RCS leakage into t and a liquid sample is not needed. This portion of the DCPR is non-Q. Two new sample valves 2CVC-111 and 2BM-37 will be added at sample panel 20116 to sa vacuum degasifier independent of the new hydrogen oxygen analyzer. This work is non-Q. DCPR #2 This DCPR provided component information for the vendor supplied waste gas compresso 2C75B and hydrogen oxygen analyzer panel 20102. The pressure transmitter for the vacu 2210, was also replaced with a new transmitter with a positive pressure range to reflect the vacuum degasifier tank. Instrument air and component cooling water connections for the new c are provided, and the path for nitrogen supply to the vacuum degasifier tank is changed. This work Question 1. This DCPR will not require a change in the Operating License documents. Question 2. Unit 2 SAR Table 6.2-26 is a listing of containment penetration barriers. Penetration 2P31 show that air operated outboard containment isolation valve 2CV-2400 2 and spectac deleted and replaced with a welded cap. Also inboard MOV 2CV-2401-1 will no longer will only receive an ILRT test (A). The Service / System description for penetration 2P37 is c this line is used for the quench tank and RDT vent rather than a quench tank liquid sample. SAR Table 15.1.34-1, a listing of safety related air operated valves, is being revised to remove 2, the outboard containment isolation valve at 2P31. SAR Table 5.2-2, a list of RCS boundary valves, is being changed to show that one of the sam valves,2SV-4632, will be changed to a MOV,2CV-4632. SAR Table 11.214, a list of waste gas system instrumentation, is being revised to show that pressure no longer controls the waste gas compressors (page 11.8-41). The waste gas surge used for relief volume for 2T18 relief valves and compressor skid relief valves. This control function was changed with the original DCP issue. Page 11.8-40 will be changed to show the new operating vacuum degasifier tank. This tank will now be operated at positive pressure. Page 11.8-43 is revised to sho that the vacuum degasifier vessel (tank) level no longer contro!s the degasifier vacuum pumps. The
amnANaAh NUCLEAR ONE - FORM TITLE: .
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FORM No. REE 1eCPR8a.a0 REVEW CONTINUATION PAGE 1000.131C 2 DCP 948012D201 DCPR Fi _ Paas /A/2 Rev.0 ""' dogasifier pumps are being sumoved by this DCPR. Page 11.5-45 is change compressor discharge is monitored for hydrogen and oxygen. Since the vacuum removed by this DCPR, Page 11.8-44 is changed to show the deletion of instrum level. Page 11.8-44 is also revised to show the new instrument ranges and operatin transmitter 2 FIT-2430. Table 9.31, Primary Sampling System Sample Points, is revised to show the d sample point. This sample point is not used and there is no requirement for a quenc Temperature and level instrumentation in the quench tank provides adequ the quench tank. This line will be used in the new waste gas system design to tank and RDT gas spaces to the new waste gas compressors. This gas will oxygen analyzers. Table 9.3-3, Waste Gas System Sample Points, is revised to show that the wa vent header will not be sampled at the new hydrogen oxygen analyzer. The surge gas in the new design. The tank will only be used for relief volume in the ev compressor relief valves lifting. A manual grab sample can be taken if sampling of th In the new system design, the CVH is used to route waste gases from the quen containment building to the waste gas compressors. The seldom used regenerati connected to the CVH, but is only us'ed to vent small amounts of waste gases during normally isolated. The new hydrogen oxygen analyzer samples gases on the dischar compressors. The sample points for the quench tank and RDT are being adde deleted in the table to more accurately describe the sampling capability for processe SAR Figures are being changed to reflect system P&lD revisions and cable and tray la _ Question 3. These changes will not result in any new or revised test or experiment. Question 4. These changes will not result in any change in radiological release practice and . impact to the environment. See checklist. Question 5. No Radiological Safety Evaluation per section 6.2.4.A is needed since this de processing of radioactive material outside of the Aux. Bldg., Reactor Bldg., , nor creates a new pathway outside of the monitored ventilation or drainage pathways. Question 6. This change does not affect Ventilated Storage Cask equipment or facilities.
I FORM TITLE: FORM NO. REV. 10CFRes.se EVALUATION 1ees.131a 2 Page / of1 OO 10CFR50.5g Eval. No. F p re h 051 (Assigned by PSC) Document No. Mitgp201 DCPR 3 Rev/ Change No. 9,
- Title Containment Vent Headerfff_1 Gas St.-.. "se; - -;;ew A WRITTEN RESPONSE PROVIDING THE BASIS FOR THE ANSWER TO E ATTACHED. EACH QUESTION MUST BE ANSWERED SEPARATELY. A SI CONCLUSION IS NOT SU'71CIENT. A1TACHMENT 2 PROVIDES GUIDANC If the answer to any question on this form is "Yes," then an unreviewed safety question is inv to all questions is "No," then the proposed change does not involve an unreviewed safety question.
1. _Will the probability of an accident previously evaluated in the SAR be increased?
' Yes O No @
This modificNions provided by this DCPR will not increase the probability l SAR. The Unit 2 Containment isolation System (CIS) is identified in the Accident Analysis Chapter 15 as a ESF system necessary to protect the health and safety of the public in the ev LOCA. This system is not an accident initiator and the modifications provided by thi the function of this system. The only safety related portion of the CIS that will be affected by penetration 2P31, the containment vent header. This DCPR will modify the containment vent h penetration piping by removing the outboard containment isolation valve 2CV 2400 2 and penetration piping just outside the containment wall. The valve operator, SOV, handswitch switches associated with this air operated valve will also be removed. The position switche 2A/B are the only environmentally qualified (Regulatory Guide 1.97) components removed The CVH has been isolated with a spectacle flange since 2Rg because the penetratio requirements. The CVH will not be used in the new waste gas system design, so the unus header located outside of containment are being removed. containment isolation integrity'. The outboard welded cap will assure i The motor operated gate valve 2CV-4632 that will replace the pressurizer steam space I valve 2SV-4632 will be part of the Primary Sampling System. The new valve has been ASME Section til Class 2 requirements in accordance with the original piping class specif! l Containment sample isolation at penetration 2P8. valves 2SV-5843-1 and 2SV-5833-1 which are unaffected by this D The vacuum degasifier system is not involved in the initiation of an accident evaluated in abandoned probability and unused of accident initiation.portions of the vacuum degasifier system will therefore have no Conduit the and initiation cable of any trays accident. are mounted to stardard details per E-2080 series drawings and 2. Will the consequences of an accident previously evaluated in the SAR be increased? Yes O No @ Since the modifications provided by this DCPR will maintain containment integrity at the C 2P31, the offsite dose consequences of an analyzed accident will not be increased. The C isolated since 2R9 with a spectacle flange to assure containment integrity. The outboard p wall. will now be removed and the penetration will be sealed with a welded cap just outside CVH
FORM TITLE: NUCLEAR ONE- - FORM No. REU. seCPResAS sVALMATloN
. . . 10es.131a 2 DCP-948012D201 DCPR #3 PAGE /0 N nev. o
- The modificahon to the containment vent header penetration and the replace radioactivity. None of the other equipment affected by t the changes to the conduit and tray layouts will no .
3. < Will the probability of a malfunction of equipment important to safety be
. Increased?
Yes O No @
- The two safety related portions of this DCPR, the CVH modification at p replacement of SOV 2SV-4832 with MOV 2CV-4632, meet the original desig systems (i.e. *Q' and ASME Section ill Cians 2), so the probability of a malfun l
equipment will not be increased. Safety related outboard containment isolatio maintained at penetration 2P31'by the new we.lded pipe c The impact of the additional maintain weight qualification. of the MOV has been analyzed and the piping support for l None of the equipment associated with the removal of portions of the related. i. which are seismically qualified, the probability of these 4. Will the consequenries of a malfunction of equipment important to safety be increased? ' Yes O No @ Since this modification will maintain both containment isolation integrity at th penetration 2P31 and the ASME Section 111 Class 2 pressure boundary at 2CV-i malfunction of equipment important to safety will not be increased. The do accidents will not be affected by this change and no new pathways for the r! created. There are no safety related tie-ins to the portions of the vacuum degasifier sy the piping termination points will be capped in accordance with pipirg class requirements 5. Will the possibility of an accident of a different type than any previously 1 I evaluated in the SAR be created? Yes O No @ The modifications pmvided by this DCPR do not create any new accidents.
. Initiators or failures are introduced by this modification. penetration 2
{ 6. Will the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR be created? Yes O No @ located outside of containment and the pressurizer ste
FORM TULE:
- - .w aww w une FORM NO. REV.
- 10CFR80.59 EVALUATM3N
- 1000.1313 2 DCP-946012D201 DCPR #3 PAGE /015 nev. O Page d of1 building. These requirements changes so no new have malfunctions been designed are created. to Seismic Class 1 requirements c 7.
Will the margin of safety as defined in the bases for any technical specification be reduced? Yes O No @ The Unit 2 Technical Specifications do not identify any marg!n of safety for contain 2CV-2400 2 that will be removed by this DCPR. Containment integrity will be maintained at the containment vent header by this modification. No margin of safety is identified either steam space sample valve, the vacuum dogasifier equipment, nor conduit and trays. h . Stechen J. Lynn Certifiep Reviegs Sipature 3/30/97 Printed Name Date Reviewer's certification expiration date: 5/26/97 Assistance provided by: Printed Name Scope of Assistance Date PSC review byOM Date: th\%1 e I i i
annamaxa muuttan une FORM TITLE: FORM NO. REV. 10CFR40.80 DETFRMINATION 1000.131A 2 PC-2,3 Pagelof 3 Document No. ER 962006D202 Rev1 Change No. 0 Title Elect. Penetration Uporade 2VvR42-3 PAGE ' REV.O Will the proposed Activity:
- 1. Require a change to the Operating License including:
Technical Specifications (ex:luding the bases)? YesO No2 Operating License? YesO No@ W Confirmatory Orders? YesO No@ 2. Result in information in the following SAR documents (including drawings and text) being (a) no longer true or accurate, or (b) violate a requirement stated in the document: Core Operating Limits Report YesO No@ SAR (multi-volume set for each unit)? Yes@ NoO CAMO?* YesO No@ ! E-Plan?* 4 YesO No@ FHA YesO No2 Bases of the Technical Specifications? YesO No@ NRC Safety Evaluation Reports? YesO No@
- 3. Involve a test or experiment not described in the SAR?
YesO No@
- 4. Result in a potentialimpact to the environment? (Complete Environmental impact Checklist of this form.)
YesO No@
- 5. Result in the need for a Radiological Safety Evaluation per section 6.2.4.A?
YesO No@
- 6. Result in any potential impact to the equipment or facilities utilized for Ventilated Storage Cask activities per Section 6.2.4.87 YesO No@
Basis for Determination: See Continuation Face. Changes to these documents require an evaluation in accordanc.e with 10CFR50.54. See Section 6.2.1.B. < - - - - - - - - - - - {
ARKANSAS NUCLEAR ONE FORM TULE: . FORM NO. REV. 10CFRSO.89 DETERMINATION 1000.131A 2 PC-2.3 PAGE 5 m.. Page 2 of.3 Document No. ER 962006D202 Rev/ Change No. O
References:
List sections reviewed in the Licensing Basis Documents, specified in questions 1,2 and keyword search was done on LRS, "all" may be entered under "Section" with the in parentheses. Controlled hard copies of the documents shall be reviewed as compu searches such as LRS are not controlled and search text only, not figures or drawings completed LDCR if LBD changes .:re required. Document Section ANO-2 Tech. Spec. All ANO-2 Operating License All i ANO-2 Confirmatory Orders All _ANO-2 SAR All QAMO All E-Plan All FHA All ANO-2 Bases of Tech. Spec. All ANO-2 NRC SERs All (LRS Keyword Search:
" Electrical Penetration", EPA, "Elec. Penetration", " Electrical Pen.", I "Elec. Pen", " Elect. Pen" m - -
Certi Reviewers Signature DARYL BARNHOUSE 7/30/97 Printed Name Date Reviewers certification expiration date: 4/3/98 Assistance provided by: Printed Name Scope of Assistance N/A Date J
-vous auct.aAn one FORM TITLE.
FORM NO. REV. SecRtse.de DETERMINATMIN 1ees.131A 2 PC.2.3 Page } of.2 ENVIRONMENTAL IMPACT CHECKLIST (UNIT 1 and UNIT 2) PAGE REV.O Document No. ER 962006D202 RevlChange No. O Complete the following checklist. If the answer to any checklist item is "Yes", an Environmental Evaluation is required. See Section 6.2.1.E for additional guidance. Willthe Activity being evaluated: Yes No > O E Disturb land that is beyond that initially disturbed during constructian (i.e., new construction of buildings, creation or remova! of ponds, or other terrestrial impact)? See Unit 2 SAR Figure 2.5-17. This applies only to areas outside the protected area. O E increase thermal discharges to lake or atmosphere? O E increase concentration of chemicals to cooling lake or atmosphere through discharge canal or tower? O E increase quantity of chemicals to cooling lake or atmosphere through discharge canal or tower? O E Modify the design or operation of cooling tower which Mil change drift characteristics? O E Install any new transmission lines leading offsite? O E change the design or operation of the intake or discharge structures? O E Discharges any chemicais new or different from that previously discharged? O E Potentially cause a spill or unevaluated discharge which may effect neighboring soils, surface water or ground water? O E involve burying or placement of any solid wastes in the site area which may effect runoff, surface water or ground water? O E Involve incineration or disposal of any potentially hazardous materials on the ANO site? O E Result in a change to nonradiological effluents or licensed reactor power level? O E Potentially change the type or increase the amount of non-radiological air emissions from the ANO site.
ARMANaAs NUCLEAR ONE FORM TITm i FORM NO. REV. i secense.se mannEw coornNUATION PAGE 1000.131C 2 Page 1 of,2 Document No. ER 962006D202 Rev1 Change No. 0_ #7 PAGE I REV. O 10CFR50.59 Review Continuation Paae SYNOPSIS OF DESIGN CHANGES
'This DCP involves the replacement of existing Amphenol-Sams electrical penetration feed module assemblies with new Conax Buffalo feedthrough adapter module assemblies for containmen building electrical penetration 2WR42-3. The affected penetration contains approximately 630 mod pigtail splicos to de-terminate and re-terminate.
Most terminations will be made using Grayboot connectors or coaxial connectors with Raychem sleeves or kits installed over the coaxial connections After installation of the new modules, all affected circuits must be tested to insure that all connect have been made correctly. This modification is being made because the presently installed Amphenol-Sams module seals deteriorating due to their age and environment which has caused some of them to leak. All of the se will soon be at the end of their calculated life span. i Amphenol-Sams is no longer in the penetration module business, therefore the new EQ modules ' beir'g procured from Conax Buffalo. The Conax feedthrough modules will require adapters whic allow the mounting of the Conax modules in an existing Amphenol Sams header plate. The basic function of the feedthrough module assemblies is to maintain containment integrill design basis accident, while maintaining electrical continuity and separation through the modules. modules were designed and manufactured according to IEEE Std. 317-1983, " Standard for Electri Penetration Assemblies in Containment Structures for Nuclear Power Generating Stations". The m design was subjected to applicable Design and Qualified-Ufe Tests as defined in IEEE Std. 31 modified by NRC Regulatory Guide 1.63, Rev. 3. Feb.,1987. Results of the testing performed demonstrate functionability as ' required under specific plant service conditions for 40 years install without loss of electrical function or containment integrity capabi!! ties in accordance with requi IEEE Std. 317-1983. Response to Determination Questions Bases For Answer To Determination Question 1 Electrical Penetration Assemblies (EPAs) and/or modules are not specifically referenced Operating documents. Ucense Ucense. Therefore, the modifications to be implemented by this DCP do not impa Bases For Answer To Determination Question 2 A search of the Unit 2 SAR, Core Operating Umits Report, QAMO, E-Plan, FHA, Bases of the Te Specifications, and the NRC Safety Evaluation Reports was performed using LRS. All hits were r using LRS and hard copies of these documents. It was determined that this DCP will caus SAR Section 8.3.1.1.13G. This section states that Conax-Buffalo Electrical Penetration Assemb require continuous Nitrogen pressurization of 15-25 PSIG. The Conax Buffalo Tech. Manual sta the 15-25 PSIG pressure was provided during shipment and that the modules do not requi pressurization to meet their design intent. An Ucensing Document Change Request has be to correct this information in the SAR. No other change were required.
._ _ _ . _ _ _ _ _ __.______m
swww.=a= nuu. man one FORM TITLE: FORM NO. REV. 10CFMG.80 K.ECONTDAAATIoN PAGE 1eet.131C 2 Page 2 of,2 Document No. ER 962006D202 Rev1 Change No. A p f 3.0 Bases For Answer To Determination Question 3 This DCP will not involve any tests or experiments not identified in the SAR. The only testing involved witn this DCP will be the normal post maintenance / outage testing as defined in approved ANO procedures. No new tests or experiments will be performed as a part of this DCP. 4.0 Bases For Answer To Determination Question 4 This DCP will not result in any adverse impacts to the environment. The operation of the plant will n changed in any way which will result in changes to the air, water, or soil conditions of the site, 5.0 Bases For Answerto Determination Question 8 This DCP does not involve the processing of radioactive material and does not create a new p outside of the monitored ventilation or drainage pathways, thus a Radiological Safety Evaluation is not required. 6.0 Bases For Answer to Determination Question 6 This DCP does not involve the Ventilated Storage Cask (VSC) or any components that might be us VSC movement and does not involve any work activities in the spent fuel pool area, train bay, ra tracks, or areas adjacent to the VSC storage pad. l 1 I
AfuuussAs NUCLEAR ONE FORM TITLE: FORM NO. REV. 10CMt00.Se sVALUATIoM 1000.131s 2 Page1 of,2 10CFR50.5g Eval. No._ Fe(V m a G (Assigned by PSC) Document No. ER 962006D202 Rev> Change No. 2 TN.!c ElectricalPenetr*Mn Um;.t 2WR42-3 A WRITTEN RESPONSE PROVIDING THE BASIS FOR THE ANSWER TO EA ATTACHED. EACH QUESTION MUST BE ANSWERED SEPARATEl.Y. A SIMPLE CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES GUIDANCE FO If the answer to any question on this form is "Yes," then an unreviewed safety question is invol to all questions is "No," then the proposed change does not involve an unreviewed safety question. 1. Will the probability of an accident previously evaluated in the SAR be increased? Yes O No @ The probability of a accident previously evaluated in the SAR will not be increased. The el accidents in the SAR. This DCP will not create any new conditio events which are credited with initiating an evaluated accident. 2. Will the consequences of an accident previously eve'uated in the SAR be increased? Yes O No @ The consequences of an accident previously evaluated in the SAR will not be increased. Th will not affect offsite conditions following an accident. The replacement feedthrough module of a superior mechanical design to the existing Amphenol-SAMS modules. Therefore, the c an accident, such as a LOCA ,should actually be reduced due to the greater probabilr.y of sust design. Therefore, the consequences of accidents are unaffected. containm 3. Will the probability of a malfunction of equipment important to safety be increased? Yes O No @ The probability of a malfunction of equipment important to safety will not beThe increased. replaced. The modules interface with safety related equipment conductors which serve this equipment, and by penetrating containment. The electrical parame new modules are comparable or superior to the existing modules. containment building integrity. replacement modules is superior to the existing Am Therefore, the probability of a malfunction of safety related equipment sh . 4. Will the consequences of a malfunction of equipment important to safety be increased? Yes O No @ DCP. There are no new conditions or events created by this DC
- consequences to the plant or public due to a malfunction of s&fety related equipment. As stated l
modules. The modules are intended to limit the consequencesmodules (CONTINUED ON NEXT PAGE)
l 1 i ( FORM TITLE: ARKANSAS NUCLEAR ONE 10CFR50.5,9 EVALUATION FORM NO. REV. 1000.131B 2 l Pagelof 2 \ 10CFR50.59 Eval. No. FF WH50f (Assigned by PSC) Document No. ER 962006D202 Rev) Change No. A PAGE /N REV.O of various events due to their containment boundary function. containment integrity associated with a LOCA sh modules. Since the modules' electrical properties are comparable or superior is no potential for an increase in the consequences of the malfunction of relat . 3 5. Will the possibility of an accident of a different type than any previously evaluated in the SAR be created? i Yes O No @ ! l created by this DCP. There are no new conditions or p! which could cause a new or different type accident than those already evalual being added are of equivalent form, fit and function as the modules being replaced. Therefore, no new l failure modes have previously been reviewed. been identified which would result in thr erotion of an accident which has not . 6. Will the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR be created? Yes O No @ evaluated in the SAR will not be created by this DCP. T an unevaluated malfunction of equipment important to safety. important to safety of a different type is not created. equivalent or su 7. Will the margin of safety as defined in the bases for any technical specification be reduced? Yes O No @ penetration assembly modules. The modules have bee requirements for inside ANO-2 containment. Consequently, the replacement of pen not decrease the margins of safety as specified in Tech. Spac. basis for containme integrity given in paragraphs 3/4.6.1.2 and 3/4.6.1.5. Therefore this DCP will n safety as defined in the bases for any technical specifications.
. an Daryl Barnhouse F
Certifs RiFviewers Signature ^ 7/30/97 Printed Name Date Reviewe s certification expiration date:_ 4/3/98 Assistance provided by: Printed Name N/A Scope of Assistance Date C review by: D y} M
, Date: \
FORM TITLE: FORM NO. REV 10CFR50.89 DETERMINATION 1800.131A 3 PC.1 Document No. ER 963230D201 Page 1 ofj RevJChange No. g Q' Title PAGE 8tEV. O 2E11A & 2E11B sa*L!N CONDENSER TUBE BUNDLE REPLACEMENT Brief description of proposed change: This modification will replace the orioinal main condenser tube bundles with shod fabricated titanium tube bundI65. The i&d@ishad CG6danser has 13% m result in a 0.8% ' reduction in circulatino water flow and knuiuve condenser har*arassure by Will the proposed Activity: 1. Require a change to the Operating License including: Technical Specifications (excluding the bases)? Operating License? YesO No@ Confirmatory Orders? YesO No@
- 2. YesO No@
Result in information in the following SAR documents (including drawings and text) being (a) no longer true or accurate, or (b) violate a requirement stated in the document: SAR (multi-volume set for each unit)? Yes@ NoC Core Operating Limits Report Fire Hazards Analysis? YesO No@ YesO No@ Bases of the Technical Specifications? Technical Requirements Manual? YesO No@ YesO No@ NRC Safety Evaluation Reports?
- 3. YesO No@
Involve a test or experiment not described in the SAR? YesO No@ (See Attachment 2 for guidance) 4. Result in a potential impact to the environment? (Complete the Environmentalimpact Determination of this form.)
- 5. YesO No@
Result in the need for a Radiological Safety Evaluation per section 6.1.57
- 6. YesO No@
Result in any potentialimpact to the equipment or facilities utilized for Ventilated Storage Cask activities per Section 6.1.67
- 7. YesO No@
Involve a change under 10CFR50.54 for the following SAR documents per Section 6.1.7: QAMO? E-Plan? YesO No@ YesO No@ Basis for Determination (Questions 1. 2 & 31: This 50.59 determination addresses the ANO-2 Main Condenser Tube Bundle Replace Search of the 50.59 - Common or Unit 2 Index and a review of the above licensing basis doc that a revision to the ANO-2 SAR is required by this modification. This modification w License, Confirmatory Orders, Core Operating Limits Report, QAMO, E-Plan, FHA, SERs o Technical Specifications. This modification does not involve a test or experiment not describe Neither a radiological safety evaluation not an environmental evaluation was required by th moditication does not impact any equipment or facilities utilized for Ventilated Storage Cask activities O Proposed change does not require 10 CFR 50.59 Evaluation per(if Attachment appropnate item #. send LDCR to Licensing). checked, note 1, item I
swouwmAs NUCLEAR ONE FORM 7TTLE: 1 FORM NO. REV-
' 0CPR40.89 DETERMINATION 1000.131A 3 PC 1 Page 2 of,3 Document No. ER 963230D201 Rev1 Change No. 9 p /
O
. s Search Scope:
List sections reviewed in the Licensing Basis Documents specified in Question 1,2 and 3 was done on LRS, "all" may be entered under "Section" with the keyword (s) used in parenth Attach and distribute a completed LDCR per Section 6.1.2 If LB Q2GLm.g.nj n Section LRS: All Documents contained under LRS All (2E11*, Circ *, Copper, Condenser-) ! 50.59 - Unit 2 All Documents Contained in LRS All(AAC, Attemate AC, Station Blackout) 50.59 - Common MANUAL SECTIONS: i U2 SAR
- Section 10.4.1.2
- Table 10.4-1 1
- Table 10.4 2
- Table 10.4-3
- Section 10.4.1,10.4.2,10.4.5
- Section 15.1.7,15.1.18,15.1.28 Section 9.2.1, 9.3.1, 8.3 FIGURES:
U2 SAR Figure 9.21 (M2211 Sh.1) - Auxiliary Cooling Water
- Figure 9.3-1 (M2218 Sh. 2) instrument Air Figure 3.2-2 (M2220 Sh. 2) Plant Heating
.- Figure 10.4.4 (M2405 Sh. 3) Circulating Water System Logic Diagram W . < /r 9 <
DOUGLAS EDGFI f Certified Reviewers Af5 nature 8/29/98 Pnnted Name Date Reviewers certification expiration date: 03/17/99 Assistance provided by: Printed Name Scope of Assistance Date 0 fl< 3.; Scope Review, Acceptability
.f / l~f.' / (NA. If performed by Technical Review per 1000.006) t'n , . i+n a %c.una C;rtified Reviewers Signature Printed Name Oft l4b
' Date
r i PORM TITLE FOCM NO. RElf 10CFR80.88 DETERMINATION 1a00,131A 3 PC4 Page } of 3 ENVIRONMENTAL IMPACT DETERMINATION Document No. ER 963230D201 (UNIT 1 and UNIT 2) g Rev1 Change No. R * { l is required. See Section 6.1.4 for additional guidance. Complete the follow Will the Activity being evaluated: Yes No O @
- Disturb land that is beyond that initially disturbed during construction (i.e., ne 2.5-17. This applies only to areas outside the protected area i O @
increase thermal discharges to take or atmosphere? O @ increase tower? concentration of chemicals to cooling lake or atmosphere through 3 i O @ increase tower? quantity of chemicals to cooling lake or atmosphere through discharge c O @ Modify the design or operation of cooling tower which will change drift characterist O @ install any new transmission lines leading offsite? O @ Change the design or operation of the intake or discharge structures? O @ Discharges any chemicals new or different from that previously discharged? O @ ot Potentially water or groundcause water? a spill or unevaluated discharge which may/ffect neighbo O @ involvewater surface burying or placement or ground water? of any solid wastes in the site area which may effe O @ involve incineration or disposal of any potentially hazardous materials on the ANO O @ Result in a change to nonradiological effluents or licensed reactor power level? O @ Potentially ANO site. change the type or increase the amount of non-radiological air em
FORM MLE: N NucMAR ME . , , , _ , , , FORM No. 10cFR$0.88 EVALUATION REV. l- 1000.131B
. 3 Pc4 '
PAGE REV.O Page 1 f.2
- 10CFR50.59 Eval. No.NYY 0.5 (Assigned by PSC)
Document No. ER 963230D201 Rev1 Change No. 2 Title 2E11A & 2E11B MAIN CONDENSER TUBE BUNDLE REPLACEMENT ATTACHED. EACH QUESTION MUST BE AN CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES . If the answer to any question on this form is "Yes," then an unreviewed safety questio to all questions is "No," then the proposed change does not involve an unreviewed . If the answer safeti 1. Will the probability of an accident previously evaluated in the SAR be increased? Yes O No @ fabricated titanium tube bundies.This modification suplaces the existing p 90 modes of any components, system or structures. All previousl reviewed for applicability. No additional failure modes that were not considere introduced by this modification. This modification will not increase the prob accidents 2. Will the consequences of an accident previously evaluated in the SAR be increased? ' Yes O No @ accident in the SAR. Nor does it change, degrade o in the SAR. This modification does not change the function, operation omponents, or fa I accident in the SAR. The dose consequences and a result of this modification. Therefore, there is no change to the consequence accident in the SAR as a result of this modification. 3. Will the probability increased? of a malfunction of equipment important to safety be Yes O No @ The main condenser is not safety related and does not affect safety relate modification does not change the function or failure modes of any componen replacement tube bundles will improve plant efficiency, eliminate copper in the . preparation for steam generator replacement, and are designed for the 107.5% This modification will not increaar the probability of a malfunction of eq .
ARKANSAS NUCLEAR ONE FORM TeTLE: !. FORM NO. REV-10CFRSO.89 EVALUATION 1000.131B 3 PC-2 Page 2 of 2 4. Will the consequences of a malfunction of equipment important to safety PAGE O REV.O be increased? Yes O No @ None of the equipment replaced, rerouted or installed by this design change is important to safety mairi condenser is not safety related and does not affect safety-related components or systems. This modification does not change the function or failure modes of any components, systems, or structure , does not change, degrade, or prevent any action described in an accident discussed in the SAR. The consequences of a malfunction of equipment important to safety will not be increased as a result of this modification. 5. Will the possibility of an accident of a different type than any previously evaluated in the SAR be created? Yes O No @ The redesigned condenser is functionally equivalent to the original design. This modification does change the function or failure modes of any components, systems, or structures. l failure or accident conditions. It will not create any new evaluated in the SAR will not be created.The possibility of an accident of a different type than an 6. Will the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR be created? Yes O No @ This modification involves replacement of equipment in the secondary system that is not safety
- has no impact on equipment important to safety, it does not change the function or failure m components, systems, or structures. The possibility of a malfunction of equipment important to safety different type than any previously evaluated in the SAR will not be created as a result of this modifi 7.
Will the margin of safety as defined in the basis for any technical specification be reduced? Yes O No @ There are no technical specification limits or basis associated with the main condenser. The modification does not affect the margin of safety as defined in the bases for any Technical S bases of the LCO's or surveillance requirements. l DOUGLAS EDGFI I 8/29/98 Csrtified Rdiewers Signliture Printed Name Date Rsviewers certification expiration date: 3/17/99 Assistance provided by: Printed Name Scope of Assistance Date PSC review by: hIM Date: S G6 i ER-9632300201 I i
ARKANSAS NUCLEAR ONE FORM TITLE; FORM NO. REV. 10CFR50JS DETERMINATION 1000.131A 2 PC 2,3 Page 1 of 12 Document No. DCP 963242D201 RevlChange No. O Title Unit Two Battery Charoer Reofacement Wi!! the proposed Activity: 1. Require a change to the Operating License including: Technical Specifications (excluding the bases)? YesO No@ Operating License? YesO No@ Confirmatory Orders? YesO No@
- 2. .
Result in information in the following SAR documents (including drawings and text) being (a) no longer true or accurate, or (b) violate a requirement stated in the document: Core Operating Limits Report YesO NoS SAR (multi-volume set for each unit)? Yes@ nod QAMO?* YesO No@ E Plan?* YesO No@ FHA Yes@ nod Bases of the Technical Specifications? YesO No@ NRC Safety Evaluation Reports? YesO No@
- 3. Involve a test or experiment not described in the SAR?
YesO No@
- 4. Result in a potentialimpact to the environment? (Complete Environmental impact Checklist of this form.)
YesO No@ 5. Result in the need for a Radiological Safety Evaluation per section 6.2.4.A? YesO No@ , 6. Result in any potentialimpact to the equipment or facilities utilized for Ventilated Storage Cask activities per Section 6.2.4.87 YesO No@ Basis for Determination: ' S;e Attached Ch:nges to these documents require an evaluation in accordance with 10CFR50.54. S:e Section 6.2.1.B. PAGE- REV.O
RNANsAs NUCLEAR ONE FORM TITLE:. FORM NO. REV. 10CPR00.00 DETERMINATION 1000.131A - 2 PC-2,3 Page 2 of 12 Document No. DCP g632420201
- Rev1 Change No. O R:ferences:
Ust sections reviewed in the Licensing Basis Documents, specified in questions 1,2 and keyword search was done on LRS, "all" may be entered under "Section" with the ke in parentheses. Controlled hard copies of the documents shall be reviewed as computer searches such as LRS are not controlled and search text only, not figures or drawings. Attac completed LDCR if LBD changes are required. Document Section U-2 T.S., U-2 0.P., U-2 Con. Orders, U-2 SAR, ALL (battery 2033, 2034, charger, chargers, charger / eliminator, charger, charger's, 2031, 2 U-2 Core Operating eliminators, regulatory guide 1.32, eight hours, safety guide 32,1.32, l Limit Report, QAMO, transfer switch, isolation device, seperation)
~
. FHA, E-Plan, Bases of T.S., SERsq David Alan Robinson
~
CIrtified Reviewers Signature 8/13/97 Printed Name Date R:: viewers certificat:On expiration date: 1/29/99
- Assistance provided by:
Printed Name Scope of Assistance Date l O PAGE-REV.o e
ARKANSAS NuCt. EAR ONE FORM TITLE: FORM NO. REV. 10CPR80.50 DETERMINATION 1000.131A 2 Pc-2.3 Page 3 of 12 ENVIRONMENTAL IMPACT CHECKLIST (UNIT 1 and UNIT 2) D:cument No. DCP 963242D201 Rev1 Change No. O Complete the following checklist. If the answer to any checklist item is "Yes", an Environmental Eva r; quired. See Section 6.2.1.E for additional guidance. Will the Activity being evaluated: 1res No O E Disturb land that is beyond that initially disturbed during conrtruction (i.e.. new construction l buildings, creation or removal of ponds, or other terrestrialimpact)? See Unit 2 SAR Fig 2.5-17. This applies only to areas outside the protected area. O E- Increase thermal discharges to take or atmosphere? O @ l increase tower? concentration of chemicals to cooling lake or atmosphere through discharge ca O E increase quantity of chemicals to cooling lake or atmosphere through discharge canal or tower? O E Modify the design or operation of cooling tower which will change drift characteristics? ! O E install any new transmission lines leading offsite? O @ Change the design or operation of the intake or discharge structures? O @ Discharges any chemicals new or different from that previously discharged? O B Potentially cause a spill or unevaluated discharge which may effect neighboring soils, surfac l water or ground water? O S- Involve burying or placement of any solid wastes in the site area which may effect runoff, surface water or ground water? O E involve incineration or dim at of any r,tentially hazardous materials on the ANO site? i , O O ! ! Result in a change to cr ediological effluents orlicensed reactor powerlevel? l ! O E Potentially change the type or increase the amount of non-radiological air emissions from the ANO site. PAGE REV.O ! L
~
ARKANSAS NUCLEAR ONE FORM TITLE: FORM NO. REV. 10CFRSO.89 EVALUATION 1000.131B 2 Page 4 of 12 10CFR50.59 Eval. No. fM47-Old (Assigned by PSC) Document No. 963242D201 RevlChange No. O Title Unit 2 Battery Charner Reolacement ATTACHED. EACH QUESTION MUST BE ANSWER CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES GUIDANC If the answer to any question on this form is "Yes," then an unreviewed safety question is involved
. If the answer to all questions is "No," then the proposed change does not involve an unreviewed safety question.
1. Will the probability of an accident previously evaluated in the SAR be increased? Yes O No @ See Attached 2. Will the consequences of an accident previously evaluated in the SAR be increased? Yes O No @ See Attached 3. Will the probability of a malfunction of equipment important to safety be increased? Yes O No @ See Attached 4. Will the consequences of a malfunction of equipment important to safety be increased? Yes O No @ See Attached 5. Will the possibility of an accident of a different type than any previously evaluated in the SAR be created? Yes O No @ See Attached 6. Will the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR be created? Yes O No @ See Attached 7. Will the margin of safety as defined in the bases for any technical specification be reduced? Yes O No @ See Attached PAGE REV.O .
FORM 71TLE: 10CRtstapEVAL13ATION PottM N 3. ItEV. 1ese.uis 2 Pa0e 5 of12
)
1
< Wr- David A* RceWh.
Cait.M Reviewers Signature Eft 3/97 Pdnted Name Date Reviewers cettdicadon expirabon date: 1/29#ff M 2e.f-98 Assistance provided 5 i' edmed Name Soope of Asasunoe
- Dee PSC remw$ Dhh Dae* NI -
f f I 1 1 i Dowal &. 463M2D.1bl ! PAag l O !
FORM TITI.E: A "AN8A8Nuct.EARONE FORM NO. 10cPR80.SS REVIEW CONTINUA 7 ion PAGE REY. 1000.131C 2
.i-Page 6 of 12 Document No. DCP 9632420201 RevjChange No. 0 10CFR50.59 Review Continuation Paae The purpose of DCP 9632420201 is to replace the Unit 2 battery chargers. mp red and green train battery chargers. 2D34 is the existing 200 Amp swing batte voltage to the 2001 DC bus,2D32 supplies DC voltage to the 2002 DC bu swing battery charger for either bus.
This DCP will delete the 2D34 swing battery charger and replace the 2031 an
. our new Q 400 Amp battery chargers will be installed. Two chargers (2031A and 2D31 and two chargers (2032A and 2D328) will be installed in the green train. The de train one battery charger will supply DC power to the DC bus while the other battery dedicated spare. This will ensure separation of the powertrains and facilitate a ery on-l chargers. The out-of-service battery charger should be normally de-energize considerations. The bettery chargers are designed, however, for load sharing a the chargers. Live load transfers will reduce battery cycling resulting from us to removin - change chargers. During maintenance on the out-of service charger, administrative de-Inergization of the charger if an operational event places the EDG in service o created. ,
The new chargers are sized to carry the DC bus load and recharge a fully disch . rating for the new chargers is the same as the existing red and green train chargers a cxisting swing charger. The power sources for the battery chargers will also be div power fmm separate Motor Control Centers (MCC's) to each charger. Existing sw provides the capability during refueling outages to supply a battery charger to a DC MCC wheneverload center 285 or 286 is taken out of service. in order to maintain th , transfer switch 2S21 will feed charger 2D318 and manual transfer switche 2S22 will f . . transfer switches will be used to transfer the charger power supply between red and during outages. The transfer switches are break-before-make, which prevents a po ' bus:s. The transfer switches are manually operated only, and will be locked . so in th the Cpposite train AC MCC breakerwill be tagged open during normal sources can not be tied together. oweroperation. l Additionally, the existing 200 Amp black battery charger 2D33 will be replaced bt chirg:r.2031 due to maintenance concems with the existing charger. The power source] PAGE 8tEV. O _ l
L FORM TITI.E: ansumsAs e one- - .- FORM NO. REV. - 10CFR80.50 Review CONTINUATION PAGE 1000.131c 2 Document No. DCP 963242D201
- RevlChange No. 0 moved from the red train MCC 2854 to the green train MCC 2B64 in order to provide additional train Emergency Diesel Generator (EDG).
j Basis for Determination: 1.) Will the proposed modification require a change to the Operating Ucense, including Tcchnical Specification (excluding the bases) ? NO Operating Ucense? NO C nfirmatory Orders? NO Technical Specifications - Dice 9ssion and Response The changeTechnical Specification's to the Technical Specifications. were reviewed to determine if the modification made by this The Battery Chargers are discussed in the Tech. Spec. but this modification will not require any statements made in the Tech. Spec.. Operatino License - Discussion and Response Facilitya change require Operating to the Ucense license. No. NPF-6 was reviewed to determine if the modifications made by this D The the batteryLicense. Operating chargers are not discussed in the Operating Ucense. Therefore, the DCP will not Confirmatory Orders - Di=ce=sion and Resannse Tha Confirmatory Orders (CO's) associated with Unit Two were reviewed to determine if modification of t battLry chargers would make any of the CO's information untrue. The CO's do not address the battery chargers. Therefore, this DCP will not require a change to the CO 2.) Will the proposed modification result in information in the following SAR documents (including
. trxt) being (a) no longer true or accurate, or (b) violate a requirement stated in the document?
Core Operating Umits Report NO SAR (multi-volume set for each unit) YES QAMO NO E-Pl:n NO PAGE- REV.O I
g ARKANSAS NUCLEAR ONE FORM TITLE. FORM NO. REV.
' 10CPR$0A0 REVEWCONTINuATION PAGE 1000.131C 2
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Document No. DCP 9632420201 Page 8 of 12
. RevJChange No. O FHA- f YES Bases of the Technical Specifications NO -
NRC Safety Evaluation Report NO Core Operatino Limits Report - Discussion and Resoonse Changes made by this DCP will not require any changes to the Core Operating Limits Report. SAR - Die == ion and Ra=nanse The reason for the YES answer is that the DCP will require revisions to SAR sections 1.3.3 [ Comp Drsign with Regulatory Guide Recommendations), Figure 1.2-4 [ Equipment Location Intermediate F
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2004], 3.10.2.2.7 [ Battery Chargers], 3.10.2.2.7.1 [ Equipment 2031, 2D33, 2034), 3.10.2.2.7.2 [ E 2D32), 3.10.2.2.8.1 [ Test Procedure], Table 3.2-6 section 4.5 [ Quality Assurance Summary L [ Analysis) Isolation Device - AC Power Circuits (4160 volt and 480 volt), DC Power Circuits, 8 System Description],8.3.2.1.2 [ Battery Chargers],8.3.2.2.1 [ Compliance with Des!gn Criteria an 8.3 2 [ Rating of Class 1E Electrical Distribution Equipment], Table 8.3-4A [125-Volt Oc Batteries 2011 L' Ch:,it), Table 8.3-48 [125 Volt DC Batteries 2012 Load Chert], Table 8.3-11 [125-Volt DC Engi , Ferture System Single Failure Analysis), Figure 8.3-6 [ Single Line Diagram E-2006], Figure 8.3-8 Diagram E-2014 sh.1], Figure 8.3-11 [ Single Line Diagram E-2014 sh. 4], Figure 8.3-12 [ Singl 2015 sh.1), Figure 8.3-15 [ Single Line Diagram E 2015 sh. 4], Figure 8.3-16 [ Single Litse Meter Di: gram E-2017 sh.1], Figure 8.3.66 [ Conduit & tray Layout E 2867 sh.1], and section 15.1.31 [ Loss System]. SIction 1.3.3 is being revised to remove the 'within eight hours' statement that is part of the Regula 1.32 (8/11/72) discussion. This statement was not in the FSAR. Based on a review of previous amendme th3 SAR it was determined that this statement was added before Amendment 6 to the SAR. However, th am ndment or any documentation supporting this statement could not be found. Safety Guide 32 (dated 8/11/72), which later became Regulatory Guide 1.32, states in section C.b (Regulatory Positio Supply) "The capacity of the battery charger supply should be based on the largest combined demands v ri:us steady-state loads and the charging capacity to restore the battery from the design minimum to the fully charged state, irrespective of the status of the plant during which these demands occur." Th tim 3 requirements evoked by this Safety Guide. Section 8.3.2.2.1 'DC Power System: Complian CritIria" which demonstrates compliance with Regulatory Guide 1.32 is a basic restatement of Safe with no reference to a time limit. EiC Design Engineering believes that the "within eight hour" statement was M PAGE gV.O
ARKANaAa NUCLEAR ONE- a --' - - * ~ *
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FORM NO. REV. 10CFRSS.58 MEVEW CONTINUATION PAGE 1000.131c 2 Document No. DCP 963242D201 Page 9 of 12 RevlChange No. O based on the calculated DC loading and recharge times at the time this SAR statement was added. Ho increases in DC loads, since this statement was added to the SAR, have resulted in an increase in th recharge times. Therefore, the *within eight hour" statement can no longer be met and since there is no bIses requirement for this capability, the eight hout reference will be deleted. l . l Stction 3.10.2.2.7, 3.10.2.2.7.1, 3.10.2.2.7.2, and 3.10.2.2.8.1 are revised to clarify the seismic documentation and requirements of the new battery chargers. !~ Tcble chargers3.2-6 is being component revised to change je existing battery charger component numbers to the new ba numbers. , Section 8.3.1.2 (Analysis) *lsolation Device
- Statement A.1, AC Power Circuits (4160 voit and 480 vo revised to add the manual transfer switches that feed battery chargers 20318 and 2032B, and A.2, DC Power i Circuit, is being revised because the existing DC cross train tie is 2D34 is being eliminated.
Srction 8.3.2.1 (DC Power System: Description) is revising the description to remove the reference to th chtrger and adding the description of one spare charger to each 1E DC bus (l.E. four charger replacing th Section 8.3.2.1.2 (Battery Chargers) is being revised to change the component numbers and the charger d;scription. Section spare 8.3.2.2.1 charger (Compliance to each train. with Design Criteria and Guides)is being revised to add the description of the Table 8.3-2. 8.3-4A, 8.3-4B, and 8.3-11 are revised to change component numbers and charger descri SAR Figures 8.3-6,8.3-8,8.3-11,8.3-12,8.3-15,8.3-18, and 8.3-66 are being revised to reflect the drawin changes required by this modification. Szction 15.1.31 (Loss of One DC System) is being revised to change the description of the standby char QAMO - Disen== ion and Reseanse
- i ThaManual.
the OA Manual does not contain sufficient details of the battery chargers, therefore, no changes are requ ' E-Plan - Dimen== ion and Response l-The E-Plan does not specify sufficient details on the battery chargers, therefore, no changes are required to t E-Plan.
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*m' H2V. O ARKANSAS NUCt. EAR ONE -
FORM TITLE: FORM NO. REV. 10CFRSS.SS REVIEW CONTINUATION PAGE 1000.131C 2 Document No. DCP g632420201 Page 10 of 12 Rev1 Change No. 0 FHA - Di===lan and Resnanse Section 5.8.4.B. and 5.8.4.6 (Bases for Components of interest List [ Unit 2) - Electrical P 13 change the component numbers and the list of MCC's which feed the battery chargers. S:;ction numbers. 5.g.5.8 (Components of Interest [ Unit 2] - Electrical Power) is being revised to changI Figure FP-2103 component numbers. (Fire Zones intermediate Floor Plan EL. 368 0 and 372-0) is being revised t Bases of the Technical Specification - D!se9ssion and Response This DCP does not alter any of the information stated in the Tech. Spec. Bases. Therefore, the being made by this DCP will not require any change to the Tech. Spec. Bases. NRC SERs - Discussion and Response 1 This DCP does not alter any of the information stated in any of the SERs. Therefore, the m made by this DCP will not require an change to the SERs. 3.) Will the proposed modification involve a test or experiment not described in the SAR? NO This DCP does not perform any test or experiment. A post modification test plan and 50.59 will be w cddress installation and post mod testing. 4.) Will the proposed modification result in a potential impact to the environment? NO As indicated impact by the NO answers on the Environmental Evaluation checklist this modification has on the environment. S.) Will the proposed modification result in the need for a Radiological Safety Evaluation per Sect NO Th's modification does not involve the processing of any radiological material. 6.) Will the proposed modification result in any potential impact to the equipment or facilities for Ventila Storage Cask activities per Section 6.2.4.B? NO This system.modification is replacing the Unit 2 battery chargers and will have no effect on the Ventila mL REV.O
ARKANSAS NUCLEAR ONE FORM TITLE: FORM NO. REV. 10CPR80.80 REVEW CoNTaAAATIoN PAGE 1ste.131C 2 Document No. DCP 963242D201 . Page 11 of 12 Rev/ Change No. O EVALUATION QUESTIONS: 1.) Will the probability of an accident previously evaluated in the SAR be increased? NO Chapter 15 of the SAR does evaluate "The Loss of One DC System". However, the cause of this ev bus fault or cable fault between the bus and the ESF distribution panel. This DCP will have no impa accident initiators or on any other accident initiators as evaluated in the SAR. This modification will system reliability by adding a spare 400 Amp battery charger to each DC train in place of the 200 battery charger. 2.) Will the consequences of an accident previously evaluated in the SAR be increased? NO Appendix R Components of Interest (battery chargers) and their cabling are changing as a result of this modification. However, revision to the Safe Shutdown Capability Assessment has resulted in no new th3 safe shutdown capability for any assumed exposure fire, This modification does not degrade, or prevent actions as described in the SAR nor does it alter used to evaluate the consequences of an accident described in the SAR. The offsite dose consequenc analyzed accident are unchanged by this modification. 3.) Will the probability of a malfunction of equipment important to safety be increased? NO The battery chargers are provided with redundant and independent power supplies. The battery consistent with the 125 VDC single failure analysis, SAR Table 8.3-11. The swing battery charger, which co supply either DC bus, has been removed. The new battery chargers design provides a dedicated s battsry charger for each DC train. This design helps to ensure separation of the DC power trains. The new chargers are seismically qualified and mounted. Based on these design features the probabili.ty of a malfu of Equipment important to safety will not be increased. 4.) Will the consequences of a malfunction of equipment important to safety be increased? NO Th3 consequences of the loss of a battery charger has been analyzed in SAR Table 8.3.11 and found to be acc:ptable. The fact that each DC bus will have a dedicated spare battery chargerwill increase the system reliability over the spare swing charger design. This modification causes no increase in the offsite dose consequences as a results of the malfunction of any equipment important to safety. 5.) Will the possibility of an accident of a different type than any previously evaluated in the SAR be creat NO The new battery chargers serve the same purpose as the existing battery chargers, to supply DC pow buses during normal and accident conditions as long as AC power is available. The loss of a batter been cvaluated in the SAR. Therefore, this modification does not create any mechanisms that would cau accident of a different type than what has been previously evaluated in the SAR. PAGE REV.O
1 t*t%3C ti d d. 'Y I ARKANSAS NUCLEAR ONE -- FORM TITLE: FORM NO. REv. 10cFRSO.88 REVIEW CONTINUATION PAGE 1000.131C 2 Document No. DCP 963242D201 Page 12 of 12 RevdChange No. 0 6.) Will the possibility of a malfunction of equipment important to safety of a different type than an evaluated in the SAR be created? NO The failure modes of the new battery chargers are the same as the existing battery chargers. As state failum of a battery charger has been evaluated in SAR Table 8.3-11. The new battery chargers meet the ) industrial codes and standards for battery chargers. Therefore, this modification does not create a malfunction of equipment important to safety of a different type previously evaluated in the SAR. 7.) Will the margin of safety as defined in the Bases of any technical specification be reduced? NO There are no margin of safety discussed in the Technical Specification Bases conceming the b PAGE ' REV.Oe
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ARKANSAS NUCLEAR ONE F07,M TITLE: FORM NO. REV. 10CFR80.59 DETERMINATION 1000.131A 2 PC 2.3 Page 1 of,3,
. Document No. DCP 943254D201 Rev./ Change No. A PAGE REV. O Title ANO-2 in-Mast Fuel Slopino Modifications Willthe proposed Activity:
1. Require a change to the Operating License including: Technical Specifications (excluding the bases)? YesO No@ Operating License? YesO NoS Confirmatory Orders? YesO NoS 2. Result in information it: the following SAR documents (including drawings and text) being (a) no longer true or accurate, or (b) violate a requirement stated in the document: Core Operating Limits Report YesO No@ SAR (multi-volume set for each unit)? . Yes@ NoO QAMO?* YesO No@ E-Plan?* YesO No@ FHA YesO No@ Bases of the Technical Specifications? YesO No@ NRC Safety Evaluation Reports? YesO No@
- 3. Involve a test or experiment not described in the SAR7 YesO No@
4. Result in a potentialimpact to the environment? (Complete Environmental Impact Checklist of this form.) YesO No@ 5. Result in the need for a Radiological Safety Evaluation per section 6.2.4.A? YesO No@ 6. Result in any potential impact to the equipment or facilities utilized for Ventilated Storage Cask activities per Section 6.2.4.B7 YesO No@ Basis for Determir=*3en: Thb DCP consists of modifications to the Refueling Machine end the Breathing Air System. in-Mast Sip equipment will be permanently installed on the hoist assembly of the Refueling Machine. A manual gate valv will be installed in an air line leading to the Refueling Machine control console. Only the SAR, listed above, (Figure 9.3-1) will be affected by this modification The remaining documents listed under items 1 and 2, above do n:t contain any details discussing the scope of activities in this DCP. This change does not involve a test or experiment. An Environment impact Checklist has been completed with no impact to the environment noted. This DCP does not involve the pmressing of radioactive material or create a new pathway for radioactive mat; rial. This DCP does not involve orimpact the spent fuel Ventilated Storage cask.
- Changes to these documents require an evaluation in accordance with 10CFR50.54.
See S;ction 6.2.1.B.
l i ARKANSAS NUCLEAR ONE FORM TITLE: FORM NO. REV. 10CFR80.88 DETERMINATION 1000.131A 2 PC-2.3 PAGE REV.O Page 2 of.a Document No. DCP 9632540201 f Rev> Change No. 0
References:
List sections reviewed in the Licensing Basis Documents, specified in questions 1 i l in parentheses. Controlled hard copies of the docume 1 searches such as LRS are not controlled and search text only, not figures or drawings. completed LDCR if LBD changes are required. Document . Section i { Core Op. Limits Report All Sections with keywords l Unit-2 SAR Unit-2 T/ Specs (Fuel Handling, Fuel Mast, in-Mast Sipping, ) Unit-2 SER Fuel Sipping, Breathing Air, Refueling, Machine control panelisolation) QAMO k6 fT-8f4LO- TD Sbkm h*4L4WE 9 d - l -A ff A Q ~ 5, an b LFO MkkNM- Nh
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2-/.r/91 Certified Rev%ers signature Saif U. Khan 09-10-97 Printed Name Date R0 viewer's certification expiration date: 06 05-99 ' Assistance provided by: Printed Name I Scope of Assistance Date l 1 1
I ARKANSAS NUCLEAR DNE FORM TULE: FORM NO. REV. 10CFR50J9 DETERMINATION 1000.131A 2 PC-2.3 Page2of_2 ENVIRONMENTAL IMPACT CHECKLIST l (UNIT 1 and UNIT 2) PAGE C ' l D cument No. DCP 963254D201 Rev1 Change No. O REV.O Complete the following checklist. If the answer to any checklist item is "Yes", an Environmental Eva required. See Section 6.2.1.E for additional guidance. l Will the Activity being evaluated: Yes No O E Disturb land that is beyond that initially disturbed during construction (i.e., new construction o buildings, creation or removal of ponds, or otherterrestrialimpact)? See Unit 2 SAR Figur 2.5-17. This applies only to areas outside the protected area. O E Increase thermal discharges to lake or atmosphere? O E Increase concentration of chemicals to cooling lake or atmosphere through discharge canal tower? O O increase quantity of chemicals to cooling lake or atmosphere through discharge canal or tower? O @ Modify the design or operation of cooling tower which will change drift characteristics? O @ install any new transmission lines leading offsite? O E Change the design or operation of the intake or discharge structures? ! O E Discharges any chemicals new or different from that previously discharged? O 3 Potentially cause a spill or unevaluated discharge which may effect neighboring soils, surface water or ground water? t O @ involve burying or placement of any solid wastes in the site area which may effect runoff, surface water orground water? O -E involve incineration or disposal of any potentially hazardous materials on the ANO site? O E Result in a change to nonradiological effluents or licensed reactor power level? O @ Potentially change the type or increase the amount of non-radiological air emissions from the ANO site. t l l 1 l
ARKANSAS NUCLEAR ONE FORM TITLE: FORM NO. j 10CFR80.89 EVALUATION REV. 1000.131B 2 Page ,1, ofj 10CFR50.59 Eval. No. E Pd-M -OCS (Assigned by PSC) Document No. DCP 983254D201 RevdChange No. 0 Title ANO-2 in-Maat Fuel Slocino Modifications PAGE 10 REV.O ATTACHED.. EACH QUESTION MUST BE A CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES l 4 to all questions is "No,"then the proposed change does . er no{ 1. ( Will the pinbability of an accident previously evaluated in the SAR be : increased? Yes O No @ l to provide isolation to the Refueling Machine pneumat - safety valve. isolation related system. Normal operation of the Breathing Air System is not affec Therefore, the probability of an accident previously evaluated in . 2. Will the consequences of an accident previously evaluated in the SAR be increased? Yes O No @ ) No. The Breathing Air System does not serve to mitigate the consequence evaluated an in theevaluated accident previously SAR. Therefore, in the SAR. the addition of an isolation valve will not increas 3. Will the probability increased? of a malfunction of equipment important to safety be Yes O No @ interface with any safety related equipment.No. The Breathing Air System important to safety will not be increased. Therefore, the probability of a malfunction of equipment 4. Will the consequences beincreased? of a malfunction of equipment important to safety Yes O No @ No. The Breathing Air System, as stated above, in response to Question 2, doe consequences of an accident. Also, the new valve will not interface with any safety r Therefore, important the to safety addition of this new valve will not increase the consequences of 5. Will the possibility of an accident of a different type than any previously evaluated in the SAR be created? Yes O No @ No. This change does affect the rewits of any analysis or cause the plant to be ope manner. SAR will not Therefore, be created. the possibility of an accident of a different type than any previously I
ARKANSAS NUCLEAR ONE FORM TITLE:
' FORM NO. REV.
10CMO.88 EVALUATION 1000.131B 2 6. Will the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR be created? Yes O No @ system. Also, as mentioned above, the new valve will not Therefore, the possibility of a malfunction of equipment important to safety of a previously evaluated in the SAR will not be created. 7. W'i ff.e margin of safety as defined in the bases for any technical specification be reduced? Yes O No E safety related. No Technical Specification bases are impactsd Certified Rev Mer's signature Saif U. Khan Printed Name 91097 Date Reviewer's certification expiration date: 06-05 99 Assistance provided by: Printed Name Scope of Assistance Date PSC review by: 1 _k / Date:_ 2- 10 Y DOCUMENT NUMBER PAGE-ER-963254D201 8tEV. O
1 Page 1
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ARKANSAS NUCLEAR ONE FORM NO, REV. FORM TITLE: 1000.131 A 2 PC-2,3 10CFR50.59 DETERMINATION This Document contains 8 Pages. Rev1 Change No. O Document No. DCP 96-3523-D202 MOV MOD FOR 2CV-1026-2, 2CV-1036-2, 2CV-1037-1, 2CV-1038 2, 2CV-1039-1 Title 1076-2 Willthe proposed Activity:
- 1. Require a change to the Operating License including:
YesO NoS Technical Specifications (excluding the bases)? YesO NoS , I Operating License? YesO No@ Confirmatory Orders? l 2. Result in information in the following SAR documents (including drawings and text) being ; (a) no longer true or accurate, or (b) violate a requirement stated in the document: YesO No@ Core Operating Limits Report Yes@ NoO SAR (multi-volume set for each unit)? YesO No@ l QAMO? YesO No@ l E Plan? YesO No@ FHA YesO No@ Bases of the Technical Specifications? YesO No@ NRC Safety Evaluation Reports? YesO No@
- 3. Involve a test or experiment not described in the SAR? l
- 4. Result in a potential impact to the environment? (Complete Environmental YesO No@
Impact Checklist of this form.) YesO No@ 5. Result in the need for a Radiological Safety Evaluation per section 6.2.4.87
- 6. Result in any potential impact to the equipment or facilities YesO No@
utilized for Ventilated Storage Cask activities per Section 6.2.4.87 Basis for Determinatio.t (See Continuation Sheets)
- Changes to these documents require an evaluation in accordance with 10CFR50.54 See Section 6.2.1.B.
PAGE REV.O DCP 96-3523-D202, Page - of __.
F I i Page ARKANSAS NUCLEAR ONE FORM NO. REV. FORM TITLE: 1000.131 A 2 PC 2.3 l 10CFR50.59 DETERMINATION See Attached Pages 1 thru 8 Rev./ Change No. O Document No. DCP 96-3523-D202
References:
List sections reviewed in the Licensing Basis Documents, specified in questions 1,2 and 3. If a keyword search was done on LRS, 'all" may be entered under "Section" with the keywor in parentheses. Controlled hard copies of the documents shall be reviewed as computer-searches such as LRS are not controlled and search text only, not figures or drawings. Attach completed LDCR if LBD changes are required. Section Document 2CV-1026, 2CV-1036, 2CV-1037, 2CV-1038, 2CV-1039, 2CV 1076. All Unit 2 LBDs (Refer to continuation EFW, EFAS, MSIS,2P35,2P65, stroke time,2P7A,2P7B,2P75 EF sheet, Page 8 of 8 for listing) l l discharge, containment w/10 valves, LOCA. l 1 i l 1 i 7/30/97 L J. Greg Hines
. Date l Cedified R9 viewers Signature Printed Name Reviewers certification expiration date: 1/30/96 1
i Assistance provided by: Date Printed Name Scope of Assistance 7/30/97 Electricalimpacts of Modification Eugene Miller l l PAGE REV.O 1 l N'o on u?'t n?f)2. Paae _ of
l Pace 3 ARKANSAS NUCL. EAR ONE FORM ND. REV. FORM TITLE: 1000.131A 2 PC-2.3 10CFR50.59 DETERMINATION ENVIRONMENTAL IMPACT CHECKLIST { (UNIT 1 and UNIT 2) 1 i RevlChange No. O Document No. DCP 96-3523-D202 f Completc the following checkilA. If the answer to any checklist item is "Yes", an Environmenta required. See Section 6.2.1.E for additional guidance. I Willthe Activity being evaluated: Y.es Ng Disturb land that is beyond that initially disturbed during construction (i.e., new constr O E buildings, creation or removal of ponds, or other terrestrialimpact)? See Unit 2 SAR F 2.5-17. This applies only to areas outside the protected area. Increase thermal discharges to take or atmosphere? O S Increase concentration of chemicals to cooling lake or atmosphere through discharge O S tower? . E increase quantity of chemicals to cooling lake or atmosphere through discharge cana O tower? E Modify the design or operation of cooling tower which will change drift characteris O E Install any new transmission lines leading offsite? O E Change the design or operation of the intake or discharge structures? O E Discharges any chemicals new or different from that previously discharged? O E Potentially cause a spill or unevaluated discharge which may effect neighborin O water or ground water? Involve burying or placement of any solid wastes in the site area which may effe O S surface water or ground water? Involve incineration or disposal of any potentially hazardous materials on the AN O E Result in a change to nonradiological effluents or licensed reactor power level? O E 4 E Potentially change the type or increase the amount of non-radiological air em O ANO site. PAGE /l
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REV.O DCP 96-3523.D202, Page _ - of .__.
Pace 4 of 7 ARKANSAS NUCLEAR ONE FORM NO. REV. FORM TITLE: 1000.131C 2 10CFR50.59 REVIEW CONTINUATION PAGE RevJChange No. 0 Document No. DCP 96-3523-D202 10CFR50.59 Review Continuation Paoe Syne 9 of Modification This Design Change Package (DCP) involves the actuator replacement and enhancing the valve d 2CV-1026 2,2CV-1036-2,2CV-1037-1,2CV-1038-2,2CV-1039-1, & 2CV-1076-2. The actuators for 2CV-and 2CV-1038-2 will be upgraded from the existing Limitorque SMB-00-15 (460 VAC) actuators to Limit SMB-0 25 (460 VAC) actuators. The actuators for 2CV-1026-2. and 2CV-1076-2 will be upgraded from existing Limitorque SMB-00-7.5 (125 VDC) actuators to Limitorque SMB 0-15 (125 VDC) actuators. The actuators for~2CV-1037-1, and 2CV-1039-1 will be upgraded from the existing Limitorque SMB-00-15 ( actuators to LimNorque SMB-0-15 (125 VDC) actuators. The disks, yokes, and yoke clamps for the exis Anchor-Darling gate valves will be refaced with parts which will be capable of sustaining higher thrust l modification to the valve bodies or piping are required as a result of this DCP Existing power and contr can be utilized but thermal overloads and circuit breakers will need to be replaced. Stroke times for the assemblies will be within the 35 second allowable stroke time provided by NED analysis under IRF 7873 DESIGN BASIS FUNCTIONS for 2CV-1036-2. 2CV-1037-1. 2CV-1038-2. and 2CV
- 1. During emergency operation, the valves feeding the intact steam generator are required to fully open in order to provide EFW flow to the respective steam generator. This action i by receipt of an EFAS signal.
- 2. To provide isolation of EFW to a ruptured steam generator, the affected MOVs receiv signal to close (EFAS signal to feed good steam generator overrides the MSIS closure
- 3. To maintain intact steam generator level post accident, the respective MOVs are attemately opened and closed admitting and isolating EFW flow. This function is maintained by E
- 4. These valves are not outboard containment isolation valves for penetrations 2P35 and 2P65 though they are the first remote / manual valve outside of these penetrations. "These pe (2P35 and 2P65) are associated with the secondary side of the steam generators and subject to GDC-57 since the containment barrier integrity is not breached during DBA conditions. The containment boundary or barrier against fission product leakage to the environment is the inside surface of the steam generator tubes, the outer surface of the line above the emanating from the steam generator, and the outer surface of the steam generat bottom tube sheet." (Table 6.2-26 Note 9)
DESIGN BASIS FUNCTIONS FOR 2CV-1026-2. and 2CV-1076-2 ARE AS FOLLOWS:
- 1. During emergency operation, the valves are required to open in order to provide EFW intact steam generator (s). This action is initiated by EFAS logic.
- 2. To provide isolation of EFW to a ruptured steam generator, the affected MOVs signal to close (EFAS signal to feed good steam generator overrides the MSIS clo
- 3. These valves are redundant valves to 2CV-1037-1 and 2CV-1039-1 (respectively) to ensure isolation of EFW flow to a ruptured S/G.
PAGE REV. 0 , BASIS FOR RESPONSE TO QUESTION 1 The safety functions potentially affected by this modification are Emergency Feedwater. It i changes to the Technical Fpaification! or Bases are required as outlined below: MD ORM7'4-tFDS@2. Pme of ;
= -
Page 5 of 7 ARKANSAS NUCLEAR ONE FORM NO. REV. FORM TITLE: 1000.131C 2 10CFR50.59 REVIEW CONTINUATION PAGE During Emergency Feedwater injection, the modified MOVs will be used to assure adequate feedwater the EFW Pumps to the S/Gs. Changes made to the operators and valves per this DCP will maintain the and quality requirements of the systems and components, and will pmvide enhanced reliability, thus 3.7.1.2 requirements. No changes to these specifications are required. These valves are not containment isolation valves per Table 6.2 26 Note ., therefore Technical Specification 3/4.6.3 requirements are not impacted by this change. Technical Specification 3.8, electrical power systems was reviewed. Surveillance requirements as dire T/S 4.6.2.1 and 4.8.2.3 will not be impacted by this change. Changes in the battery loading have been ana under engineering calculation 963523D202-03/0 and the additional loads for 2CV 1026-2 and 2CV-10 supplied by Battery 2D12 with no reduction in design margin. MOVs 2CV-1037-1 and 2CV1039-1 existing DC motors. No changes to these specifications are required. T/S Table 3.3-5 lists the limiting T/S EFW system response time of 97.4 seconds. IRF #7873 was respo Nuclear Safety which concluded that an acceptable response time for the EFW valves was 35 seco ensure adequate margin between the allowed and the field tested stroke time,35 seconds was used stroke time value for this modification. The calculated stroke time for two AC powered MOVs,2CV-1036 2CV-1038-2, is 29.5 seconds. The calculated stroke times for the four DC powered MOVs, 2CV 102 1037-1,2CV 10391, and 2CV-1076-2, are 33.7. 34.2,33.5 and 33.1 seconds respectively. The calc time is based on full valve travel and motor RPM at degraded voltage conditions with full DP loading du duration of the stroke. Therefore the calculated stroke times represent a worst case scenario. Actual f are expected to yield stroke times that will be less than calculated stroke times. The Operating Ucense and Technical Specifications including their bases and COLR sections h reviewed and no impact was found, the design change controls used are in compliance with the QA E Plan is not affected because this DCP does not involve the emergency plan and equipment. FHA is not affected even though the DCP involves Components of IntereM and related power sup identifies the component tag numbers but specifies no ratings, so the DCP does not alter the exis addition, the DCP does not require changes in cables and raceways that could affect the existing analysis. No changes to the Operating License nor to any Confirmatory Orders No willConfirmatory be required as a re Orders This was verified by review of the LBD's using the Licensing Research System. regarding EFW were noted, however Confirmatory Order 2N-80-156 regarding Environm Safety-Related equipment is assured to be met by EQ Program review of the new actuators. BASIS FOR RESPONSE TO QUESTION 2 No SAR 3.1.2 CRITERlON 17 and 18 discusses the electrical power system design, inspections, and changes to the text are required. SAR 3.1.5 CRITERION 52 and 53, discusses containment leak rate testing and containment inte not Containment boundary valves No changes to the text are required. SAR 3.1.5 CRITERlON 54,55,56, and 57 discusses the piping systems penetrating containment. Re also made to Table 6.2-26 and SAR Section 6.2.4, which lists reactor building isolation valv valve arrangement scheme for each penetration. No changes to the text or table are required. SAR 3.6.4.1 discusses high energy pipe breaks outside containment. No changes to the text are SAR 3.6.4.2.5 discusses emergency feedwater pipeline break points. No changes to the text are PAGE b REV.O DCP 96-3523-D202, Page of_ --
ARKANSAS NUCLEAR ONE Page 6 of 7 FORM TITLE: FORM NO. REV. 10CFR50.59 REVIEW CONTINUATION PAGE 1000.131C 2 SAR Table 6.2 26 lists Reactor Building isolation Valve Information. Penetration 2P35 is the EFW to S/G 2E24A, , and penetration 2P65 is the EFW to S/G 2E248. Per Note (9) to the table. "These penetrations are associated with the secondary side of the steam generators and are not subject to GDC-57 since the containment bamer integrity is not breached during DBA LOCA conditions. ... Valves associated with these penetrations are not containment isolation velves. No changes to the table are required. SAR 7.1.1.2 and 7.4.1.2 identifies the emergency feedwater system as one required for safe shutdown and available in the event normal power sources are unavailable. No changes to the text are required. SAR 7.3 discusses the Engineered Safety Features Actuation System, in particular it refers to the Emergency Feedwater (7.3.1.1.11.8) and to the Main Steam isolation System (7.3.1.1.11.4). Changes made to the MOVs per this DCP will maintain the design and quality requirements of the system and will provide enhanced reliability, thus meeting these SAR requirements. No changes to the text are required. SAR Figure 7.3-12 shows the Functional Description & Logic Diagram for the Emergency Feedwater System. In particular, it lists by component tag number 2CV-1026-2, 2CV-1036-2, 2CV-1037-1, 2CV-1038-2, 2CV-1039-1, and 2CV-1076-2. No changes to the text are required. SAR Figures 8.314 and 8.3-110 will be changed as a result of this DCP. The changes reflect the larger motor i sizes for 2CV-1026-2,2CV-1036-2,2CV-1038-2, and 2CV 1076-2 for the new actuators. The function and design basis of the MOVs will not be impacted by this change. The required LDCR is attached to this 50.59 and is included as part of the DCP. SAR 9.3.6.1.2, General Design Criterion 34 requires that a residual heat removal system capable of performing its design function in the event of a single failure be provided. Both the redundant safety related EFW system, described in Section 10.4.9, and the Shutdown Cooling System are designed to provide this residual heat removal capability. The operating relationship between these two systems in meeting the General Design Criterion 34 requirement is described in Section 9.3.6.2.1, No changes to the text are required. SAR Section 1F.4.9 discusses the design basis operational requirements for the Emergency Feedwater System. 2CV-1026 2,20V-1036-2,2CV-1037-1,2CV-1038-2,2CV-1039-1, and 2CV-1076-2 are discussed. No changes to the text are required. The Single Failure Analysis done for the EFW in SAR Table 10.4-11 will not be affected by this DCP. The single failure of either 2CV-1026-2, 2CV-1036-2, 2CV-1037-1, 2CV-1038-2, 2CV-1039-1, or 2CV. 1076 2 to open in the event of an EFAS will not affect the functioning of the system since the redundant parallel path will still be operable. If one of the two series valves in the supply lines fails to close during steam generator isolation (upon receipt of an EFAS) the other valve will close (valve uses a separate redundant power supply from the other valve that it is in series with.) If EFAS is present and EFW bus 2001 fails, motor driven EFW pump 2P7B will not start. However turbine driven EFW pump 2P7A can supply either steam generator. SAR Figure 10.4-2 shows 2CV-1026-2,2CV-1036-2,2CV-1037-1,2CV-1038-2,2CV-1039-1, and 2CV-1076-2 by tag number. No changes to the text are required. 1 SAR Table 14.1-1 states that the EFW system must be capable of performing per Section 10.4.9. Changes j made to the MOVs per this DCP will maintain the design and quality requirements of the system and will provide l ' enhanced reliability, thus meeting these SAR requirements. No changes to the text are required. The changes made by this DCP will not alter the purpose, function, or operation of the subjects valves but will instead improve the valve reliability. The modification will not introduce a new potential failure mode. PAGE REV.O nem-u- ew a
ARKANSAS NUCLEAR ONE PeGe 7 of 7 FERM TITLE: FORM N3. REV. 10CFR50.59 REVIEW CONTINUATION PAGE 1000.131C 2 Review of other LBD's indicates that no changes are required. References Used for Evaluation Document Sections Reviewed Applicable to Question ANO-2 T/S and Bases 2.1, 3/4.3.2, 3/4.7.1.2, 4.4.5.3, 3.4.6.2, 3/4.6.3, 1 q 3/4.6, 6.9.6, Table 3.3-6, Table 4.3-2, Table 3.3-3, Table 3.3-4 ANO-2 SAR 3.1.2, 3.1.5, 3.5, 3.6.1, 3.6.4, .6.4.1, 3.6.4.2, 2 l 5.2.1.6, 6.2.4, 7.1.1.2, 7.3, 7.3.1.1.11.4, I 7.3.1.1.11.8, 7.4.1.2, Figure 7.3-2b, Figure 7.3412, Figure 8.3-110, 9.3.6.1.2, 9.3.6.2.1, 10.4.9,15.1.8,15.1.14.1.2, Table 3.3-3, Table 3.3-4, Table 6.2-26. Table 6.2-33, Table 10.4-11, Table 14.1.-1, Figure 10.2-3. Figure 10.4- )
- 2. Figure 15.1.14-6 Figure 15.1.14 46 ANO-2 SER and All 2 Amendments FHA Appendix 98 - Section 5.8 and 5.9 2 QAMO Section 3.0 2 l
1 1 i PAGE /5 R EV. O DCP 96-3523-D202, Page of
Page 40 of 43 Page 1 of '/ Document No. DCP 9/o 35'23Dzo2 Rev./ Change No. O 10CFR50.59 Eval. No. ((h N ll (Assigned by PSC) Title NO\/ }!)Db A%' 2cV-to Zb-2 , 2c V-to u-Z. 2C V-w37-1,2G?-/031-Z , 2ct)-tub 9-t , f 2cv to A WRITTEN RESPONSE PROVIDING THE BASIS FOR THE ANSWER TO EACH QUESTION MUST BE ATTACHED. EACH QUESTION MUST BE ANSWERED SEPARATELY. A SIMPLE STATEMENT OF CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES GUIDANCE FOR RESPONSE. If the answer to any question on this form is "Yes," then an unreviewed safety question is involved. If the answer to all questions is "No," then the proposed change does not involve an unreviewed safety question.
- 1. Will the probability of an accident previously evaluated in the SAR be increased? Yes No /
- 2. Will the consequences of an accident previously evaluated in the SAR be increased? Yes No /
- 3. Will the probability of a malfunction of equipment important to safety be increased? Yes No (
- 4. Will the consequences of a malfunction of equipment important to safety be increased? Yes No I
- 5. Will the possibility of an accident of a different type than any previously evaluated in the SAR be created? Yes No I
- 6. Will the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated No -j in the SAR be created? Yes
- 7. Will the margin of s'afety as defined in the bases for any 7 tech ' cal sp cification be reduced? Yes No
- 3. Ow, N wis 8 ll 97 Certifiec Riviewer's Signature DrAnted Name Date Reviewer's certification expization date: //30/94 l Assistance provided by:
Scope of Assistance Date Printed Name PSC review by: M Date: Gh l PAGE REV.O OcP %-3523-02o2 P- 3 FORM NO. REV, FORM TITLE. 1000.1318 2 10CFR50.59 SAFETY EVALUATION _ _ --
ARKANSAS NUCLEAR ONE Paes 't of G F5RM TITLE: FORM NO. REV. 10CFR50.59 REVIEW CONTINUATION PAGE 1000.131C 2 Document No. DCP 96-3523-D202 Rev/ Change No. 9 10CFR50.59 Review Continuation Page For 10 CFR 50.59 Safety Evaluation
- 1. The probability of an accident previously evaluated in the SAR will not be increased.
The system being modified (EFW) is an accident mitigation system which is not involved with the initiation of any accident previously evaluated in the SAR. The modifications do not affect the probability of failure of any component and 'even if they did, those components affected are not accident initiators.The 2SAR sections listed in the Determination references and 2SAR section 15 were reviewed for accidents relevant to the modifications made by this DCP. The modified MOVs have active safety functions as defined below: DESIGN BASIS FUNCTIONS FOR 2CV-1036. 2CV-1037. 2CV 1038. AND 2CV-1039 are as follows:
- 1. During emergency operation, the valves feeding the intact steam generator are required to be fully open in order to provide EFW flow to the respective steam generator. This action is initiated by receipt of an EFAS signal.
- 2. To provide isolation of EFW to a ruptured steam generator, the affected MOVs receive an MSIS signal to close (EFAS signal to feed good steam generator overrides the MSIS closure signal).
- 3. To maintain intact steam generator level post accident, the respective MOVs are attemately opened and closed admitting and isolating EFW flow. This function is maintained by EFAS logic.
- 4. These valves are not outboard containment isolation valves for penetrations 2PS5 and 2P65 even though they are the first remote / manual valve outside of these penetrations. *These penetrations (2P35 and 2P65) are associated with the secondary side of the steam generators and are not subject to GDC-57 since the containment banier integrity is not breached during DBA LOCA conditions.' The containment boundary or barrier against f!ssion product leakage to the environment is the inside surface of the steam generator tubes, the outer surface of the fines emanating from the steam generator, and the outer surface of ths steam generator above the bottom tube sheet (Table 6.2-26 Note 9).
DESIGN BASIS FUNCTIONS FOR 2CV-1026-2. AND 2CV-1076-2 are as fo!!m:
- 1. During emergency operation, the valves are required to open in order to provide EFW flow to the intact steam generator (s). This action is initiated by EFAS logic.
- 2. To provide isolation of EFW to a ruptured steam generator, the affected MOVs receive an MSIS signal to close (EFAS signal to feed good steam generator overrides the MSIS closure signal).
3.- These valves are redundant valves to 2CV-1037-1 and 2CV-1039-1 (respectively) to ensure isolation of EFW flow to a ruptured S/G. Nothing in the modifications of this DCP will increase the probability or consequence of a LOCA, MSLB, or other accident. The valves' safety-related functions and safety significant capabilities are improved by the new valve parts and actuators which will enable the MOVs to function with enhanced reliability. Nothing in the DCP will adversely affect the EFWS or cause it to operate outside of its design or testing limits. The modification will not impose unanalyzed loads on the piping or supports, cause system vibration, water hammer, fatigue, corrosion, thermal cycling or degradation in the environment of the system or the piping and piping components. y PAGE / =I REV.O DCP 96 35 3-D202, Page of
ARKANSAS NUCLEAR ONE Pace 3 of 4 FORM TITLE: FORM NO. REV. 10CFR50.59 REVIEW CONTINUATION PAGE 1000.131C 2 Failure modes and effects analysis (FMEA) and single failure criteria (SFC) were adequately the Mechanical Design Change Summary of the DCP. The DCP does not change the capability of meeting SFC for protecting the EFW pressure boundary. The replacement actuators and va not change the function or failure inodes of EFW components. Replacing the actuators an will maintain systems, component, and structural quality standards of design and materials. The modification increases the capability of the affected systems and valves to perform their safety The modifications component. do not add or delete automatic or manual features of a safety system, stru No new system interactions or failure modes are introduced. function, quality and failure modes of components, systems and structures will rema improved existing by the DCP, the probability of an accident previously analyzed remains at least the design. 2. The consequences of an accident previously evaluated in the SAR will not be increased. The modifications are in accordance with the requirements of the codes of record. The modif not change the reliability of the systems involved and do not alter the consequences of any e accident. The valve stroke times are within the allowed stroke time. No accidents evaluated in the 2SAR have their radiological consequences altered as a resul The modified valves and actuators will not change, degrade or prevent the EFW syste its safety function capabilities as described or assumed in any accident discussed in the 2SAR. The modifications improve the capability of the MOVs to operate with enhanced reliability that wil that the valves can be opened or closed as required during design basis conditions. The stroke the Designnewvia actuators IRF 7873. is within the 35 second allowable stroke time as provided by Nuclea There is no added dose consequence due to change in stroke time from the existing actuators to the new actuators. 3 The probability of the malfunction of equipment important to safety will not be increased. The function and mode of operation of the valve will not change due to this DCP. Nothing in modifications significant adversely impacts the capability of equipment or systems to perform their sa functions,
- 4. )
The consequence of a malfunction of equipment important to safety will not be increased. f Modifications made per this DCP maintain the single failure criterion for these valves. Eachi each steam generator is provided with redundant series connected valves, in accordance wj i V Jre criterion. This ensures isolation of the steam generators and feeding of the remaining in gnerator(s) as required during an EFAS operation of the EFWS following a postulated MSLB or FWL {1 The modifications made by this DCP will not increase the consequences of a malfunction of th The valves covered by this DCP are not containment boundary valves, and the modifications h impact on the radiological consequences or affect dose limits stated in the LDBs or in 10 CFR 5. The possibility of an accident of a different type than previously analyzed in the SAR will not be created. Changes made to the valve and operators will not affect the RCS boundary or the conta or the valves listed as RCS or containment boundary valves. Replacing the valve parts and o does not change the function or failure modes of any component, system, or structure. standard are maintained. Changes to the piping analysis have been reviewed and no new seismic restraints will be needed. No new missiles or seismic 11/1 interactions are created.
- 6. \
The possibility of a malfunction of equipment important to safety of a different type than previously evaluated in the SAR will not be created. PAGE /g REV.O DCP 96 3523-D202, Page .. of _. J
ARKANSAS NUCLEAR ONE Pagi 4 of 4 FORM TITLE: FORM NO. REV. c 10CFR50.59 REVIEW CONTINUATION PAGE 1000.131C 2 Thefunction and operating mode of systems or components are not being revised by this DCP, and no new-failure sysMs modes different are introduced. from those TheSAR. evaluated in the modified valves and actuators will not cause a m Therefore the possibility of malfunction of any equipment important mog(ications of this DCP. to safety beyond what has already been evaluated is not created by the 7. The dia gin of safety as defined in the bases for Technical Specifications will not b 5 , The. Survelliance Requirements and Bases for the Unit 2 Technical Specifications were reviewed. Replacing valve parts or operators does not reduce the margin of safety in the bases of the Technical Specifications. The existing quality standards are maintained with these modifications, therefore preserving the single failure criteria implicit in the Technical Specifications and Bases. Impacts to Batt 2D12 loading due to changes in motor sizes for MOVs 2CV-1026-2 and 2CV-1076-2 have been ana under;calculation 963523D202-03/0, defined in the bases will therefore not be reduced by these modifications.and there is I a s L 4 h r i 1 PAGE /7 REV. O l DCP 96-3523-D202, Page of
runM t e s t.r.. i FORM NO. REtf 10cFR50.89 DETERMINATION 1000.131A 2 Pc.2.3 , ( Page _ of Document No. ER963559D301 i RevlChange No. ! l Title Computer and Telephone Rooms Power & AC Uparade PAGE REV.O Will the proposed Activity: 1. Require a change to the Operating License including: Technical Specifications (excluding the bases)? YesO No@ { Operating License? YesO No@ Confirmatory Orders? YesO No@ 2. Result in information in the following SAR documents (including drawings and text) being (a) no longer true or accurate, or (b) violate a requirement stated in the document: Core Operating Limits Report YesO No@ SAR (multi-volume set for each unit)? Yes@ nod QAMO?* YesO No@ E Plan?* YesO No@ ) FHA YesO No@ Bases of the Technical Specifications? YesO No@ NRC Safety Evaluation Reports? YesO No@ 3. Involve a test or experiment not described in the SAR? YesO No@ 4. Result in a potentialimpact to the environment? (Complete Environmental Impact Checklist of this form.) Yes@ nod 5. Result in the need for a Radiological Safety Evaluation per section 6.2.4.A? YesO No@ 6. Result in any potential impact to the equipment or facilities utilized for Ventilated Storage Cask activities per Section 6.2.4.87 YesO No@ Basis for Determination:
- 1. The changes being made to CSB, GSB, EOF, and Admin buildings do not affect the oper documents.
- 2. This modification will change the name of the incinerator Building. SAR figures 1.2 1, 2.4-8, 3.5-BA, 8.3-17, 8.3-53 and SAR Chapeter 8 table of contents for figure 8.3-17 will need to be changed to reference building name No other SAR documents are affected.
- 3. No new tests or experiments will be implemented per this modification.
a.wwena quGLhAR ONE ~ FORM TITLE:
' 10CFR40.80 DETERMINATION FORM NO REV.
1000.131A i 2 PC-2.3
- 4. S o environm:ntal checidist. Environmental review performed per ER 9635591302 S. This modification does not involve work in radiological areas or affect radiologic
- 6. This modification does not affect the cask storage activities or facilities.
SeeChanges to these Section 6.2.1.B. documents require an evaluation in accordance with 10CFR50.54 PAGE REV.O 9
3 tot <M TITLE. ~ 10CFR50JS DETERMINATION FORM NO REV 1000.131A 2 PC.3.3 Page _ of Document No. ER963559D301 Rev> Change No. 0 ~
References:
PAGE /v$ REV.O List sections reviewed in the Licensing Basis Documents, specified i , keyword search was done on LRS, "all" may be entered under'Se ue in parentheses. Controlled hard copies of the documents sha!! - be re searches completed LDCR such aschanges if LBD LRS are arerequired. not controlled and search text only, not figu Document Section LRS All; Keywords: Building" OR GSB, Admin
- Building, node 3, te numbers in modification package), backup generator *, EOF pow Unit 2 SAR Section 9.5.2.2 Figures 1.2-1,2.4-8,2.5-17,3.5-8A, 8.3-17, of Contents for Figure 8.3-17. Chapter&88.3 53; Table
. o, ,,
Certified Reviewer's S.gfiature M Kan Montoomerv 4/14/97 Pnnted Name Date Reviewer's certification expiration date: i 2/26/99 < Assistance provided by: Pnnted Name None . Scope of Assistance Date
l FORM TITLE: ARKANSAS NUCLEAR ONE 10CFRSO.58 DETERMINA730N FOCM NO. REV.
}
1000,131A 2 Pc.2.3 Page _ of ENVIRONMENTAL IMPACT CHECKLIST (UNIT 1 and UNIT 2) Document No. ER963559D301 RevlChange No. 2 PAGE O REV.O required. See Section 6.2.1.E for additional guidance. Completeon the is foll Will the Activity being evaluated: v.. ne ! O E buildings, creation or removal of ponds, or o 2.5-17. This applies only to areas outside the protected area. I O O increase thermai discharges to lake or atmosphere? O g increase tower? concentration of chemicais to cooiing iake or atmosphere I O O increase tower? quantity of chemicais to cooiing take or atmospnere tnroug O @ Modify the design or operation of cooling tower which will change drift O @ k Install any new transmission lines leading offsite? O g Cnange the design or operation of the intake or discharge structures? l' O @ oischarges any chemicais new or different from that previously disenarg
@ O Potentially water cause or ground water? a spill or unevaluated discharge which may , e effec O @
involve surface burying water or groundor placement water? of any solid wastes in the site area whic D. E involve incineration or disposal of any potentially hazardous materials on O E Result in a change to nonradiological effluents or licensed reactor power lev @ O Potentially ANO site, change the type or increase the amount of non radiologi t
FORM TITLE: FORM NO. REV. 10CFR50J9 EVALUATION
- 1000.1313 2 Page ,_, of 10CFR50.59 Eval. No. FFW9'P F10 (Assigned by PSC)
Document No. ER963559D301 RevlChange No. 0 Title Computer and Telephone Rooms Power & AC Unorade PAGE REV.O l l ATTACHED. EACH QUESTION MUST BE AN CONCLUSION IS NOT SUFFICIENT ATTACHMENT 2 PROVIDES . If the answer to any question on this form is "Yes," then an unreviewed safety question to all questions is "No," then the proposed change does not involve an .unreviewed If the answer saf 1. Will the probability of an accident previously evaluated in the SAR be increased? Yes O No @ j The changes per this modification will be performed outside the power block. These Training Center, and Admin Building. This modification accident previously evaluated in the SAR. 2. Will the consequences of an accident previously evaluated in the SAR be increased? Yes O No @ This modification does not affect any radiological systems or processes or the cause of radiological releases to the atmospnere. The offsite dose will not be affected , and therefore the consequences.of an acrident previously evaluated in the SAR will not b increased. 3. Will the probability increased? of a malfunction of equipment important to safety be Yes O No @ This modification will not be installing safety related equipment nor will it be insta and is physically and electrically isolated from safety-re of a malfunction of equipment important to safety will not be increased. 4. Will be the consequences increased? of a malfunction of equipment important to safety ! Yes O No @ This modification does not affect safety related equipment or the equipment connected to safety related components that might cause an increase to offsite dose. be The consequences of a malfunction of equip important to safety will not increased. 5. Will the possibility of an accident of a different type than any previously evaluated in the SAR be created? Yes O No @ { This modification does not install or affect safety related equipment or the potential for offsite dose increases. This modification will not prevent plant operators from
FORM TITLE ARKANSAS NUCLEAR ONE 8 D( 10CFR60.69 EVALUATION FORM NO. REV. 1000.131B 2 perforMng theirjob as required by the plant license and licensing documents. A new accident of a different type than any previously evaluated in the SAR will not be created. PAGE REV.O 6. Will the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR be created? I Yes O No @ This modification does not install or affect safety related equipment. The possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR will not be created.
- 7. \
Will the margin of safety as defined in the bases for any technical specification be reduced? Yes O No @ This mcdification does not affect the technical specification bases or any margin of safety called out in the technical specification bases.
~ ~
... t ,
Certified Reviewer'iSignatui Ken Montoomerv _ 4/14/97 Printed NEme Date Reviewer's certification expiration date:_ 2/26/99 Assistance provided by: Printed Name None S pe of Assistance Date PSC review by: - - s -
/h W Date: N i
j 7 l
AmunsAs NUCLEAR ONE Page 1 FORM TITLE: FORM N3. REV. 10CFR50.59 DETERMINATION 1000.131 A 3 PC-1,2 PAGE REV.O This Document contains 4 Pages. Document No. DCP 9739500201 Rev> Change No. 0 1 Title ANO-2 NAOH REPLACEMENT WITH TSP Brief description of proposed change: This modification replaces the existing NaOH addition system with three TSo-C (Trisodium Phosphate) baskets. A previous Technical Specifications submittal requested that the NaOH syst 3 n be eliminated and replaced with a Trisodium Phosphate (TSP) system. That submittal took care of the chang from NaOH to TSP. This 50.59 will discuss the removal and decommisioning of the current NaOH Additior. System. The existing NaOH addition system is an active system while the TSP-C baskets will be a passive system. The post-accident buffering of the building spray fluid has been shown to be acceptable and consistent with the current licensing basis. Portions of the existing NaOH system will be abandoned in place, where some of the system will be removed and capped off from the remainder of the system. Willthe proposed Activity:
- 1. Require a change to the Operating License including:
i Technical Specifications (excluding the bases)? YesO No@ { Operating License?
- YesO No@
Confirmatory Orders? YesO No@ 2. Result in information in the following SAR documents (including drawings and text) being (a) no longer true or accurate, or (b) violate a requirement stated in the document: SAR (multi volume set for each unit)? Yes@ NoO Core Operating Limits Report? YesO No@ Fire Hazards Analysis? YesO No@ Bases of the Technical Specifications? YesO No@ Technical Requirements Manual? YesO No@ NRC Safety Evaluation Reports? Yes@ NoO
- 3. Involve a test or experiment not described in the SAR?
(See Attachment 2 for guidance) YesO No@
- 4. Result in a potential impact to the environment? (Complete Environmental impact Determination of this form.)
. YesO NoS
- 5. Result in the need for a Radiological Safety Evaluation per section 6.1.57 YesO NoS
- 6. Result in any potential impact to the equipment or facilities utilized for Ventilated Storage Cask activities per Section 6.1.67 YesO NoS
- 7. Involve a change under 10CFR50.54 for the following SAR documents per Section 6.1.77 QAMO?
YesO No@ E-Plan? YesO No@
Amuussas mucWWut ONE FORM TITLE: Page 2 FORM N3. REV. 10CFR50.89 DETERMINATION 1000.131A 3 PC-1, 2 2M Document No. DCP g73950D201 Rev> Change No. O ! O Basis for Determination (Questions 1,2, & 3): The Licensing Basis Documents were researched for items which could be applicable to this DCP. Each identified section of the LBD was reviewed to determine the potential impact created by the implement this DCP. The indexes searched were: 2.,,COLR,2_LFO,2_NSE,2_SER,2_SAR,2_TS, EPLAN, and FH1 . The responses to questions 1 & 2 were formed using the results of the LBD search as a basis. The Envimnmental Impact Checklist (Form 1000,131 A) was used 10 to evaluate the potential impact on the environment and it was determined that no impact existed. This DCP does not impact the Ventil Cask, QAMO or E-Plan. A Radiological Safety Evaluation will not be required for this design modificatio Question 1: The ANO-2 Operating License (including Technical Specifications and Confirmatory reviewed for reference to the NaOH addition system and the impact of the TSP-C basket installation. A previous Technical Specification change (as referenced in submittal 2CAN960403) was performed to re the existing requirements for the NaOH system and replace it with requirement relating to the TSP bask This change was the only one previously identified that impacted the operating license. No oth{ Operating License were found that results from this modification. 1 Question 2: The LBD were reviewed for any potential changes required by the implementation of this The following sections were reviewed and determined to be impacted by this modification. SAR Section Section Title 3.6.4.1.g.2 to Concentrators- Safety Eval. Protection Against Dynamic Effects Associated with Postulate 3.g.2.4 Valves ASME Code Class 2 & 3 Components - Analytical & Empirical Methods for Design of Pu 3.11 Environmental Design of Mech. & Elec. Equipment 6.2.2.1 Containment Heat Removat Sys. - Design Bases 6.2.2.1.1 - *
-Containment Spray Sys.
6.2.2.2.1 *
* - Sys. Design - Containment Spray Sys.
6.2.2.3.1 - 6.2.2.4.1 *
- * - Design Evaluation - Containment Spray Sys.
6.2.2.5.1 *
* - Testing & Inspection - Containment Spray Sys.
6.2.3.1 * - Instrument Requirements - Containment Spray Sys.
* - Cont. Air Purification & Cleanup Sys. - Design Bases 6.2.3.2.1 -
- System Design - Sys. Description 6.2.3.2.2 *
- Sys. Operations 6.2.3.2.2.1 *
- Injection Mode l 6.2.3.2.2.2 *
* - Recirc. Mode 6.2.3.2.2.3 -
- Sys. Response 6.2.3.3.1 -
- 6.2.3.3.1.1. *
- * -Design Evaluation - Cont. Spray Sys. Performance Eval.
- Injection Mode 6.2.3.3.1.2 -
- 6.2.3.3.1.3 *
- Recirc. Mode
* - Sodium Hydroxide Addition 6.2.3.3.2.1 -
- 6.2.3.3.2.2 *
- Eval.* Of Analytical Assumptions-lodine Retention by Spray Sys.
6.2.3.3.2.1 *
- Mass Transfer
- Drop Size 6.2.3.3.2.1 - *
-Drop Residence Time 6.2.3.3.2.1 -
- 6.2.3.4 * -Condensation & Coalescence
- Test & Inspection 6.2.5.3.1.4 -
- 6.3.2.14 - Post Accident H2 Concem in Cont.
' Emergency Core Cooling - Sys. Design - Net Positive Suction Head 15.1.0.5.2 Radiological Parameters - Dose Model Assumption 15.1.13.2 15.1.13.3 Major
- Rupture of Pipes Containing RX Coolant - Analysis of Events and Consequences
- Results
F ARKANSAS NUCLEAR ONE FORM TITLE: Page 3 FORM No. REV. 10CFR60.69 DETERMINATION 1000.131A 3 PC-1,2 15.1.13.4.1 - *
-Recirculation Leakage 15.1.13.4.3 - -
- Major Secondary Sys. Pipe Breaks w/wo a Concurrent Loss of AC Power-
' Radiological Consequences of an ECCS Pump Seal Failure SAR Table 3.2-2 Seismic Categories of Sys., Components and Structures SAR Table 3.2-3 Equipment Code Group Classification SAR Table 3.2 6 Quality Assurance Summary Level Q-list SAR Table 3.5-8 Safety Related Components Located Outdoors SAR Table 6.218A NaOH Additive Pump Design Data SAR Table 6.219 Sodium Hydroxide tank Data SAR Table 6.2-23 Cont. Spray Sys. Single Failure Analysis SAR Table 6.2-28 Quantities of Zinc, Copper & Aluminum in Cont.
SAR Table 6.2-29 SAR Table 6.3-4 Metal Corrosion Rates Assumed for Post-Accident H2 Generation Process instrumentation Available During Post LOCA Conditions SAR Table 7.3 5 Safety hlated Sys. Instr. Ranges, Setpoints & Margins to Actuation SAR Table 7.5-2 ESFS f; doring SAR Table 7.5-3 R.G.1.0 Post Accident Monitoring Variables SAR Table 8.3-1 Diesel Load Table SAR Table 14.1-1 initialTest Program Summaries SAR Table 15.1.0-3 Physical Data For isotopes SAR Table 15.1.0-4 Breathing Rates SAR Table 15.1.13-1 Assumptions for Design Basis Loss of Coolant Accident SAR Table 15.1.13-2 Loss of Coolant Accident Doses SAR Table 15.1.13-3 lodine inventory in Cont. Sump at Time of Recire. Start SAR Table 15.1.13-5 Leakage Quantities to Auxiliary Building SAR Figure 1.2-6 Equipment Location Plan Below Grade SAR Figure 1.2-11 Equipment Location Misc. Plans & Sections SAR Figure 3.5-18 Plant Design Drwg A24 Cont. Aux. Bldg Plan @ Elev. 335' 0* to 354'-0* SAR Figure 3.5-19 * *
- A24 & 26 Cont. Aux. Bldg Plan Below Elev. 335'-0*
SAR Figure 3.5-25 - *
- A26 Cont. Aux. Bldg Plan Below Grade SAR Figure 3.6-35 * *
- A25 Cont. Bldg Section A25-A25 SAR Figure 3.6-48 * * * * *
- SAR Figure 3.8-31 Plan below Elev. 357'-0*
Temporary Construction Opening Plan @ Elev 335'-0* SAR Figure 6.2-17 Containment Spray Sys. -2 Sheets SAR Figure 6.3-12 LOCA Sequence of Events Diagram SAR Figure 7.3-7 Functional Description & Logic Diagram Cont. Spray Sys.- 3 Sheets SAR Figure 8.3-12 Single Line Diagram 480 V Mc, tor Control Center 2B61 SAR Figure 8.3-13 Single Line Diagram 480 V Motor Control Center 2B62 SAR Figure 8.3-55 Conduit & Tray Layout Aux. Bldg A24 & A26 Elev. 326'-0* to 335'-0* SAR Figure 9.2-5 Area 24 Cont. Aux. Bldg Plan Elev. 335'-0* to 354'-0* 2NSE 3.0 2NSE 6.0 in Relations to: EQ Qualifications of Safety Related Equipment- pH on Cont. Spray on Motom in Relations to: Section 6.2.3,6.3.2 -lodine & NaOH & NPSH 2NSE 15.0 in Relations to: NaOH 2SER 82.0 in Relations to: pH Reduction 2SER 90.0 in Relations to: NaOH Deliver to the Spray Lines Question 3: This DCP does not involve any tests or experiments which require addition to the SAR as related to the TSP-C baskets. ERIDIrdD201 PAGE REV.O
ARKANSAS NUCLEAR ONE Page 4 FORM TITLE: FORM NO. REV. 10CFR50.59 DETERMINATION 1000.131 A 3 PC-1,2 O Proposed change does not require 10CFR50.59 Evaluation per Attachment 1, item # m (if checked, note appropdate item W, send LDCR to Licensing). Search Scope: List sections reviewed in the Licensing Basis Documents specified in questions 1,2 and 3. If search was performed on LRS, the LRS search index should be entered under "Section" with the search statement (s) used in parentheses. Controlled hard copies of the documents shall be reviewed (LRS is not verified and searches only 1:xt, not figures or drawings). Attach and distribute a completed LDCR per Section 6.1.2 If LBD changes are required. Document Section LRS: oH. iodine remova!! containment sorav. CSS. 2P136A. 2P1368. Trisodium Phosphate. 2CV56571. 2CV56672. sodium hvdieAle. NaOH. chemical addition. 2T10. sumo w/10 sodium MANUAL SECTIONS: 3.11. 6.2. 6.3. section 7 tables & 15.1 FIGURES: 1.2.35. 3.6. 3.8.6.2. 7.3.8.3 & 9.2 Jerry W. Howell 9/3/98 Ce av rs Signature Printed Name Date Rsviewers certification expiration date: 2/24/2e30 Assistance provided by: Printed Name Scope of Assistance Date J.R. Dorman. Jr. (FTI) DCP Preparation 8/31/98 Jacque Lingenfetter Calculations & 50.59 review 9/15/98 Search Scope Review Acceptability (NA, if performed by Technical Reviewer per 1000.006) Danielle Smith 9/15/98 Cepied Revi ignature Printed Name Date ER 8730EOD201 i PAGE REV.O I
1 ARKANSAS NUCLEAR ONE Page G FORM TITLE: FORM ND. REV. 10CFR50.59 DETERMINATION 1000.131 A 3 ENVIRONMENTAL IMPACT DETERMINATION ! (UNIT 1 and UNIT 2)- D:cument No. DCP 973950D201 RevJChange No. O Crmplete the following Determination. If the answer to any item belowis "Yes", an Environmental Evaluation is required. See Section 6.1.4 for additional guidance. Will the Activity being evaluated: 111 NQ O E Disturb land that is beyond that initially disturbed during construction (i.e., new construction of buildings, creation or removal of ponds, or other terrestrial impact)? See Unit 2 SAR Figure 2.5-17. This applies only to areas outside the protected area. O 2 Increase thermal discharges to lake or atmosphere? O 2 Increase concentration of chemicals to cooling lake or atmosphere through discharge canal or tower? O @ Increase quantity of chemicals to cooling lake or atmosphere through discharge canal or tower? O E Modify the design or operation of cooling tower which will change drift characteristics? O E instati any new transmission lines leading offsite? O @ Change the design or operation of the intake or discharge structures? O @ Discharges any chemicals new or different from that previously discharged? O E Potentially cause a spill or unevaluated discharge which may effect neighboring soils, surface water or ground water? O E Involve burying or placement of any solid wastes in the site area which may effect runoff, surface water or ground water? O O involve incineration or disposal of any potentially ilazardous materials on the ANO site? O @ Result in a r.hange to nonradiological effluents or licensed reactor power level? O E Potentially change the type or increase the amount of non-radiological air emissions from the ANO site. i l l PAGE REV.O
ARKANSAS NuCt. EAR ONE Page
- FORM TITLE:
FORM NO. REV. 10CFR50.69 SAFETY EVALUATION 1000.131B 3 PC-2
% This Document contains 1 Page.
Document No. DCP 973950D201/ Rev/ Change No. 10CFR50.59 Eval. No. NN-M - NQ (Assigned by PSC) Title ANO-2 NaOH Replacement with TSP A WRITTEN RESPONSE PROVIDING THE BASIS FOR THE ANSWER TO EACH QUESTION MUS ATTACHED. EACH QUESTION MUST BE ANSWERED SEPARATELY. A SIMPLE STATEMENT OF CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES GUIDANCE FOR RESPONSE If the answer to any question on this fom11s "Yes,"then an unreviewed safety question is involved. if the answer to all questions is *No," then the proposed change does not involve an unreviewed safety question.
- 1. Will the probability of an accident previously evaluated in the SAR be increased?
YesO No @
- 2. Willthe consequences of an accident previously evaluated in the SAR be increased?
YesO No @
- 3. Will the probability of a malfunction of equipment important to safety be increased?
YesO No @ 4. Willthe consequences of a malfunction of equipment important to safety be increased? YesO No @
- 5. Will the possibility of an accident of a different type than any previously evaluated in the SAR be created?
YesO Na @ 6. Will the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR be created? YesO No @ 7. Will the margin of safety as defined in the basis for any technical specification be reduced? YesO No @
- A A .o Jerry W. Howell 9/17/98 ylfi Reviewers Signature Printed Name Date R viewers certification expiration date: 2/24/2000 Assistance provided by:
Printed Name Scope of Assistance Jacque Lingenfelter _ Review of pH & lodine retention Date 9/15/98 PSC review by: M3 -- Date:
\\ \.\okc4 PAGE- REV.O
AaKaNsAs NUCLEAR ofE - Page 1 FORM TITLE: FORM NO. REV. 10CFR80.59 REVIEW CONTINUATION PATE 1000.131C 3 Document No. DCP 973950D201 Rev1 Change No. _ 0 PAGE REV.0 10CFR50 59 Review Continuation Paoe Basis for answers to questions 1-7 on 1000.1318. 1. Will the probability of an accident previously evaluated in the SAR be increased? This DCP removes some portions and decommissions other portions of the existing NaOH addition system f the containment spray system. The containment spray function is important for containment heat removal, pressure mitigation and iodine retention. This change does not affect the probability of occurrence of the accident initiators, which result in the need for the containment spray function. Since the removal and decommissi the existing NaOH system eliminates an active system and replaces it with a new passive system, then the removal NaOH of the NAOH system will not affect the probability of any previously analyzed accident pertaining to the System. Therefore, this change does aglincrease the probability of an accident previously evaluated. 2. Will the consequences of an accident previously evaluated in the SAR be increased? The proposed change removes and decommissions portions of the existing NaOH Addition System. The proposed changes do not affect the heat removal / pressure mitigation functions of the containment spray syste since the spray flow rate and droplet size are unchanged. Since the iodine retention function of containment spray is undiminished, the proposed change also will not adversely affect the radiological doses for the D Basis Accident (DBA) Loss-of-Coolant Accident (LOCA) at the Exclusion Area Boundary, Low Population Z Centrol Room, or Emergency Response Facility. The change does not adversely affect the calculated peak tImperature for the DBA LOCA. Since the corrosion rate is lower due to the spray fluid being less caustic, there are no adverse affects on Environmental Qualification (EQ) of components located inside containment. Therefore, this change does Bg! involve an increase in the consequences of any accident previously eva 3. Will the probability of a malfunction of equipment important to safety be increased? The proposed change removes and decommissions portions of the existing NaOH Addition System; howeve rsmoval of this equipment will not cause the Containment Spray System to become inoperable or cause it to malfunction. The seismic qualifications of the abandoned equipment left in place wM remain the sane. Also, since the corrosive nature of the NaOH System is higher than the TSP System, the probability of a malfunctio cquipment important to safety is not increased. All electrical equipment that is important to safety that is locatej , in the containment building is also less impacted by the lower corrosiveness of the TSP system. No other equipment impostant to safety is adversely impacted by this change. ! ThIrefore, this change does Dglincrease the probability of a malfunction of equipment important to safet 4. Will the consequences of a malfunction of equipment important to safety be increased? The removal and decommissioning of the NaOH Addition System and the installation of the TSP System increase onsite dose consequences that restrict access to vital areas and thereby will not obstruct actions to mitigate the consequences of a malfunction of equipment important to safety. The iodine retention function of the Cantainment Spray System is undiminished. The proposed change will not adversely affect the radiolog i doses under a DBA due to a malfunction of any equipment. This change does not impact any important equipment that is required to prevent or mitigate plant conditions that could result in offsite exposuras to beco grater than limits established in 10CFR100 guidelines. Th;refore, this change does Bgj increase the consequences of a malfunction of equipment important to s
ARKANSAS NUCLEAR ONE Page 2 FORM TITLE: FORM NO. REV. 10CFR50.89 REVIEW CONTINUATION PAGE 1000.131C 3
- 5. Will the possibility of an accident of a different type than any previously evaluated in the SAR be created?
Potential malfunctions relating to the proposed modification have been evaluated for their effect on plant safety and have been found to be non-existent. Additionally, the transient pH behavior due to the removal and decommissioning of the NaOH Addition System and replacement with TSP does not adversely affect the EQ of components located inside containment. Since the replacement system consist of passive components compared to the existing NaOH Addition System use of active components, possibility of creating accidents different than previously evaluated are eliminated. No new accident initiators are developed due to this change. Therefore, this change does ng! create the possibility of a new or different kind of accident from any previously cvaluated. 6. Will the possibility of a malfunction of equipment important to safety of a different type than previously evaluated in the SAR be created? The effect of the TSP-C basket on the systems iri1portant to safety has been evaluated and presents no additional malfunction possibilities. The TSP-C is simply a passive replacement for the NaOH addition in the reactor building spray system and does not add any new equipment, which would create a new malfunction scenario.
- Rimoval and decommissioning of the existing NaOH Addition System will not cause any type of malfunction under any normal or abnormal operations. The existing NaOH Addition System will be physically disconnected from the Containment Spray System. This modification is supported by Standard Review Plan 6.5.2 Therefore, this change does agt. create a possibility of a malfunction of equipment important to safety of a different type than previously evaluated.
- 7. Will the margin of safety as defined in the basis for any technical specification be reduced?
The proposed change does not adversely affect the ability of the Containment Spray System to perform the functions of containment heat removal, pressure mitigation, and fission product (iodine) retention. The proposed ch nge does not adversely affect any equipment credited in the safety analysis. Also, the proposed change does q not increase the peak-clad temperature or the offsite doses due to the DBA LOCA. A Technical Specification l' change has been provided to replace the existing NaOH Addition System requirements with the TSP System requirements, therefore, the margin of safety as glefined in the Basis for Tech. Specs. will not be reduced. Therefore, this change does ngj involve a significant reduction in the margin of safety as defined in the basis of Tsch. Specs. 4 l
. ER 9739500201 PAGE I REV.0
FORM TITLE: -ARKANSAS NUCLEAR ONE Prot
- FORM NO. REV.
1DCFR50.59 DETERMINATION 1000.131 A 3 PC 1 PAGE b REV.O This Document contains 3 Pages. Document No. ER 975015D201 RevdChange No. O Title ANO-2 MSIS Setpoint Reduction _.m Brief description of proposed change: Amendment (TBD) to the Technical Specifications by the NRC.Th n Will the proposed Activity: 1. Require a change to the Operating License including: Technical Specifications (excluding the bases)? Operating License? YesO No@ Confirmatory Orders? YesO No@
- 2. YesO No@
(a) no longer true or accurate, or (b) violate a requirement s SAR (multi-volume set for each unit)? Yes@ NoO Core Operating Limits Report? YesO No@ Fire Hazards Analysis? YesO NoS Bases of the Technical Specifications? YesO No@ Technical Requirements Manual? YesO No@ NRC Safety Evaluation Reports?
- 3. YesO No@
Involve a test or experiment not described in the SAR? (See Attachment 2 for guidance) YesO NoS 4. Result in a potentialimpact to the environment? (Complete Environmental Impact Determination of this form.)
- 5. YesO No@
Result in the need for a Radiological Safety Evaluation per section 6.1.57
- 6. YesO No@
Result in any potential impact to the equipment or facilities utilized for Ventilated Storage Cask activities per Section 6.1.67
- 7. YesO No@
Involve a change under 10CFR50.54 for the following SAR documents per Section 6.1.77 QAMO? E-Plan? YesO No@ YesO No@
ARKANSAS NUCLEAR ONE ! FORM TITLE: Pao 1 : FORM NO. REV. 10CFR50.59 DETERMINATION 1000.131 A 3 PC-1 Document No. ER 975015D201 RevjChange No. O Basis for Determination (Questions 1,2, & 3): See Attachment 1. REV.O I l i O Proposed change does not require 10CFR50.59 Evaluation per Attachment 1, item appropriate item #, send LDCR to Licensing). Search Scope: List sections reviewed in the Licensing Basis Documents specified in questions 1,2 an was done on LRS, "all" may be entered under "Section" with the keyword (s) used in parenthe Attach and distribute a completed LDCR per Section 6.1.2 if L Document Section LRS: ALL (* Main Steam isolation" "MSIS*
- steam oenerator pressure *
" reactor trio setooint".
- steam oenerator setooint' '678" '751")
l MANUAL SECTIONS: 5.5.5.6.2.1.1.2.6.87.2.1.1.1.7.7.2.1.1.1.8.7.3.7.3.1.1.11.4.1 FIGURES;
/ 7.2 series. 7.3-2. 7.3-9: Tables 7.2-2. #7.2-4. 7.2-5. #7.3 5. 7.3-6$ 5.1.0-15.1.10-4.15.1.14-111hru 19 l/ @
Certified Reviewers Signature JB Robinson Printed Name 3f/ /h Date Reviewers certification expiration date: _5/28/98 Assistance provided by: Printed Name Scope of Assistance Tim Rush SAR Chapter 6 Review Date Bryan Daiber 3/12/98 SAR Chapter 6 & 15 Review Jacque Lingenfelter Chapter 7 nout 2l12/98 i 3/12/98 Search, Scope Review Acceptability (NA, if performed by Technical Reviewer per 1000.00i w ~
% Stu.0374DLLk '4 l Certified Reviewers Signature 3/sLlG8 ;
Printed Name Date ! l i E
ARKANSAS NUCLEAR ONE FORM TITLE: P'01 3 ! FORM NO. REV. 10CFR50.59 DETERMINATION 1000.131A 3 ENVIRONMENTAL IMPACT DETERMINATION (UNIT 1 and UNIT 2) PAGE REV.O Document No. ER 975015D201 RevlChange No O Complete the following Determination. If the answer to any item below is "Yes', an Environmental Ev required. See Section 6.1.4 for additional guidance. Will the Activity being evaluated: Yes No O @ Disturb land that is beyond that initially disturbed during construction (i.e., new construction o buildings, creation or removal of ponds, or other terrestrial impact)? See Unit 2 SAR cig 2.517. This applies only to areas outside the protected area. O E increase thermal discharges to take or atmosphere? O E increase tower? concentration of chemica!s to cooling lake or atmosphere through discharge O @ ~ increase tower? quantity of chemicals to cooling lake or atmosphere through discharge canal or O @ Modify the design or operation of cooling tower which will change drift characteristics? O E Install any new transmission lines leading offsite? O E Change the desig., or operation of the intake or discharge structures? O O Discharges any chemicals new or different from that previously discharged? O O Potentially water or ground cause water?a spill or unevaluated discharge which may effect neighboring soils, s O E Involve burying or placement of any solid svastes in the site area which may effect ru surface water or ground water? O E involve incineration or disposal of any potentially hazardous materials on the ANO site? O @ Result in a change to nonradiological effluents orlicensed reactor power level? O E Potentially change the type or increase the amount of non-radiological air emissions from t ANO site.
))WAcilmeur ro Foxa. /obo. /3lA Attachment I to 50.59 Resiew PAGE h REV.O Document ER975015D201 '
Rev.0 l Answers to Questions 1. This maM~iaa chan 4tj r documented in Amendmen,the low pressure trip serpoint (MSIS setpoint) from 751 to 712 i the Operating License were affected by this modification.to the ANO-2 Tech ' I 2. Several sections of the SAR discussed the setpoint, the analytical l pressure values that were changed in conjunction with the setpoint. These secti revised and are included in this modification as part of an Also, LDCR. the effects of the { pressure changes on the steam generator level instrumentation had to be addre The net change on the level instrumentation was negligible, but the SAR
- changed to show the new numbers.
Chapters 6 and 15 of the ANO-2 SAR have been changed as a result of th NRC as established by procedure 1062.002. Specification. Thes to Chapters 6 and 15 will be addressed in response to the LIR, and . None of the other SAR documents contained the level of detail require change. 3. This change does not involve any new tests or experiments. 4. As documented on the Environmental Checklist, there is no impact to change. 5. There is no work inside of a Radiologically Controlled Area (RCA) by this m Furthermore, there are no potentially uncontrolled release paths created b Therefore, an RSE is not warranted by this modification. 6. This modification das not alter or impact any of the VSC facilities or equipm transport or storage of the VSC's. 7-
, required This modification does not impact the QAMO or the E-Plan. They do not co!
to be impacted by this modification. l l I
ARKANSAS NUCLEAR ONE FORM TITLE: Paco 1 FORM NO. REV. 10CFR50.59 SAFETY EVALUATION 1000.131B 3 g This Document contains 1 Page. Document No. ER 975015D201 Rev) Change No. 0 10CFR50.59 Eval. No. Fro-@lns Title _ ANO-2 MSIS Setpoint Reduction (Assigned by PSC) A WRITTEN RESPONSE PROVIDING THE BASIS FOR THE ANSWER ATTACHED. EACH QUESTION MUST BE ANSWERED SEPARATELY. A S CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES GUIDAN If the answer to any question on this form is "Yes," then an unreviewed safety. question is involv If the answer to all questions is *No," then the proposed change does not involve an unreviewed safety quest 1. Will the probability of an accident previously evaluated in the SAR be increased? YesO No @ 2. Will the consequences of an accident previously evaluated in the SAR be increased? YesO No @ 3. Will the probability of a malfunction of equipment important to safety be increased? YesO No @ 4. Will the consequences of a malfunction of equipment important to safety be increased? YesO No @ 5. Will the possibility of an accident of a different type than any previously evaluated in the SAR be created? YesO No @ 6. Will the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR be created? YesO No @ 7. Will the margin of safety as defined in the bases for any technical specification be reduced? Yes O No @ J Jo Certified RevTewer's Signature Joel Brent Robinson Printed Name '
/2 7 Date Reviewer's certification expiration date: 5/28/98 Assistance provided by: ,
Printed Name Scope of Assistance Date PSC review by:A 3 Date: 1\\S
$URC/fMEHT To RRM /soo mB Backcround Answers to 10CFR50.59 Evaluation questions. P [ O The scope of this evaluation is not all encompassing of the ANO 2 MSIS Setpoint Reduction modification.
The scope of this evaluadon only covers the issue of the effects of rapid depressurization of the steam j generators on level instrttmentation. In summarizing these effects. a pornon of the reference leg will } flash to steam upon rapid'depressurization of the steam generators. The calculated values for the conditions are that the worst case scenario would be 0.36 inches, and for the new parameters. 0.48 inches. When this is added to the losses due to configuradon of the condensing pot, the total losses for the existing parameters is 2.36". and 2.48" for the new parameters. Therefore the net change in reference leg flashing is 0.12" w.c. There are two cases that need to be addressed:
- 1) the effects of reference leg flashing on pressure measurement and; 2) the effects of reference leg flashing on level measurement.
In the first case (pressure measurement), the steam generator low pressure inp is credited for Steam Line Breaks and Feedwater Line Breaks with or without loss of AC. and steam B) pass System Malfunction. When the above values are convened to percent pressure. as per calculadon 93-EQ-2001-01. Rev. 2. the effects are negligible. The instrument still responds within the acceptable limits. Therefore, the exis analyses are unaffected by tids change. In the second case (level measurement), the steam generator low level trip is credited for Steam Line Breaks and Feedwater Line Breaks with or without loss of AC, and the andcipated operational occurren (AOO) of a Loss of Feedwater event. The high steam generator water level trip is not credited for SAR Chapter 15 event. When the above values are convened to percent level. as per calculadon 2001-01. Rev. 2. the effects are bounded by the existing calculations addressing process / environmental conditions and static pressure span bias. The change in MSIS actuation / low steam generator pressure inp setpoint has been reviewed an by the NRC. As a result of this. the remaining SAR changes are exempted from the evaluation pro per ! tem C of Attachment I to Operating Procedure 1000.131. While the effects of the flashing of the reference leg was not specifically addressed in the TS change and its analysis. it is bounded by the conditions that were analyzed. 1. Will the probability of an accident previously evaluated in the SAR be increased? The probability of an accident previously evaluated in the SAR will not be increased as a result of the change in the reference leg flashing effect. The net change in process measurement crror (PME). as a result of the reference leg flashing effect. was determined to be neglig;ible for the case of pressure, and bounded by existing conditions in the case oflevel, per calculation 93-EQ-2001-01, Rev. 2. For the accidents previously evaluated in the SAR. major secondary system - pipe breaks with or without a concurrent loss of AC power, tids scenario will not increase their probability. The initiating factors are in no way tied to instrument error that might occur in the case of rapid depressurization of the steam generators. 2. Will the consequences of an accident previously evaluated in the SAR be increased? The consequences of an accident previously evaluated in the SAR will not be increased as a result of the change in the reference leg flashing effect. The net change in process measurement error (PME), as a result of the reference leg flashing effect was determined to be negligible for the case of pressure, and bounded by existing conditions in the case oflevel, per calculation 93-EQ. 2001-01. Rev. 2.. Therefore, if the effect on the PME for SG pressure is negligibic and the effect ER 975015D203
1 ff/ M HRIG N 7 TO F9 &rt M o e B/B PAGE N REV.O on SG level is bounded by existing conditions. the consequences of an accident prev evaluated in the SAR is not increased. { 3. Will the probability of a malfunction of equipment important to safety be increased? 1 The probability of a malfunction of equipment important to safety will not be increased As determined above. the reference leg flashing effect is negligible to the case of pressure measurement and bounded by existing conditions in the case oflevel measurement. A loss of 0.12 inches of water level in the reference leg does not impact the ability of the pressure 1 instruments to perfonn their safety function. Therefore, the probability of a malfunction of ! eqmpment imponant to safety is not increased. l i 4. Will the consequences of a malfunction of equipment imponant to safety be increased? The consequences of a malfunedon of equipment important to safety will not be mereas determined above, the reference leg flashing effect is negligible to the case of pressure i measurement and bounded by existing conditions in the case oflevel measurement. Therefore, the consequences of a malfunction of equipment important to safety is not increased. 5. Will the possibility of an accident of a different type than any previously evaluated in the SAR be created? The possibility of an accident of a different type than any previously evaluated in the S AR not be created. As determined above, the reference leg flasinng effect is negligible to the c pressure measurement and bounded by existing conditions in the case oflevel measurement. Rapid depressurization of the steam generators is not a new concept, and the effects a by existing scenarios. Therefore, the possibility of an accident of a different type than previously evaluated in the SAR will not be created. 6. Will the possibility of a malfunction of equipment important to safety of a different ty than any previously evaluated in the SAR be created? The possibility of a malfunction of equipment important to safety of a di1Terent typ previously evaluated in the SAR will not be created. As determined above, the refere flashing effect is negligible to the case of pressure measurement and bounded by e conditions in the case oflevel measurement. Reference leg flashing has already been consi in the ANO design analysis, and determined to be bounded by other scenarios. Th change does not change any of the previous conclusions reached by this change. There possibility of a malfunction of equipment important to safety of a different type than anl previously evaluated in the SAR will not be created. 7. Will the margin of safety as defined in the bases for any technical specification be reduc The margin of safety as defined in the bases for any technical specification will not be red The TS Bases focuses on the 2-out-of-4 logic of the protective and engineered safety f systems, the diversity of the instruments feeding the systems, .the operability of the instruments feeding these systems, and the ability to perform on-line maintenance on the systems. N these issues are impacted by this proposed change. Also, the TS Bases do not contam actua numbers when discussing the steam generator level or pressure tnps. Therefore, the m safety as defined in the bases for any technical specification will not be reduced. ER 9750150201
ARKANSAS NUCLEAR ONE PsDe 3 F&RM TITLE: FORM No. REV. 10CFR50.59 DETERMINATION 1000.131A 3 l ENVIRONMENTAL IMPACT DETERMINATION (UNIT 1 and UNIT 2) PM h REV.O Document No. ER 975015D201 RevlChange No. O Complete the following Determination. if the answer to any item below is "Yes", an Environmental Evalua required. See Section 6.1.4 for additional guidance. , Will the Activity being evaluated: Y.f.ft Hg O- E Disturb land that is beyond that initially disturbed during construction (i.e., new construction buildings, creation or removal nf ponds, or other terrestrialimpact)? See Unit 2 SAR F 2.5-17. This applies only to areas outside the protected area. O E increase thermal discharges to take or atmosphere? O E increase concentration of chemicals to cooling lake or atmosphere through discharge tower? O E increase quantity of chemicals to cooling lake or atmosphere through discharge canal or tower? O E IAodify the design or operation of cooling tower which will change drift characteristics? l O E Install any new transmission lines leading offsite? O E Change the design or operation of the intake or discharge structures? O E Discharges any chemicals new or different from that previously discharged? O E Potentially cause a spill or unevaluated discharge which may effect neighboring soils, su water or ground water? O E involve burying or placement of any solid wastes in the site area which may effect runof surface water or ground water? O E Involve incineration or disposal of any potentially hazardous materials on the ANO site? O E Result in a change to nonradiological effluents or licensed reactor power level? O E Potentially change the type or increase the amount of non-radiological air emissions from the ANO site.
ARKANIAS NUCLEAR ONE FORM TITLE: Paca FORM NO. REV. 10CFR50.59 SAFETY EVALUATION 1000.131B 3 g This Document contains 1 Page. Document No. _ ER 975015D201 Rev> Change No. 0 10CFRSO.59 Eval. No. 'lF-OCO Title ANO-2 MSIS Setpoint Reduction (Assigned by PSC) ATTACHED. EACH QUESTION MUST BE AN CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES GU If the answer to any question on this form is "Yes," then an unreviewed safety que to all questions is "No," then the proposed change does not involve an unreviewed safety . 1. Will the probability of an accident previously evaluated in the SAR be increased? Yes O No @ 2. Will the consequences of an accident previously evaluated in the SAR be increased? Yes O No @ 3. Willthe probability of a malfunction of equipment important to safety be increased? YesO No @ 4. Will the consequences of a malfunction of equipment important to safety be increased? YesO No @ 5. Will the possibility of an accident of a different type than any previously evaluated in the SAR be created? YesO No @ 6. Will the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR be created? YesO No @ 7. Will the margin of safety as defined in tha bases for any technical specification be reduced? Yes O No @
~ --L Certified Reviewer's Signature Joel Brent Robinson 9ft7 Printed Name '
I[ Date Reviewer's certification expiration date: 5/28/98 Assistance provided by: Printed Name Scope of Assistance Date PSC review byA 3 Date: _ W
p u nw inic. p s sv . -ei.u' p w. wy Backeround Answers to 10CFR50.59 Evrluatien nuestions. PAGE
!! REV.O The scope of this evaluation is not all encompassing of the ANO-2 MSIS Setpoint Reduction modification.
The scope of this evaluation only covers the issue of the effects of rapid depressurization of the steam generators on level instrumentation. In summanzing these effects, a portion of the reference leg will
. flash to steam upon rapid *depressunzation of the steam generators. The calculated values fo conditions are that the worst case scenario would be 0.36 inches, and for the new parameters. 0.48 inches. When this is added to the losses due to configuration of the condensing pot, the total losses for the existing parameters is 2.36", and 2.48" for the new parameters. Therefore, the net change in reference leg flashing is 0.12" w.c. There are two cases that need to be addressed: 1) the effects of reference leg flashing on pressure measurement and; 2) the effects of r-Jerence leg flashing on level measurement.
In the first case (pressure measurement), the steam generator low pressure trip is credited for Steam L Breaks and Feedwater Line Breaks with or without loss of AC, and steam Bypass System Malfunction. When the above values are convened to percent pressure, as per calculation 93-EQ-2001-01, Rev. 2, t effects are negligible. He instrument still responds within the acceptable limits. Therefore, the e analyses are unaffected by this change. In the second case (level measurement), the steam generator low level trip is credited for Steam Line Breaks and Feedwater Line Breaks with or without loss of AC, and the anticipated operational occurre (AOO) of a Loss ofFeedwater event. The high steam generator water level trip is not credited for SAR Chapter 15 event. When the above values are converted to percent level, as per calculat 2001-01, Rev. 2, the effects are bounded by the existing calculations addressing process / environmental conditions and static pressure span bias. The change in MSIS actuation / low steam generator pressure trip setpoint has been reviewed by the NRC. As a result of this, the remaining SAR changes are exempted from the evaluation pro per item C of Attachment I to Operating Procedure 1000,131. While the effects of the flaslAg of the reference leg was not specifically addressed in the TS change and its analysis, it is bounded by conditions that were analyzed. A 1. Will the probability of an accident previously evaluated in the SAR be increased? The probability of an accident previously evaluated in the SAR will not be increased as a result o the change in the reference leg flashing effect. The net change in process measurement error (PME), as a result of the referenc~ leg flashing effect, was determined to be negligible for the case of pressure, and hanaM by existing conditions in the case oflevel, per calculation 93-EQ 2001-01, Rev. 2. For the accidents previously evaluated in the SAR, major secondary system pipe breaks with or without a concurrent loss of AC power, this scenario will not increase their probability. The initiating factors are in no way tied to instrument error that might occur in the case of rapid depressurization of the steam generators. 2. Will the consequences of an accident previously evaluated in the SAR be increased? The consequences of an accident previously evaluated in the SAR will not be increased as a re of the change in the reference leg flashing effect. The net change in process measurement error (PME), as a result of the reference leg flashing effect, was determined to be negligible for the case ofpressure, and bounded by existing conditions in the case oflevel, per calculation 93 EQ 2001-01, Rev. 2., herefore, if the effect on the PME for SG pressure is negligible and the effect ER.9750150201
J/MCHJ1GN7 70 /M /000 BM PAGE- N REV.O on SG level is bounded by existing conditions. the consequences of an accident pr evaluated in the SAR is not increased. 3. Will the probability of a maalfunction of equipment important to safety be increased? The probability of a ==1A=" ion of equipment important to safety will not be increased A detenmned above, the reference leg flaslung efect is negligible to the case of pressur measurement and bounded by existing conditions in the case oflevel measurement. A loss of 0.12 inches of water level in the reference leg does not impact the ability of the press mstruments to perform their safety function. Therefore, the probability of a =ain=dion of { equipment imponant to safety is not increased. 4. Will the consequences of a malfunction of equipment important to safety be increased The consequences of a ==1A=" ion of equipment imponant to safety will not be incre h mined above, the reference leg flashing efect is negligible to the case of pressure measurement and bounded by existing conditions in the case oflevel measurement. Therefore. the consequences of a ==1 A=dion of equipment important to safety is not increased 5. Will the possibility of an accident of a different type than any previously evaluated in the SAR be created? The possibility of an accident of a diferent type than any previously evaluated la the not be created. As detennined above, tne reference leg flashing effect is negligible to t pressure measurement and bounded by existing conditions in the case oflevel measurement. Rapid depressunzation of the steam generators is not a new concept, and the efects a by existing scenarios. Therefore, the possibility of an accident of a diferent type th previously evaluated in the SAR will not be created-6. Will the possibility of a malfunction of equipment important to safety of a differen than any previously evaluated in the SAR be created? I The possibility of a malfunction of equipment important to safety of a diNerent ty previously evaluated in the SAR will not be created As determined above the refer flashing efect is negligible to the case of pressure measurement and bounded by e conditions in the case oflevel measurement. Reference leg flashing has already been co in the ANO design analysis, and detenmned to be bounded by other scenarios. T change does not change any of the previous conclusions reached by this change. Th; possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR will not be created. . 7. Will the margin of safety as defined in the bases for any technical specification be reduced The margin of safety as defined in the bases for any tecimical specification will not be redu The TS Bases focuses on the 2 out of 4 logic of the protective and engineered safety fe systerrs, the diversity of the instruments feeding the systems, the operability of the instruments feeding these systems, and the ability to perform on-line maintenance on the systems. No these issues are impacted by this proposed change. Also, the TS Bases do not contain ac numbers when discussing the steam generator level or pressure trips. Therefore, the
- safety as defined in the bases for any technical specification will not be reduced.
ER-975015D201
p-l' e w ARKANSAS NUCLEAR ONE FORM TITLE: FORM NO. QEV. 10CPRee.ee DETERM8 NATION 1900.131A 3 PC-1 Page 1 of_1Q Document No. Amendment 189 & 190 RevlChange No. J l Title SAR updates for RCS Flow and MSIS Tech Spec Chances Brief description of proposed change: This chance incorporates the RCS Flow and MSIS Tech Spec chance results plus covers various miscellaneous effects of up to 30% tube pluacino. , I l Will the proposed Activity: I i
- 1. Require a change to the Operating License including:
Technical Specifications (excluding the bases)? YesO No@ Operating License? YesO No@ Confirmatory Orders? YesO No@
- 2. Result in information in the following SAR documents (including drawings and text) being (a) no longer true or accurate, or (b) violate a requirement stated in the document:
l l' SAR (multi-volume set for each unit)? Yes@ NoO l Core Operating Limits Report YesO NoS
- Fire Hazards Analysis? YesO No@
Bases of the Technical Specifications? YesO No@ Technical Requirements Manual? YesO No@ NRC Safety Evaluation Reports? YesO NoS
- 3. Involve a test or experiment not described in the SAR? YesO No@
(See Attachment 2 for guidance)
- 4. Result in a potential impact to the environment? (Complete the Environmental impact Determination of this form.) YesO No@
- 5. Result in the need for a Radiological Safety Evaluation per section 6.1.57 YesO No@
- 6. Result in any potential impact to the equipment or facilities utilized for Ventilated Storage Cask activities per Section 6.1.67 YesO NoS
- 7. Involve a change under 10CFR50.54 for the following SAR documents per Section 6.1.7:
QAMO? YesO No@ E-Plan? YesO No@ Basis for Determination (Questions 1. 2 & 31: See Attached O Proposed change does not require 10 CFR 50.59 Evaluation per Attachment 1, Item # ,(if checked, , note appropriate item #, send LDCR to Licensing). l i I
muuwsAs NUCLEAR ONE FORM TITLE: FORM NO. REV. 10CFRSO.89 DETERMINATION 1000.131A 3 PC 1 Page 2 of.10 Document No. Amendment 189 & 190 RevlChange No. 9. search scope: List sections reviewed in the Licensing Basis Doc,uments specified in Question 1,2 and 3. If a keyword search was done on LRS,"all" may be entered under "Section" with the keyword (s) used in parentheses. Controlled hard copies of the documents shall be reviewed (LRS is not verified and searches only text, not fi0ures or drawings). Attach and distribute a completed LDCR per Section 6.1.*2 If LSD changes are required. Document Section LRS: All ('RCS Flow", 120.4, "322,000", 60.2, "161,000", "80,500",
" Design Flow", RCP, " reactor coolant flow", " core flow", "RCS volume", steam w/5 volume, reactor w/5 volume, reactor w/5 mass,
" coolant volume", 678, 751, MSIS, " main steam isolation signal",
venturl*, " steam generator" w/10 nozzle, CEA w/10 withdrawal, CPC w/10 filters, " loss of load", excess w/10 load, dynamic w/10 constant *, CENTS, deadband*, CEA w/10 deviation, " dead band", resistance w/10 RCS, resistance w/10 hydraulic, " pressure drop", resistance w/10 steam, resistance w/10 reactor, "line loss", " head loss", steam w/10 differential, sgtr, " steam generator tube ruptute", and activity w/10 primary. MANUAL SECTIONS: SAR Sections 4.4 (all subsections), 5.2.4.3.1, 5.2.4.3.2, 5.3 (all subsections), 5.5 (all subsections), 6.2.1.1.2.6, 6.2.1.1.4, 6.2.3.3.2.1, t 7.2.1.1.1.7, 7.2.1.1.2.4, 15.1.4 (all subsections), 15.1.5 (all ' subsections), 15.1.18 (all subsections), and 15.1.10 (all subsections) and Tables 3.94, 4.24A, 4.4-1, 4.4-2, 4.4-3, 4.4-4, 4.4-5,4.44,5.1-1,5.1-2,5.1-3,5.31,5.3-2,5.5-1,5.5-2,5.5-4,5.6-1,6.2-1 A, 6.2 1 B, 6.2 1 C, 6.24, 7.2 4, 7.34,15.1.0-1,15.1.04,15.1.4-1, and 15.1.104 FIGURES: SAR 5.31, and 5.3-2 TiertifiedTteviewers Signature Bryan Daiber bI Printed Name Date Reviewer's certification expiration date: 3-18-2000 Assistance provided by:
. Printed Name Scope of Assistance Date Search Scope Review Acceptability (NA, if performed by Technical Review per 1000.006) b Certified Reviewers Signature bd k), f~ods 7- /E-18 Printed Name Date
ARMANSAS NUCt. EAR ONE FORM TITLW FORM NO. REV. 19CPRSS.Se DETERMINATION 1800.131 A 3 PC-1 i Page 2 of 1Q ENVIRONMENTAL IMPACT DETERMINATION (UNIT 1 and UNIT 2) Document No. Amendment 189 & 190 RevlChange No. 0 Complete the following Determination. If the answer to any checklist item is "Yes", an Environmental Evaluation is required. See Section 6.1.4 for additional guidance. Will the Activity being evaluated: Yes No O @ Disturb land that is beyond that initially disturbed during construction (i.e., new construction of bur. dings, creation or removal of ponds, or other terrestrial impact)? See Unit 2 SAR Figure 2.5-17. This applies only to areas outside the protected area. O @ increase thermal discharges to lake or atmosphere? O B Increase concentration of chemicals to cooling lake or atmosphere through discharge canal or tower? O E Increase quantity of chemicals to cooling lake or atmosphere through discharge canal or tower? O 2 Modify the design or operation of cooling tower which will change drift characteristics? O E Install any new transmission lines leading offsite? O @ Change the design or operation of the intake or discharge structures? O @ Discharges any chemicals new or different from that previously discharged? O E Potentially cause a spill or unevaluated discharge which may effect neighboring soils, surface water or ground water? O @ Involve burying or placement of any solid wastes in the site area which may effect runoff, surface water or ground water? O E Involve incineration er disposal of any potentially hazardous materials on the ANO site? O O Result in a change to nonradiological effluents or licensed reactor power level? O E Potentially change the type or increase the amount of non-radiological air emissions from the ANO site. I
I ARKANSAS NUCLEAR ONE FORM TITLE: FORM NO. REtf. 14CPRSS.89 REVIEW CONTINUATION PAGE 1eet.131C 3 Page 1 of.19 Document No. Amendment 189 & 190 Rev> Change No. O j 10CFR50.59 Review Continuation Pane BackEround I As a result of steam generator tube plugging, RCS flow and steam generator pressure have degraded To accommodate these effects the RCS flow Technical Specification limit was reduced from 120.4 x 10' lbm/ hr to 108.4 x 10' lbm/hr and the MSIS setpoint was reduced from 751 psia to 712 psia. Dese Technical Specificatio change requests were approved in Amendments 189 and 190. Most of the information provided in the submittal ) packages to the NRC was already incorporated in the SAR as a result of the Cycle 13 reload effort. His SAR . change completes the implementation of these two Technical Specification Amendmaat packagsplus incorporate the results of various other calculations performed to support the overall 30% tube plugging effort. 90-E-0096-02 Boron Dilution l l 88-E-0032-08 Contamment LOCA Impacts I 96 E-0028-01 RCS Flow Coastdown Data 96 E-0028-02 Loss of Flow analysis 96-E-0053-01 CPC Filters 96-E-0025-01 j SYSFLOW 97-E-0204-01 ) SGTR 88-E-0097-04 COPATTA for MSLB Round 1 { 88-E-0097-05 MSLB Summary Calc 88-E-0097-06 COPATTA for MSLB Round 2 94-E-0093-01 MSLB Recommended Inputs 3 I 94-E-0093-02 RELAP Basedeck 94-E-0093-03 SGN-III Basedeck 94-E-0093-04 SGN-III Round 1 Runs 94-E-0093-07 SGN-III Round 2 Runs 95-E-0076-05 MSLB Dose 91-E-003 ? 08 Figure 3.6-1 conversion to Procedure Numbers Results from all of the above calculations were explicitly covered in the Technical Specification change requests expect the SYSFLOW calc and 91-E-0035-08. Discussion
%c SAR changes covered by this package can be classified as those not requiring an evaluation based on Item # C (incorporation of information submitted and approved by the commission) and Item # F.2 (Minor clarifications) ,
and other changes which require an evaluation. All of the changes made which are consistent with the information submitted to the NRC will be discussed first followed by a discussion of all other changes.
ARKANSAS NUCLEAR ONE l FORM TITLE. FORM NO. REV. 10CFR50.89 REVIEW CONTINUATION PAGE { 1000.131C 3 1 l Page 5 of_1_0 I Document No. Amendment 189 & 190 FievjChange No. _0_ j J
- Changes Not Requiring an Evaluation
- 1) Changes Consistent with the Amendment Submittals Item # C l
)
J a) - Section 4.4.1.2 has been updated to indicate that the thermal margin analyses were performed to a value of 90% of " design" flow, his statement is consistent with the analytical work presented in the RCS flow reduction Technical Specification submittal. This information is also consistent with the work performed to address the Cyc!c 13 reload analysis work covered by a separate 50.59. I b) He last paragraph of Section 4.4.3.8 has been deleted. This infonnation is old information which should have been removed when Amendment 156 was incorporated into the SAR. Amendment 156 submittal presented a revision to the LBLOCA methodology including the application of that methodology to ANO-2. i ne limiting break size was changed to a 0.6 DEG/PD break. This information with respect to the new LBLOCA analysis is presented in Section 6.3.3.2.2 of the SAR; hence, this information does not need to be repeated in Section 4.4.3.8. Although the new LBLOCA SAR description does not indicate the exact amount of energy released due to the chemical reaction of Zircaloy clad, this section does reflect the* amount of potential clad oxidation. The submittal to the NRC with respect to the application of the new LBLOCA methodology also did not discuss the energy released, only the amount of clad oxidation. Based on this, the removal of the old LBLOCA application information can be removed from the SAR as it is superseded by the current LBLOCA analysis. l c) Section 5.5.1.3 has been updated to indicate that the analyses of steady state and anticipated transients is ! performed assuming " conservative" flow rates versus "nummum design". Minimum design flow rate refers l to 322,000 gpm (120.4 x 10'lbm/hr). A new nummum flow has been defined by Amendment 190 as 90% of the mimmum design flow. Rather than changmg the "nummum design" statement to "90% of nummum design" a more bounding state.:nent indicating " conservative" flow rates are assumed. This change was made due to the fact that some of the anticipated transients are analyzed at 110% of design flow as this assumption is more conservative. This change is consistent with the analyses submitted to the NRC in Amendment 189 and 190. Most of these analyses were performed at 90% ofdesign flow; however, several exceptions were noted in the submittals where higher flow was considered more conservative. l d) Section 6.2.1.3.3.4.4 was added to the SAR to discuss the effect of a 10% reduction in RCS flow due to steam generator tube plugging. His information is consistent with the iaformation submitted to the NRC in f l the Amendment 190 request. l c) Table 15.1.0-6 was updated to reflect an analysis RCS flow range of 90% to 110% ofdesign flow. This analysis flow range is consistent with the information presented in the Amendment 190 submittal. All of the ! Chapter 15 events affected by 90% flow were specifically addressed in the submittal. I f) Section 15.1.4.4 and 15.1.4.4.1 were added to the Boron Dilution Incident section of the SAR. His information is consistent with the assessment submitted to the NRC in Amendment 190 to address the RCS volume reduction effect.s due to steam generator tube plugging. g) Sections 15.1.5.2.3,' 15.1.5.2.3.2, and 15.1.5.2.3.3, Tables 15.1.5-8 and 15.1.5-9, and Figures 15.1.5-16 and 15.1.5-17 have been added to the Loss of RCS Flow section of the SAR. The information in the text, tables and figures comes directly from the Amendment 190 submittal. Section renumbering was also undertaken to fit these sections into the SAR.
l annassas muetEan OmE FORM TITLE: FORM NO. R EV. 14CPRSS.40 REVIEW CONTINUATION PaGE 1000,131C 3 Page g of 12 Document No. Amendment 189 & 190 Rev> Change No. ,,,,,,9 h) Section 15.1.2.4.2.2 was added to the CEA Withdrawal analysis section. The first two paragraphs of this section come directly for the Amendment 190 submittals. This information was reviewed and approved by the NRC. i) Section 15.1.18.4 and 15.1.18.4.1 have been added to the SGTR event discussion in the SAR. His additional information comes directly from the submittal to the NRC in support of Amendment 190. His information was reviewed and approved by the NRC. j) Sections 15.1.10.4 and 15.1.10.4.1 have been added to the SAR. The first and last paragraphs from this new section come directly from the submittal to the NRC. As a result this information has been reviewed and approved as part of Amendment 190.
- 2) Minor Clarifications Item #F.2 a) The note on Table 5.1-3 which indicates that the information in the table is considered nominal has been expanded to further clarify that steam generator tube pluggmg impacts are not reflected in this table. He pressure drops noted in this table are dapandant upon the number of steam generator tubes plugged. As this table is presented for nommal value information, the amount of steam generator tube plugging is not
;elevant. To ensure the reader is aware of this, the additional clarifying statement has been added. This does change the intent of this table, as the table is presented for nommal information only.
b) The title to Table 5.5-4 has been clarified to reflect that these parameters are " design" parameters. His clarification is consistent with the wording in Section 5.5.3.1 which refers to Table 5.5-4 as contammg principal parameters. This is the only reference to these tables in the SAR. This title clarification is consistent with the information provided in the table and the title of the information noted in the table (Design Temperature, and Design Pressure) c) The MSIS instrument setpoint is repeated in Section 6.2.1.1.2.6 of the SAR. This information is contained in the Technical Specifications and in SAR Tables 7.3-5 and 7.2-4. He inclusion of this information in ; Section 6.2.1.1.2.6 has no relevance on the analysis assumptions and is only a repeat of the inrormation i already documented in the SAR. The sentence following the MSIS instrument setpoint information in Section 6.2.1.1.2.6 needed to be reworded for proper undermading following the removal of the setpoint information. He word " pressure"in the following sentence was replaced with " Main Steam Isolation System (MSIS) analysis setpoint". This effectively clarifies the " pressure" assumption and does not constitute a change in the intent or scope of the statement.
ARKANaAa NUCLEAR ONE FORM TITLE: FORM NO. REV. 10CPRSS.se REVIEW CONTINUATION PAGE 1980.1s1C s s. Page I of_1g Document No. Amendment 189 & 190 Rev> Change No. _1 Changes Requiring an Evaluation a) Currently, the SAR indicates that nunimal imM-- are expected due to few steam generator tubes being plugged. 'the SYSFLOW analysis was performed to quarmfy the allowable asymmetry which could exist such that the current analysis assumptions with respect to flow imbalances are still maintamed A paragraph defimng the allowable limits on tube pluggmg asymmetry has been added to quantify the current SAR statement with respect to the analysis assumptions. These restrictions are being inserted to ensure the
. analysis assumptions are protected. This statement has no impact on the current analysis results as the limits were back calculated from conservative assumptions with respect to flow imbalances which have always existed in the analysis bases b) Table 4.4-2 has been updated to reflect Cycle 13 data. This information was covered under the Cycle 13 50.59; however, this table was not updated with that effort. Most of the updates are related to the RCS flow reduction effort addressed for Cycle 13 and the reduction in the r.enber of shim rods. All of the Cycle 13 core design changes and effects of reduced RCS flow have been assessed. All of the analyses presented to the NRC as part of An=h s 189 t and 190 incorporated these design considerations. The results of the analyses presented to the NRC were considered acceptable plus all of the Cycle 13 specific analyses not presented to the NRC have also indicated acceptable results by incorporation under the Cycle 13 50.59.
The data in this .able is used in the thermal hydraulic analysis to verify that the calculational methods for determining DNBR, veri 6 cation of DNB in the plant with SCU, and the assumptions in the ECCS analysis with respect to fuel thermal hydraulics are acceptable. The Cycle 13 verification effort is.arnated that the appropriate assumptions were made in the ECCS analysis (which was part of the NRC submittal) and other DNB verification =4ade with the appropriate penalties applied for Cycle 13. c) Section 5.3.4 has been updated to reflect that power operations requires all 4 RCPs to be runmng. This is consistent with Technical Specification 3.4.1.1 requirements. Table 5.3 1 was deleted as part of this update. The original intent of this Table and operatmg description was to allow for power _ operations with less than all 4 RCPs running As these other modes of operation were not approved by the NRC, these changes update th:: SAR to remove this inforniation. As these modes of operation are not allowed by Technical Speci6 cations, this information has not been revalidated on a Cycle speci6c basis. This change effectively updates the SAR to be consistent with the safety analysis assumptions and Technical Specification limitations. No new modes of operation are being added by this change, this change only rewords the section to be consistent with the current analyzed allowable operatmg modes d) _ Two paragraphs in Section 5.3.4 refernng to the COAST code have been removed as historical data. COAST was noted here for essentially 2 purposes lhe first was for predicting steady state flow with 4 RCPs operatmg and the other is to determme steady state flow with less than 4 RCPs runmng Consistent with item e above, only 4 pump operation _ is allowed by the Technical Specifications; therefore, the use of COAST for determimag steady state flow for the other power operatmg modes is not used. The use of COAST for deternumns the initial 4 pump steady state flow is also being removed as historical. The 4 pump flow verification is essentially a moot point as Technical Specifications requires a verification of the minimum flow.' The COAST predicted flow value is not used in any analyses. Additionally, the COAST code was submitted to the NRC as a method to predict flow coast down following a loss of flow or seized rotor event, not as a method for determmmg steady ete flow. Removal of the COAST code discussion has
- no impact on the analysis results or changes the allowable operatmg modes. Steady state RCS flow
Anx4NsAs Nuca. EAR ONE FcRM TITLE: FORM NO. REV. 10CPRSS.88 MEVIEW CONTINUATION PAGE iget.131C 3 Page 1 of_1g Document No. Amendment 189 86 190 RevdChange No. ,,,,,,g predictions from COAST are not used in the safety analyses. De safety analyses use their own methods for determmmg RCS flow and are adjusted as necessary to acco.. ..cdde h=ading mmimum and maximum RCS flow input assumptions. As this information with respect to COAST is being removed so must Reference 1. i e) Some of the volumes in Tables 5.1-1,5.1-2, and 5.5-2 were updated to reflect what are currently believed l be the original design volumes. Most of these changes are very minor and are based on ABB-CE references that date back to 1975 and 1980. These volumes are believed to be the latest best estimate volumes. These values are consistent with the values in Groundrules and the RCS volume c 92-E-0050-XX. Rese values were used in the analyses submitted to the NRC as part of A= adments 189
)
and 190. Changes in these volumes have no impact on the plant operation or accident analyses. De { corrected volumes are believed to be the actual volumes used in most of the original design analysis work { and current analysis efforts. He SAR already refers to the calculation series 92-E-0050-XX for the latest RCS volume numbers. Most of these numbers have not changed over the years; hence, the more correct original design numbers are being incorporated into these tables. His change has no effect on the plant, it is only updatmg RCS volume estimates in the SAR to what are believed to be more accurate numbers. 1 f) ne last paragraph of Section 15.1.2.4.2.2 was added to the SAR as additional information which describes this event input assumptions. His event description and results were presented to the NRC in Amendment 190 submittals; however, this additional information with respect to the input assumptions were not provided to the NRC. This information is being added to be consistent with the current level of detail provided in the SAR currently for this event. These input parameters are considered conservative inputs assumptions for this analysis and are consistent with the limits allowed by Technical Specifications, where the Technical Specifications apply. These input assumptions bound the currently allowed operating conditions; thereby, ensure the analysis results are conservative. g) The second paragraph of Section 15.1.10.4.1 has been added to the SAR to present additional information consistent with the other excess heat removal events already presented in the SAR. His information was not directly provided to the NRC in the Ann = i=wt 190 submittal. This information is being added to be consistent with the current level of detail provided in the SAR currently for this evem. These input parameters are considered conservative inputs assumptions for this analysis and are consistent with the limits allowed by Technical Specifications, where the Technical Speci6 cations apply. %cse input assumptions bound the currently allowed operatmg conditions; thereby, ensure the analysis results are conservative. Basis for Determination:
'I) Require a change to the Operating License?
This effort does not result in a change to the Technical Specifications, Operating Licenc or Confirmatory Orders. The purpose of the change is to implement the information submitted to the NRC in support of Amendments 189 and 190, which are the MSIS setpoint and RCS flow Technical Specification change paA=>an. The majority of the information covered by this deternunation is related to the ar.m'ir.cr.t submittal information. Some additional information relating to the 30% steam generator tube pluggmg effort is also included in this package. His i additional information does not require a change to the Operatmg License. ' l
l ARKANaAa NuCLaAR ONE FORM TITLE: FORM NO. REV.
' 1eCFR88.88 REVIEW CONTINUAT1oN PAGE 1000,131C 3 Page 1 of,lg Document No. Amendment 189 & 190 Rev/ Change No. _g
- 2) Result in information in the followmg SAR documents (including drawmgs and text) being (a) no longer true or accurate, or (b) violate a requirement stated in the document:
The main purpose of this effort is to update the SAR to be consistent with the information presented to the NRC as a result of the RCS flow and MSIS setpoint Technical Specifications. During this update process various other SAR updates were noted associated with other mi-H=a~ms 30% tube pluggmg efforts. These other SAR changes are being incorporated by this 50.59. As most of the information being updated by this 50.59 is related to the Aw='=ent submittals, that information has already been reviewed and approved by the NRC, Other SAR updates include a clean-up of an outdated LOCA statement, several minor clarification efforts relatung to nommal information not accounting for tube plugging, design data clarification, repeated information with respect to the MSIS setpoint, added information with respect to flow imbalance allowances, Cycle 13 thermal hydraulic data, less than 4 pump operation considerations, removal of COAST code reference, minor RCS volume corrections, and the inclusion of additional information for the CEA withdrawal and excess heat removal events which was not include in submittals to the NRC. No changes or impacts to the COLR, FHA, Technical Specification bases, or Technical Requirements Manuals have been identified as a result of the hmaats covered by this 50.59. All of the analyses performed in the above noted calculations and those performed to support the NRC submittals were performed consistent with the current limitations and requirements of the documents, with the exception of RCS flow and MSIS setpoint which were changed by Amaadmaats 189 and 190. All of the analyses submitted to the NRC as part of Amaadmaats 189 and 190 amend the sinular analyses presented to the NRC for various other occasions and included in the associated SERs. As these new analyses were reviewed and approved by the NRC, the prior analyses results noted in prior NRC SERs are considered to be appropriately amand~t
- 3) Involve a test or expenment not described in the SAR7 No tests or experiments are proposed or affected by this change. These changes on!y relate to SAR updates consistent with the analysis assumptions to support the Amendment 189 and 190 submittals. 'Ihe various other SAR updates being proposed by this effort also relate to the other analyses efforts that have in some way been updated or affected by 30% tube pluggmg, non of which relate to test or expenments.
- 4) Result in a potential impact to the environment?
No physical plant modifications or conditions are being proposed by the change. 'Ihese changes are related only to safety and design analysis assumptions and results. No physical modifications to the plant are being implemented by this change.
- 5) Result in the need for a Radiological Safety Evaluation per Section 6.1.57 No activity involving the processing of radioactive material is being proposed by this change; hence, a Radiological Safety Evaluation is not needed 'Ihis package relate only to safety and design analyses.
' 6) - Result in any potential impact to the equipment of facilities utilized for Ventilated Storage Cask actisities per Section 6.1.67 t-
anw===== eeucuman cme FORM TULE: FORM NO. REV. 14CFRSS.89 REVIEW CONTINUATION PAGE 1000.131C 3 Page 10 of 10 Document No. Ardendment 189 & 190 Rev./ Change No. J There are no proposed activities by this change which would involve any aspect of the VSC is limited to RCS safety and design basis analysis considerations. . 7) Involve E-Plan? a change under 10CFR50.54 for the following SAR documents per Section 6.1.7: a) QAM There are no proposed activities by this change which would affect the QAMO or E-Plan. This limited to SAR updates related to RCS safety and design analysis efforts in support of the RCS flow and Technical Specification changes, and 30% tube pluggmg related analyses. l I
ARKANSAS NUCLEAR ONE FORM TITLE: FORM NO. REE 10CFR50.58 EVALUATION 1000.1318 3 l 1 Page 1 of,j! 10CFR50.59 Eval. No. Wf@WCOO (Assigned by PSC) ! Document No. Amendment 189 & 190 Rev> Change No. O i f Title SAR Updates for RCS Flow and MSIS Tech Spec Chances A WRITTEN RESPONSE PROVIDING THE BASIS FOR THE ANSWER TO EACH QUESTION MUST BE ATTACHED EACH QUESTION MUST BE ANSWERED SEPARATELY. A SIMPLE STATEMENT OF CONCLUSION IS NOT SUFFl : LENT. ATTACHMENT 2 PROVIDES GUIDANCE FOR RESPONSE. If the answer to any question on this form is "Yes," then an unreviewed safety question is involved. If the answer l to all questions is "No," then the proposed change does not involve an unreviewed safety question.
- 1. Will the probability of an accident previously evaluated in the SAR be increased?
Yes O No @
- 2. Will the consequences of an accident previously evaluated in the SAR be increased? Yes O No @
- 3. Will the probability of a malfunction of equipment important to safety be 1
increased? Yes O No @ l ? l l 4. Will the consequences of a malfunction of equipment important to safety be increased? Yes O No @
- - 5. Will the possibility of an accident of a different type than any previously evaluated in the SAR be created? Yes O No @
l l
- 6. Will the possibility of a malfunction of equipment important to safety of a l different type than any previously evaluated in the SAR be created? Yes O No @
l l
- 7. Will the margin of safety as defined in the bases for any technical
. specification be reduced? Yes O No @
e AnnaNaas NUCLEAR ONE FORM TITLE: FORM NO. REV. tecFRee.se EVALUATION 1000,1313 3 PaQe2of.E' Document No. Amendment 189 & 190. Rev> Change No. 1 J-) Bryan Dalber 7" ll- i T' MfiedMeviewers Signature Printed Name Date Reviewers certification expiration date: 3-18-2000 Assistance provided by: Printed Name cope of Assistance Date PSC review by: I
' Date:
( (
)
i 2 i 1 1
AmuussAs NUCUEAR ONE FORM TITLE: FORM NO. GEV. SOCPRSS.88 HEVIEW CONTINUATION PAdBE 1SSS.131C 3 t Page 2 of g Document No. Amendment 189 & 190 Rev/ Change No. g 10CFR50.59 Review Continuation Pace A summary of the changes being incorpomted by this evaluation has been made in the determination section. The ! information in the detennination section delineates the need for this evaluation. Refer to the determina discussion of the changes and background with respect to these changes. l 1. Will the probability of an accident previously evaluated in the SAR increased? No I i ne changes proposed by this package are mostly related to safety and design analysis input assumptions, methods, and results. Limitations with respect to steam generator tube plugging asymmetry was added to ensure the safety l analysis input assumptions are maintained, assumptions with respect to the Cycle 13 thennal hydraulic data l effectively related to a 10 % reduction in RCS flow and fewer shim rods were added consistent with the Cycle 13 l analysis assumptions, less than 4 pump operation considerations cunently dermed in the SAR were removed as historical data due to these modes of operation not being allowed by Technical Specifications, reference to the COAST code was removed as used as an input to these less than 4 pump operating modes consistent with the removal of the information discussing these modes of operation, minor RCS pnmary volume corrections consistent with design and safety analysis considerations were made, and the inclusion of additional input parameter information for the CEA withdrawal and excess heat removal events which was not included in the Amendment 189 and 190 submittals to the NRC. l Here are no new systems, components, substructures, design changes, physical alterations, nor operating procedure l changes being proposed by this change. The information incorporated into the SAR by this change relates only to the safety and design analysis input assumptions, methods, and results so the probability of an accident presiously l evaluatelin the SAR will not be increased
- 2. Will the consequences of an accident previously evaluated in the SAR be No increased?
The changes proposed by this package are mostly related to safety and design analysis input assumptions, methods, and results. Limitations have been added with respect to' steam generator tube plugging asymmetry allowances to l- ensure the safety analysis input assumptions are maintamed for analysis consequences which have already been l approved by the NRC or addressed under the Cycle 13 50.59. Dese limitations will not increase the presiously l evaluated consequences, rather they will help ensure the current analysis results are maintamed Assumptions with
. respect to t eh Cycle 13 thermal hydraulic data effectively related to a 10 % reduction in RCS flow and fewer shim rods were added consistent with the Cycle 13 analysis assumptions. These assumptions also define the limits to which the already approved Cycle 13 analysis has been assessed to, helping to ensure the analysis limits are ;
maintamed. Dese assumptions will not affect the consequences of the previously evaluated SAR accidents. SAR l discussions with respect to less than 4 pump operation considerations and the use of the COAST code were removed as historical data. He removal of this information is consistent with the current Technical Specification limits. Removing this information does not impact the consequences of an accident previously evaluated in the SAR. Minor RCS pnmary volume conections consistent with design and safety analysis considerations were made. The corrections to these RCS volumes are consistent with the current analysis assumptions; hence, support the current accident analysis consequences presented in the SAR. Additional input parameter information for the CEA withdrawal and excess heat removal events which was not included in the Amendment 189 and 190 submittals to the { i
}
ARKANSAS NUCLEAR ONE FORM TITLE: FORM NO, REV. 10CFR50.89 REVIEW CONTINUATION PAGE 1000.131C 3 Page 4 of g Document No. Amendment 189 & 190 Rev/ Change No. 0 NRC is being added by this effort. He addition of this information to the SAR does not affect the result of these analyses. De input parameters added tu these sections are considered to be conservative input assumptions wit respect to the limits allowed by Technical Specifications, where applicable. As these input assumptions are l considered to be conservative with no impact on the accident consequences. The SAR changes resulting from the calculations and other efforts covered by this evaluation deal with the input parameters, methods and results presented in the SAR. As these analyses are performed with input assumptions which are considered to be conservative and appropriate for the current plant configuration, with no increase in the consequences of the analyses previously evaluated in the SAR. 3. Will the probability of a malfunction of equipment important to safety be increased? No The purpose of the change is to update the SAR to reflect the safety and design analysis assumptions and rem SAR statements with respect to power operations with less than 4 pumps runnmg. With respect to the ana assumptions, these facets of the safety analyses do no affect the way in which the plant equipment operates or functions. There are no new operating modes being imposed, no new equipment, no changes in the current procedures nor physical changes to the pt:nt being proposed. Dese analysis assumptions are conservative with respect to the current plant operation. He results from these analyses have already been approved by the NRC o under a separate 50.59. The changes in the RCS primary volumes are due to more accurate estimates of these volumes which are consistent with the current analyses assumptions with respect to these volumes. With res the removal of the SAR considerations of power operation with less than 4 RCPs runnmg, this change is not a change in the plant operating conditions as it is consistent with the current Technical Specification limits. None of these changes result in a change to the original design specifications for material and construction practices. N of these SAR updates propose an activity which degrades safety system component reliability. Based on these considerations, the probability of a malfunction of equipment important to safety is not impacted. l 4. Will the consequences of a malfunction of equipment important to safety be No l increased? There are no physical impacts on equipment which could result in an increase in consequences proposed by th change. No new plant operating modes nor changes in plant operating conditions or physical design are bein proposed. Limitations with respect to steam generator tube plugging asymmetry was added to ensure the safety analy assumptions are maintained, these limitations are consistent with the current analysis assumptions and results thereby not increasing the consequences of a malfunction of equipment important to safety. Assumptions with respect to the Cycle 13 thermal hydraulie data effectively related to a 10 % reduction in RCS flow and fewer shim rods were added consistent with the Cycle 13 analysis assumptions, these updates are consistent with the Cycl analysis work for which the results of the accident analyses were submitted to the NRC in Amendments 189 and 190 or covered by a separate 50.59. He Cycle 15 thermal hydraulic data does not increase the consequence of a malfunction of equipment important to safety. Removing the less than 4 pump operation considerations curre defmed in the SAR does not affect the consequences of a malfunction of equipment. Removal of this information is consistent with the Tech Spec limitations. Removal of the reference to the COAST code for determmmg RCS flow does not effect the consequences of a malfunction ofequipment important to safety as the mmimum RCS flow value
p-assummas Nuctman aNe FORM TITLa: FORM NO. REV. 10CPRse.se REVIEW CONTINUATION PAGe 10st.131C 3 Page l of,1 Document No. Amendment 189 & 190 RevlChange No. 9. which is verified as required by Tech Spec (or maximum expected RCS flow) is used to deternune the accident consequences. Minor RCS pnmary volume corrections consistent with design and safety analysis considerations were made, these volumes changes do not impact the consequences of a malfunction of equipment important to safety. 'Ihe inclusion of additional input parameter information for the CEA withdrawal and excess heat removal events which was not included in the A.T.cr.d.T. cat 189 and 190 submittals to the NRC does not affect the consequences of a malfunction of equipment important to safety, as this input parameter information is considered to be conservative analysis assumptions ensuring conservative deternunation of the impact of these events. 5.- Will the possibility of an accident of a different type than any previously evaluated in No the SAR be created? l l No new ist ators or accidents are caused by this change This change relates to the update of the SAR to reflect the i safety and design analysis assumptions and removal of the SAR statements with respect to power operation with less than 4 RCPs nmning No plant modifications, new components, physical alterations, or operating conditions are being implemented by this change. No new accidents are created and no currently non-limiting events becomes more limiting.
- 4. Will the possibility of a malfunction of equipment important to safety of a different No type than any previously evaluated in the SAR be created?
1 There are no new systems, components, substructures, physical design changes, physical alterations, or operating procedure changes being proposed or required by this change. 'Ihis change affects the SAR safety and design analysis assumptions. As there are no physical changes to the plant and the results of these analyses are approved by the NRC or other 50.59, the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR will not be created.
- 7. Will the margin of safety as defined in the bases for any technical specification be No reduced?
The safety and design analyses efforts covered by this evaluation have been performed consistent with the Technical Specification requirements. The analyses which utilized the Technical Specifications or COLR limits are considered ! acceptable with respect to the bases for these specifications. The removal of the SAR discussion of power I operations with less than 4 RCPs runmng is consistent with the Technical Specifications. 'Ihe margin of safety as l defined by the bases for the Technical Specifications are not reduced by these changes. i 1
- j i
I i
)
ARKANSAS NUCLEAR ONE FORM TITLE: FORM NO. REV. 10CFR60.69 DETERMINATION 1000.131A 3 PC 1
\
l Page 1 of_5 i Document No. 98-00641 Rev/ Change No. 0 l l Title Chance to M-2220. Sheet 1 i Brief description of proposed change: Chance valve position of 2PH-1 from open to closed on P&lD { M-2220. sheet 1. l Will the proposed Activity:
- 1. Require a change to the Operating License including:
Technical Specifications (excluding the bases)? YesO No@ l Operating License? YesO No@ Confirmatory Orders? YesO No@
- 2. Result in information in the following SAR documents (including drawings and text) being s (a) no longer true or accurate, or (b) violate a requirement stated in the document:
SAR (multi-volume set for each unit)? YesO NoO Core Operating Limits Report YesO No@ Fire Hazards Analysis? YesO No@ Bases of the Technical Specifications? YesO No@ Technical Requirements Manual? YesO No@ NRC Safety Evaluation Reports? YesO No@
- 3. Involve a test or experiment not described in the SAR? . . ] No@
(See Attachment 2 for guidance)
- 4. Result in a potential impact to the environment? (Complete the EnvironmentalImpact Determination of this form.) YesO No@
- 5. Result in the need for a Radiological Safety Evaluation per section 6.1.57 YesO No@
l 6. Result in any potential impact to the equipment or facilities utilized for Ventilated Storage Cask activities per Section 6.1.67 YesO No@ l 7. Involve a change under 10CFR50.54 for the following SAR documents
. per Section 6.1.7:
QAMO? YesO No@ E-Plan? YesO No@ l'
ARKANSAS NUCLEAR ONE FORM TULE: FORM NO. REV. 10CFR50.55 DETERMINATION 1000.131A 3 PC-1 Page 2 of_5 Document No. 98-00641 RevdChange No. O Basis for Detennination (Questlo'ns 1. 2 & 31: The proposed activity changes the indicated position of 2PH-1 from open to closed on M-2220, sheet 1, A-8 (SAR figure 3.2-2). Therefore, a 50.59 evaluation is required. This change does not affect the LBD's in any other way and does not involve a test or experiment. O Proposed change does not require 10 CFR 50.59 Evaluation per Attachment 1, item # . (If checked, note appropriate item #,, send LDCR to Licensing).,, Search Scope: List sections reviewed in the Licensing Basis Documents specified in Question 1,2 and 3. If a keyword search was done on LRS, "all" may be entered under "Section" with the keyword (s) used in parentheses. Controlled hard copies of the documents shall be reviewed (LRS is not verified and searches only text, not figures or drawings). Attach and distribute a completed LDCR per Section 6.1.2 if LBD changes are required. Document Section LRS: UNIT 2 50.59 All (plant heating system; 2pht; 2t*22; plant heating w/25 day tank; boiler w/25 day tank; plant heating; heating) MANUAL SECTIONS: UNIT 2 SAR 3.6.4.4.3; 9.2.4; 9.4 FIGURES: UNIT 2 SAR 3.2-2 d t ifA Kent Fancher 03/23/98 Cerfified Reviewers Signature Printed Name Date Reviewers certification expiration date: 09/26/98 Assistance provided by: Printed Name Scope of Assistance Date Search Scope Review Acceptr.bility (NA, if performed by Technical Review per 1000.006) @ -i Y*% tw A wtsal Vtv/ftf Certified Reviewers Signature Printed Name ~ Date i
ARKANSAS NUCLEAR ONE FORM TITLE: FORM NO. REV. 10CFR50.59 DETERMINATION 1000.131A 3 PC-1 t l Page 3 of.! ENVIRONMENTAL IMPACT DETERMINATION (UNIT 1 and UNIT 2) Document No. 98-00641 Rev1Cnange No. 0 l l Complete the following Determination, if the answer to any checklist item is "Yes" an Environmental Evaluat is required. See Section 6.1.4 for additional guidance. Will the Activity being evaluated: Yes No O @ Disturb land that is beyond that initially disturbed during construction (i.e., new construction of buildings, creation or removal of ponds, or other terrestrialimpact)? See Unit 2 SAR Figure 2.5-17. This applies only to areas outside the protected area. O @ Increase thermal discharges to lake or atmosphere? O O Increase concentration of chemicals to cooling lake or atmosphere through discharge cana , - tower? . O E increase quantity of chemicals to cooling lake or atmosphere through discharge canal or tower? O E Modify the design or operation of cooling tower which will change drift characteristics? j O O Install any new transmission lines leading offsite? O E Change the design or operation of the intake or discharge structures? O @ Discharges any chemicals new or different from that previously discharged? 1 0 @ Potentially cause a spill or unevaluated discharge which may effect neighboring soils, surface water or ground water? O O involve burying or placement of any solid wastes in the site area which may effect runoff, surface water or ground water? ! O O Involve incineration or disposal of any potentially hazardous materials on the ANO site? O E Result in a change to nonradiological effluents or licensed reactor power level? O @ Potentially change the type or increase the amount of non-radiological air emissions from the ANO site.
lll:- 1b-00b4I y09e "I or IV ARMANsAs NUCLEAR oNE FORM TITLE: FORM fdO. REV. 10CFR40.69 EVALUATION 1000.1318 3 Page 4 of! 10CFR50.59 Eval. No. T-6 -M4T- Oti'l (Assigned by PSC) Document No. 98-00641 RevJChange No. O Title Chance to M 2220. Sheet 1 A WRITTEN RESPONSE PROVIDli4G THE BASIS FOR THE ANSWER TO EACH ATTACHED. EACH QUESTION MUST BE ANSWERED SEPARATELY. A SIMPLE S CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES GUIDANCE FOR If the answer to any question on this form is "Yes," then an unreviewed safety question is involved. to all questions is "No," then the proposed change does not involve an unreviewed safety question. 1. Will the probability of an accident previously evaluated in the SAR be increased? Yes O No @ This activity chances the valve position of 2PH-1 from open to closed on the P&lD (M-2220, SH1 The valve is already maintained closed by the procedure, except while fillino the Unit 2 Plant Heatino Boiler Fuel Oil Day Tank (2T-22). and does not chance the operatina characteristics.This chance does not affect system desian or fu The probability of an accident previousiv evaluated in the SAR will not be increased. 2. Will the consequences of an accident previously evaluated in the SAR be increased? Yes O No @ This chance causes the P&lD to acree with actual plant confiauration. The Plant Heatino System is not s ~ed related or risk sianificant and does not involve any type of accident. The consecuences of an accident creviousiv evaluated in the SAR will not be increased. 3. Will the probability of a malfunction of equipment important to safety be increased? Yes O No @ This chance does not involve any oculoment important to olant safety. The chance provides acreefr,ent Mt;;;;a the P&lD and the procedure and it ensures that the day tank will not be inadvertentiv filled while tinit 1 is supolvino other eauipment with fuel oil. The Plant Heatino gygtem is not reouired to mit!at the consecuences of any analyzed accident condition. The probability of a malfunction of souloment important to safety will not be increased. 4. Will the consequences of a malfunction of equipment important to safety be increased? Yes O No @ 2PH-1 is the last valve opened when fillina the day tank and the first valve closed when securino the fill. Beceeee it is located near the day tank level indication, it provides the best positive control and monitorina of the day tank fill ossration. Failure of 2PH-1 in the closed position will not affect any safety related csiasonant and there is no equipment important to safety located in the area. The consecu6nces of a malfunction of souloment important to safety will not be increased.
, y w w w 7, m .v . ..
ARKANSAS NUCLEAR ONE
, FORM TITLE:
FORM NO. REV. 10CFRSS.59 EVALUATION 1900.1313 3 Page 5 of_1 Document No. 98 00641 Rev1 Change No. 2 5. Will the possibility of an accident of a different type than any previously evaluated in the SAR be created? Yes O No @ The Plarff Heatina 8v m is not safety related or risic sianificant. This chance in no way affects the ability oFthe system to ise!a+3 clant hest;ns to the contain; neat buildina on a Containment isolation Acte ^_':-s Siansi. No accid at condition previousiv evale":d on Unit 2 will be affected or initiated by this chance. The possib!!?v of an accident of a different tvoe than previously evaluated in the SAR will not be created. 1
- 6. Will the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR be created?
Yes O No @ No new failure modes are intredeced by maintainina 2PH-1 in the closed position. The only safety function of the Plant H : tina Svei-m is to isolate the centsinment buildina on a CIAS and the position of 2PH-1 ec=s nct eLct that function at all. The poss!b!!ity of a malfunction of eauipment imsoitant to safety of a different tvss than previousiv evaluated in the SAR will not be created. 2
- 7. Will the margin of safety as defined in the bases for any technical specification be reduced?
Yes O No @ Eeither 2PH-1 nor the Plant Heatina System is included in the Unit 2 Technical Specifications, therefore the marain of safety as defined in the bases for any technical specification will not be reduced. ;
- i. s:
4d Kent Fancher CeAlfied Reviewers Signature 03/23/96 Printed Name Date Reviewers certification expiration date: 09/26/98 Assistance provided by: Prirped Name Scope of Assistance Date PSC review by:
~
k _ Date: % ~L - 4
f ., ARKANSAS NUCLEAR oNE FORM TITLE: FORM NO. REV. 10CFR50.80 DETERMINATION 1000.131A 3 PC.1 Page 1 of.) ' Document No. P&lD M-2204 sheet 1 Rev1 Change No. 39 Title Condensata and F;e?;;;; : Pinina and instrr:.t Dissisin i 1 Brief description of proposed change: Modify indicated valve oositions to reflect positions at full cower Will the proposed Activity:
- 1. Require a change to the Operating License including:
l Technical Specifications (excluding the bases)? YesO No@ Operating License? YesO No@ Confinnatory Ordens? YesO NoS
- 2. Result in information in the following SAR documents (including drawings and text) being (a) no longer true or accurate, or (b) violate a requirement stated in the document:
SAR (multi-volume set for each unit)? Yes@ NoO Core Operating Limits Report YesO No@ Fire Hazards Analysis? YesO No@ Bases of the Technical Specifications? YesO No@ Technical Requirements Manual? YesO No@ NRC Safety Evaluation Reports? YesO No@
- 3. Involve a test or experiment not described in the SAR?
YesO No@ (See Attachment 2 for guidance)
- 4. Result in a potential impact to the environment? (Complete the Environmentalimpact Determination of this form.) YesO No2
- 5. Result in the need for a Radiological Safety Evaluation per section 6.1.5?
YesO No@
- 6. Result in any potential impact to the equipment or facilities utilized for Ventilated Storage Cask activities per Section 6.1.67 YesO No2
- 7. Involve a change under 10CFR50.54 for the following SAR documents per Section 6.1.7:
QAMO7 YesO No@ E-Plan? YesO NoS
ARKANSAS NUCLEAR ONE FORM TITLE: FORM NO. REV. 10CFRSS.80 oETERMalATIoM 1000.131A 3 Pc.1.2 Page 2 of,1 i Document No. P&lD M-2204 theet 1 Rev> Change No. jg Basis for Determination W-- ^ir- .s 1. 2 & Sh Chanoina the ir***d valve ananians 'on M-2204 sheet i for 2CA-7. 2CA-9. 2CA 19 and 2CA
=-a and cher.s:Te the ;r t- ^ 1 =-T=i for 2CV.nna? and 2CV.nna3 ficm Gsen to ch==: will raanire the same chs.=== to be ire = on BAR F'aa u 10.4 2. Tt-- v=lvc:; are NOT ssfe's rd-* -1 and.have no imred on the Unit 2 Grese".;ie i' = nee. Tech Ar m or i:r=nse p--'- decumen's citer than ch&aG;na the indicsted valve iiiiniijigs on SAR Floura 10.4-2. Tha== cher- s will make the SAR fioure serse witti the system alicament as dagcP=-1 in SAR ==2 -as 10.3.5 and 10 A.6 O Proposed change does not requite 10 CFR 50.59 Evaluation per Attachment 1, item,(if# checked, note appropriate item #, send LDCR to Licensing).
~
~__ . _ __,_._ __._..
Search Scope: List sections reviewed in the Licensing Basis Documents specified in Question 1,2 and 3. If a search was performed on LRS, the LRS search index should be entered under "Section" with the search statement (s) u parentheses. Controlled hard copies of the documents shall be reviewed (LRS is not verified and searches text, not figures or drawings). Attach and distribute a completed LDCR per Section 6.1.2 If LSD changes arei required. ' I Document Section LRS: 80.69 Unit 2: hydrazine, amine, recirculation valve, condensate flow, condensate pump J MANUAL SECTIONS: Section 10 FIGURES: i Figure 10.4-2 Ib[O Certified Reviewdr's SigtfaturF Thomas K. Mosby 11/23/98 Printed Name Date Reviewers certification expiration date: January 10.1999 Assistance provided by: Printed Name Scope of Assistance Michael Prock Date Reviewed Environmentalimpact Determination 11/24/98 Search Scope Review Acceptability (NA, if performed by Technical Review per 1000.006) h Ce
- _W D .At-v' & ~ A. W /2 ~ ME Revfs Sign'sture /
Printed Name Date
~
amansas muci. man our FORM MLE: FORM NO. REV. 10cFRse.se DETERMINATION 1998.131A 3 Page 1 of.1 , ENVIRONMENTAL IMPACT DETERMINATION (UNIT 1 and UNIT 2) , Document No. M-L3M Rev) Change No, B Complete the following Determination, if the answer to any checklist item is "Yes", an Environmental Evaluation is required. See Section 6.1.4 for additional guidance. . Willthe Activity being evaluated: Yes No O E Disturb land that is beyond that initisily distuttied during construction (i.e., new construction of buildings, creation or removal of ponds, or other terrestrial impact)? See Uni'. 2 SAR Figure 2.5-17. This app!ies only to areas outside the protected area. O E Increase thermai discharges to iake or atmosphere? O E increase concentration of chemicais to cooiing take or atmosphere through discharge canal or tower? O E increase quantity of chemicals to cooling lake or atmosphere through discharge canal or tower? . O E Modify the design or operation of cooling tower which will change drift characteristics? O E install any new transmission lines leading offsite? O E Change the design or operation of the intake or discharge structures? t l O E Discharges any chemicais new or different from that previously discharged? O E Potentially cause a spill or unevaluated discharge which may effect neighboring soils, surface l water or ground water? O E involve burying or placement of any solid wastes in the site area which may effect runoff, I surface water or ground water? O E Involve incineration or disposal of any potentially hazardous materials on the ANO site? O E Result in a change to nonradiological effluents or licensed reactor power level? O E Potentially change the type or increase the amount of non-radiological air emissions from the ANO site.
ARKANSAS NucuEAR ONE FORM TITut FORM NO. GElf. 10CFR80.89 EVAL.UATION 1000.131s 3 PC.3 Page 4 of,1 10CFRSO.59 Eval. No. FA) % oa 5 (Assigned by PSC) Document No. M-2204 sheet i Rev> Change No. 19 Title Conder-* :nd F::f:;2^ r :";e:n; cr.d !ne;.c.T.;d Di;w:ss A WRITTEN RESPONSE PROVIDING THE BASIS FOR THE ANSWER TO EACH Q ATTACHED. EACH QUESTION MUST BE ANSWERED SEPARATELY. A SIMPLE STAT CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES GUIDANCE FOR RE If the answer to any question on this form is "Yes," then an unreviewed safety question is involved. If th to all questions is "No," then the proposed chenge does not involve an unreviewed safety question.
- 1. Will the probability of an accident previously evaluated in the SAR be increased?
Yes O No E Section cer '- 10.4.8 of the Unit 2 MA_R descr!t-;= the process of iniect;ns hvdi zine end amine into tha hvd.4..e andsystem at the d!?rhers of the esaden5*** ournee. This is accomE!!E'.ed by the
;c.:i.; ch_.--::
21 vih;ch am neer.;:'s es.n v.t. E! iniection
' nu.T.ss d!schars;ns throueh 2CAJ. 2CA-9. 2CA-19 and 2CA-s at full power. P&lD M 2204 Rev. 90 shcw; the va =; as closed.
Ch;r ' .; the :n "= ^ ' nae"!en of th=; ass-safety rel;^;d val=; will m=ke the P&lD and SAR
^
fleu o match the ^ sci'-- ;ss of sv.^-.7. csei.ison. In add;; ion. the Ccade.= 9 oc.T.s recire valves 2CV4SS2 :nd 2CV ^"*3 are shewr, as -:-n=a when thev should be shcwr. as clGsed at full scw;i. The rec;ic v;t;;; fall to the c e;;d position. Ther. fore esir.ctino the indie ^rd r sosit'sn for these valves will be in sces-dance with assisved siscedures and SAR infcrinet'Ga and will not ine the probability of an acci:nt Lieviousiv eve!usted by the SAR.
- 2. Will the consequences of an accident previously evaluated in the SAR be increased?
Yes O No C The chemical inime*len lines are rpeda of 3/8" stainiggs steel tubino. If any of these lines w; e to break off downstream of the n e-:^;sa !=s'E!;on valves. two situ =*isas would arise. First. an un!=- -M leak fi-.i. the associated conder.; ate l't: 'er would be e ;;;rt. "cw;ver. the ;;n.; are ao La '" the consecuen- +E waas,e be hassded by the sievisusiv sve uated esader.ests"=dw ";.- line break in the Unit 2 MA_M. Secsad. hev'as the che E'-re! iniectisn v; x; csen ..^;si than ciceed waa'd reaeM in a ct.62!-:E! ==!!" to the tuit;..; bt!!d:..a sumo. The chein!-:E! Es!!" could be te..r. 7 -^-d by r!=E; .; the i=ei=2 95 ve ve or secGi;ns the e==9-:::^;d nG.T.s. This scensiis is be:sw the threeh=?d of =r=I': 4. eve:aated in the BAR and dses not h:ve any safety s'snificance wlth reserd to a cendes- ^ ." :d;._^_;line br=9_ The position of the coredensate ee- s r. circ v=:=; dses not have any s 7 ;s sionificance whetter the ;;t;= are er: . or closed. The =:=; cee!d be scened at full sc;;r if needed. The rss=s*=d J._ ;;; ct.;nse is ;.J-nded to ..Tsct that the v; =; are normally c eeed at full scwn the ===2e Therefsie. en of the =:;;; h;ve no ::nna-t on any of the ccasesuances of acciderae described in the AAB. 3. Will the probability of a malfunction of equipment important to safety be increased? Yes O No @ o
ARKANSAS NUCLEAR ONE FORM TITLE: FORM NO. RQf. 10CFRSS.80 EVALUATION 1000.1313 3 PC-2 Neae of the -:r .sen...t. listed sh are safety related nor due thev aussc-Et any souisraent
*c :-Etest to safety. Therefore. chene:a; the sesit ea of th;;; va == will not incr=ame the erEM""; of safety aa %nt inalfunction.
- 4. Will the consequences of a malfunctior? of equipment important to safety be increased?
Yes No G None of sthe components !!ed ebeve are safety related nor due th-v susssit any eauls, neat important to safety Therefore. chanoina the aae8*!on of th==c val=; v/ill not increame the consecuenceE of a malfune*!on of aae!ement :.; :-eit nt to - 'e's
- 5. Will the possibility of an accident of a different type than any previously evaluated in the SAR be created? Yes No 5
&,ggad;.---^-T;;f;__^__.- line break has been orova ees!v =v=!uated in the Unit 2 *A*. This would bound the ce..;;;aane== of condensate leak at a che.T' =! la:ect:sa line breek at the csadea==*=
header. The chen-!cs! se!!! which would eccur until eee.-ter ec4c-a =; ^="=a to stas the ee!!! is be:ew the threshe:d of the Unit 2 *A*. Aaamorlate actions for the ch;.TJce! sc-!!! would be dliscted fie.a the Unit 2 Cest..: Room in accordance with siecedure 1052.030 Scill Cea;.el and Countermeasure Plan.
- 6. Will the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR be created? Yes No @
None of the components !!=**d above are safety related nor due they suonort any eauipment important to safety. Therefore. chanoina the ind!cr^ d position of tt :: v= =; will not cause a malfunction of eauloment important to safety in a manner which has not h;;a sieviously evaluated in the SAR.
- 7. Will the margin of safety as defined in the basis for any technical specification be reduced?
Yes No 5 i None of the components I!=*ad ebeve are related to technical specif!-:=*!ons nor due thev succort any sou:e...ent related to techn!ce! specifications. Thereiere. the marain of s !;;< as des;aed in the basis for tect.a ;al e==cifie=*lons is unaffected. Thomas K. Mesby 11/23/98 /tertified Reviebrs Spl(pnp' Printed Name Date Reviewers certification expiration date: January 10.1999 Assistance provided by: Printed Name Scope of Assistance Date PSC review by: Ar% - Date: 3 @
ARKANSAS NUCLEAR ONE FORM TITLE:
- FORM NO. REV.
1eCFR80.88 DETERMINATION 1eet.131A 3 PC 1 Pagel of G Document No. P&lD M-2239 sheet i Rev1 Change No. It Title Nitronen Addition System Ploina and Instrument Diasrein Erlef description of proposed change: Modify indicated valve position to reflect position at full power Willthe proposed Activity:
- 1. Require a change to the Operating License including:
Technical Specifications (excluding the bases)? YesO No@ Operating License? YesO No@ Confirmatory Orders? YesO No@ 2. Result in information in the following SAR documents (including drawings and text) being (a) no longer true or accurate, or (b) violate a requirement stated in the document: SAR (multi-volume set for each unit)? Yes@ NoO l Core Operating Limits Report I YesO No@ Fire Hazards Analysis? YesO No@ Bases of the Technical Specifications? YesO NoS Technical Requirements Manual? YesO No@ NRC Safety Evaluation Reports? YesO No@
- 3. Involve a test or experiment not described in the SAR?
(See Attachment 2 for guidance) YesO No@
- 4. Result in a potential impact to the environment? (Complete the Environmental Impact Determinatior, of this form.)
YesO No@
- 5. Result in the need for a Radiological Safety Evaluation per section 6.1.57 YesO No@
- 6. Result in any potential impact to the equipment or facilities utilized for Ventilated Storage Cask activities per Section 6.1.67 YesO No@
- 7. Involve a change under 10CFR50.54 for the following SAR documents per Section 6.1.7: .
1 QAMO? YesO No@ E-Plan? YesO No@ l
amaransene Nuct.aAR ofE renM TrTLE:- FORM NO. REV. 10CMt00Je DETEnMelATIoN 1000.131A 3 PC-1,2 Page 2 of,1 Document No. P&lD M-2239 sheet 1 Rev> Change No. ji
,,,a ,,, p -- .~._ _ y _ _..:- .e 1. 2 & 31:
Che.x;..a the . 'i- -4 v=!va = 'k-is on M-2239 sheet 1 for 2CV4213-2 from ==n to car =1 will renuire th
;;s..e ct.e.- -
to cGate: ..T.en' = to:=2=^'= be re=t on SAR Fiaa o 3.2 5 and m__AR T=M= 6.2-2d.2CV4213-2 is the low cr- , e a;irssem v:!v: and T -Ed.^:: a '-- -' -! If a Safety laed n Adna'ian Sicr.&; (S!AS) or Cop *=L Ten; =- Ada=2Lr Sisas: i "1-" _Adr^Lr Sv;M.T. tKA8AS) to Liev;de ceaie:. T. eat fret the Unit 2 orne_=-m
'r2=2'=i Eac r.;;.ed Safsn=d.
i.(OLA8) Famences is Gener 2CV4213-2 is asiri; :P. ' ==_' wt.;;s at full cower that is saGis conservative. fis,T a safetv Ger- =-3_!ve. itEn sLerilins w iti the valve in the seen r. O Proposed change does not require 10 CFR 50.59 Evaluation per Attachment ,(if checked.1, item #_ note appropriate item #, send LDCR to Licensing). Search Scope: List sections reviewed in the Licensing Basis Documents specified in Question 1,2 and 3. If a search was performed on LRS, the LRS secrch index should be entered under *Section" .dth the search stateme parentheses. Controlled hard copies of the documents shail be reviewed (LRS is not verified and s text, required. not figures or drawings). Attach and distribute a completed LDCR per Section 6.1.2 If L8D cha Document Section LRS: 80.89 - Unit 2: nitrogen, nitrogen isolation, containment isolation valve, 2CV4213 2, nitrogen blanket, layup MANUAL SECTIONS: Section 10, Section 6, Section 15 FIGURES: Figure 3.24, Table 4.2 26
/Iwe -[N w The-- K. Ma=hy 12/4/98 / Certified Reviewdr's Sign 6 tug Printed Name Date Reviewer's certification expiration date: January 10.1999 Assistance provided by:
Printed Name Scope cf Assistance I Date Sea Sco Review Acceptability (NA, if performed by Technical Review per 1000.006)
/N JNrtified'RevWignature u A %~ ih/n Printed Narde Date
r
~ aawanasna Nuct2.AR osE FORM TITLE
- FORM NO. REv.
10CFRet.88 oETaRMSGATIoN gees.131A 3 Page 2 off ENVIRONMENTAL IMPACT DETERMINATION l (UNIT 1 and UNIT 2) Document No. M-2239 sheet 1 Rev./ Change No ji
- Complete the following Determination. If the answer to any checklist item is "Yes", an Environmental Evaluation ' is required. See Section 6.1.4 for additional guidance. . Will the Activity being evaluated:
1res a m O E Disturt> innd that is beyond that initially disturbed during construction (i.e., new construction of buildings, creation or removal of ponds, or other terrestrial impact)? See Unit 2 SAR Figure 2.5-17. This applies only to areas outside the protected area. O E increase thermal discharges to take or atmosphere? O E- incr==se concentration of chemicais to cooling lake or atmosphere through discharge canal or tower? O @ increase quantity of chemicals to cooling take or atmosphere through discharge canal or tower? . O E Modify the design or operation of cooling tower which will change drift characteristics? O E instali any new transmission lines leading offsite? O E Change the design or operation of the intake or discharge structures? i O E Discharges any chemicals new or different from that previously discharged? O E Potentially cause a spill or unevaluated discharge which may effect neighboring soils, surface water or ground water? O 2 involve burying or piacement of any solid wastes in the site area which may effect runoff, surface water or ground water? i ' O E involve incineration or disposal of any potentialir hazardous materials on the ANO site? O E Result in a change to nonradiological effluants or licensed reactor power level?
.O @ Potentially change the type or increase the amount of non-radiological air emissions from the ANO site.-
m . ; ARKANSAS NUCLEAR ONE FORM TITLE: FORM NO. REV. 10CPR$0.80 EVALUATION 1000.1313 3 PC4 Pagef of,1 10CFR50.59 Eval. No. FFA).q 4.ogy (Assigned by PSC) ! Document No. M-2239 sheet 1 Rev> Change No. H I Title _. Mitronen Addition Svstem Plaina and Instru.T. eat Di;w:as A WRITTEN RESPONSE PROVIDING THE BASIS FOR THE ANSWER TO EACH QUES ATTACHED. EACH QUESTION MUST BE ANSWERED SEPARATELY. A SIMPLE STATE
- CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES GUIDANCE FOR RESP If the answer to any question on this fonn is "Yes," then an unreviewed safety question is involved, if the an to all questions is "No," then the proposed change does not involve an unreviewed safety question. '
- 1. Will the probability of an accident previously evaluated in the SAR be :
increased? Yes O No @ Containment feew e=*** ;av;ses..T.;aten in v;:x; ensure the ;;ent of athat r: the esatslr..T. eat .^...csches will be !ee'e+ad freg],jlg M E of rd!c diva material to the cGat. an.ent e...ssst. sis or areaEsA 2ss of the cen^ &: .... eat. 2CV4213-2 race:v;; a class sianal fraai been an JdAS and O!?*. " ' .^ :a:as 2CV4213-2 in the c: seed position adscss the arsbsbllltv of a G"res freia the cea's 7.. .;at s;nce the vstes wee!d not have to no cic-ssd on a SIAS or C!A* sianal. Themisre. cerisct:as the r-di-~^ d E-:=";ss for this valve is in accordance with assieved procedure 2106.008. Steam C;r.emier CzereGsas and will not incraeme the sic-t-eb!!ity of an accident sievisusiv pvaluated by the SAR.
- 2. Will the consequences or en accident previously evaluated in the SAR be increased?
Yes O No @ The consecuences renardino a failure of contsia. Tant intesiity would be unchanced since the accident an;Pisis assu.T.e; csate:n. Teat isolation v=iv;; to be clssed. Thersisis. chanoina the ind;-- ^ d se:1;ea of it-;:-e ve:=: to closed will have no irneed on any of the conseausacss of acciden^e desci;;,sd in the SAR.
- 3. Will the probability of a malfunction of equipment important to safety be increased?
Yes O No @ Maintainina 2CV4213 2 in the closed eeshion whlie at full sc;;r rsdur== the orebebility of a
===un
--.. of ===t..ent irnee tant to ;;;;i fi.e. them is '=== chance of 2CV4213-2 f; llas to cissa since it ;;;eede ric :dt The efes. chans:as the indicated position of this v=ivs to cis;;d will not increase the orchabilltv of s f;;< eauioment malfunction.
- 4. Will the consequences of a malfunction of equipment important to safety be increased?
Yes No @ if 2CV4213-2 malfunctier&d while at**motina to close the valve. as it is normally alisasd. actions ce"!d be initiated in scesideace with the Unit 2 Tech Saacs. If the valve malfunctissed while closina from an ESFAS sisaal. the ceae.susacss coe!d be viciss. Treisfore. chanslas the indicated seE:Gea of this v;te; to closed will not increena the ceasesuances of a malfunction of ee d - .;at important to enfatv
ARKANGAs NUCLEAR ONE FORM TITLE: ; FORM NO. REV. 14CFRSS.84 EVALUATION 1e00.1313 3 PC-2 1 5. Will the possibility of an accident of a different type than any previously evaluated in the SAR be created? Yes No 2 The notential for a laeE oath from the conti.:na. eat to the savlrsa.nea; thrt-Joh systems which sen in-i the cent;li.s.es; walls has been eve!"-^ed in the BAR. Therefere. channina the indicated Essition for 2CV4213-2 to c c;;d will not cr ^r a new tvse of accident 2lch has not b;;a evaluated in the SAR, 6. Will the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR be created? Yes No ti 2CV4213-2 was desica d to fall cle==. The valve sce:v;; ESFAS slsseis to c!ss=. The valve is normally kept cic;dd elle at full sc;;;.. If the v;lr. malfunctisas. e"!dence is sisvids-d in the Unit 2 Tech Snecs for =^ic-se to be tskea in the event a csnte: La.;a !=e8=*'en v; ve becsroes inoperable. Nothina is beina done to char'oe the physical operational char =e*=ri=*!cs of 2CV4213-2. Therefore. chanoina the indicatejutg,g[fion of 2CV4213-2 to closed will not ce"a= a malfunction of eauioment important to safety in a mer.r ,r which has not t;;a sievisusiv eve >:^ d in the SAR.
- 7. Will the margin of safety as defined in the basis for any technical specification be reduced?
Yes No G l Maintainino 2CV4213-2 in the closed aasl* ion desrlas full scre;r osei.;;sa maintains or imoroves the mars ln of safety relative to the 10CFR100 llm.l s d!scussed in the t- :: for Tech Spec 3.8.1.1. Since the valve is nstraally c c;;d. andssm*!c act'sas l6lt:sted by a SlAS or CIAS to close the valve are not he'r.s r;lled unsa. Thereisie. chanoina the irad!r^:d oosition of 2CV4213-2 to clceed will not reduce the marain of seis;< as der.Ged in the bas:s for technical specifications. F eI Thomas K. Mosby 1,.2/43) Certified Revieners Sl6 nap Printed Name Date Reviewers certification expiration date: January 10.1999 Assistance provided by: Printed Name Scope of Assistance Date I P5C review by: Wu Date: 35
AHMANSAS NUCLEAR ONE FORM TITLE:
- FORM NO. REV.
14CFR84.40 DETERMINATION it00.131A 3 PC 1 Page1off Document No. P&lD M-2202 sheet 2 Rev> Change No. Af Title Main Steam Pinino and Instru.T.e.7; Disoram Brief description of proposed change: Modify indicated valve oositions to reflect oositions at full oc;;;.- Will the proposed Activity:
- 1. Require a change to the Operating License including:
Technical Specifications (excluding the bases)? YesO No@ Operating License? YesO NoS Confinnatory Orders? YesO No@ 2. Result in information in the following SAR documents (including drawings and text) being (a) no longer true or accurate, or (b) violate a requirement stated in the document: SAR (multi-volume set for each unit)? Yes@ nod Core Operating Limits Report ' YesO NoS Fire Hazards Analysis? YesO No@ Bases of the Technical Specifications? YesO No@ Technical Requirements Manual? YesO No@ NRC Safety Evaluation Reports? YesO NoS
- 3. Involve a test or experiment not described in the SAR?
(See Attachment 2 for guidance) YesO NoS
- 4. Result in a potential impact to the environment? (Complete the EnvironmentalImpact Determination of this form.)
YesO No@
- 5. Result in the need for a Radiological Safety Evaluation per section 6.1.57 YesO No@
1
- 6. Result in any potentialimpact to the equipment or facilities utilized for Ventilated Storage Cask activities per Section 6.1.6?
YesO No@ }' 7. Involve a change under 10CFR50.54 for the following SAR documents per Section 6.1.7: QAMO? YesO No@ E-Plan? , YesO No@
Amuussas NuCt. EAR ONE FORM TITI.E FORM NO. REV. 10CFpt00.80 DETWtMINATION 1908.131A 3 PC-1.2 Page 2 of_E Document No. P&lD M-1202 sheet 2 Rev> Change No. 11 Baala for Determirwien (Oceusas 1. 2 & 31: c Charoina the Irwne=*M valve na*81ons on M-2202 sheet 2 for 2CV-0314 and 2 r= awe the same char +E to be m=da on SAR Floure 10.2-4. These valves are NOT safeiv related and have no imnar* on the Unit 2 Cesistina Ucessa. Tech Stea or Licanse Baele documents irt.;a:ad valve E= tons on SAR Flours 10.2-4. The alianment of these valves is not refsrsaced in any Licensa n==1= Docurr.en^a O Proposed change does not require 10 CFR 50.59 Evaluation per Attachment ,1,(ifitem # checked, note appropriate item #, send LDCR to Licensing). Search Scope: List sections reviewed in the Licensing Basis Documents specified in Quostion 1,2 and 3. If a sear performed on LRS, the LRS search index should be entered under "Section" with the search s parentheses. Controlled hard copies of the documents shall be reviewed (LRS is not verified an text, not figures or drawings).' Attach and distribute a completed LDCR per Section 6.1.2 If LB required. Document Section LRS: 50.89 - Unit 2: drain valve, steam drain, feedwater pump turbine MANUAL SECTIONS: Section 10 l FIGURES: Figure 10.2-4 fb[*O / Certified ReflowersfSig9dfGre Thomas K. Mosby 12/3/98 Printed Name Date Reviewers certification expiration date: January 10.1999 Assistance provided by: Printed Name Scope of Assistance 1 Date Searc Sco Review Acceptability (NA, if performed by Technical Review per 1000.006)
%ified'Reviewersbure M b Y NAlt)eu '
l~ lN Printed Name/ Date i 4
meauwe=Ah MMk.AM ONE FORM TITLE: FORM NO. MEV. 10cFnes.Se OETERMINATIoN l 100s.131A 3 Page 1 of.) ENVIRONMENTAL IMPACT DETERMINATION (UNIT 1 and UNIT 2) Document No. M-2202 sheet 2 Rev> Change No, ji Complete the following Determination. If the answer to any checklist item is "Yes", an Environmental Evaluation is required. See Section 6.1.4 for additional guidance. Will the Activity being evaluated: 1res mo O 2 Disturb land that is beyond that initially disturbed during construction (i.e., new construction of buildings, creation or removal of ponds, or other terrestrialimpact)? See Unit 2 SAR Figure
- 2.517. This applies only to areas outside the protected area.-
O E increase thermai discharges to lake or atmosphere? O 3 increase concentration of chemicals to cooling lake or atmosphere through discharge canal or tower? O' 2 increase quantity of chemicals to cooling lake or atmosphere through discharge canal or tower? , O E Modify the design or operation of cooling tower which will change drift characteristics? O E instati any new transmission lines leading offsite? O E change the design or operation of the intake or discharge structures? O E Discharges any chemicals new or different from that previously discharged? ! O .S > Potentially cause a spill or unevaluated discharge which may effect neighboring soils, surface water or ground water? l t j O 2 involve burying or placement of any solid wastes in the site area which may effect runofr. surface water or ground water? ! I O E involve incineration or disposal of any potentially hazardous materials on the ANO site? O E Result in a change to nonradiological effluents or licensed reactor power level? O_ E Potentially change the type or increase the amount of non-radiological air emissions from the l ANO site. I
ARKANSAS NUCLEAR ODE FORM TITLE: FORM NO. REV. 10CMtSS.80 EVALUATION 1000.131] 3 PC.2 Page 4 of.) 10CFR50.59 Eval. No. 'M N (Assigned by PSC) Document No. M-2202 sheet 2 Rev1 Change No. H Title t'.= n r - Pinina and la.i..mnent Dr -:..a A WRITTEN RESPONSE PROVIDING THE BASIS FOR THE ANSWER TO EACH QUES ATTACHED. EACH QUESTION MUST BE ANSWERED SEPARATELY. A SIMPLE STAT CONCLUSION IS NOT SUFFICIENT, ATTACHMENT 2 PROVIDES GUIDANCE FOR RESPONSE. If the answer to any question on this form is "Yes," then an unreviewed safety question is involved. If the an to all questions is "No." then the proposed change does not involve an unreviewed safety question.
, 1. Will the probability of an accident previously evaluated in the SAR be increased?
Yes O No @ if either 2CV.0314 or 2CV-0317 were -::: ; teh' e at fulf ;;;;;; *8-!n Steam we .mid have a 4..ct ssth to the main cerd: .;e7 tr. sees %" alc-E. lOJ:r.e the vet;;; c Grad isolates the csadeneer from lWa:n r tr+ Steam t'.er.ta MErins the sisF=E"!rs of an aarr eat main -*-- t line leak throuah T'.;;efer. cerisct:_.s the indicated ==sMien for it-;:-; v;: ;; will be in acco dance with
===reved Liscadeo 2100.007. ""isin :"eed; _:_: Pumo and FWC5 Oc-Eisi;sas. and will not incr=ame the probability of an sce:d r.: LieJ'seelv eveh-"d by the SAR.
- 2. Will the consequences of an accident previously evaluated in the SAR be increased?
Yes No @ The nositicri of the hich or===u o **ee valve below seat 4..in v;!v;; is bounsd by the accidents included Irm the BAR for the "r'r. Steam System '"J; tt.er the v;:v;; are sceaed or cic;;d. the small cine tr . would limit stessi ":ew to cr.te a small fraction of the des!en ba=!s -*-- !i leak evaluated I;t the Unit 2 8.*R. There;ere. ch;aslas the ir. dice;ed position of Pese valves to closed will have h;;e no impact on any of the ccr.;escer. css of accidenis described in the SAP.3 3. Will the probability of a malfunction of equipment important to safety be increased? Yes O No @ None of the cos=Mnents n.*M =5-:-we ars =-!_;; i=l=^-d nor due thev sesssit any eau;Lir.ent la +:-is.r. to safetv Tr re ae. ct:ss:..s the :nd!=nd ssEE;cs ofit.s; ;;t;;; will not incF eee the eier i;as of safety =ar'n a: r sacties. 4. Will the consequences of a malfunction of equipment important to safety be increased? Yes No @ None of the commonents listed ebeve are s ;e;< i;;;;ed nor due they susasst any sou;sireat
;ir.ee; test to safstv T'.ere;er.. ct.;.. :as the nd;r-Md position ofit;:-; v; v;; will not incr Err the cer.:::::nces of a r. "unc^ c-a of =ar'- i.ea: in restant to safety.
\
ARKANSAS NUCLEAR ONE FORM TITLE: FORM NO. REV. itCFR54.68 EVALUATION 1000.1313 3 PC-2 5. Will the possibility of an accident of a different type than any previously evaluated in the SAR be created? , Yes O No @ A main -^- = line break l E,Wn oreviousiv eve!e**ed in the Unit 2 SAR. This is the only event with any safety sianificance vehlch coe!d be ret:ted to the valve: lis'_ad Et-sys. Thereisis. chanoino l i the ind! =^ ' MeM!en to c -:::d will not crs=tc a r.;.;tvse of accidea; not sysle ":d in the SAR.
- 6. I Will the possibility of a malfunction of equipment important to safety of a '
different type than any previously evaluated in the SAR be created? Yes No C None of the comeenents !!-+-d ebeve are ;;;;< rel:ted nor due they suneort any souisment imssitant to s;;.;<. Therefsss. ch;as no the ind! td nosition of th;n valves will not cause a malfunction of eauis.r..at imscitsat to s.ie;v in a manner which has not L;;a sievicusiv evaluated in the SAR.
- 7. Will the margin of safety as defined in the basis for any technical specification be reduced?
Yes O No @ None of the components listed above are related to technical specifications nor due they suonort any enulement related to technical specificatiens. Therefore. the maroin of ssieiv as defined in the basis for technical specifications is unell.cted. C I N Thomas K. Mosby 12/4/98 1 / ertified RevieWfers Signaturg7 Printed Name Date Reviewers certification expiration date: January 10.1999 Assistance provided by:
- Printed Name Scope of Assistance Date PSC review by: h& Date: D *1b l
i
~
X Li$una FORM TIRE:
, FORM NO. REV.
teCPRse.as DETERMINATION 1000.131A 3 PC 1.2 Page1 of.! Document No. ER963555E204 Rev/ Change No. A Title ANO-2 Baa"--._ ;; :'2;. Missile Shlald B!aak Rs.ws: Brief description of proposed change: Removal of r==e+er huisd!na ee=" T.er.: b=ti elWM shi:3 blocks Willthe proposed Activity: 1. Require a change to the Operating Ucense including: Technical Specifications (excluding the bases)? YesO No@ Operating Ucense? YesO NoS Confirmatory Orders? YesO NoS 2. Result in information in the following SAR documents (including drawings and text) being (a) no longer true or accurate, or (b) violate a requirement stated in the document: SAR (multi-volume set for each unit)? YesO No@ Core Operating Umits Report YesO No2 Fire Hazards Analysis? YesO NoS
' Bases of the Technical Specifications?
YesO NoS Technical Requirements Manual? YesO No@ NRC Safety Evaluation Reports? YesO No@ 3. Involve a test or experiment not described in the SAR7 (See Attachment 2 for guidance) YesO No@ 4 Result in a potential impact to the environment? (Complete the Environmental Impact Determination of this form.) . YesO No@ 5. Result in the need for a Radiological Safety Evaluation per section 6.1.57 YesO No@ 6. Result in any potential impact to the equipment or facilities utilized for Ventilated Storage Cask activities per Section 6.1.67 YesO No@ 7. Involve a change under 10CFR50.54 for the following SAR documents per Section 6.1.7: QAMO? YesO No@ E-Plan? YesO No@ Basis for Determin=*3en l'h'::^;c-r.s 1. 2 & 31: See attached 1000.131C form. O Proposed change does not require 10 CFR 50.59 Evaluation per Attachment , (if checked. 1, item # note appropriate item #, send LDCR to Ucensing). i FW
FORM TITLE. AsuumaAs-astone FORM NO. REV. 10CFRES.00 DETER 8ANATION 1000.131A 3 PC-12 Page 2 of,3 Document No. ER963555E204 Rev> Change No. 9 Search scope: List sections reviewed in the Licensing Basis Documents specified in Question 1,2 and 3 performed on LRS, the LRS search index should be entered under *Sedion" with the parentheses. Controlled hard copies of the documents shall be reviewed (LRS is text, not required. figures or drawings). Attach and distribute a completed LDCR per Section Document LRS: Section 50.59 Unit 2 all (" equipment hatch", [ tornado and missile], " reactor building integrity", " containment integrity", " missile shield", ) MANUAL SECTIONS: Unit 2 TS 1.8, 3/4.6.1, 3.6.1.2, Table 1.1 Unit 2 SAR 3.8.1, 3.3.2, Criterion 4, Table 14.1 1 (Pg.14.345) FIGURES: Unit 2 SAR 3.5 3 thru 3.5-8A DD
,mae .O t Certiflfd llteviewers Signalbfb Herbert R. Rideout , 11/03/98 l Printed Name Date }
}
Reviewers certification expiration date: 02-11 2000 1 Assistance provided by: { Printed Name Scope of Assistance l Date l
}
l Search Scope Review Acceptability (NA, if performed by Technical Review per 1000 0 . Ne" _ boyce G. Asnm s Certified Geviewers Signature Printed Name
///3l98
' \
Date
FORM TITLE: FORM NO. REV. 10CHISS.59 DETERMINATION 1000.131A 3 PC4.2 Page 2 of 4 ENVIRONMENTAL IMPACT DETERMINATION (UNIT 1 and UNIT 2) Document No. ER963555E204 Rev1 Change No. 9 Complete the following Determination. If the answer to any checklist item is "Yes", an En is required. See Section 6.1.4 for additional guidance. Will the Activity being evaluated: Yes No O @ Disturb land that is beyond that initially disturbed during construction (i.e., new con buildings, creation or removal of ponds, or other terrestrial impact)? See Unit 2 2.5-17. This applies only to areas outside the protected area. O E increase thermal discharges to lake or atmosphere? O E increase tower? concentration of chemicais to cooling take or atmosphere through discha O E increase quantity of chemicals to cooling lake or atmosphere through discharge cana tower? O @ Modify the design or operation of cooling tower which will change drift characteristics? O @ install any new transmission lines leading offsite? D' @ Change the design or operation of the intake or discharge structures? O @ Discharges any chemicals new or different from that previously discharged? O -@ Potentially water cause or ground water?a spill or unevaluated discharge which may effect neighboring soi D. E Involve burying or placement of any solid wastes in the site area which may effec surface water or ground water? O E involve incineration or disposal of any potentially hazardous materials on the ANO site? O E Result in a change to nonradiological effluents or licensed reactor power level? O E Potentially change the type or increase the amount of non-radiological air emissions ANO site. l
JueoussAs NucLEARONE Ir0R M TITLE: FORM NO. REV. secense.se amnewcostnNWATIoM PAGE 1000.131c 3 Page f of,1
- Document No. ER963555E204 Rev/Ch'ange No. A 10CFR50.59 Review ContinWlon Pac-e which coffelstes to an RCS temperature greater than 200 d Criterion 4 requires structures important to safety to be designed to withstand the effects one of which is a tomado, without loss of capability to perform their safety functions. Also, ,
states that Seismic Category 1 structures have been analyzed for tomado driven missi structures to be appropriately proteded against dynamic effects, including missiles, that ma and conditions outside the nuclear power unit. Although not explicitly stated in the SAR, th containment equipment hatch missile shield blocks is to provide additional protection for will not occur as a result of a tomado missile impact.against tomado genera result from events and conditions caused by natural phenom , to any safety-related equipment necessary to safely place o Additionally, no tomado missile penetration will occur ossed on evaluations performed und . results indicate that the existing hatch will maintain containment integrtty. Therefore, the SAR result Mode 5. of removing the equipment hatch missile shield blocks subsequent to subcritica The following delineates compensatory measures to be taken if there are no tomadoes pr , the shield blocks in plarce. Should severe weather threaten shield blocks shall not be removed priorto reaching Mode 5, cold shutdown, if the equipm blocks are removed between Modes 3 and Mode 5, appropriate measures shall be tak t:mado missile hit the equipment hatch, that the hatch leak requirements are met prior to (startup). LCO requirements as identified in Section 3.6.1.1 of the Tech Specs would be ini Existing condition of the reactor, reaching Mode 5, will not be changed and does not impos not addressed. The requirements for an LOO are delineated in the Tech Specs and do not co requirement. Damage to the equipment hatch door due to tonadoes is a highly unlikely event. administrative controls are in place to determine if leaks in excess of those established by T If those limits cannot be obtained, then repairs to the hatch must be performed, or m&ures ensure containment integrity, prior to the containment being considered to meet its desis,. enction. Therefore, svaluation performed under ER963555E204. removal of the equipment hatch miss Procedure 2504.037 has been developed to ensure control of this effort meets the criteria set forth ir, the ' Engineering evaluation. The procedure provides the controls and guidance required f reinstallation of the ANO-2 reactor building missile shield blocks for the imposed conditions. Due to the nature of this request a 50.59 evaluation is also attached to this review.
FoRuTnu: FORM NO. REV.
. SOCPRe4Je EVALMA11oM 1000.131a 3 PC.2 Page 1 of,2 10CFR50.59 Eval. No blN M (Assigned by PSC)
Document No. ER963558E204 Rev/ Change No. 2 Title ANO.2 saa" 1.;..; Ha+eh Miae!Ie Shn.sd Bisck P.:- ;;;: A WRITTEN RESPONSE PROVIDING THE BASIS FOR THE ANSWER TO ATTACHED. EACH QUESTION MUST BE ANSWERED SEPARATELY. A SIMP CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES GUIDANCE If the answer to any question on this form is "Yes." then an unreviewed safety question is in to all questions is "No," then the proposed change does not involve an unreviewed safety que 1. Will the probability of an accident previously evaluated in the SAR be increased? Yes O No @ The current accident sieb=&"'^v ersaariss Edds==M in the SAR cor .ac s av** mal ;;es.^e. su tsiando missiles. The removal of the m'==' =F' 'I his"-- will not ch ;.se the sisS=S""i of an ace!dsat sccuri;..e W: in the SAR asest' ^ I w;;h this ;_...sval will ;;.T.;;r. the same.2 *eM8v on th.ir r;..-.s 2. Will the consequences of an accident previously evaluated in the SAR be increased? Yes O No @ Consecuences missile of anv shield b!st" r dui;..e the accid:s:s serisd ider.;;;edd icd. essed in the SM will not be incr-- 1 b All Eas!!==hE =rc'd:M Ersamf.ss dui;..e the seriod be:sw hot sha*dcwn will not be ch=ased nor will lr.;sar=iien ===sc!M - d w:t. tt::: scenarios be coneMrse u.a..se. 3. Will the probability of a malfunction of equipment important to safety be increased? Yes O No @ j Eoulement csais;i... r. malfune*3en is not i;r.sacted by the ie.T.sisi of the r!:E!b sh'; d hiscks.The i h*^':2: .s fun-^;ea durino this ser.sd will not be altered and it is still ====he of l Its :..;_ad:1 safety a"___d funct;sa. Ev=?=^1sa sie?!-t2 by E;E:Ma=204 Edd e=::: l 1.T==-M to the === ' 1.; .; t-22 :. due to a torr *ede .rJE=!h ;nd csar!P=E the; cente:a.. at will be me;a; ;aed sheaand =ech an ;;;at occur. "c.. ses. no safety systems ; s:_:..d to brino the unit to cs;d shand-:r .. are challenced by the n.T.sys: of the r!==2'
-- sh' 'd b ss/n. end the - ;--Eh!!sv to siever.;
exposure is unaffected. or mittaate the csaa.:sacts of oisai csad;1;sas that coe!d result in as^ese;;al eis.i 4. Will the consequences of a malfunction of equipment important to safety ' be increased? Yes No @ Consecuences of a malfun ^'sa of eauis.T.en: ';r.ssr: sat to safety will not be inci =wd as a re the seinaval of the .r'=E!h sh; id b!ack. t.e:sw hot shadec;;a. The =t'.:';w of the r=Edsr be!!d;as p.esisaa it des;ea funct;sa is u: '";ed by the sedv i;.T.sval and the sea ' T.;;; b= G is e==Ebh of
- ..;.;a;..s cea^e;n.T.;at li.; sdiv. The r== der be!;d; ass fun-:^isa duf.as th;s ser;sd is not ::; sed.
AfeEANsAs naar e oan'omE FORM Tm.E. FORM NO. REV. secens0JS EVALunTIoM 1eGe.1313 3 PC.2 Page 2 of.2 5. Will the possibility of an accident of a different type than any previously evaluated in the SAR be created? Yes O No @ No new accident or w - t4: "-- d...'...; the ^ mu*v of an scr* frat will be created des to the removal of the rai=-!M shield f time sested. The cui...t sct;-f: . eeer-:'^ I w th esteinsi mie-!h L,- ^_ -Men is the same se. ^^L:- .e Edd--En f in the 8LtAR and are unchanced as a r=seM of g.asvins the missile shield blocks. 6. Will the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR be created? Yet, No @ No - "ur 't a of = a
- T.e.t 'r nea^ to safety of a d: tere.t tvse than tite =- Edd. ==ed in the SAR will be created
===c- The fu.'-'*sa of the ceate:n...ent bd' dine is Eddre==e1 and ecci-fea's
'" I w;;L lt are ce.==M:_~ed to be the same condit;cas as siis;asily ide..i; sed in the SAR. No new accident will be created.
- 7. Will the margin of safety as defined in the basis for any technical specification be reduced?
Yes O No @ The marnins identified in techr4ce! specifie-'!sas for the csa^6;..... eat ht!! dias are ' der.;ii' sed for scecific sne"== dus;..s ner.na: sce Ei;c-as or for ser;sds wt.ea esat.;i ..en; c!eeers is resulted. The 7;.T.sve; of the missile shield t'eclis do no alter tr.;;e ccaditisas and will not result in it.;e; scanarias he;as huisde or ;sd2ced. _1 1
,, pN MA vN Herbert R. Rideout 11 03-98 Certified Reviewers Signature Printed Name Date ;
Reviewers certification expiration date: 02-11-2000 Assistance provided by: Printed Name Scope of Assistance Date PSC review by: '\.tN-- - Date:_ \\ \%
- 1 ARKANSAS NUCLEAR ONE Page 1 FORM TITLE:
FORM NO. REV. 10CFR50,59 DETERMINATION 1000.131 A 3 PC-1, 2 Document No. ER974550E201 Rev/ Change No. Title Brief description of proposed change: Downgrade of CVCS injection line components from 'Q* to "S" or "N" and change of QFUNC for those components that remain "Q" as part of a new #Q* boundary. Will the proposed Activity:
- 1. Require a change to the Operating Ucense including:
Technical Specifications (excluding the bases)? YesO No@ > Operating License? YesO NoS Confirmatory Orders? YesO No@ 2. Result in information in the following SAR documents (including drawings and tcxt) being (a) no longer true or accurate, or (b) violate a requirement stated in the document: SAR (multi-volume set for each unit)? Yes@ NoO Core Operating Limits Report? YesO No@ Fire Hazards Analysis? YesO No@ Bases of the Technical Specifications? YesO No@ Technical Requirements Manual? YesO No@ NRC Safety Evaluation Reports? YesO No@
- 3. Involve a test or experiment not described in the SAR?
(See Attachment 2 for guidance) YesO No@
- 4. Result in a potentialimpact to the environment? (Complete Environmental impact Determination of this form.)
YesO NoQ
- 5. Result in the need for a Radiological Safety Evaluation per section 6.1.57 YesO No@ l
- 6. Result in any potential impact to the equipment or facilities utilized for Ventilated Storage Cask activities per Section 6.1.67 YesO No@
- 7. Involve a change under 10CFR50.54 for the following SAR documents per Section 6.1.7?
QAMO? YesO No@ l O E Plan? YesO' No@ l 1
ARKANSAS NUCLEAR ONE FORM TITLE: Page 2 FORM NO. REV. 10CFR50s59 DETERMINATION 1000.131A 3 PC-1,2 Document No. _ ER974550E201 Rev/ Change No. Basis for Determination (Questions 1, 2, & 3): None of these documents discuss the "Q' designation of components except for the SAR experiments being proposed of any kind, previously described in the safety analysis report or nl 3.2-6 lists, by description, some of the components being downgraded from 'Q"in a list of j{e . ( O Proposed change does not require 10CFR50.59 Evaluation per Attachment (if checked, note1, item # appropriate item #, send LDCR to Licensing). Search Scope: 1 List sections reviewed in the Licensing Basis Documents specified in questions 1,2 and 3. If sea performed on LRS, the LRS search index should be entered under "Section" with the search parentheses. Controlled hard copies of the documents shall be reviewed (LRS is not verifie text, not figures ordrawings). Attach and distribute a completed LDCR per Section 6.1.2 If L required. Document Section LRS: MANUAL SECTIONS: ANO-2 SAR 669.3.4 and 3.13 FIGU S: d L. Daniel H. Williams 7/21/98 Certified Reviewers Signature Printed Name Date Reviewers certification expiration date: _11/18/99 Assistance provided by: Printed Name Scope of Assistance Date Search Scope Review Acceptability (NA, if performed by Technical Reviewer per 1000.006)
.Y Certified Reviewers Signature h,aJtL, Euh Sl17l49 Printed Name Date
ARKANSAS NUCLEAR ONE Page 3 FORM TITLE: FORM NO. REV. 10CFR50.59 DETERMINATION 1000.131A 3 ENVIRONMENTAL IMPACT DETERMINATION (UNIT 1 and UNIT 2) Document No. ER974550E201 RevjChange No. Complete the following Determination. If the answer to any item below is "Yes", an Environmental Evalua required. See Section 6.1.4 for additional guidance. Will the Activity being evaluated: Y.n.s Ng O E Disturb land that is beyond that initially disturbed during construction (i.e., new construction of buildings, creation or removal of ponds, or other terrestrial impact)? See Unit 2 SAR Figure 2.5-17. This applies only to areas outside the protected area. O E increase thermal discharges to lake or atmosphere? O E Increase concentration of chemicals to cooling lake or atmosphere through discharge canal or tower? O E increase quantity of chemicals to cooling lake or atmosphere through discharge canal or tower? 4 O E Modify the design or operation of cooling tower which will change drift characteristics? O E Install any new transmission lines leading offsite? O E Change the design or operation of the intake or discharge structures? O E Discharges any chemicais new or different from that previously discharged? O E Potentially cause a spill or unevaluated discharge which may effect neighboring soils, surface water or ground water? O E involve burying or placement of any solid wastes in the site area which may effect runoff, surface water or ground water? O E involve incineration or disposal of any potentially hazardous materials on the ANO site? O E Result in a change to nonradiological effluents or licensed reactor power level? O E Potentially change the type orincrease the amount of non-radiological air emissions from the ANO site.
ARKANSAS NUCLEAR oNE FORM TITLE: Pop 1 FORM NO. REV. 10CFR50.89 CAFETY EVALUATIEN 1000.1318 3 PC-2 Document No. erg 74550E201 RevJChange No. 10CFR50.5g Eval. No. FN-M-1N Title (Assigned by PSC) A WRITTEN RESPONSE PROVIDING THE BASIS FOR THE ANSWER TO EACH ATTACHED. EACH QUESTION MUST BE ANSWERED SEPARATELY. A SIMPLE ST CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES GUIDANCE FOR If the answer to any question on this fonn is "Yes," then an unreviewed safety question is involved. If to all questions is "No," then the proposed change does not involve an unreviewed safety question. 1. Will the probability of an accident previously evaluated in the SAR be increased? YesO No @ The charging pumps, pump motors and pump suction and discharge dampeners cannot initiate accidents. 2. Will the consequences of an accident previously evaluated in the SAR be increased? YesO No @ The charging pumps, pump motors and pump suction and discharge dampeners are not credited forthe mitigation of any accidents. 3. Willthe probability of a malfunction of equipment important to safety be increased? YesO No @ The charging pumps, pump motors and pump suction and discharge dampeners have been shown to not be equipment important to safety. The fact that they are no longer classified as "Q" does not affect any other equipment important to safety. 4. Will the consequences of a malfunction of equipment important to safety be increased? YesO No @ i The charging pumps, pump motors and pump suction and discharge dampeners are not credited I for the mitigation of any malfunctions of equipment important to safety. 5. Will the possibility of an accident of ; different type than any previously evaluated in the SAR be created? YesO No @ The "Q" status of the charging pumps, pump motors and pump suction and discharge dampeners cannot initiate accidents of any kind. 6. Will the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR be created? YesO No @ The charging pumps, pump motors and pump suction and discharge dampeners have been shown to not be equipment important to safety. The fact that they are no longer classified as "Q" does not aNect any other equipment important to safety. 7. Will the margin of safety as defined in the basis for any technical specification be reduced? I YesO No @
ARKANSAS NUCLEAR ONE FORM TITLE: Page 2 FORM NO. REV. 10CFR5049 SAFETY EVALUATION 1000.131B 3 PC-2 There are no margins of safety defined in the basis for any technical specification that are rela in any way discharge to the "Q" classification of the charging pumps, pump motors or pump suction an dampeners. b mi Daniel H. Williams Certified Reviewers Signature 7/21/98 Printed Name Date Reviewers certification expiration date: 11/18/99 l Assistance provided by:
)
Printed Name Scope of Assistance ( Date l 0 I I i i
ER9803o2E1os REV o ARKANSAS NUCLEAR ONE FERM TITLE: Peoe 13
> FORM NO. REV.
10CFR50.59 DETERMINATION 1000.131 A 3 PC-1
. Document No. ER980302E108 & JO-00965479 Rev/ Change No. O Title HPl Full Flow Test Gage Installation Temporary Alteration Brief description of proposed change:
This ER and Jo will install, control, and remove test gages for the purpose of monitoring differe across MU-66A, MU-66B, MU-66C, and MU-66D via upstream and downstream vent or drain stacks. Appropriate personnel will open and close these valves to obtain pressure data. This data will be tak conjunction with 1104.002 Supplement 8, the MU & Purification System Check Valve and CV Stroke test is run with the RCS in a cold shutdown condition, i.e. RC temperature less than 150 degrees F, a pressure less than SG psig. Under these operational constraints, HPI is not TS required. For th cold shutdown conditions are credited as the plant mode. Since JO controls are utilized, this is conside maintenance activity. The gages are installed for a specific monitoring purpose on a temporary no SAR figures need to be changed for this maintenance activity. Will the proposed Activity: 1. Require a change to the Operating License including: l Technical Specifications (excluding the bases)?
, YesO No@
Operating License? YesO No@ Confirmatory Orders? YesO No@ 2. Result in information in the following SAR documents (including drawings and text) being (a) no longer true or accurate, or (b) violate a requirement stated in the document: ) SAR (multi-volume set for each unit)? Yes@ nod Core Operating Limits Report? YesO No@ Fire Hazards Analysis? YesO No@ Bases of the Technical Specifications? YesO No@ Technical Requirements Manual? YesO No@ NRC Safety Evaluation Reports? YesO NoE
- 3. Involve a test or experiment not described in the SAR?
(See Attachment 2 for guidance) YesO No@ 4. Result in a potentialimpact to the environment? (Complete Environmental impact Determination of this form.) YesO No@ 5. Result in the need for s Radiological Safety Evaluation per section 6.1.5? YesO No@
- 6. Result in any potential impact to the equipment or facilities utilized for Ventilated Storage Cask activities per Section 6.1.67 YesO No@
- 7. Involve a change under 10CFR50.54 for the following SAR documents per Section 6.1.7?
QAMO? YesO No@ E. Plan? YesO No@ ,
err- -ME108. REV. o ARKANSAS NUCLEAR ONE Page 14 FORM TITLE: FORM NO. REV. 10CFR80.89 DETERMINATION 1000.131A 3 PC-1
' Document No. ER980302E108 RevJChange No. O Basis for Determination (Questions 1,2, & 3):
Question 1 The TS and SAR are predominately concemed with full power operations. Maintenance are referenced, but specific requirements as to when " maintenance" can be performed is gener Conservative efforts are taken in outage planning / scheduling such that work is performed in a safe a RCS is in a cold shutdown condition, there are no OL document than specified time periods.pages. The OL documents force the plant to cold shutdown when Question 2 figure 4-1 and 7-20 were made untrue for the test gage installation required for this activity as there is no exception for maintenance activities besides those given in N inspection Manual 9800. ANO 50.59 information does not address this difference. Question 3 This activity does not affect safety related equipment which is being relied upon to main the adequacy of DH or degrade the plant response to an accident or m Question 4 result of this activity.The environrriental impact determination form did not reveal any impact on the Question 5 outage activities, an RSE is not required.JO planning and implementation will require H
- Question 6 <
This activity does not impact any VSC equipment or activities. Question 7 No E-plan or QAMO search hits were identified that need to be changed for this act O Proposed change does not require 10CFR50.59 Evaluation per Attachment 1, Jtem # (if checked, note appropriate item #, send LDCR to Licensing). L - - - - - _ _ _
--- ] ~
ER980302Eto8. REV o ARKANSAS NUCLEAR ONE FCRM TITLE: Psoe 1G FORM NO. REV. 10CFR50.59 DETERMINATION 1000.131A 3 PC-1 Search Scope: List sections reviewed in the Licensing Basis Documents specified in questions 1,2 and 3. If a keyword sea , was done on LRS, "all" may be entered under "Section" with the keyword (s) used in parentheses. Controlled hard copies of the documents shall be reviewed (LRS is not verified and searches only text, not figures or dra Attach and distribute a completed LDCR per Section 6.1.2 If LSD changes are required. Document Section LRS: Unit 1. Afl. (LTOP). (HPl w/10 test *). (full flow). (" cold shutdown"). (' cold shutdown" w/10 refuelino) MANUAL SECTIONS: TS section 1. section 3.2. 3.3. 3.8. SAR 14.2.2.3. 9.1.1. 6.1.2.1.1. FIGURES: 4-1,6-1, & 7-20 James J. Souto 4/8/98
%Ified Revihers Signature Printed Name Date
~
Reviewers certification expiration date: _ 2/24/99 Assistance provided by: Printed Name Scope of Assistance Date S: arch Scope Review Acceptability (NA, if performed by Technical Reviewer per 1000.006)
"C
_ T ified Reviewers Signature
- ~
boy AI O cew dT 78
* ~ Printed Name Ope
ER9803o2E106, REV. o ARKANSAS NUCLEAR ONE FORM TITLE: Pace 16 FORM NO. REV. 10CFR50.59 DETERMINATION 1000.131A 3 ENVIRONMENTAL IMPACT DETERM; NATION (UNIT 1 and UNIT 2) Document No. ER980302E108 RevlChange No. O Complete the following Determination. If the answer to any item below is "Yes", an Environmental Ev required. See Section 6.1.4 for additional guidance. Willthe Activity being evaluated: Yes No O @ Disturb land that is beyond that initially disturbed during construction (i.e., new constructi buildings, creation or removal of ponds, or other terrestrial impact)? See Unit 2 SAR 2.5-17. This applies only to areas outside the protected area. O @ Increase thermal discharges to lake or atmosphere? O @ increase tower? concentration of chemicals to cooling lake or atmosphere through discharge ca O @ increase quantity of chemicals to cooling lake or atmosphere through discharge canal or tower? O E Modify the design or operation of cooling tower which will change drift characteristics? O E instati any new transmission lines leading offsite? O @ Change the design or operation of the intake or discharge structures? O E Discharges any chemicals new or different from that previously discharged? O E Potentially cause a spill or unevaluated discharge wh:ch may effect neighboring soils, surf water or ground water? O O Involve burying or placement of any solid wastes in the site area which may effect runoff, surface water or ground water? O E involve incineration or disposal of any potentially hazardous materials on the ANO site? O 3 Result in a change to nonradiological effluents or licensed reactor power level? O @ Potentially change the type or increase the amount of non-radiological air emissions from t ANO site.
I kNWe0302E108. REV o IRKANSA3 NUCLEAR ONE FORM TITLE: Page 17 FORM NO. REV. 10CFR50.59 SAFETY EVALUATION 1000.131B 3 Document No. ER980302E108 RevjChange No. 0 10CFR50.59 Eval. No. FFM4 c53 Title HPl Full Flow Test Gage Installation Temporary Alteration (Assigned by PSC) A WRITTEN RESPONSE PROVIDING THE BASIS FOR THE ANSWER TO EAC ATTACHED. EACH QUESTION MUST BE ANSWERED SEPARATELY. A SIMPLE CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES GUIDANCE F If the answer to any question on this form is "Yes,' then an unreviewed safety question is involve to all questions is *No," then the proposed change does not involve an unreviewed safety question . 1. Will the probability of an accident previously evaluated in the SAR be increased? YesO No @ 2. Will the consequences of an accident previously evaluated in the SAR be increased? YesO No @ 3. Will the probability of a malfunction of equipment important to safety be increased? YesO No @ 4. Will the consequences of a malfunction of equipment important to safety im increased? YesO No @ 5. Will the possibility of an accident of a different type than any previously evaluated in the SAR be created? YesO No @ 6. Will the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR be created? YesO No @ 7. Will the margin of safety as defined in the bases for any technical specification be reduced? YesO No @ [/ CertifiedlReviewer's Signature James J. Souto 4/8/98 Printed Name Date Reviewer's certification expiration date:
- 2/24/99 Assistance provided by:
Printed Name Scope of Assistance Date
- /
. /) & \ / /_ s PSC review by:
//
/ \ f b Date:
/ /
d
ER;r8108. REv. O ARKANSAS NUCLEAR ONE FORM TITLE: Proe 18 FORM NO. REV. 10CFR80.89 REVIEW CONTINUATION PACE 1000,131C 3 Document No. ER980302E108 and JO-00965479 RevlChange No. 0 10CFR50.59 Review Continudon Paoe Evaluation Question 1 With the unit in cold shutdown conditions, the installation of test gages on the HPI system for pressure monitoring is not likely to increase the probability of any accident c No direct ties could be found with the gage installation and any cold shutdown acci accident. In consideration of double isolation valves, the loss of inventory from any of thes is unlikely.-. As such, there will not be an increase in the probability of this accident. Evaluation Question 2 in the event of a fuel handling accident, RB integrity is the primary dose mitigatio function to restore. Since RCS water inventory is not affected by this activity, there woul offsite dose consequences as a result of this test gage installation and removal activity. Evaluation Question 3 The RCS primaty function while in cold shutdown conditions is to maintain water le With the RCS in a cold shutdown condition, DH is being us activity does not affect any equipment associated with RCS inventory or cooling, there is no in probability of a malfunction of equipment important to safety. Evaluation Question 4 For the installation and removal of these test gages on the HPl system, RCS level inventory, DH, and fuel handling activities are not affected. As such, the consequences of a equipment important to safety will not be increased. under the authorized conditions of usage. While inadverten inventory, the pipe size is small enough that the drain down rate would be very slow suc can be taken. This assumes that personnel forget to close two valves and the verifier misses considered very unlikely. As such, a different type of accident will not be created. Evaluation Question 6 noted, RCS inventory, DH, or fuel handling activities are not affect
.-time possibility of usage. of a malfunction of equipment important to safety as no such equipment is im Evaluation Question 7 This activity is performed in a specific plant condition that has no defined marg safety. ANO utilizes Administrative controls and the SOPP under these plant conditions. As reductions in any TS bases margin of safety for this activity.
Based on the esponses above, this activity is not a USQ.
. Filed as 0302E108. doc in h:ysouto%vinword
____ --- -- - - - ---~ m- % . 7 ----- ANMANSAs NuCLaAR OW FORM mLE: N, 6
. FORM N3. REV.
10CFR60J9 DETERMINATION 1000,131A 3 PC-1
~
Document No. ER960579E201 Rev1 Change No. 0 - Title Satpoint Change Package for2TRS-1223 Brief description of proposed change: Change still provide the setpoint adequate pumpfor 2TRS-1223 protection to prevent nuisance alarm on Circ Water Pump Bearing Temper Will the proposed Activity: 1. Require a change to the Operating License induding: Technical Specifications (excluding the bases)? YesO NoX Operating License? YesO No X Confirmatory Orders? YesO NoX 2. Result in information in the following SAR documents (including drawings and text) being (a) no longer true or accurate, or (b) violate a requirement stated in the document: SAR (multi-volume set for each unit)? YesX NoO Core Operating Limits Report? YesO No X Fire Hazards Analysis? YesO NoX Bases of the Technical Specifications? YesO No X Technical Requirements Manual? YesO No X NRC Safety Evaluation Reports? YesO No X
- 3. Involve a test or experiment not described in the SAR?
(See Attachment 2 for guidance) YesO No X 4. Result in a potentialimpact to the emvironment? (Complete Environmental impact Determination of this form.) YesO No X 5. Result in the need for a Radiological Safety Evaluation per section 6.1.57 YesO No X 6. Result in any potential impact to the equipment or facilities utilized for Ventilated Storage Cask activities per Section 6.1.67 YesO No X 7. Involve a change under 10CFR50.54 for the following SAR documents per Section 6.1.77 QAMO? YesO No X E-Plan? YesO No X t in,,
F +-M amuussas mar'2e or, FORM inLE: paa. 7 FORM NO. REV. 10CFR80.89 DETERMINATION 1000.131A 3 PC-1 Document No. ER980579E201 Rev> Change No. O Basis for Determination (Questions 1,2, & 3):
- 1. The circulating water system is below the level of scope of the Operating License.
- 2. Thewillsystem change change SAR description figure 10.4-4. of the circulating water system is contained in the SA
- 3. This setpoint change is not considered a test or experiment.
O Proposed change does not require 10CFR50.59 Evaluation per Attachment 1, item # (if checked, note appropriate item #, send LDCR to Licensing). Search Scope: List sections reviewed in the Licensing Basis Documents specified in questions was done on LRS, " pit" may be entered under "Section" with the keyword (s) used copies of the documents shall be reviewed (LRS is not verified and searches o Attach and distribute a completed LDCR per Section 6.1.2 ff LBD changes are required. . Document Sectiou LRS: ALL C'Circulatino Water". 203a) MANUAL SECTIONS: Unit 2 SAR Section 10.4 FIGURES: 10.4-4 n/ M John Harvey prtified ReviewersMure Printed Name 5/11/98 Date RLviewers certification expiration date: _12/11/99 Assistance provided by: Printed Name N/A Scope of Assistance Date Search Scope Re-A; Acceptability (NA, if performed by Technical Reviewer per 1000.0 A m /swANn C:rtified Reviewers Signature SKMdE FKAMfl/N 6lW9I Printed Name Date 1
FCRM mLE: N MM ME %8
. FORM N3. REV.
10CFR50.59 DETERMINATION 1000.131A 3 ENVIRONMENTAL IMPACT DETERMINATION (UNIT 1 and UNIT 2) Document No. ER980579E201 RevJChange No. O Complete the following Determination. If the answerto any item below is "Yes", an Env required. See Section 6.1.4 for additional guidance. Will the Activity being evaluated: Yes No o i O x
. oisturb land that is beyond that initially disturbed during construction (i.e., new con buildings, creation or removal of ponds, or other terrestrial impact)? See Unit 2.5-17. This applies only to areas outside the protected area.
O x increase thermal discharges to take or atmosphere? O' x Increase tower? concentration of chemicals to cooling lake or atmosphere th:Dugh disch O x increase tower? quantity of chemicais to cooling lake or atmosphere through discharge ca O X Modify the design or operation of cooling towerwhich will change drift characteristics? O x instati any new transmission lines leading offsite? O x change the design or operation of the intake or discharge structures? O x oischarges any enemicals new or different from that previously discharged? O X Potentially water cause or ground water?a spill or unevaluated discharge which may effect neighboring soi O X Involve burying or placement of any solid wastes in the site area which may effect surface water or ground water? O x Involve incineration or disposal of any potentially hazardous materials on the ANO site? O X Result in a change to nonradiological effluents or licensed reactor power level? O x Potentially ANO site, chang 9 the type or increase the amount of non-radiological air emissions fr i
EResuo r=__"1 FtJRM TITLE: ARKANSAS NUCLEAR ONE Ps je 9
. FORM NO. REV.
10CFR50.59 SAFETY EVALUATION 1000.131B 0 Document No. erg 8057gE201 Rev) Change No. 0 10CFR50.5g Eval. No. F A09i O'Ks Title Setpoint Change Pa@2e for 2TRS 1223 (Assigned by PSC) ATTACHED. EACH QUESTION MUST BE AN CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES . to all questions is "No," then the propor,ed change does n . 1. Will the probability of an accident previously evaluated in the SAR be increased? YesO No X has no impact on any of the postulated accidents an System functions will remain the same and the probability of a loss of circulating unchanged. 2. Will the consequences of an accident previously evaluated in the SAR be incasased? Yes O No X The failure of this alann function will not change failure considerations or change th consequences of failure previously evaluated in the SAR. This is because the Circula accident evaluated in the SAR. System does not involve any safety rela 3. Will the probability of a malfunction of equipment important to safety be increased? YesO No X Because the Circulating Water System is not essential to safety, the implementati setpoint enange will not change the probability of a malfunction of equipmen 4. Will the consequences of a malfunction of equipment important to safety be increased? YesO No X This setpoint change affects only the Circulating Water System, which has no e
. of equipment important to safety. components important to safety. Thi
- 5. I Will the possibility of an accident of a different type than 1 any previously evaluated in the SAR be created? l Yes O No X The changing of the setpoint does not change the function or the failure mo Water System. Since the failure of the Circulating Water is already evaluated in ,
of Condenser Vacuum), this change will not alter SAR accident type considerations. Th ! possibility created. of an accident of a different type than those previously evaluated in the SAR 6. Will the possibility of a rnalfunction of equipment important to safety of a different type than any previously evaluated in the SAR be created? YesO No X The function and the operating modes of the Circulating Water System are n This setpoint change does not affect any equipment important to safety. T of a malfunction of equipment important to safety beyond those previously evaluated created by this change.
ER980579E201 ARKANSAS NUCLEAR ONE FORM nile: Peae 10 FORM NO. REV. 10CFR50.69 SAFETY EVALUATION - 1000.131B 3 7. Will the margin of safety as defined in the bases for any technical specification be reduced? YesO No X There are no Technical Specifications limits or bases defined for the Circulating Water Sys therefore no such margins will be reduced. - d/L m
/ Certifiedhrs Signature sohn Narvey sisii.s Printed Name Date Reviewers certification expiration date: 12/11/99 Assistance prov'ided by:
Printed Name Scope of Assistance N/A Date n p PSC review by: h Date: (, js M 4 A l
-~ ARKANSAS NUCLEAR OBE " FORM TITLE: Page t' FORM NO. REV. 10CFR50.59 DETERMINATION 1000.131 A 3 PC-1,2 This Document contains 3 Pages. Document No. ER 980898 E201 Rev/ Change No. O Title PEAKING OPERATIONS USING THE AAC GENERATOR Brief description of proposed change: The subject Engineering Evaluation discussed the acceptability of using the Altemate AC peaking Unit. In the past this generator was used only in a stand-by capacity and sta purposes. Based upon the Subject Engineering Evaluation it is acceptable to operate the AAC peaking unit. The effect is an increase in the run time on the machine that affects main an increased risk associated with grid distustiences that may initiate protective action for reduce ANO's ability to protect the health and safety of the public. pro l Willthe proposed Activity: ! 1. Require a change to the Operating Ucense including: Technical Specifications (excluding the bases)? YesO No@ Operating License? YesO No@ Confirmatory Orders?
- 2. YesO No@
Result in information in the following SAR documents (including drawings and text) be (a) no longer true or accurate, or (b) violate a requirement stated in the document: { SAR (multi-volume set for each unit)? YesO No@ Core Operating Limits Report? YesO No@ Fire Hazards Analysis? ' YesO No@ i 9Ms of t;;; Technical Specifications? YesO No@ Technical Requirements Manual? YesO No@ NRC Safety Evaluation Reports?
- 3. YesO No@
Involve a test or experiment not described in the SAR? (See Attachment 2 for guidance) YesO No@ 4. Result in a potential impact to the environment? (Complete Environmental impact Determination of this form.)
- 5. Yes@ nod Result in the need for a Radiological Safety Evaluation per section 6.1.57
- 6. YesO No@
Result in any potential impact to the equipment or facilities I utilized for Verallated Storage Cask activities per Section 6.1.6?
. 7.
YesO NoS i Involve a change under 10CFR50.54 for the following SAR documents per Section 6.1.77 - QAMO? E-Plan? - YesO rio@ ! YesO No@
FORM TITLE: ARMANsAs NuCLaAR oNE Page f FORM NO. REV. 10CFR50.89 DETERMINATION 1000.131A 3 PC-1,2
- Document No. ER 980898 E201 Rev1 Change No. 0-Basis for Determination (Questions 1,2, & 3):
The utilization of the Altemate AC generator (AACG) as a peaking ur,it was discuss and approved in the sunsequent SER. Because of this, the subject of peaking was incl both units during the installation of the Altemate AC system. The Licensing basis doc negative response to questions 1,2, & 3. references to the acceptability of ope! { Although, the peaking issue was reviewed in moderate detail by the NRC, th indicate this. The regulators approval in the SER appears to be tacit, as it does no probabilities or a reduction in margin (no matter how sm I O Proposed change does not require 10CFR50.59 Evaluation per, Attachment (if checked, note 1, item # appropriate item #, send LDCR to Licensing). Search Scope: List sections reviewed in the Licensing Basis Documents specified in questions 1,2 an parentheses. Controlled hard copies of the documents text, not required. figures or drawings). Attach and distribute a completed LDCR perre Secti Document Section LRS: C 5059. C88 throuch C93 (Altemate AC) (Station Blackout) (PaaH MANUAL 0a089203: SECTIONS: 2SAR 8.3: 2SAR 8 A 2SAR 1.2.2.10.f: 1SAR 8 On129208 . FIGURES: Floures were not rat;;;,d due to the nature of the chanoe. This chsnae does not involv channes ssnereier. to any chvalem! cGnTscisisens: only the isnsin and ft:.ssacy of stsitire a
&- Richard A. Bames Certified Reviewers S$riature 10/06/98 Printed Name Date
' Reviewers certification expiratN a de: 06/27/99 ' Assistance provided by: Printed Name Dennis Calloway Scope of Assistance Nonradiological Enviommental evaulation Date 9/21/98 1 Search Scope Review Acceptability (NA, if performed by Technical Reviewer per 1000 - (2k-les CIrtified Reviewers Signature hb Scwem. /o/n,/9fr 1 Printed Name ' Date
runm a a a Le: . . . - 1 F3RM NO. REV. 10CFR50J9 DETERMINATION 1000.131A 3 i ENVIRONMENTAL IMPACT DETERMINATION . (UNIT 1 and UNIT 2) Document No. _ER 980898 E201 Rev> Change No. O
. required. See Section 6.1.4 for additional guidance. Complete the follo Will the Activity s' eing evaluated:
.Y.tB 119 O E' bulidings, creation or removal of ponds, or oth 2.5-17. This applies only to areas outside the protected area.
O E increase thermal discharges to take or atmosphere? O E increase tower? concentration of chemicais to cooling take or atmosph.. a thr O E increase tower? quantity of chemicais to cooiing take or atmosphere through disch O 2 Modify the design or operation of cooling tower which will change drtft chara O E install any new transmission lines leading offaite? . O E change the design or operation of the intake or discharge structures? O @ Discharges any chemicais new or different from that previously discharged? O @ ) Potentially water or ground cause water? a spill or unevaluated dischage which may effect neigi O E involve surface burying water orwater? orground pacement of any solid wastes in the site si== which ma O 3 . involve incineration or disposal of any potentially hazardous materials on t O E
. Result in a change to nonradiological affluents or licensed reactor power level E O' Potentially ANO site. change the type orincrease the amount of non-radiological air G
w_ . ...m
\
Page cf I Doc #: E E ffe fhf 9' Rev #: d'7 Title ()TilI2 *n 4* $N e Al'ftf**e. t Y l $dte/A b'< *4) s4 b d E'**
- 4' V
NONRADIOLOGICAL ENVIRONMENTAL EVALUATION If the answer to any question is "Yes", then an Unreviewed Environmental Question is involved. If the answer to all questions is "No", then the proposed change does not involve an Unreviewed Environmental Question. A written response providing the basis for the answer of each question must be provided. Attach addit,4- M Jages as necessary. A simple statement.of conclusion is not sufficient. 2.1 Does the proposed activity result in a Yes significant increase in any adverse No e environmental impact previously evaluated by the NRC in References 3.2.3-3.2.9? D:tse t. CA on 'Tda< A ,(fecee1/2diasAh
? YA e. l1 w $ eor/s/2/so s a / f/< J7s /le n. $kcf.Gei Discussion:
- a. , ,t yt a_ J.g /e c f.sas2sa w A L D,c ,,,,J 4D 1/47 no t Ator db <s./% 'ttb 2, t4 tJe D:es*are~YJ 2.2 Does the proposed activity result in a Yes significant adverse environmental impact not No t-previously evaluated in References 3.2.3-3.2.9?
Discussion: ~)"b 4 $NfA/,4.s e ul Ils f /ch.us A}Acd'. & WY .D;essi l*.4 an .e ,4 hest
~7/n < L .~s l*/ & vira,eme.n(n l .heessera. air weis / drJlJLeo. Tke /Xs.Wo~ AJAs.k d.< T h;*
15 A n toeraD ~To e/adef<. unebel AFAs Aix hn=2 9 % GG10-M . / eta;+ 3see; f:c4fie Alla ' ' fas DAdPA1,**r s# To JV A A. A D e ./ "I D a.11 A t X . J s/ ts,#. fed f un A.e t Ces b e e,HfAf**' AJ "A Pedki"J *** i 7 * ' ' ' 2.3 Does the proposed activity r.ssult in a Yes significant change in nonradiological effluents No "' or licensed reactor power level? I Discussion: 'Tle JYsl:*r AJAc,k'-4s.Y b~ea e I ($/tae lJ k.de e b e*=e. [Ve-ls a. e
*T4 ?) /~.n l 1 i :e s SArt c t '3 ft/ene.*11tf f., 7Le AffheJA3 rdp/se fare rl co- fg)/ rf. aa. O !dfl f
4D /soJorv. TA t> se n i+ dad." .*De s fes de >- A.s.:t AO . Doe's sto f fen,.i.?e A .es C k A e Z 'i;,u l-+'o d Are e4 L ,e s / s . " ' l Evaluator: WS Date: N.2/-f[ Supt., Chem.: Date: b[-- PSC Review: Date: (O \% FORM TITLE. " FORM NO. REV. NONRADICLOGICAL ENVIRONMENTAL EVALUATION FORM 1052.034A 0
' FORM TITLE: ~
FORM NO. REV. 10CFR50.59 2AFETY EVALUATION 1000,1310 3 PC-2 This Document contains 2 Pages. Document No. ER 980898 E201 RevlChange No. 0 10CFR50.59 Eval. No. }:t-d QT.lf (Assigned by PSC) Title PEAKING OPERATIONS USING THE AAC GENERATOR A WRITTEN RESPONSE PROVIDING THE BASIS FOR THE ANSWER TO EACH QUESTION MUST B ATTACHED. EACH QUESTION MUST BE ANSWERED SEPARATELY. A SIMPLE STATEMENT OF ' CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES GUIDANCE FOR RESPONSE. ' If the answer to any question on this form is "Yes," then an unreviewed safety question is involved. If the answer to all questions is "No," then the proposed change does not involve an unreviewed safety question.
- 1. Will the probability of an accident previously evaluated in the SAR be increased?
YesO No @
- 2. Will the consequences of an accident previously evaluated in the SAR be increased?
YesO No @
- 3. Will the probability of a malfunction of equipment important to safety be increased?
YesO No @
- 4. Will the consequences of a malfunction of equipment important to safety be increased?
YesO NO @ l S. Will the possibility of an accident of a different type than any previously evaluated in the SAR be created? YesO No @
- 6. Will the possibility of a malfunction of equipment important to safety of a different typa than any previously evaluated in the SAR be created?
YesO No @ 7. Will the margin of safety as defined in the basis for any technica! specification be reduced? YesO No @ r- N A m- Richard A. Bames 10/06/98 (;ertified ReviesPs Signature Printed Name Date Reviewer's certification expiration date: 6/27/99 Assistance provided by: Printed Name Scope of Assistance Date PSC review by: hO-- . Date: \Ok\'1b
Document Ns. ER 980898 E201 Page 2 of 5 Rev/Ch:nga No. R:v 0 10CFR50.59 Review Continuation Pace
- 1. Will the probability of an accident previously evaluated in the SAR be increased?
The operation of the Altemate AC generator as a peaking unit is no different during testing with the exception ofincreased frequency of starts and tota the AAC generator is operated less then 50 hours per year, although the environmen discharge permit allows 4032 hours of full power operations per year. As a p , it is anticipated that the generator will be operated approximately 800 ho evaluated power accident (LOOP). that could be impacted by the increase in operating time is a By design, the Altemate AC system does not contain a single point vulnerabilit the simultaneous failure of the preferred power (Off-Site) source, the eme sources and the Attomate AC source. Because the AAC system serves both point vulnerability criteria applies to the interface with either ANO-1 or 2. C that ma.y occur within the AAC system will not produce significant elfacts system or the emergency onsite systems. During the post lastallation testing of the system several full and partial load rejection test were conducted without noticea upon the on-site or off site power distribution systems. During peaking operations, the AAC generatorinterfaces only with one of the 41 electrical bus 1A1 or 2A1. Even postulating a complete catastrophic failure of the generator in a manor where all protective design features also failed the maximu could occuris the loss of bus 1A1 or 2A1. The effect of a failure or frequency fluctuations if not significant enough to action will not have a measurable affect on the off-site power grid or the on site distribution system. The failures anticipated, up to including a protective t not initiate a Loss of Off Site power. A loss of offsite power without additiona initiate any other accidents. Therefore, the increased operating time (4 the Altemate AC does notincrease the probability of the occurrence of evaluated.
- 2. Will the consequences of an accident previously evaluated in the SAR be increas During peaking operations, the AAC generator interfaces only with one of the!
electrical bus 1A1 or 2A1. Even postulating a complete catastrophic failure of th generatorin a manor where all protective design features also failed the maximum could occuris the loss of bus 1A1 or 2A1. The effect of a failure , assumption in several of the evaluated accidents. T{ were determined based upon an assumed LOOP. Therefore, the offsite dos remain unchanged from those previously evaluated in the SAR. FORM TITLE: FORM NO. REV. 10CFR50.59 REVIEW CONTINUATION PAGE 1000.131C 3
D:curnent No. ER 980898 E201 ' Rev> Change N3. R:v 0 10CFR50.59 Review Continuation Pace
- 3. Will the probability of a malfunction of equipment important to safety be increased?
The use of the AAC generator as a peaking unit was part of the design basis. operations of the AAC are no different from routine surveillance testing except f the run period and the frequency of the evolution. The generator is rated for conti with all the design features of a peaking unit. It can be connected to either A1 or2 displace the unit house loads, thereby increasing the not output of either ANO 1 or 2 design contains sufficient controls and protection to allow long parallel runs with little monitoring. The govemor automatically changes into a " load share" mode upon synchronization and connection to 2A1 or 1A1. The design includes a power factor con feature that will adjust the generator output to maintain a constant kW output contains sufficient protective features to isolate and/or trip if a major grid disturbance unit trip or a loss of offsite power would not cause damage to the AAC genarator. Th effect would be the protective tripping of the AAC generatorlockout relay requir . Operator reset prior to retuming the generator to service. The overall vulnerability to grid disturbances will be increased due to the anticipa operating time. According to NUMARC 8700 Rev1, which forms the design basis for system, there shall be minimal potential for common cause failure of the station A sources. Seven design criteria are contained within NUMARC 8700 that assure a " min potential for common cause failures." The current configuration complies with each o seven criteria. The criteria related most closely to peaking operations is single point vulnerability criteria. This criteria states that no single point vulnerability sha likely weather-related event or single active failure could disable any portion of the emergency power source. AC power sources or the preferred power sources and simultaneously fa The AAC system configuration does not contain a single point vulnerability that c simultaneous failure of the AAC power source and the onsite emergency AC sou offsite preferred source. The vulnerability identified associated with the operation o generator lockout is a designed AAC protective function and not considered a failure. limiting scenario of concem is a loss of offsite power during peaking operations. scenario the reactor would trip, followed by a main generator trip. These action an approximate 300% overload on the AAC generator. Review of the protectiv inconclusive as to which protective action will occur first, the tripping of 2 or the 2K9 overcurrent relays that initiate a generatorlockout. The tripping of th eliminates the single point vuinerability of concem portion the SBO commitment. 4 The overall design of the AAC system was submitted to the NRC and ap evaluation report (SER). Deviations from the SER were evaluated by Safety arid Environmental evaluations and 50.59s for the associated Design Change Pa submittals, SER, and 50.59s addressed the use of the AAC generator as a pe major concems expressed verbally by the regulator were LOOP initiators, and the ability to place the machine in operation within 10 minutes from the control Peaking service was discussed on several occasions and approval is docume The current operating procedure for the AAC generator includes a precaution lockout relay may actuate during a loss of offsite power. The procedure goes o lockout should be reset immediately to ensure the diesel can be started within 10 m FORM TITLE: FORM NO. REV. 10CFR50.59 REVIEW CONTINUATION PAGE* 1000.131C 3
Document N?. ER 980898 E201 Prge 4 of 5 Rav> Change No. Rsv 0
$UCFR50.59 Review Continuation Paae declaring a station blackout." It is intended that a non-control room operator would be accountable to reset the AAC generator lockout relay while the control room operators performing the steps in the EOPs. This would be completed prior to the declara blackout and the beginning of the committed 10 minute period. The actions ne AAC generator and align it to the selected safety bus could then be accomplished control room. The inclusion of the precaution in operations procedure 2104.037 sufficient assurance that we will and remain in compliance with the regulatory reqI i regarding station blackout while operating the AAC generator as a peaking unit.
A review of the potentialimpacts on the Probabilistic Safety Analysis Model (PS that the use of the AAC Generator as a peaking unit adds an AAC trip condition t l actuation that requires a local manual reset. This may I assumed in the PSA model for AAC starting. However, the action to reset the rela performed well within the required 10 minute AAC initiation on declaration of a s The will relay procedure revision prevent delay recommended to have a non control room operator r in AAC start. The interface with the on site emergency diesel generator system will not be result of the increase in operating time for peaking. The maintenance intervals for generator were found acceptable with only minor changes in oil sampling frequency In Summary, the increased operating time is well within the design of the Alte source, regulatory compliance, original design submittals, and the SER. Peaking ope will not degrade existing safety systems, in particular the on site emergency pow addressed above the function of the AAC generatoris to mitigate the effects of a Stat Blackout. The increased operating time associated with peaking does not s the ability to accomplish this. Therefore, although the activity will increase the ope for the AAC system, it does not represent an increase in the probability of malfunc equipment important to safety.
- 4. Will the consequences of a malfunction of equipment important to safety be increa During peaking operations, the AAC generator' interfaces only with one of the 4160 electrical bus 1A1 or 2A1. Even postulating a complete catastrophic failure of the AAC generator in a manor where all protective design features also failed the maximum eff could occuris the loss of bus 1A1 or 2A1. The effect of a failure of eith enveloped by a complete loss of off site power (LOOP). A loss of offsite power assumption in several of the evaluated accidents. The consequences from these accide were determined based upon an assumed LOOP. Therefore, the dose consequences unchanged from those previously evaluated in the SAR.
- 5. be Will the possibility of an accident of a different type than any previously evaluate created?
FORM TITLE: FORM NO. REV. 10CFR50.59 REVIEW CONTINUATION PAGE 1000.131C 3
Doc:<aent N3. erge o or b ER 980898 E201 Rw)Ch nga No. R:v 0 { 10CFR50.59 Review Continuation Psae During peaking operations, the AAC generatorinterfaces only with one of the 4160 v electrical bus 1A1 or 2A1. Even postulating a complete catastrophic failure of the AA , generator in a manor where all protective design features also failed the maximum effec) could occuris the loss of bus 1A1 or 2A1. The effect of a failure of e{l enveloped by a complete loss of off site power (LOOP). A loss of offsite powI assumption in several of the evaluated accidents. Because of this single interface with th safety AAC systems, generator. an accident of a different type can not be created by the increased I n ltnp nt to safety of a different type than vio sly e a sted in R be c a d? During peaking operations, the AAC generatorinterfaces only with one of the 4160 electrical bus 1A1 or2A1. Even postulating a complete catastrophic failure of the AAC generator in a manor where all protective design features also failed the maximum could occur is the loss of bus 1A1 or 2A1. The effect of a failure of enveloped by a complete loss of off site power (LOOP). A loss of offsite po assumption in several of the evaluated accidents. Therefore, the increase operatio single interface point associated with peaking will not create the possibility of a r equipment important to safety of a different type then previously evaluated.
- 7. Will the margin of safety as defined in the basis for any technical specification be The Alternate AC power source is not referenced in the technical specification o the technical specifications. Consequently, there can be no reduction the m defined from peaking.in the basis for a technical specification associated with the increased l
j 4 i FORM TITLE: FORM NO. REV. 10CFR50.59 REVIEW CONTINUATION PAGE 1000.131C 3
[ - l- 98103sE201. Rev. o ARKANSAS NUCLEAR ONE Page 1 FORM TITLE: FORM NO. REV. 10CFR50.59 DETERMINATION 1000.131 A 3 PC-1 This Document contains 5 Pages. j Document No. 981035E201 Rev/ Change No. 0 l Title RELOCATE EMERGENCY LIGHT BATTERIES 2EL-71 AND 2EL-72 Brief description of proposed change: The Exide B200 emergency lighting units (ELU) 2EL-71 and 2EL-72 are located in the upper elevations of the Unit 2 penthouse area. The units are subjected to temperatures in excess of 120*F. This elevated temperature is detrimental to the batteries and electronic charging board inside the units. The battery units are to be moved to a lower elevation to l increase their reliability and the ability to perform maintenance on the units. 2EL-71 and 2EL-72 are required per 10CFR50 Appendix R Section IIIJ to illuminate valves 2CV-1001 and 2CV 1050-2 for access by Operations to execute manual alternate shutdown actions. This evaluation will provide details to relocate the ELU's and install their lamps at the locations previously illununated by the ELU. The new remote lamps will be fed from the same ELU at new locations and the ELU will be powered from the same power supply. The de. sign function of the ELU's is not changed as the same number and wattage lamps will be located where previously they were installed directly on the B200 unit. l Will the proposed Activity:
- 1. Require a change to the Operating License including:
Technical Specifications (excluding the bases)? YesO No@ Operating License? YesO NoS Confirmatory Orders? YesO No@
- 2. Result in information in the following SAR documents (including drawings and text) being (a) no longer true or accurate, or (b) violate a requirement stated in the document:
l SAR (multi-volume set for each unit)? YesO No@ l Core Operating Limits Report? YesO No@ Fire Hazards Analysis? Yes@ NoO Bases of the Technical Specifications? YesO NoS Technical Requirements Manual? YesO No@ NRC Safety Evaluation Reports? YesO No@ l
- 3. Involve a test or experiment not described in the SAR? YesO No@
(See Attachment 2 for guidance)
- 4. Result in a potential impact to the environment? (Complete Environmental l Impact Determination of this form.) YesO No@
- 5. Result in the need for a Radiological Safety Evaluation per section 6.1.57 YesO No@
- 6. Result in any potential impact to the equipment or facilities utilized for Ventilated Storage Cask activities per Section 6.1.67 YesO NoS l
- 7. Involve a change under 10CFR50.54 for the following SAR documents per Section 6.1.77 QAMO? YesO No@
E-Plan? YesO No@ i l
ARKANSAS NUCLEAR ONE P' age 2 FORM TITLE: FORM No. REV. 10CFR60.59 DETERMINATION 1000.131 A 3 PC-1,2 Document No. 981035E201 Rev/ Change No. 0 Basis for Deterunination (Questions 1,2, & 3):
- 1. No Operaung License documents were found which addressed the conAguration of 2EL-71 or 2EL-72. As such. no changes to the OL willbe requued to i , lement the relocation of the emergency light units.
- 2. FHA Sgure 5-28 (drawing FP-2314) and Fire Zone Analysis for Fire Zone 2155-A were found to be affected by changes of ER 981035E201. A 10CFR50.59 Evaluation has been prepared to address the identi6ed conflicts. No other SAR documents were found that would be made untrue, nor were any found that would be violated as a result of this change.
- 3. Testing associated with the relocation of 2EL-71 and 2EL-72 will be performed on equipment that is out of senice.
Therefore, testing directed by this ER is not within the scope tests or experimente requinns evaluauon for an USQ. O Proposed change does not require 10CPR50.59 Evaluation per Attachment 1, Item # m (Ifchacirari noteappropriate item #, send LDCR to Licensing) Search Scope: List sections reviewed in the Licenstag Basis D=.u a specified in questions 1,2 and 3. If search was performed on LRF. the LRS search index should be entered under "Section" with the search statement (s) used in parenthanas Controlled hard copies of the ?'-? -2 shall be scru (LRS is not venfied and searches only text, not figures or drawings). Attach an ' distribute a completed LDCR per Sartian 6.1.2 Af LBD changes are required. Document Egge LRS: 50.59-Unit 2 (- -- -- < lieht*: ------ -{- r w/2 lieht*) MANUAL SEClTIONS: FHA Fientes I Fire 7ana 2155-A: Tablel.7-2 FIG S: Fieure 5 28
. Ronald D. Hendnx 1/12/99
'Certined Reviewer's Signature Printed Name Date Reviewer's cert:6 cation expiration date: 3/19/99 Assistance provided by: Printed Name Scope ofAssistance Date Sea Scope Review A--;=id"'y (NA, if performed by Technical Reviewer per 1000.006) W $mA v }ALk&2 l -/s~-1 aj Cent:6ed ewer's Signature Date
/' Printed Name
ARKANSAS NUCLEAR ONE Page 3 FCRM TITLE: FORM NO. REV. 10CFR60.59 DETERMINATION 1000.131 A 3 ENVIRONMENTAL IMPACT DETERMINATION (UNIT 1 and U191T 2) Document No. 981035E201 Rev/ Change No. O Complete the following Determination. If the answer to any item below is "Yes", an Emironmental Evaluation i See Section 6.1.4 for additional guidance. Will the Activity being evaluated: XSE HQ O B Disturb land that is beyond that initially disturbed during construction (i.e., new constmetion of buildings, creation or removal of ponds, or other terrestrial impact)? See Unit 2 SAR Figure 2.5-17. This applies only to areas outside the protected area. 5 Increase thermal discharges to lake or atmosphere? Increase concentration ofchemicals to cooling lake or atmosphere through discharge canal or tower? O Increase quantity of chemicals to cooling lake or atmosphere through discharge canal or tower? E Modify the design or operation of cooling tower which will change drift characteristics? E ' Install any new transmission lines leading offsite? O E Change the design or operation of the intake or discharge structures? O 5 Discharges any chemicals new or difrerent from that previously discharged? O E Potentially cause a spill or unevaluated discharge which may efrect neighboring soils, surface water or ground water? O E Involve burying or placement of any solid wastes in the site area which may effect runoff, surface water or ground water? O E Involve incineration or disposal of any potentially hazardous materials on the ANO site? 1 0 5 Result in a change to nonradiological effluents or licensed reactor power level? 5 Potentially change the type or increase the amount of non-radiological air emissions from the ANO site.
981035E201. Rev. o ARKANSAS NUCLEAR ONE Page 4 FORM TITLE: FORM NO. REV. 10CFR50.59 SAFETY EVALUATION 1000.131B 3 PC-2 Document No. 981035E201 Rev/ Change No. ' 0 10CFR50.59 Eval. No. N- TI dO (Assigned by PSC) Title RELOCATE EMERGENCY LIGHT BATTERIES 2EL-71 AND 2EL-72 A WRITTEN RESPONSE PROVIDING THE BASIS FOR THE ANSWER TO EACH QUESTION MUST BE ATTACHED. EACH QUESTION MUST BE ANSWERED SEPARATELY. A SIMPLE STATEMENT OF CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES GUIDANCE FOR RESPONSE. If the answer to any question on this form is "Yes," then an unreviewed safety question is involved. If the answer to all questions is "No," then the proposed change does not involve an unreviewed safety question.
- 1. Will the probability of an accident previously evaluated in the SAR be Yes O No @
increased?
- 2. Will the consequences of an accident previously evaluated in the SAR be Yes O No @
increased?
- 3. Will the probability of a malfunction of equipment important to safety be Yes O No @
increased?
- 4. Will the consequences of a malfunction of equipment important to safety be Yes O No @
increased?
- 5. Will the possibility of an accident of a different type than any presiously Yes O No @
evaluated in the SAR be created?
- 6. Will the possibility of a malfunction of equipment important to safety of a Yes O No @
different type than any previously evaluated in the SAR be created?
- 7. Will the margin of safety as defined in the basis for any technical Yes O No @
specification be reduced? 0 -
/ .
Certified ReviewerTSignature ' D Ronald D. Hendrix Printed Name
/2 Date M
Resiewer's certification expiration date: 3/19/99 Assistance provided by: Printed Name Scope of Assistance Date PSC review by: Dgh Date: \\\% l 1 I i
901035E2ol. Rev. o ARKANaAa NUCLEAR ONE Page 5 FORM TITLE: FORM NO. REV. 10CFR50.89 REVIEW CONTINUATION PA2E 1000.131C 3 Document No. 981035E201 Rev/ Change No. O e 10CFR50 59 Reviewraatin='ian Pane
- 1. Will the probability of an acculent previously evaluated in the SAR be increased?
There are no accidents in the SAR attributed to 12VDC emergency lighting. Relocation of emergency lighting batteries without afectirs illommation levels will therefore not result in a change of frequency class (or significantly change within a class) such that the probability of an accident previously evaluated will not be increased.
- 2. Will the consequences of an aceirient previously evaluated in the SAR be increased?
There are no accidents in the SAR attributed directly or indirectly to Appendix R emergency lighting. Changes of { the emergency lighting configuration will tlerefxe not increase the dose consequences of any previously analyzed beyond the ANOlicensed limit.
- 3. Will the probability of a malfunction of equipment important to safety be increased?
Appendtx R emergency lighting has no afect on any equipment important to safety. As such, the probability of a malfunction of equipment important to safety will not be increased by the proposed configuration changes.
- 4. Will the consequences of a malfunction of equipment important to safety be increased?
Appendix R emergency lighting has no afect on any equipment important to safety. No increase in the offsite dose resulting from a malfunction of equipment important to safety will result from the proposed configuration changes.
- 5. Will the possibility of an accident of a diferent type than any presiously evaluated in the SAR be created?
Emergency lighting has no direct or indirect impact on plant systems and as such its failure cannot initiate a different type accident than any previously evaluated in the SAR as a result of the proposed configuration changes 6. Will the possibility of a malfunction of equipment important to safety of a diferent type than any previously evaluated i the SAR be created? Emergency lighting has no direct or indirect impact on equipment important to safety. Therefore, a nialfunction of equipment important to safety of a different type than any previously evaluated cannot be created by the proposed a, q lighting configuration changes.
- 7. Will the margin of safety as defined in the basis for any technical specification be reduced?
No margin of safety as defined in the Technical Specifications involves Appendix R emergency lighting. Therefore, no margin of safety will be reduced by the proposed configuration change.
es ... . nuu.cm vac FORM TITLE; FORM NO. REV. 10CFR50.59 DETERMINATION 1000.131 A 3 PC.1 Pageje of M Document No. ER981059E201 RevlChange No. O Title Ul2 Main Turbine Quarterly Valve Stroke Test Deferral and SAR Minor Discrepancy Resolution Brief description of proposed change: This chanae consists of addina a provision to the TRM to allow i deferral of U/2 Main Turbine Co'ntrol Valve testina if proper lustification is provided in an ER evaluation. The provision to defer the test will be used at this time to defer the next CV test. It is desired to defer this test beg _ause a control problem results in oscillations of the CV servo's which leads to hammerino in the EHC pipina. The ability of the CV's to operate or perform properly in the test is not auestioned, but the hammerina of the pipina is undesirable so deferral of the test is preferable. Resolution of several SAR discrepancies identified by SDID-2-97-0225 and research of the testina issue are included in this assessment. Will the proposed Activity:
- 1. Require a change to the Operating License including:
Technical Specifications (excluding the bases)? YesO No@ Operating Lice.ise? YesO No@ Confirmatory Orders? YesO No@ 2. Result in information in the following SAR documents (including drawings and text) being i (a) no longer true or accurate, or (b) violate a requirement stated in the document: SAR (multi-volume set for each unit)? Yes@ NoO Core Operating Limits Report YesO No@ Fire Hazards Analysis? YesO No@ Bases of the Technical Specifications? YesO No@ Technical Requirements Manual? Yes@ NoO NRC Safety Evaluation Reports? YesO No@
- 3. Involve a test or experiment not described in the SAR?
(See Attachment 2 for guidance) YesO No@
- 4. Result in a potentialimpact to the environment? (Complete the Environmentalimpact Determination of this form.)
YesO No@
- 5. Result in the need for a Radiological Safety Evaluation per section 6.1.57 YesO No@
- 6. Result in any potential impact to the equipment or facilities utilized for Ventilated Storage Cask activities per Section 6.1.67 YesO No@
- 7. Involve a change under 10CFR50.54 for the following SAR documents per Section 6.1.7:
i
ARKANSAS NUCLEAR ONE FORM TITLE: FORM NO. REV. 10CFR50.59 DETERMINATION 1000.131A 3 PC-1 Document No. ER981059E201 Pgt. :: # 24 RevlChange No. O l Page _ of QAMO? YesO No@ E-Plan? YesO No@ Basis for Determination fQuestions 1. 2 & 3):
- 1) This valve stroke test is not discussed in the Operating License. Therefore the change will statements in these documents nor is it necessary to add any new information as a result of this c of GL9510, the turbine generator overspeed protection system is not considered a primary p design basis accident. Based on this philosophy the turbine overspeed requirements have already from the TS.
- 2) Revisions will be necessary to the TRM and SAR as a result of this change. The TRM has bee the provision to defer control valve testing if an engineering evaluation is done to justify the deferral. Th provision is added via a note at the bottom of the TRM page. The revision also extends the action ti more appropriate values. The current values of 4 hours are shorter than what is warranted for this acti current TRM requires a valve that falls the test be restored to operable or closed within 4 hours. The 4 is being changed to a more reasonable duration of 72 hours. If the valve cannot be restored to ope ,
then the turbine must be isolated from the steam supply. The current requirement is for turbine isolation steam supply be achieved in 6 hours; this duration is being changed to 12 hourt to support smoother shut operation. There is also a requirement that if the overspeed protection system is otherwise inoperable that turbine be isolated from the steam supply in 6 hours. This duration is extended to a more realistic duratio hours. The changes to the SAR are editorial and typographical corrections. The description of the valve t be revised to reflect the current quarterly valve testing approach as opposed to the former weekly t affected section is 3.5.2.2.3. A gramtnatical change is also being made to section 10.2.1, last paragraph. The change removes the wor
- inservice
- because it does not accurately describe the type of testing done. A typographical erro corrected in section 3.5.2.2.2.1, where the LP '8" monoblock rotor P1 value should be 1.0x10E-8 rather than 10x10E 8 (decimalis currently missing). Another correction in the same section consists of
- refe corrected to
- reference 53*.
l The SER includes discussion of historical changes regarding valve testing, but this change does not inva information contained in the SER. Therefore no changes to the SER are warrented.
7 ww.... ..wou.m w,ir. l I- FORM TITLE: 1 FORM NO. REV. 10CFR50.59 DETERMINATION 1000.131 A 3 Pc.1.2 ER981059E201 R/0 Page t2 of 2 4
\
- 3) Turbine valve stroking is mentioned in the SAR although not discussed in detail. The stroke test does involve placing the plant in any unanalyzed condition. The testing is controlled via procedure OP2106.0 Based on these observations, this change does not involve an IPTE.
- 4) The Environmental Impact Determination was completed, it was found that this change has no im environment.
- 5) This change does not involve radioactive material or handling of radiological material.
- 6) Missile generation is an area of concern for this change. The missile analysis located in the SAR does consider the Ventilated Storage Casks as a potential target. Since the casks are moved through areas tha targets identified, a concern was raised that the casks should be considered as a potential target. Two 10CFR72.48 reviewers were contacted to assist in assessing this concem. The key point to as to why missiles weren't a concem for the VSC's has to do with the missile path and the cask strength. The turbine missiles are postulated to travel in a radial direction with a range of up to 25 degrees (reference the SAR 3.5.2.2.2.2). The casks do not enter this path until after the cask assembly is completed. Thus the maximum protection is provided prior to the casks entering the postulated area. Once the assembly is completely it is designed to withstand missiles that envelop the turbine missiles. Therefore, the turbine missile by the existing analysis although it is not specifically discussed in the SAR. Based on this conclusion, it is necessary to perform 10CFR72.48 review.
- 7) The QAMO and E-Plan are unaffected by this change.
O Proposed change does not require 10 CFR 50.59 Evaluation per Attachment 1, item #
. (if checked, note appropriate item #, send LDCR to Licensing).
ARKANSAS NUCLEAR ONE FORM TITLE: FORM NO. REV. 10CFR50.59 DETERMINATION 1000.131 A 3 PC.1.2
' ER981059E201 R/0 Page 13 of ZG Search Scope:
List sections reviewed in the Licensing Basis Documents specified in Question 1,2 and 3. If a performed on LRS, the LRS search index should be entered under "Section" with the search parentheses. Controlled hard copies of the documents shall be reviewed (LRS is not verifi text, not figures or drawings). Attach and distribute a completed LDCR per Section 6.1.2 if L required. Document Segli.gn LRS: 50.59-common All- keywords: overspeed w/20 turbine, missile w/20 turbine, turbine w/20 (valve stroke) MANUAL SECTIONS: 1SAR 10.2,14.1.2.9 2SAR TRM 3.5.2,10.2,15.1.7,15.1.29,15.1.33,15.1.10, Table 3.5-2 & 3.5-3 3/ 4.3.4 SER FIGURES: Amendment 174,191,192 2SAR 3.5-1 M hlcVE.lAnel E A 50eVet-Certified Reviewers Signature 5[97 Printed Name Date Reviewers certification expiration date:
/
M Assistance provided by: Printed Name Scope of Assistanc' Date Search Scope Review Acceptability (NA, if performed by Technical Review per 1000.006) I A db m _ LLCh oEp di: rT6 Certified 3eviewers Signature Printed Name II/M/M Date l
annant*> Nuu.e.An uhe FORM TITLE: FORM NO. REV. 10cFR50.59 DETERMINATION 1000.131 A 3 Page d of 26 ENVIRONMENTAL IMPACT DETERMINATION (UNIT 1 and UNIT 2) Document No. ER98105',1201 Rev./ Change No. O Complete the following Determination. If the answer to any checklist item is "Yes", an Environmental Evaluation is required. See Section 6.1.4 for additional guidance. Willthe Activity being evaluated: Yes No O E Disturb land that is beyond that initially disturbed during construction (i.e., new construction of buildings, creation or removal of ponds, or other terrestrialimpact)? See Unit 2 SAR Figure 2.5-17. This applies only to areas outside the protected area. O E increase thermal discharges to lake or atmosphere? O M Increase concentration of chemicals to cooling lake or atmosphere through discharge canal or tower? O E increase quantity of chemicais to cooling lake or atmosphere through discharge canal or tower? O E Modify the design or operation of cooling tower which will change drift characteristics? O 2 install any new transmissior. lines leading offsite? O 2 change the design or operation of the intake or discharge structures? O E Discharges any chemicals new or different from that previously discharged? O @ Potentially cause a spill or unevaluated discharge which may effect neighboring soils, surface water or ground water? O @ Involve burying or placement of any solid wastes in the site area which may effect runoff, surface water or ground water? O O involve incineration or disposal of any potentially hazardous materials on the ANO site? O E _ Result in a change to nonradiological effluents or licensed reactor power level? O @ Potentially change the type or increase the amount of non-radiological air emissions from the ANO site, , l
ARKANSAS NUCLEAR ONE FORM TITLE: FORM NO. REV. 10cFR50.59 EVALUATION 1000.131B 3 PC.3 Page(of 26~ 10CFR50.59 Eval. No. F FN cf 8- N 8 (Assigned by PSC) Document No. ER981059E201 Rev./ Change No. O Title U/2 Main Turbine Quarterly Valve Stroke Test Deferral and SAR Minor Discrepancy Resolution A WRITTEN RESPONSE PROVIDING THE BASIS FOR THE ANSWER TO E ATTACHED. EACH QUESTION MUST BE ANSWERED SEPARATELY. A SIMP CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES GUIDANCE If the answer to any question on this form is "Yes," then an unreviewed safety question is invo to all questions is "No," then the proposed change does not involve an unreviewed safety question. 1. Will the probability of an accident previously evaluated in the SAR be increased? Yes O No @ Several accidents were deemed to be worthy of review. They are: Loss of Extemal Load and Turbine Trip Turbine Trip with Coincident Failure of Turbine Bypass Valves to Open Turbine Trip with Failure of Generator Breaker to Open Excess Heat Removal Due to Secondary System Malfunction The first three accidents listed are concemed with a turbine trip. The changes in test fre change the probability of the turbine tripping. Moreover, the probability of the turbine to trip is best described as an incident of Moderate Frequency. This means that there is an expectat will occur periodically in the course of plant life. If there is some minute change in the pro turbine tripping, it certainly doesn't significantly change the current probability. l The last accident listed (Excess Steam Removal) is concerned with rapid opening of the tu admission valves when they are supposed to remain close1 The proposed change i method that is done when the valves are open/ stroking. Therefore this accident is not applicabl The minor analysis. discrepancies that are being resolved in this change are not associated with the 2. Will the consequences of an accident previously evaluated in the SAR be increased? Yes O No @ The change involves deferral of the control valve test. The control valves are not critica mitigating radiological exposure to the public. Accident, it is assumed that the radioactive material travels from the steam g . The control valves are not designed to actuate in order to prevent radioactive material f other areas of the plant and potentia!!y re:,ching the public. The stop valves are assumed in some ; accidents to close during the transcient but the reliability of the stop valves is not impac The minor discrepancies that are being resolved in this change are not ae,sociated with t an accident.
- 3. I Will the probability of a malfunction of equipment important to safety be increased?
Yes O No @ The primary concern with this change is that reliability of the control valves to trip m impacted. The main turbine control valves, as well as the stop valves and reheat stop/ inte ,
~ m~. ~ m%, m vac FORM TITLE:
FORM NO. REV. 10CFR60.59 EVALUATION 1000.131B 3 PC 2 ER981059E201 R/0 Page N of 2.5h normally stroked quarterly to ensure that they are not bound up such that they would not trip when demanded to do so. Since the control valves will not be stroked on their normal interval, there is some diminished confidence in the ability of these valves to trip, Ultimately the concern is that the probabil the generation of a missile would increase, which would increase the probability of a malfunction of other equipment (i.e. the target) important to safety. In order to address this concern, the missile generation analysis and other documents were examined to determine the effect of the valve testing. SAR section 3.5 contains the missile analysis for the Main Turbine. The analysis shows that there is a very low probability of missile generation. The value calculated is 6.3E 10. There is a significant source of conservatism in this value due to the fact that the LP turbines were replaced with monoblock rotors in the late 1980's. The missile analysis was not revised to show the decrease in probability associated with the LP monoblocks. A revised P1 probability value was quoted in the SAR as a result of the monoblocks, but the analysis was not revised to include the revised P1 value. The P1 value is one of three values (P1, P2, P3) which are multiplied together to calculate the final probability value P4. When the monoblocks were added, General Electric (GE) provided new information on the probabilities of missile generation. The GE memo's state that some components in the turbine generator (primarily the generator) would fail without causing a missile, but the failed components would cause the turbine generator to slow down. These failures that slow the turbine down occur at speeds icwer than the speed 1 which would result in a missile being generated. GE contends that the installation of the monoblock rotors
- eliminates the potential for any turbine missiles." Reference GE memo from J. Swenson to J. Levine !
dated 6-17-87 contained in DCP-87-2062 R/0 page 51. Certainly, a detailed PSA analysis would calculate a ! specific very low. revised missile probability value for the rmblocks, but the conclusion is that the probability is ! To gain further perspective on these probability values, NRC guideline probability values were compared to the current conservative values. The NRC guideline value for shutting down the plant in 60 days is 1E-3; the conservative curTent value is 6.3E-10. This shows that significant margin exists. Sixty days was considered because this is the time the test will be deferred (i.e. time from schedule tes NRC values were retrieved from DCP 87-2062 R/0 page 89. I The question remains, is there an increase in probability of a turbine missile? GE has been contacted to assist in this question. GE's current position echo's the historicalinformation discussed above. The GE position is that the installations of the monoblocks results in missile generation not being plausible, even if the steam admission valves remain open on a loss of generator load. in addition, it is also noted that the ! control valves are not suspected to have a problem. The reason they will not be stroked is concems with oscillations in the EHC piping. There is no reason to believe that the control valves would not stroke. Another point is that the control valves do move slightly when required to maintain the balance of the turbine and reactor. This is an indicator that the valves are free to move. Lastly, there is redundancy in isolation of the turbine from the steam source. The main stop valves are rcdundant to the control valves with regard to isolating the turbine. The main stop valves will be stroked on the quarterly schedule, so there is hign confidence in the stop valves working. The sum of these observations is that missile generation is highly unlikely, and deferral of the control valve quarterly stroke does not significantly alter the probability. The conclusion is that there is not a significant change in the probabi:ity of missile generation. Likewise the probability of a malfunction (missile strike) of equipment important to safety is unchanged. The minor discrepancies that are being resolved in this change have no affect on the malfunction of equipment important to safety.
- 4. Will the consequences of a malfunction of equipment important to safety be increased?
Yes O No @ If a malfunction of equipment important to safety occurred, the consequences would not be increased as a result of this change. The change involves deferral of the control valve test. As previously mentioned, the
l ARKAMsAs NUCLEAR ONE FORM TITLE: FORM NO. REV. 10CFR50.59 EVALUATION 1000.131B 3 PC.3 I
- ER981059E201 R/0 Page 17 of th in the Steam Generator Tube Rupture Accident, it is assum steam Denerators to the condensers. The control valves are not designed to actuate in o radioactive material from spreading into other areas of the plant and potentially reaching th
- stop valves valves are assumed is not impacted in some accidents to close during the transient, but the reliab by this change.
l The minor discrepancies that are being resolved in this change are not associated with the an accident. 5. Will the possibility of an accident of a different type than any previously i evaluated in the SAR be created? Yes O No @ The change being implemented is simply a change in test interval. This is no physical change to the equipment and no change in the manner that it is operated. It is not plausible that another acc scenario would result. discussed in section 3.5 cf the SAR.The current accidents consider that the turbine trips an The only possibility of a new accident is the potential of a control valve and stop valve sticking open causing excessive steam to be removed from the S/G, but enveloped by consideration of secondary steam line breaks. 6. Will the possibility of a malfunction of equipment important to safety of a . different type than any previously evaluated in the SAR be created? ) Yes O No @ l l The nature of this change is such that the plant is not modified. The form, fit and function o have not changed. There are no changes in interfaces with other systems. Since system functi performance remains the same, an alternate malfunction is not plausible. The applicable equipment malfunction is turbine missile generation that is already discussed in the SAR.
- 7. i Will tne margin of safety as defined in the basis for any technical specification be reduced?
Yes O No @ The valve stroke requirements are no longer included in the technical specifications. Likew of safety are included in the TS basis that depends on the control valves. An integral key in remov valve stroke tests from the TS was Generic Letter 95-10. In this document, the NRC states that turbi overspeed protection is not a primary success path to mitigate design basis accidents. Based on omission of any margins, there is no change to the margins as a result of this activity, bh M 0kWP A& EASONEr 5
~ Certified Reviewers Signature ~
Printed Name Date Reviewers certification expiration date:_ bf Assistance provided by: Printed Name Scope of Assistance Date PSC review by: Date: Ui #d'
~~
FORM TITLE: ARKANSAS NUCLEAR ONE _
- Paga 1 FORM NO. REV.
10CFR50.59 DETERMINATION 1000.131 A 3 PC-1 This Document contains 3 Pages. Document No. _ ER991335E201 Rev> Change No. O Title SKID MOUNTED VALVES AND AUXIL1ARY Brief description of proposed change:
\
The Unit 2 Plant Heating Boiler burner and associated controls were replaced under ER98071
. Some of the vendor Assign components, additional component numbers, update applicable i a da s
Will the proposed Activity: 1. Require a change to the Operating License including: Technical Specifications (excluding the bases)? Operating License? YesO NoS
- l YesO No@ l Confirmatory Orders?
- 2. YesO No@
Result in information in the following SAR documents (including drawings and text) (a) no longer true or accurate, or (b) violate a requirement stated in the document: SAR (multi-volume set for each unit)? Yes@ NoO Core Operating Limits Report? YesO No@ Fire Hazards Analysis? i YesO No@ Bases of the Technical Specifications? YesO No@ Technical Requirements Manual? YesO No@ ! NRC Safety Evaluation Reports? ! 3. YesO No@ Involve a test or experiment not describeo in the SAR? (See Attachment 2 for guidance) YesO No@ 4. Result in a potential impact to the environment? (Complete Environmental impact Determination of this form.) 5. YesO No@ i Result in the need for a Radiological Safety Evaluation per section 6.1.57 3. YesO No@ Result in any potentialimpact to the equipment or facilities utilized for Ventilated Storage Cask activities per Section 6.1.67 7. YesO No@ Involve a change under 10CFR50.54 for the following SAR documents per Section 6.1.77 OAMO? YesO No@ E-Plan? YesO No@
r 1 y /\ 77 /2 32 C M/ WW // \ FORM TITLE: ARKANSAS NUCLEAR ONE Page : FORM NO. ! REV. 10CFR50.59 DETERMINATION 1000.131 A 3 P C-1, 2 Document No. _ ER991335E201 l Rev./ Change No. O Basis for Determination (Questions 1,2, & 3): ER 980711C201 replaced the Unit 2 Plant heating Boildr burnerence andancass supplied with the original boiler. ER 991335E201 a classification and revises the affected P&lD M 2220 Sheet 1. The Ope , and Confirmatory Orders are not affected by the addition of theseagons comp Unit 2 SAR Figure 3.2-2 and thus requries a SAR revision to update , sa the af on of these the environment as noted in the attached Environmen . e o a non-pact to radioactive the VSC Facilities,system the QAMOand or thetherefore, Emergency Plan. do not require a RSE review. The ect ad O appropriate Proposed change does not require 10CFR50.59 Evaluation item #, send LDCR to Licensing). per Atta (If checked, note Search Scope: List sections reviewed in the Licensing Basis Documents specified in questions 1 2
. if search was parentheses. Controlled hard copies of the documen e n es only text, not required. figures or drawings). Attach and distribute a completed LDCR ..
per Sec changes are Document Section LRS: U-2 5059 Index
" Plant Heatino" "Heatino Boiler" "2M-25" "2C-151" "Atomizino Ai MANUAL SECTIONS: SAR Section 3.2. SAR Section 9 4 FIGURES: 32-2
[444d4.- [ _ Bruce Franklin i Certified Reviewers Signature 2/22/99 Printed Name Date Reviewer's certification expiration date. _ 8/24/00 Assistance provided by: Printed Name N/A Scope of Assistance Date Search Scope Review Acceptability (NA, if performed by Technical Reviewer Md
/ /
John Harvey Certified ReviewerTSignature . J2 7f
. Printed Name
- Date
1 I FORM TITLE: A.RKANSAS NUCLEAR ONE
- Pagi 3 10CFR50.59 DETERMINATION FORM NO. REV.
1000.131 A 3 1 ENVIRONMENTAL IMPACT DETERMINATION (UNIT 1 and UNIT 2) , Document No. ER991335E201 RevlChange No. O required. See Section 6.1.4 foradditionalguidance. Complete the , uationfollowi is Willthe Activity being evaluated: I Xidi .No O E Disturb land that is beyond that initially disturbed during construction (\ ! 2.5-17. This applies only to areas outside the protectedj O 8 increase thermal discharges to lake or atmosphere? O E Increase tower? concentration of chemicals to cooling lake or atmospher '.hro O B Increase tower? quantity of chemicals to cooling lake or atmosphere through disc O B Modify the design or operation of cooiing tower whien wiii change drift chara O @ install any new transmission line; leading offsite? O E change the design or operation of the intake or discharge structures? I O 2 Discharges any chemicals new or different from that previously discharged? O @ l Potentially water or groundcause water? a spill or unevaluated discharge which may effect neig i { O E involve surface burying water or water? or ground placement of any solid wastes in the site area which may e O E Involve incineration or dispos &l of any potentially hazardous materials on th O E j Result in a change to nonradiological effluents or licensed reactor power level? ! O E i Potentially ANO site. change the type orincrease the amount of non-radiological air em I i i I
. c li 7 7 /.3 :> D b M MWE U FORM TITLE: ARKANSAS NUCLEAR OteE Page ? ~
10CFR50.59 SAFETY EVALUATION FORM NO. REV. , 1000.1318 3 PC 2 ? This Document contains 1 Page. _ ! Document No. ER1335E201 ! Rev1 Change No. 0 Title 10CFR50.59 Eval, No. % C603 (Assigned by PSC)
. MOUNTED VALVES AND AUXILIARY POWER CABINET ID A WRITTEN RESPONSE PROVIDING THE BASIS UST FOR BE TH i
ATTACHED. EACH QUESTION MUST BE ANSWERED . OF SE CONCLUSION IS NOT SUFFICIENT. ATTACHMENT SE. 2 PR If the answer to any question on this form is "Yes." then an unreviewed safety to all questions is "No," then the proposed change does nvolved. not involve uestion. If the answer an un 1. Willthe probability increased? of an accident previously evaluated in the SAR be YesO No @ 2. Will the consequences of an accident previously evaluated in the SAR be increased? YesO No @ l 3. Will the probability increased? of a malfunction of equipment important to safety be YesO No @ 4 Willthe safety consequences be increased? of a malfunction of equipment important to ! YesO No @ 5. Will the possibility of an accident of a different type than any previously k evaluated in the SAR be created? YesO No @ 6. Willthe possibility of a malfunction of equipment important to safety of a ! different type than any previously evaluated in the SAR be created? YesO No @ . 7. Will the margin of safety as defined in the basis for any technical specification be reduced? YesO No @ d/6 [ Bruce Franklin Certified Reviewer's Signature 2/22/99 Printed Name Date Reviewer's certification expiration date: 8/24/00 l Assistance provided by: Printed Name N/A Scope of Assistance _ Date PSC review by: O Date: $ , hD
FORM TITLE: ARKANSAS NUCLEAR ONE Ptge 1
,10CFR50.59 REVIEW CONTINUATION PAGE FORM NO. REV.
1000.131C 3 Document No. ER991335E201 RevlChange No. 0 10CFR50.59 Review Continuation Pace 1. during normal operation and during periodsepsystem of sh ant the administrative control of manual isolation valves, wh
. The addition of an affect the system's ability of providing containmen previo.usly evaluated in the SAR will not be increased.
~
2. The only safety function of the plant heating system is to provide containm . requir6d to be manipulated during unit outages and the. system to provide containment isolation. evaluated in the SAR will not be increased. v ously There 3. { The et]anges and addition of new components ves. to on of the cabinet to safety. menual containment isolation valves. The installation to the plant heating boiler will not increase the probability of a ary power of tw 1 t 4. The installations of additional components to the heating boiler are e to heatingto safety. important system. These additions can not increase the consequences ent of 5. heating boiler will not create the possibility of an evaluated in the SAR. 6. The installation of a new auxiliary power cabinet and two atomizing er can air l 7. safety of a different type than any previously eva The plant heating system is a non-safety-related system and is not a inclu Specifications. not be reduced. Therefore, the margin of safety as defined in the basis n will for a O em 6
FORM TITLE: FORM NO. REV.
, 10CFRSO.59 DETERMINATION 1000.131A 2 PC-2.3 Pagelof 4 ,
Document No. LCP 95-8078L201 Rev/ Change No. 0 Title ANO-2 Replace 2T10 level transmitters 2LT-5658. 2LT-5674. 2LT 5659-1 and 2LT-5668-2. Will the proposed Activity: - 1. Require a change to the Operating License including: Technical Specifications (excluding the bases)? YesONo@ Operating License? YesONo@ Confirmatory Orders? YesONo@ 2. Result in information in the following SAR documents (including drawings and text) being (a) no longer true or accurate, or (b) violate a requirement stated in the document: Core Operating Limits Report YesONo@ SAR (multi-volume set for each unit)? Yes@ nod QAMO?* ' YesONo@ E-Fian?* YesONo@ l FHA l YesONo@ l Bases of the Technical Specifications? I YesONo@ NRC Safety Evaluation Reports? YesONo@ 3. Involve a test or experiment not described in the SAR? YesONo@ 4. Result in a potentialimpact to the environment? (Complete Environmental Impact Checklist of this form.) ! YesONo@ ' 5. Result in the need for a Radiological Safety Evaluation per section 6.2.4.A? i YesONo@
- 6. 1 Result in any potentialimpact to the equipment or facilities utilized for Ventilated Storage Cask activities per Section 6.2.4.8 YesONo@ l Basis for Determination:
See continuation page. Changes to these documents require an evaluation in accordance with 10CFR50.54. See Seetion 6.2.1.B. t 9
}
FORM TITLE: ARKANSAS NUCLEAR ONE FORM NO. REV. 10CFRSO.58 DETERMINATION 1000.131A 2 PC-2,3 Page 2 of 4 Document No. LCP 05-8078 RevjChange No. O
References:
List sections reviewed in the Licensing Basis Documents, specified ,
. If a in que in parentheses. Controlled hard copies of the do searches such as LRS are not controlled and search text only, not figures completed LDCR if LBD changes are required.
Document Section U2 Tech. Specs U2 OP. License ALL (NAOH, Tank 2T10, Transmitters), including sections 3.6.2.2 and 4.i ALL (NAOH, Tank 2T10, Transmitters) U2 Confirm. Orders U2 SAR ALL (NAOH, Tank 2T10, Transmitters) ALL3. (NAOH, and Tank 2T10, Transmitters), including section 6.2.3 and . QA Manual E-Plan ALL (NAOH, Tank 2T10, Transmitters) FHA ALL (NAOH, Tank 2T10, Transmitters) U2 TS Bases ALL (NAOH, Tank 2T10, Transmitters) NRC SER ALL (NAOH, Tank 2T10, Transmitters) ALL (NADH, Tank 2T10, Transmitters) Certified Reviewers Signature NICK MEHTA 05-31-97 Printed Name Date Reviewers certification expiration date: 05-05-99 Assistance provided by: Printed Name Scope of Assistance Date i 9 __W
I FORM TITLE: FORM NO. REV.
, 10CFR50.59 REVIEW CONTINUATION PAGE 1000.131c 2 PC.2,3 l
ENVIRONMENTAL IMPACT CHECKOST (UNIT 1 and UNIT 2) Document No. LCP 95-8078L201 RevdChange No. O Complete the following See Section 6.2.1.E checklist. for additional guidance.If the answer to any checklist item is "Yes", an Environmenta . Willthe Activity being evaluated: Yes No O E Disturb land that is beyond that initially disturbed during construction (i.e., new construction of l buildings, creation or remeval of ponds, or other terrestrial impact)? See Unit 2 SAR Fig This applies only to areas outside the protected area. O a increase thermal discharges to lake or atmosphere? O @ Increase concentration of chemicals to cooling lake or atmosphere through discharge cana O E Increase quantity of chemicals to cooling lake or atmosphere through discharge canal or to O B Modify the design or operation of cooling tower which will change drift characteristics? O E Install any new transmission lines leading offsite? O @ Change the design or operation of the intake or discharge structures? O O Discharges any chemicals new or different from that previously discharged? O @ Potentially cause a spill or unevaluated discharge which may effect neighboring soils, surface or ground water? O @ Involve burying or placement of any solid wastes in the site area which may effect runoff, surf water or ground water? O E involve incineration or disposal of any potentially hazardous materials on the ANO site? O E Result in a change to nonradiological effluents or licensed reactor power level?
.O E Potentially change the type or increase the amount of non-radiological air emissions from the ANO site.
FORM TITLE: ARMANsAs NUCLEAR ONE FORM NO. REY. 10CFR80.50 REvlEW CONTINUATION PAGE 1000.131C 2 PC.2,3 Page 4 off Document No. LCP 958078L201 Rev> Change No. 0 10CFR50.59 Review Continuation Pace This design change will replace existing obsolete Foxboro sodium hydroxide j transmitters 2LT-5658,2LT-5674 with Rosemount model 3051CD transmitters a 3 T-5659-1,2LT-5668-2 with Rosemount model 1154DPSRB transmitters. The
.voquently found out of calibration. The wide range loops are used for technical spec
- verification of 2T10 and also stop the sodium hydroxide pumps and close tank event of an accident. these loops are also used for space heater interlocks. The and tank heater interlocks. For the narrow range transmitters 2LT-5658 and by the fact that the amount of zero suppression required exceeds that allowed 91-E-0003 01 Rev.1 (calibration and level setpoints analysis)'provides the basi will be calibrated and changes to level heater interlock setpoints. Loop error ca has been revised for the new transmitters by this design change.
for any functions of these levelloops. None of the cont spray are derived from these loops. commercial grade and dedicated to Q application by Procureme 5659-1 and 2LT 5668 2 are Q-active and procured under Q requirements. mounted seismically, All the new transmitters will be replacement of the existing Foxboro transmitters with the Ro the heater not affect interlock the system functions. setpoints due to the error increases with the new Rosem . could not be met due to calibration discrepancy. This transmitters 2LT-5659-1 and 2LT-5668-2 to be in agreement with Tech. Spec. requirements. The roplacement transmitters will operate within the existing technical specifications Rosemount model 3051CD and 1154DP transmitters will operate the sameThe . as the BASIS OF DETERMINATION: 1 The ANO-2 Technical Specifications, Operating License and Confirmatory Orders the level of detail to address this modification. specific requirements for 2. This modification will have impact on Unit-2 SAR. Figure 7.3.7 Sh. 3 (M-2422 Sh. 3), Fu Description and Logic Diagram Containment Spray System. Setpoint for the heater by this modification. High and low level alarms were revised fer clarity on the drawi of setpoint evaluallon was willbe not altered. Since SAR figure is affected by this design change, an 10CF performed. 1 1
- 3. i The testing required by this plant change consist of normal instrument calibration a; {
or experiments not described in the SAR are involved in this plant change. - 4. The environmental checklist 'was reviewed and this plant change does not have a . spill or unevaluated discharge" was considered. Howev transmitter can not cause a spill and an overflow of this tank is already evaluated . 5. This plant change will not affect the requirements of section 6.2.4.a of Procedure # 1000 6.'
- This plant change will not affect the requirements of section 6.2.4.b of Procedure # 1000,13 i
m
, FORM TITLE; ARKANSAS NUCLEAR ONE FORM No. REV.
10CFRSO.88 EVALUATION 1000.131B 2 ! l 1.C P 9 58 07 81.201 Page 1 ofl PAGE " REV 0 10CFR50.59 Evar. No. F FS W } T (Assigned by PSC) Document No. LCP 958078L201 RevJChange No. 0 i Title ANO-2 Reclace 2T10 Level Transmitters 2LT4658,2LT4674. 2LT4659-1 and 2LT4668-2 A WRITTEN RESPONSE PROVIDING THE BASIS FOR THE ANSWER TO E ATTACHED. EACH QUESTION MUST BE ANSWERED SEPARATELY. A SIMP : CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES GUIDANCE! if the answer to any question on this form is "Yes," then an unreviewed safety question is involv to all questions is "No," then the proposed change does not involve an unreviewed safety question. 1. Will the probability of an accident previously evaluated in the SAR be increased? Yes O No @ The change to the heater interlock setpoints on the narrow range transmitters (2LT-5674 & 5658) on the Sodium Hydroxide Tank (2T10) will not increase the probability of an accident l evaluated in the SAR. These transmitters are Q-passive (i.e. pressure retaining bounda since the safety analysis does not take credit for any functions of these level loops. Since th narrow range transmitters are not credited for any safety function except pressure boundary th change in setpoint will not have any affect on an accident probability. New wide range Rosemount 1154DP transmitters will operate the same as the existing Foxboro transmitters safety functions of the containment spray system will not be affected by this replacements. 2. Will the consequences of an accident previously evaluated in the SAR be increased? Yes O No @ The change to the %4 der interlock setpoints on the narrow range transmitters (2LT-5674 5658) on the Sodio Hydroxide Tank (2T10) will not increase the consequences of an accid evaluated in the SAR. The only safety function that these transmitters serve is as a press retaining boundary. The new Rosemount transmitters have been evaluated to meet the pr retaining requirements and thus will not potentially leak any more than the current Foxboro transmitters. Therefore the off-site dose consequences due to this change will not be increased. 3. Will the probability of a malfunction of equipment important to safety be increased? Yes O No @ Changing the heater interlock setpoints on the narrow range transmitters (2LT-5674 will not change the probability of an equipment malfunction. Also the new Rosemount transmitters have been evaluated and commercially dedicated to meet the requirements of pressure retaining equipment. This evaluation determined that the equipment will maintain its integrity under_ design conditions. Therefore the probability of equipment malfunction does not enclosure. exist. These transmitters will be mounted seismically in side existing weather proof
FORM TITLE: ARKANSAS NUCLEAR ONE 10CFR80.89 EVALUATION FORM NO. REV. 1000.131B 2 4. Will be the consequences increased? of a malfunction of equipment important to safety Yes O No @ The new Rosemount transmitters have been evaluated and commercia requirements of pressure retaining equipment. This evaluation determin maintain its integrity under design conditions. With no additional potentialI transmitters (2LT-5674 & 2LT-5658) will not cause . 5. dose consequences due to malfunction of equipment is associated wi . Will the possibility of an accident of a different type than any previously evaluated in the SAR be created? Yes O No @ The changes to heater interlock setpoints on the narrow range transmit in the narrow range setpoints is not an accident initiator. pos Also the change in transmitters themselves not created. does not initiate an accident. Therefore the possi
- 6. PAGE .'n REV 0 Will the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated l' 'he SAR be created?
Yes O No @ The new Rosemount transmitters are used extensively throughout transmitterstransmitters These are Q-Passive. are more reliable than the Foxboros they are rep enclosures. The new transmitters will be calibrated per m a malfunction of equipment important to safety of a in the SAR will not be created,
)
i I 7. Will the margin of safety as defined in the bases for any technical specification be reduced? Yes O No @ fluid. Therefore, the margin of safety as defined in not be reduced. (b . W < % M . Certified Reviewers Signature Nick Mehta 05-31-97 Printed Name Date RGviewers certification expiration date: 05-05-97 Assistance provided by: Printed ames Scope of Assistance Date PSC review by A Date: (* i
ARKANSAS NUCLEAR ONE Page 2 FORM TITLE: FORM No. REV. 10CFR60.59 DETERMINATION 1000.131 A 3 PC-1,2 Document No. DRN 98-02670 Rev./ Change No. O Title Add spectacle flanges to P&ID Drawing M-2220 Sheet 1 ( Brief description of proposed change: ANO Engineering Programs personnel initiaterl ER %2018 documenting recuning difficulties in the performa testing of Unit 2 Containment isolation valves 2PH-44 and 2PH-23 on the Plant Heating System. Design group responded to the ER by developing me.fification package 962018L201 installing spectacle flanges piping 2JBD 105-1 and 2JBD-106-1. Engiacering evaluated the design requirements during the modific development and set the following design basis. Installation of the spectacle flanges would permit rea testing isolation and still maintain normal system configuration. Resiew of design general engineering stand System Piping and Instrument Diagram, determined it was not necessary to depict the spectacle flanges drawings. DRN 98-2670 adds the spectacle flanges to the affected P&ID drawin;l due to the system being l configuration flanges are blankedthan off. identified in the change package. A note of clarification is also added to the P&ID dra Will the proposed Actisity: 1
- 1. Require a change to the Operating License including:
Technical Specifications (excluding the bases)? . YesO No@ Operating License? YesO NoS Confirmatory Orders? YesO NoS 2. Result in information in the following SAR documents (including drawings and text) being (a) no longer true or accurate, or (b) violate a requirement stated in the document: SAR (multi-volume sat for each unit)? Yes@ NoO t Core Operating Limits Report? YesO No@ Fire Hazards Analysis? YesO No@ Bases of the Technical Specifications? YesO No@ Technical Requirements Manual? YesO No@ NRC Safety Evaluation Reports? YesO NoS
- 3. Involve a test or experiment not described in the SAR?
(See Attachment 2 for guidance) YesO No@
- 4. Result in a potentialimpact to the environment? (Complete Environmental Impact Determination of this form.)
YesO No@ !
- 5. Result in the need for a Radiological Safety Evaluation per section 6.1.S?
YesO No@
- 6. Result in any potentialimpact to the equipment or facilities utilized for Ventilated Storage Cask activities per Section 6.1.6?
YesO No@
- 7. Involve a change under 10CFR50.54 for the following SA R documents per Section 6.1.77 QAMO?
YesO No@ E-Plan? YesO NoS
~ - -
FCRM TITLE: ARKANSAS NUCt J( ONE Page a FORM No. REV. 10CFR50.89 DETERMINATION 1000.131A 3 PC-1,2 Dannent No. DRN 98-02670 Rev) Change No. O BJs for Deteradaation (Qa=atlana 1,2, & 3):
%c ANO Unit 2 Plant Heatmg System is Non-Q, Non-Safety Related. He only safety func for the isolation of the cantninment building via the containment penetration isolation valves. R Ito show the inse=Had spectacle flanges will not cause a change to the operating license. Chan information in the SAR documents being no longer true or accurate as P&lD drawing pceM-2220 in the SAR as Figur* 2.0 2. Blanking the spectacle flanges does not impact the plant heatin as the portiori locatS :.L/a of containment is normally isolated by containment isolation valves. A no P&lD willidentify tra need to realign the flanges should the system need to be operated. An LDC
' Initiated to revise the affected SAR Figure 3.2 2. Adding the spectacle flanges to the a does not involve any test or experiment not discussed in the Unit 2 SAR. Reusing the Plant H P&lD will not cause a potential impact to the envimnment nor willit result in '.no need fo evaluation. Drawing M-2220 Sheet i revision will not result in any potentia' impact to the VSC equ facilities. The drawing revision does not affect the QAMO orthe E-Plan.
O Pmposed change does not require 10CFR50.59 Evaluation per Attachment 1, Item # ited, send LDCR to Li-'a$ (Ifchecked, note appropriate Semich " :;2 List sections reviewed in the Licensmg Basis Documents specified in questions 1, 2 and 3. the LRS search index should be entered under "Section" with the search statement (s) used in paren copies of the documente shall be reviewed (LRS is not verified and searches only text, not figures o distribute a completed LDCR per Section 6.1.2 If LBD changes are required. Document Engigm LRS: 5059 Unit 2 Tah "AT T " with keywords (" Plant I!=:ia=". "C;=*=en: Temo*" "C.- ac.,=a--- gg,Ja*" "P == Bni1Aine Environ *" "P
=;=.t IE :ina" Isolation" 2 ter Buildia ". "Caa'ainment At===Me" "Cantnin=sa:
MANUAL SECTIONS: U-2 SAR 9.4. 9.4.5 FIGURES: 1,2-2 8MI C noed Reviewer's Signature Bruce Frankhn 9/24/98 Printed Name Date Reviewer's cemfication expiration date: 8/24/2000 Amaissance provided by: ( Prmted Name Scope of Assistance N/A Date Search I Review Am,,tability (NA, if g. formed by Technical Reviewer per 1000.006) [/ M John Harvey 9/24/98
,0tiMlfledR., J(ufignature Pnnted Name . Date
l ARKANSAS NUCLEAR ONE Page 4 FORM TITLE:
- FORM NO. REV.
10CFR50_59 DETERMINAllON 1000.131 A 3 ENVIRONMENTAL IMPACT DETERMINATION (UNIT 1 and UNIT 2) I i Document No. DRN 98-02670 Rev/ Change No. O Complete the following Determmation. If the answer to any item below is "Yes", an Emironmental Evaluation is r See Section 6.1.4 for additional guidance. Willthe Activity being evaluated: Ytf En O E Disturb land that is beyond that initially disturbed during construction (i.e., new construction of buildings, creation or removal of ponds, or other terrestrialimpact)? See Unit 2 SAR Figure 2.5-17. This applies only to areas outside the protected area. O E Increase thermal discharges to take or atmosphere? O 2 Increase concentration of chemicals to cooling lake or atmosphere through discharge canal or tower? O E Increase quantity of chemicals to cooling lake or atmosphere through discharge canal or tower? O 2 Modify the design or operation of cooling tower which will change drift cluracteristics? O E Install any new transmission lines leading ofrsite? O E Change the design or operation of the intake or discharge structures? O @ Discharges any chemicals new or different from that previously discharged? O 2 Potentially cause a spill or unevaluated discharge which may effect neighboring soils, surface water or ground water? i O E Involve burying or placement of any solid wastes in the site area which may effect runoff, surface water i or ground water? O S Involve incineration or disposal of any potentially hazardous materials on the ANO site? O 2 Result in a change to nonradiological effluents or licensed reactor power level? O G Potentially change the type or increase the amount of non-radiological air emissions from the ANO site.
FORMb LE: ARKANSAS NUCLEAR ONE Pagi 5 FORM No. REV. 10CFR50 59 SAFETY EVALUATION 1000.131B 3 PC-2 Document No. _ DRN 98-02670 RevlChange No. 0 10CFR50.59 Eval. No. Fc'tdc12-two Title Add spectacle flanges to P&ID Drawing M 2220 Sheet 1 (Assigned by PSC) A WR111 r.N RESPONSE PROVIDING THE BASIS FOR 'IIIE ANSWER TO EA ATTACHED EACH QUESTION MUST BE ANSWERED SEPARATELY A SIMPLE CONCLUSIONIS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES GUIDANCE F If the answer to any question on this form is "Yes," then an unreviewed safety question is invo questions is "No," then the proposed change does not involve an unreviewed safety question. 1. Will the probability of an accident previously evaluated in the SARbeincreased? - Yes O No @ 2. Will the consequences of an accident previously evaluated in the SARbe increased? Yes O No @ 3. Will the probability of a taalfunction of equipment important to safety beincreased? Yes O No @ 4. Will the consequences of a ==Waaion of equipment important to safety beincreased? Yes O No @ 5. Will the possibility ofan accident of a different type than any previously evaluated in the SAR be created? Yes O No @ 6. Will the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SARbe created? Yes O No @ 7. Will the margm of safety as defined in the basis for any technical specification be reduced? Yes O No @ 8A4ME Im Bruce Fr=Hin 9/24/98 Certified Reviewer's Signature Printed Name Date ~! Reviewer's certification expiration date: 8/24/2000 Assit*ance provided by: Printed Name l Scope of Assistance N/A Date i PSC reviewby: htb-- Date: 3 M Il
Document No. _ DRN 98-02670 RevlChange No. _0 10CFR50.59 Review CantinaWn Pane 1. The ANO Unit 2 Plant Heating System is a non-safety related Seismic Class II system. The o system has is to prmide contamment isolation. Limited Change Package LCP 962018L201 add the plant heatmg supply and return lines inside contamment for LLRT testing isolation purpo flanges to the affected P&ID drawing will not increase the probability of an accident previously ev 2. The blankest flanges inside contamment provide addition margin for contamment isolation. D flanges on the affected P&ID will not affect the system safety function of providing contammen the consequences of an accident previously evaluated in the SAR will not be increased 3. The Plant Heating System has no safety function other than providing contamment isolation. is normally isolated in the contamment building. Isolating the plant heatmg supply and retum lines insi contamment does not affect system operation. The blanking of the piping inside of contamment will n operation of the contamment isolation valves. Dus the probability of a malfunction of equipme will not be increased. 4. Instalhng blank flanges in the Plant Heating System supply and return lines inside of containmen system isolation through the contamment penetration. He spectacle flanges can not impact th contamment isolation valves. Therefore, the consequences of a malfunction of equipment important mereased. ' 5. The plant heating system has no safety function other than providing containment isolation; in th olant heatmg lines inside of the contamment building can not initiate an accident ofr :ident of.a different type than any previously evaluated in the SAR will not be created. ) l 6. The taankmg of spectacle flanges in the plant heating system inside of containment has been 10CFR50.59 evaluation in design change package 962018L201. De plant heating system has no s Herefore, the possibility of a malfunction of equipment important to safety of a different type than in the SAR will not be created. 7. De plant heatmg system has no safety function and is not utilized to mitigate the consequences o condition. The blanking off of the plant heatmg piping inside of containment can not affect the contamme function. Herefore, the margin of safety as defined in the basis for any technical specification will not
ARKAN #AS NUCLEAR ONE Page 2
' FORM TITLE:
FORM NO. REV. 10CFR50.59 DETERMINATION 1000.131 A 3 PC-1 Document No. PAGE s$ . _ REV.O LCP963205L201 RevlChange No. O Title Removal Of U-2 EDG Exhaust Rain Hoods And installation Of Drains On Silence Brief description of proposed change: This LCP will remove the Rain Hoods from the Emergency Diesel Generator Exhaust Stacks, add around the top section of each stack and install drain lines for the removal of any water that accumulates in the exhaust silencers. The Rain Hoods are not capable of sustaining a Tomado Missile without potentially compromising the EDG exhaust capability. The stiffener rings will maintain the functionality of the exhaust stacks when hit by tomado missiles. Removing these Rain Hoods will also allow the EDG exhaust gas t vertically at a higher velocity thus reducing the occurrence of diesel exhaust fumes entering the unit 2 contr room ventilation system. (Ref. CR 2-95-0097, CR C-97-0275 and Calc No. 88-E-0040-41) Will the proposed Activity:
- 1. Require a change to the Operating License including:
Technical Specifications (excluding the bases)? YesO No@ Operating License? - YesO No@ Confirmatory Orders? YesO No@ 2. Result in information in the following SAR documents (including drawings and text) being (a) no longer true or accurate, or (b) violate a requirement stated in the document: SAR (multi-volume set for each unit)? Yes@ NoO Core Operating Limits Report? YesO No@ Fire Hazards Analysis? YesO No@ Bases of the Technical Specifications? YesO No@ Technical Requirements Manual? YesO No@ NRC Safety Evaluation Reports? : YesO No@
- 3. Involve a test or experiment not described in the SAR?
(See Attachment 2 for guidance) YesO No@
- 4. Result in a potential impact to the enviro'nment? (Complete Environmental Impact Determination of this form.)
YesO No@ 5. Result in the need for a Radiological Safety Evaluation per section 6.1.57 YesO No@
- 6. Result in any potentialimpact to the equipment or facilities utilized for Ventilated Storage Cask activities per Section 6.1.6?
YesO No@
- 7. Involve a change under 10CFR50.54 for the following SAR documents per Section 6.1.77 QAMO?
YesO No@ E-Plan? YesO No@ I i
ARKANSAS NuCLEAS ONE FORM TITLE: Pace G FORM NO. REV. 10CFR50.59 DETERMINATION 1000.131 A 3 PC-1 Document No. LCP963205L201 RevdChange No. O Basis for Determination (Questions 1,2, & 3): pggg - ggy g Based on electronic and manual searches, modifying the diesel exhaust stacks by remoI installing stiffener rings and adding drain lines will not affect the plant license. No Technical Sl Confirmatory Orders are impacted by the changes involved in this Modification. A 50.59I based on the LBD changes noted in the Search Scope section. This Modification will not change the design or function of the Emergency Diesel System Two SAR figures (3.5-26 and 9.5-8) will be revised to incorporate the diesel silencer drain ! of the exhaust stack rain hoods. The changes to the EDG exhaust stacks do not involve any! Specification Bases. This modification has no impact on any other SAR documents. i These changes to the Emergency Diesel Exhaust System do not involve a test or exper The activities attached associated Environmental with this Impact Checklist. LCP do not create a potential impact to the environ O Proposed change does not require 10CFR50.59 Evaluation per Attachment(If1,checked, item #note appropriate item #, send LDCR to Licensing). I Search Scope: List sections reviewed in the Licensing Basis Documents specified in questions 1,2 and 3 was done on LRS, "all" may be entered under "Section" with the keyword (s) used in parenth Attach and distribute a completed LDCR per Section 6.1.2 If L ' l Document Section LRS: ALL (EMERGENCY DIESEL GENERATOR), (DIESEL EXHAUST), (MISSILE *), (TORNADO) MANUAL SECTIONS: ANO-2 SAR Sections 3.3.3.5.3.5.7.1.1.2.8.3.9.4.2.9.5.15.1.9 ANO-2 Tech Specs. Table of contents ALL. 3/4 8 FIGURES: Fio 3.5-3. 3.5-4. 3.5-6. 3.5-7. 3.5-7. 3.5-8. 3.5-8A. 3.5-26. 9.5-8 C. Joseph C. King Jr. Ce$ified l$ viewers Signa)ur 1-12 98 Printed Name Date Reviewers certification expiration date: 11-14-99 Assistance provided by: Printed Name Scope of Assistance Date Search Scope Review Acceptability (NA, if performed by Technical Reviewer per 1000.006) J l C rtifieJ Reviewers Signature J4,, ,s Q em 2-24-9s Printed Name Date
ARKANSAS NUCLEAR ONE Page 4 FORM TITLE: FORM NO. REV. 10CFR50.59 DETERMINATION 1000.131 A 3 ENVIRONMENTAL IMPACT DETERMINATIONPAGE- 5 - asy, o (UNIT 1 arid UNIT 2) Document No. LCP963205L201 _ RevdChange No. O Complete the following Determination. If the answer to any item below is "Yes", an Environmental Evalu required. See Section 6.1.4 for additional guidance. Will the Activity being evaluated: Yes .N_g I O 3 Disturb land that is beyond that initially disturbed during construction (i.e., new construction of buildings, creation or removal of ponds, or other terrestrial impact)? See Unit 2 SAR Figure 2.5-17. This applhs only to areas outside the protected area. O @ increase thermal discharges to lake or atmosphere? O E increase concentration of chemicals to cooling lake or atmosphere through discharge canal or
\
tower? O @ increase quantity of chemicals to cooling lake or atmosphere through discharge canal or tower? l O E Modify the design or operation of cooling tower which will change drift characteristics? O O install any new transmission lines leading offsite? O O Change the design or operation of the intake or discharge structures? O E Discharges any chemicals new or different from that previously discharged? O 3 Potentially cause a spill or unevaluated discharge which may effect neighboring soils, surface water or ground water? O @ Involve burying or placement of any solid wastes in the site area which may effect runoff, surface water or ground water? O @ involve incineration or disposal of any potentially hazardous materials on the ANO site? O B Result in a change to nonradiological effluents or licensed reactor powerlevel? O @ Potentially change the type or increase the amount of non-radiological air emissions from the ANO site.
1 j FORM TITLE: ARKANsks NUCLEAR ONE Page 1 FORM NO. REV. 10CFR50.59 SAFETY EVALUATION 1000.131B s 3 !
/
W, This Document contains 2 Pages. PAGE 8tEV. O ' Document No. LCP 963205L201 Rev> Change No. 0 10CFR50.59 Eval. No. N ' C C l*)I Title (Assigned by PSC) Removal Of U-2 EDG Exhaust Rain Hoods And installation Of Drains O _ A WRi it:N RESPONSE PROVIDING THE BASIS FOR THE ANSWER ATTACHED. EACH QUESTION MUST BE ANSWERED SEPARATELY. A CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES GUID . to all questions is *No," then the proposed change does not .
- 1. !
Will the probability of an accident previously evaluated in the SAR be increased? YesO No @ This LCP will remove the Emergency Diesel Generator Exhaust Stack Rain Hoods the top section of the Stacks and install drain lines on Silencers 2M-67A and B. With the the EDG Exhaust Stacks can meet the challenge of a SAR defined Tomado Missile. A removal of the Rain Hoods will reduce the problem of diesel exhaust fumes entering the contr ventilation system. SAR Figures 3.5-26 (Dwg M-2065) and 9.5-8 ( P&lD Dwg M-2217) will incorporate these changes. These modifications to the diesel exhaust system do not imp performance, operation or function of the EDG system. Therefore, these improvements to th exhaust system will not increase the probability of an accident previously evaluated in the SA 2. Will the consequences of an accident previously evaluated in the SAR be increased? YesO No @ This LCP modifies the EDG exhaust system in order to meet the challenge of a SAR def i Missile and reduce the occurrence of the intrusion of diesel exhaust f{ These changes will not increase the consequences of any a 3. Will the probability of a malfunction of equipment important to safety be increased? ' YesO No @ The modifications to the EDG exhaust system by this LCP will be in accordance with design requirements. The removal of the exhaust stack rain hoods and installation of s eliminate the potential of loss of function of the EDGs due to Tomado Missiles. Also, the rem Rain Hoods willincrease the exit velocity of the exhaust gases from the stacks which will redu potential for the intrusion of diesel exhaust fumes into the control room. These improvements to th exhaust piping configuration willinsure the availability of the Emergency Diesel Generator LCP will not increase the probability of a failure of equipment important to safety. 4. Will the consequences of a malfunction of equipment important to safety be increased? YesO No @ The scope if this LCP is limited to improvements to the EDG exhaust piping. These impro impact the operation or function of any other safety equipment or safety systems. Be will not impact any other systems or equipment, it is concluded that the consequences of equipment important to safety are not increased by the installation of this LCP. l DOCUMENT NUMBER ER-963205L201 l l
ARKANSAS NUCLEAR ONE I FORM TITLE: Pace 2 FORM NO. REV, 10CFR50.59 SAFETY EVALUATION 1000.1318 3
- 5. Will the possibility of an accident of a different type than any previously evaluated in the SAR be created?
YesO No @ This LCP will only remove the EDG exhaust stack rain hoods, add stiffener rings at the top of the stacks and install drain lines. This modification does not change the performance or function of the EDG sys No existing SAR safety evaluations are impacted by this modification. No new accidents will be introduce by this modification,
- 6. Will the possibility of a malfunction of equipment important i to safety of a different type than any previously evaluated in the SAR be created?
YesO No @ l The changes provided by this LCP meet the original design specifications for material and construction practices. This LCP does not create any new connections or interrelations between safety related syst or equipment. On this basis, the possibility of a malfunction of equipment important to safety of a different i l type than previously evaluated in the SAR is not created.
- 7. Will the margin of safety as defined in the bases for any technical specification be reduced?
YesO No @ This modification will insure that the EDG exhaust stacks can meet the challenge of a SAR defined Tomado Missile. Therefore, the improvements to the EDG exhaust system will maintain the availability of the Emergency Diesel Generators and maintain the margin of safety for the EDG Units.
/M( -x Joseph C. King Jr. 1-12-98 pertifie( Reviewer's Sig re Printed Name Date Reviewer's certification expiratiorrdate: 11-14-99 Assistance provided by:
Printed Name Scope of Assistance Date PSC review by: Date: l 1 l l l l DOCUMENT NUMBER ER-963205L201 PAGE
/[
REV.O 1 l 1 l i 1 1
974379L201, Rev. o ARKANSAS NUCLEAR ONE Page 1 FORM TITLE: FORM NO. REV. 10CFR50:59 DETERMINATION 1000.131 A 3 PC-1,2
, 3 eo This Document contains a eaoes.
Document No. 974379L201 RevdChange No. O Titic U-2 VENTS AND DRAINS MODIFICATIONS Brief description of proposed change: This . modification package is to resolve the remaining ANO Unit-2 vent / drain line code compliance issues in response to CR-C-97-0084. The four remaining problem vent / drain lines are double valve isolation type D as shown in M-2555, Appendix C. These four vent / drains will be corrected by removing the outboard valves (2F 2003B,2EFW-3034B,2FW-2009 and 2EFW-10678), installing threaded caps and enchancing the existing sockets welds. Stress Calculation 974379L201-03 qualified these piping modifications in accordance with j ASME Section ill, Class 2 Code. The LCP provides the basis for the use of the threaded caps, which is l acceptable per ASME Subsection NC-3671-3. These changes will significantly increase the pipe resistance to vibration and resolve these code compliance problems. Will the proposed Activity:
- 1. Require a change to the Operating License including:
Technical Specifications (excluding the bases)? YesO No@ Operating License? YesO No@ Confirmatory Orders? YesO No@ 2. Result in information in the following SAR documents (including drawings and text) being (a) no longer true or accurate, or (b) violate a requirement stated in the document: 1 SAR (multi-volume set for each unit)? Yes@ NoO Core Operating Limits Report? YesO No@ Fire Hazards Analysis? YesO No@ Bases of the Technical Specifications? YesO No@ Technical Requirements Manual? YesO No@ IGC Safety Evaluation Reports? YesO No@
- 3. Involve a test or experiment not described in the SAR?
(See Attachment 2 for guidance) YesO No@
- 4. Result in a potential impact to the environment? (Complete Environmental i
Impact Determination of this form.) ! YesO No@
- 5. Result in the need for a Radiological Safety Evaluation per section 6.1.5?
YesO NoS
- 6. Result in any potential impact to the equipment or facilities utilized for Ventilated Storage Cask activities per Section 6.1.67 YesO No@ I
- 7. Involve a change under 10CFR50.54 for the following SAR documents l per Section 6.1.77 l QAMO?
YesO No@ E-Plan? YesO No@ l
974379L201. Rev. O ARKANSAS NUCLEAR ONE Pace 2 F&RM TITLE: FORM NO. REV. I 10CFR60.59 DETERMINATION 1000.131 A 3 PC-1,2 Document No. 974379L201 RevJChange No. O { Basis for Determination (Questions 1,2, & 3): 1. h Based on electronic and manual searches, the changes to these vent / drain valves provided by this Modification will not impact any of the Licensing Basis Documents. { 2. SAR Figure 10.2 3 (P&lD M-2206 SHT 1) will have to be revised to show the changes to these vent / drain valves. These changes meet all code requirements and will not change the basic function of these valves These changes will significantly increase the pipe's resistance to vibration. j 3. The changes provided for these vent / drain valves do not involve any test or experiments. 4. This LCP does not involve any potential impacts to the environment as determined by the environmental review. 5. The piping involved in this LCP is inside the Containment Building. The installation package and the RWP for work in this area will control all activities related to this LCP. 6. This LCP does not involve any potential impacts to any VSC equipment or facilities.
- 7. The modifications of these vent / drains will not impact the QAMO or the E-Plan.
j O Proposed change does not require 10CFR50.59 Evaluation per Attachment 1, item # (if checked, note appropriate item #, send LDCR to Licensing). Search Scope: List sections reviewed in the Licensing Basis Documents specified in questions 1,2 and 3. If search was performed on LRS, the LRS search index should be entered under "Section" with the search stri.ement(s) parentheses. Controlled hard copies of the documents shall be reviewed (LRS is not verified and searche text, not figures or drawings). Attach and distribute a completed LDCR per Section 6.1.2 if LBD changes a required. Document Section l LRS: ALL (efw. emeroency feed *. containment inteority. Dioe cao*. threaded cao. screwed cao. 2DBB*. 2FW-20038. 2EFW-30348. 2FW-2009. 2EFW-10678. ILRT) MANUAL SECTIONS: Tech Soecs ALL (Table Of Contents). 3/4.1.2. SAR 3.5. 3.51. 3.5. 3.6.2. 3.6.4.2.10.4.7.10.4.9. (Tables 3.5-1. 3.5-5A. 3.6-1. 3.6-11. 3.6-12. 3.6-15. 3.6-16) FIGURES: 3.6-39.3.6-40.3.6-44.3.6-45.10.2-3 b P Joseph C. King Jr. 9/28/98 Ce(fied RWviewers Signattire Printed Name Date Reviewers certification expiratibn date: 11/14/99 Assistance provided by: Printed Name Scope of Assistance Date > 4 Search Scope Review Acceptability (NA, if performed by Technica! Reviewer per 1000.006) b- David E. Torgerson Certified Reviewers Signature 10-1-98 Printed Name Date.
I 974379L2ot Rev o ARKANSAS WUCLEAR ONE Page 3 FORM TITLE:
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FORM NO. REV. 10CFR50.69 DETERMINATION 1000.131 A 3 ENVIRONMENTAL IMPACT DETERMINATIOlibAGE 8 REV.0 (UNIT 1 and UNIT 2) Document No. 974379L201 Rev/ Change No. O Complete the following Detemiination. If the answer to any item below is "Yes", an Environmental Evaluation is required. See Section 6.1.4 for additional guidance. Will the Activity being evaluated: 121 Ng O E Disturb land that is beyond that initially disturbed during construction (i.e., new construction of buildings, creation or removal of ponds, or other terrestrial impact)? See Unit 2 SAR Figure 2.5-17. This applies only to areas outside the protected area. l l 0 B Increase thermal discharges to lake or atmosphere? O B increase concentration of chemicals to cooling lake or atmosphere through discharge canal or tower? O 3 Increase quantity of chemicals to cooling lake or atmosphere through discharge canal or tower? l 0 E Modify the design or operation of cooling tower which will change drift characteristics? O 8 instali any new transmission lines leading offsite? O 2 Change the design or operation of the intake or discharge structures? O G Discharges any chemicals new or different from that previously discharged? O B Potentially cause a spill or unevaluated discharge which may effect neighboring soils, surface water or ground water? O E involve burying or placement of any solid wastes in the site area which may effect runoff, surface water or ground water? O E involve incineration or disposal of any potentially hazardous materials on the ANO site?
.O @ Result in a change to nonradiological effluents or licensed reactor power level?
O @ Pote'itially change the type or increase the amount of non-radiological air emissions from the ANO site. ' i l
974379L201 Ret o ARKANSAS NUCLEAR ONE Pace 1 FORM TITLE: ! FORM NO. REV. 10CFR50.59 8AFETY EVALUATl3N 1000.131B 3 PC-2 { This Document contains 1 Page. Document No. 974379L201 Rev./ Change No. 0 10CFR50.59 Eval. No. %) -@ IQ Title (Assigned by PSC) U-2 Vents And Drains Modifications A WRITTEN RESPONSE PROVIDING THE BASIS FOR THE ANSWER TO EACH QUESTI ATTACHED. EACH QUESTION MUST BE ANSWERED SEPARATELY. A SIMPLE STATEMEN CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES GUIDANCE FOR RESPONS If the answer to any question on this form is "Yes," then an unreviewed safety question is involved. If the answ to all questions is *No," then the proposed change does not involve an unreviewed safety question.
- 1. Will the probability of an accident previously evaluated in the SAR be increased?
YesO No @ The changes provided by this LCP will resolve vibration and Code compilance issues by removing the outboard valve on four double vent / drain valves, installing threaded pipe caps and the enhancement of the existing socket welds. The valves to be removed are 2EFW-10678,2EFW-30348,2FW-2003B and 2FW-2009. These are 1" or 3/4" valves on FW and EFW lines in the Containment building. Section 3.6.2.2 of the SAR states that no pipe breaks were considered for line sizes one inch and smaller. The possibility of a threaded cap or a double vent / drain valve becoming a high energy missile has been evaluated in the SAR. These changes do not impact ony accidents analyzed in the SAR. Therefore, there is no increase in the probability of an accident previously evaluated in the SAR. 2. Will the cons 6quences of an accident previously evaluated in the SAR be increased? YesO No @ The changes to be performed by this modification package will not affect the function or operati components or systems used to mitigate a postulated accident. No new pipe break locations are introduced by the modification of these vent / drains. Because any potentialleakage from these vent / drains would be contained, no new pathways for the release of radioactive materials will be created. Therefore, there is no change to the consequences of an accident previously evaluated in the SAR.
- 3. Will the probability of a malfunction of equipment important to safety be increased?
YesO No @ These vent / drain valves are located on the FW and EFW piping inside containment. The equipment important to safety potentially impacted by this change are the Emergency Feedwater flow paths to the su am generators. These changes will not impact the function, operation or flow path of the EFW system. The installation of these changes will be made in accordance with the appropriate design co specifications and standards. Since these changes will be installed to codes and standards consistent with importantthe to originalinstallation, safety. there will be no increase in the probability of a malfunction of equipment 4. Will the consequences of a malfunction of equipment important to safety be increased? Yes O No @ The equipment important to safety potentially impacted by this change are the Emergency Feedwater flow paths to the steam generators. These vent / drain valves being modified by this LCP perform the safety function of maintaining the pressure boundary of the EFW and FW piping inside containment. If a failure were to occur in the vent / drain piping modified by this LCP, it would do so in a manner similar the originalinstallation and yield the same effects. Because any potentialleakage would be in the containment building, there would be no increase in the off site release of radioactive materials, and the consequences of a malfunction of equipment important to safety will not be increased.
m 974379L201,Rev.0-ARKANSAS NUCLEAR ONE Psoe 2 FsRM TITLE: ' FORM NO. REV. 10CFR50.59 SAFETY EVALUATION 1000.131B 3 PC-2 5. Will the possibility of an accident of a different type than l any previously evaluated in the SAR be created? ) YesO No @ The changes to be installed by this LCP will involve the modifications to the configuration of vent / drains l ' and the installation of threaoed pipe caps. The type of accident that this change could present would be a line tireak. High energy line breaks of one inch and less were specifically excluded from the analyi SAR se,ction 3.6. SAR section 3.5.2.3 lists credible missiles resulting from failure of pressurized ! components. Unrestrained vent / drain valves were evaluated as potential missiles. Since the modified line and components are the same size orless that the original, the changes in this LCP are the existing analysis. Therefore, the possibility of an accident of a different type than previously evaluated is not created.
- 6. Will the possibility of a malfunction of equipment important l to safety of a different type than any previously evaluated in the SAR be created?
YesO No @ The malfunctions of equipment important to safety in this LCP are the EFW supply lines and portions of the Feedwaterlines inside containment. Failures related to these lines have been evalu ' No new credible malfunctions could be introduced by this LCP, Therefore, a malfunction of equipme important to safety of a different type than previously evaluated will not be created. 7. Will the margin of safety as defined in the basis for any technicalspecification be reduced? YesO No @ The modifi' cations provided by this LCP involve changes to the configuration of vent / drain valves in the Feedwater and Emergency Feedwater systems. Based on a review of the bases for technical specification 3/4.7.1.2
- Emergency Feedwater System", the EFW system is designed to supply sufficie water to the steam generators to remove decay heat with the steam generator pressure at the MSSV setpoint. This change does not impact the operation of the EFW system nor impact any EFW system surveillance requirements. Therefore, this change will not reduce the margin of safety as defined in the basis for any technical specifications.
(. Joseph C. King Jr. 9/28/98 Cpified yviewer's Sign $ur ' Printed Name Date Reviewer's certification expiration dhte: 11/14/99 Assistance pro'vided by: Printed Name Scope of Assistance Date PSC review by: \6.~.-- Date: \ W% PAGE 7 REV O
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ARKANSAS NUCLEAR ONE FORM TTrLE: FORM NO. REV. 10CFR60.58 DETERMINATION l 1000,131A 2 PC.2.3 I Page1of.5 Document No. SDID 2-97-0226 Rev1 Change No. N/A Tit;a Turbine Buildina Exhaust Flow Rates Listed in the Unit 2 SAR
- Will the proposed Activity:
- 1. Require a change to the Operating License including:
Technical Specifications (excluding the bases)? YesO No@ Operating License? YesO No@ Confirmatory Orders? YesO No@ 2. Result in information in the following SAR documents (including drawings and text) being (a) no longer true or accurate, or (b) violate a requirement stated in the document: Core Operating Limits Report YasO No@ SAR (multi-volume set for each unit)? Yes@ nod QAMO?* YesO No@ E-Plan?* YesO No@ FHA YesO No@ Rases of the Technical Specifications? YesO No@ NRC Safety Evaluation Reports? YesO No@
- 3. Involve a test or experiment not described in the SAR?
YesO No@
- 4. Result in a potential impact to the environment? (Complete Environmental impact Checklist of this form.)
YesO No@
- 5. Result in the need for a Radiological Safety Evaluation per section 6.2.4.A?
YesO No@
- 6. Result in any potential impact to the equipment or facilities utilized for Ventilated Storage Cask activities per Section 6.2.4.B7 YesO No@
Basis for Determination: See continuation sheets, paces 4 and 5. Changes to these documents require an evaluation in accordance with 10CFR50.54. See Section 6.2.1.B.
ARKANSAS NUCLEAR ONC FORM TITLE: FORM NO. REV. 10CFRSO.59 DETERMINATION 1000.131A 2 PC-2,3 Page 2 of.5 Document No. SDID 2-97-0226 RevdChange No. N/A
References:
List sections reviewed in the Licensint Basis Documents, specified in questions 1,2 and
- 3. If a keyword search was done on LetS, "all" may be entered under "Section" with the keyword (s) used in parentheses. Controlled hard copies of the documents shall be reviewed as computer-based searches such as LRS aro not controlled and search text only, not figures or drawings. Attach a completed LDCR if LBD changes are required.
Document Section LRS Search (60.69 U-2) All sections (666* 660*.688*. 2VEF3*. 2VEF-3A. 2VEF-38 turbine build RLL K077end (?!qM,684W,2.Wy,2VEF-9, 2VKF-26) mouirino channes Section 12.2.2.4. Table 11.3-6 and Floure 9.4-3 (P&lD M2260 Sheet 1) Hard copies reviewed Unit 2 SAR: Sections 12.2.2.11.6.6. and 15.1.18: Table 11.3-6 and Floure in Certified Reviewers Signature Bruce Franklin Printed Name
~ h2/!9 ~7
' Date Reviewers certification expiration date: 2 Y Assistance provided by:
Printed Name Scope of Assistance Julie D. Jacks Date 60.59 research 04-10-97 Tom Scott 50.59 research 04-10-97 l t h
ARKANSAS huCLEAR ONE FORM TITLE:
- FORM NO. REV.
10CFR40.59 DETERMINATK3N 1900.131A 2 PC 2.3 Page 3 of 5 ENVIRONMENTAL IMPACT CHECKLIST (UNIT 1 and UNIT 2) Document No. SDID 2-97-0226 Rev> Change No. N/A Complete the following checklist. If the answer to any checklist item is "Yes", an Environmental Evaluation is required. See Section 6.2.1.E for additional guidance. Will the Activity being evaluated: Yes No O E Disturb land that is beyond that initially disturbed during construction (i.e., new construction of buildings, creation or removal of ponds, or other terrestrial impact)? See Unit 2 SAR Figure 2.5-17. This applies only to areas outside the protected area. O E increase thermal discharges to lake or atmosphere? O O Increase concentration of chemicals to cooiing lake or atmosphere through discharge canal ortower? O @ increase quantity of chemicals to cooling lake or atmosphere through discharge canal or tower? O B Modify the design or operation of cooling tower which will change drift characteristics? O. O Install any new transmission lines leading ofrsite? O E change the d.sia'n or operation of the intake or discharge structures?
~O E Discharges any chemicals new or different from that previously discharged?
O @
~ Potentially cause a spill or unevaluated discharge which may effect neighboring soils, surface water or ground water?
O E involve burying or placement of any solid wastes in the site area which may effect runoff, surface water or ground water? O E- Involve incineration or disposal of any potentially hazardous materials on the ANO site? . O @ Result in a change to nonradiological effluents or licensed reactor power level? O O Potentially change the type or increase the amount of non radiological air emissions from the ANO site. n , . .
l ARMANsAs NUCLEAR ONE FORM TITLE: FORM NO.
~
REV. 10CFR50J9 REVIEW CONTINUATION PAGE 1000.131c 2 Page 4 ofl Document No. SDID 2-97-0226 RevJChange No. 0 10CFR50.59 Review Continuation Paae Basis for Determination (continues) Responses to questions on form 1000.131 A (see page 1 of 5): 1 & 2. The numbers given in different Unit 2 SAR sections for the Turbine Building total exhaust f agree. P&lD M-2260 Sheet 1 (SAR Figure 9.4-3) shows a combined exhaust flow of 666,6 2VEF-1s at 60,000 each; 2 2VEF-3s at 25,000 cfm each; and 2VEF-4 at 16,600 cfm). Th with the numbers given for Turbine Building exhaust flow in SAR Section 12.2.2.4 (660,600 cfm) or in Table 11.3-6 (688,600 vendor drawings cfm). The SAR numbers (including the P&lD) need to be revised to ag for the fans: 60,000 for the 2VEF-is (no change needed); 24,000 for the 2VEF-3s; and 16,000 for 2VEF-4. Engineering review of the Turbine Building Ventilation System has determin flow of 2VEF-26 should be included in the total exhaust flow values as this fan supports open between the Turbine and Auxiliary Extension Buildings. P&lD M-2250 Sheet 1 shows 2VEF-26 22,100 cfm. This value needed to be revised to agree with the vendor drawing for the fan,22,00 This gives a total flow of 686,000 cfm for the Turbine Building exhaust. j For the 2VEF-1s, the roof-mounted fans, the flow rate on the purchase requisition (6600-Ml with the flow rate on the vendor drawing and on the P&lD. All three documents have 60,000 cfm fo fan. This gives a total flow rate for the 10 2VEF-1s of 600,000 cfm. For the 2VEF-3s, the flow rate in the requisition (6600-M-2057) is 25,000 cfm, which is the same as P&lD. However, the vendor drawing (6600-M-2057 9-1) has a flow rate of 24,000 cfm. (Thi is within the standard tolerances of 10% for these fans). l For 2VEF-4, the flow rate in the requisition (6600-M-2060) is 16,000 cfm. This agrees with the vend drawing (6600-M-2060-12) but not with the P&lD, which shows a flow rate of 16,600 cfm. For 2VEF-26, the flow rate in the requisition (6600-M-2060) is 22,000 cfm. This agrees with the vendo drawing (6600-M-2060-1-2) but not with the P&lD, which shows a flow rate of 22,100 cfm. The flow rates for these fans on P&lD M-2260 (Figure 9.4-3) have not been changed since Bechtel tum over the drawings. Table 11.3-6 was revised by Amendment 24 to the FSAR to change the Turbine Building exhaust flo from 686,000 cfm to 688,600 cfm. t in f:: 9:: :! tx . ..i 2 - . ;m.,J. The riiscrepancy actually stems from the differences between the vendor drawings (666,000 cfm) and the P&lD dr (688,600 cfm). The only other document found which is close to these numbers is a Bechtel interoffice memo dated June 6,1974, which mentions 'an exhaust quantity of 638,600 cfm' for exhaust fans 2VEF-1s,2VEF-4 and "VEF-26". The addition of 2VEF-3A and 3B at 25,000 cfm each (P&lD values) woutr 688,600 cfm. (Revision 01 of Bechtel drawing M-2260 Sheet 1 relocated VEF-26.) "VEF-26'is a identified as 2VEF-26 and is located on drawing M-2259 Sheet 1 with an indicated flow rate of 22 Section 12.2.2.4 in the FSAR listed a flow rate of 666,600 cfm (same as P&lD M-2260 Sheet 1). In of the SAR, this was changed to 660,600 cfm. There was no revision bar by the changed number and n record of change. This appears to be a typographical error made when the FSAR became the SAR.
AMKANSAS NUCLEAN ONti FORM TITLE:
. FORM NO. REV.
10CFR60.89 REVIEW CONTINUATION PAGE 1000.131c 2 Page 5 ofl The total flow using the P&lD (M-2260 Sheet 1) numbers is 666,600 cfm; the total flow using the p re ;uisition number is 666,000 cfm; and the total flow rate using the vendor drawing numbers is 664 cf',n. Review of original sta.1up test data tumover files (microfilm 1056-1210) shows 2VEF-26 was included in the SU-2-35A test boundarler, since the fan is shared between the Turbine Building Auxiliary Extension Building. This adds 22,000 cfm to the 664,000 cfm for a total flow rate of 686,000 ; Given the uncertain history of the SAR numbers, the best thing to do is reconcile all the SAR references the vendor drawings which document the fans actually received by the plant. Th6efore, Section 12.2.2.4 and Table 11.3-6 should both be changed to show a Turbine Building exhaust flow rate of 686,000 cfm. DRN will be issued to change SAR Figure 9.4-3 to show a flow rate of 24,000 cfra for the 2VEF3s and ; 16,000 cfm for 2VEF-4. A DRN will be issued to change PalD M-2259 Sheet to show a flow rate of 22l cfm for 2VEF-26 exhaust fan.
)
Where Turbine Building Vent Exhaust flow rates are discussed in the SAR, the context is potential off-sit releares. Section 12.2.2.4 mentions steam generator tube leaks, but no specific analysis is done.The Steam Generator Tube Rupture analysis in Chapter 15 does not discuss Turbine Building exhaust air flow. Section 11.6.6, in-plant Effluent Monitoring, bounds the analysis with an approximate turbine building exhaust of 1E6 cfm. Therefore, changing the numbers in Section 12.2.2.4, Table 11.3-6, and Figure 9.4 will not affect any analysis in the Licensing Basis Documents. No other licensing basis documents are affected. 3. This change does not constitute a test or experiment not described in the SAR-
- 4. This change has no potentialimpact on the environment.
5. Since this change does not involve processing of radioactive material outside the Auxiliary Building, Reactor Building, or Low Level Radwaste Building and does not create any new pathway, there is no need for a radiological Safety Evaluation. 6. Since iMs change does not involve any aspect of spent fuel ventilated storage casks, there is no need for a 10CFR 72.48 Review. I l; I l
AR ONE FORM TITLE: FORM NO. REV. 10CFRSO.SS EVALUATION 1000.1318 2 Page 6 off
- 10CFR50.59 (Assigned by PSC)
Eval. No._ NhD - b l Document No. SDID 2_-97-0231 Rev1 Change No. 0,,, Title Turbine 4eildina Exhaust Flow Rates Listed in the Unit 2 SAR ATTACHED. EACH QUESTION MUST BE ANSW CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES GUIDAN . If the answer to any question on this form is "Yes," then an unreviewed safety question is invo to all querlons is "No " then the proposed change does not involve an unreviewed . If the answer safety questio 1. Will the probability of an accident previously evaluated in the SAR be increased? Yes O No @ The heatino. ventilation. and air conditionino systems for the turbine buildino are des provide and a suitable personne!. environmsat Airborne radioactivitytemperature betweer 60 decrees F and 105
, doorees F for e auipment sienificant (see 28AR *=ction 12.2.4). I:v;;; inside the turbine buildino are not considered
.;..-esst.ers w#ricut any ===cial treatment.The exhaust air from this area is released into the v&atilators and thiseeh exhae=* fans in various areas (switchoesr. lobe oi the turbine be!!&.a. The turbine buildina HVAC system is no function the SAR willin notan accideae be increased. condition. Therefore. the probability of an accident previousiv 2.
Will the consequences of an accident previously evaluated in the SAR be increased? Yes O No @ retiis in tfw turbine buildina Uoon indication of steam oene air wil! be seiss@=M!v monitored for radioactive material with portable sampline instr Steam Generster t@ le=%ee will be detected by the radiation monitor on the condens lines. main stssai line radiation monitors. or the steam cenerator sample monitor
. Ventilation air is exhausted threesh 10 roof mounted and three .
wallr mounted Thus, the consecuences of an accident previousiv evaluated in the SAR will not be in creased. 3. Will the probability increased? of a malfunction of equipment important to safety be Yes O No @ in addressi6n any sa ;aed ace! dant condition.The Turbine Buildina' Ventila effluent sca;;si;as by assionina an eneroxirnate flow rate value of 1.000.000 c failure of the EE_- . is bound-Ed by this analysis. Therefore. the probability of a malfenc equisa ht ;nssitsnt to safety will not be increased,
ARKANSAS NUCLEAR ONE FORM TITLE: FORM NO. REV. i 10CFR$0.89 EVALUATION 1000.131B 2
\
Page Z off
- 4. l Will the consequences of a malfunction of equipment important to safety be in:',reased?
\
Yes O No @ The Turbine Buildino Ventilation System is non-safety related and can not impact any safety related equipment. Failure of the system or any component within the system will not cause an increase in the consecuences of a malfunction of eauipment important to safety. S. Will the possibility of an accident of a different type than any previously evaluated in the SAR be created? Yes O No @ The Turbine Buildina System is desianed to remove the excess heat from area oculoment and lichtina heat loads and to maintain suitable environment conditions for equipment and personnel. l Failure of components in the turbine buildina ventilation system is not considered an initiator of any type accident. The SAR revision (SDIDi chanaes the exhaust flow values listed in various sections of SAR. No chances are made to the physical plant eauipment. Therefore, the possibility of an accident of a different type than previousiv evaluated in the SAR will not be created. l
- 6. I Will the possibility of a malfunction of equipment important to safety of a different type than any previously eva!uated in the SAR be created?
Yes O No @ The Turbine Buildina Ventilation Cystem is desianed and constructed utilizine multiple components. plant Failure of one or more of these components will not cause a maior impact to normal oDerations. The system has no impact on any safety related system nor component. Consecuentiv. the turbine buildina ventilation system can not cause an increase in the possibility of a malfunction of eauipment important to safety of a different tvDe than any DreviouslV evaluated in the SAR. 7. Will the margin of safety as defir5ed in the bases for any technical specification be reduced? Yes O No @ Since the turbine buildina ventilation system is desianed with no safety function and havina no potential for impactina equipment important to safety. these chanaes will not reduce any maroin of safety as defined in the bases for any technical specification. Bruce Franklin 8/21/97 Certified Reviewers Signature Printed Name Date Reviewers certification expiration date: 8/26/98 ' I Assistance provided by: ' 1 Printed Name Scope of Assistance N/A Date PSC review by: h _\ f N- U Date: 9-
\
h\ 4 d Mp y% 4 A ~g4 dm
avowvenan rua. Lean usee FORM TITLE: FORM NO. REV. 10CFRSS.89 DETERMelATION 1000.131A 3 PC 1 l Page lof 4 Document No. ODCM Rev/ Change No. 11 Title Offsite Dose Calculation Manual Brief description of proposed change: Allow dilution flow rate less than 2 Unit 1 Circulatina water pumps. Will the proposed Activity:
- 1. Require a change to the Operating License including:
Technical Specifications (excluding the bases)? YesO No2 Operating Ucense? YesO No@ Confirmatory Orders? YesO No@ 2. Result in information in the following SAR documents (including drawings and text) being (a) no longer true or accurate, or (b) violate a requirement stated in the document: SAR (multi-volume set for each unit)? , Yes@ nod Core Operating Limits Report YesO No@ Fire Hazards Analysis? YesO No@ Bases of the Technical Specifications? YesO No@ Technical Requirements Manual? YesO No@ NRC Safety Evaluation Reports? YesO No@ 3, involve a test or experiment not described in the SAR? (See Attachment 2 for guidance) YesO No@
- 4. Result in a potential impact to the environment? (Complete i the Environmentalimpact Determination of this form.)
YesO No@
- 5. Result in the need for a Radiological Safety Evaluation per section 6.1.57 YesO No@
- 6. Result in any potential impact to the equipment or facilities utilized for Ventilated Storage Cask activities per Section 6.1.6?
YesO No@
- 7. Involve a change under 10CFR50.54 for the following SAR documents per Section 6.1.7:
QAMO? ' YesO No@ E Plan? YesO No@ Basis for Determination (Questions 1. 2 & 31: See Pace 4 O Proposed change does not require 10 CFR 50.5g Evaluation per Attachment 1, item # ,(if checked, note appropriate item #, send LDCR to Licensing).
A6 mas NUCLEAR ONE FORM TITLE: FORM NO. REV. 10CFRSO.89 DETERMINATION 1900,131A 3 PC 1 Page 2 of,4 Document No. ODCM Revjchange No. 11f ' Search Scope: List sections reviewed in the Licensing Basis Documents specified in Question 1,2 and 3. If a keyw was done on LRS, "all" may be entered under"Section" with the keyword (s) used in parentheses. Controlle copies of the documents shall be reviewed (LRS is not verified and searches only text. not figures or Attach and distribute a completed LDCR per Section 6.1.2 If LBD changes are required. Document Section LRS: 50.59 Common All (Circ., Circulating w/5 water, liquid w/10 release *, liquid w/10 effluent *, liquid w/10 waste) MANUAL SECTIONS: 1TS and bases 4.29.1, 4.30, 1.10.4,, 3.25.1, 3.25.3, 3.5.6, 6.8, 6.9, 6.14, Table 3.5.6-1
.QA Manual 16.4 1SAR 2.8.1,1.4.51,1.4.55,11.0, all '
chapter 11 tables 2TS & -bases 3/4.11.1, 3/4.11.3, 3/4.3.3.10, 1.30, 6.9.2, 6.9.3, Tables 4.11 1,4.3-13,4.29-1 2SAR 2.4.12, 3.1.6,11.0, all chapter 11 tables FIGURES:
/> L Mb Certified Reviewers Signturiii ReNae Partridae 1/8/98 Printed Name Date Reviewers certification expiration date: 10/27/99 Assistance provided by:
Printed Name Scope of Assistance Steve Bennet Date Evaluate aoolicable portions of chapter 11 1/8/98 Search Scope Review Acceptability (NA, if performed by Technical Review per 1000.006) . Certified Reviewers Signature Printed Name Date I 1
FORM TITLE: FORM NO. REV. 10CMtSt.88 DETERMINATION 1000.131A 3 PC.1 1 Page 3 of,4 l ENVIRONMENTAL IMPACT DETERMINATION . (UNIT 1 and UNIT 2) Document No. gggi Rev> Change No. j.1 Complete the following Determination. If the answer to any checklist item is "Yes", an Environmental Evaluation is required. See Section 6.1.4 for additional guidance. Will the Activity being evaluated: Yes No O O Disturb land that is beyond that initially disturbed during construction (i.e., new construction of buildings, creation or removal of ponds, or other terrestrial impact)? See Unit 2 SAR Figure 2.5-17. This applies only to areas outside the protected area. O E Increase thermal discharges to lake or atmosphere? O E increase concentration of chemicals to cooling lake or atmosphere through discharge canal or tower? O E increase quantity of chemicals to cooling lake or atmosphere through discharge canal or tower? O E Modify the design or operation of cooling tower which will change drift characteristics? O 9 Install any new transmission lines leading offsite? O O Change the design or operation of the intake or discharge structures? O O Discharges any chemicals new or different from that previously discharged? O @ Potentially cause a spill or unevaluated discharge which may effect neighboring soils, surface water or ground water? O E involve burying or placement of any solid wastes in the site. area which may effect runoff,
. surface water or ground water?
O O involve incineration or disposal of any potentially hazardous materials on the ANO site? O @ Result in a change to nonradiological effluents or licensed reactor power level? O @ Potentially change the type or increase the amount of non-radiological air emissions from the ANO site. I
Page 4 of 4 ODCM 50.59 Determmation Discussion The liquid effluent release limits for the dose commitment to a memW of the public are pro in the Technical Specifications for both units. The Technical Specifications limits provide assurance that effluent releases will result in concentrations far below the limits dose and concentration. Currently, ANO has set admmistrative effluent goals at 1% of the Technical Specifications dose imuts Prior to the discharge of any radioactive liquid release
. calculation is performed to detemune the dose, dose rates, and radmonitor setpoints associa the release. Calculational vanables include tank activity, tank volume, release flow rate flow rate, and quarterly dilution volume. 'Ihe decision to release radioactive liquid is based on results of the calculations.
The change to the ODCM is to allow release of radioactive liquid from the site with less than tw circulating water pumps in operation. Although two or more circulatmg water pumps in operation, there are rare occasions when less than two pumps are in operation. Effluent r calculations have determmed that release oflow level activity, with discharge flow due t service water and cooling tower blowdown flow, will have muumal impact on dose and dose rate Release of radioetive liquid with less than two circulating water pumps in operation will occur on rare occasione nie ader strict controls. Ouestion 1 The Operating i Sr.se does not address the number of circulating water pumps or specify discharge flow rate required to release radioactive liquid from the site. 'The change made to ODCM does not conf!N with the information contained in the Operating License Ouestion 2 -
. Unit i SAR 11.1.2.4 mentions that " liquid waste effluent is diluted in the discharge canal to '
concentrations far below those stipulated in 10CFR20." Section 11.1.3 describes the design ba for the liwid effluent system. The unit 1 SAR does not provide specific requirements for circulattog water pump operation during normal operation of the plant. 1 Unit 2 SAR 11.2.6.4 also describes the design basis for the liquid. effluent system S ti ec on 11.2.8 states that "The amount of flow available for diluting the waste is a function of the number of unit i circulatmg water pumps in operation. The maximum dilution flow will be 766,000 gpm wit i four circulatmg water pumps in operation and the minimum flow will be ~383,000 gpm with { circulating water pumps in operation." Tables 11.2-16A&B contain expected liquid release vj) . during normal operation and unanticipated operational occurrences Section i1.2.8 needs to be i changed to address off'-normal operation ofless than two circulatmg water pumps. Ouestion 3 - , 1 - This change does not involve a test or experiment.-
FORM TITLE: FORM NO. REV. 10CFR$0.80 EVALUATION 1000.1313 3 ' Page 1 of2 10CFR50.59 Eval. No._ FVN T -003 (Assigned by PSC) Document No. SEGM RevlChange No. 11, Title Offsite Dose CalcaMion Manual A WRITTEN RESPONSE PROVIDING THE BASIS FOR THE ANSWER TO EACH QUESTION M ATTACHED. EACH QUESTION MUST BE ANSWERED SEPARATELY. A SIMPLE STATEMENT CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES GUIDANCE FOR RESPONSE If the answer to any questior on this fonn is "Yes," then an unreviewed safety question is involved. If the answer to all questions is "No," then the proposed change does not involve an unreviewed safety question.
- 1. Will the probability of an accident previously evaluated in the SAR be increased?
Yes O No @ Reducino the dilution flow rate when performino a monitored and known low level liould radisactive rese dess not have any affect on the accidents m the SAR. Therefore. the probability of an accident previousiv evaluated in the SAR is not increased. . 2. Will the consequences of an accident previously evaluated in the SAR be increased? Yes O No @ fleducina the dilution flow rate when performino a monitored and known low level hould radioactive release does not have any affect on the accidents in the SAR. Therefore. the consasuences of an accident previousiv evaluated in the SAR is not increased. , 3. Will the probability of a malfunction of equipment important to saft4y be increased? Yes O No @ This chance involves allowino release of radioactive liould with less than two ANO-1 circulatino ".;_^ r numos in operation. The chance does not direct the operation of any souissant ;mssitant to safety; therefore. this chance does not increase the probability of a malfunction.of ecuipment important to safety. 4. Will the consequences of a malfunction of equipment important to safety be increased? Yes O No @ This chance involves allowino release of radioactive liould with less than two ANO-1 circulatino .._ ^ r stemss in operation. The chance does not direct the operation of any seules&;it imssitant to safety: therefore. this chance does not increase the consecuences of a malfunction of eculoment important to safety. l. l l
ARKANSAS HUCLEAR ONE
- FORM TITLE:
FORM NO. REV. 10CFR50.89 EVALUATION 1000.1313 3 1 Page 2 of 2 5. Will the possibility of an accident of a different type than any previously evaluated in the SAR be created? Yes O No @ This chance involves allowine i:t::: of radiame+ive liouid ##, !=95 than two ANO. 1 circe 8=W.s water DumDs in seei.iion. Re! a::: are allowed after Dreliminary ca!cadMer.s -f:^__ cc.:r.s suffic s..';< low dose and de = r=*e= for discharoe. This char.ss d955 not create the Dossibility of an accident of a different tvDe than any j Dreviousiv evaluated in the SAR. l 6. Will the possibility of a malfunction of equipment important to safety of a j different type than any previously evaluated in the SAR be created? Yes O No @ This 1 chance involves allowine release of radioactive licuid with less than two ANO-circulatino
.._..r DumDs in speration. The chance does not direct the operation of any equipment imssitent to safety: thereforel this chance does not create the Drobability of a malfunction of eeu!9 ment imDottant to safety of a different tVDe than any DreviouslV evaluated in the SAR.
7. Will the margin of safety as defined in the bases for any technical specification be reduced? Yes O No @ The Technical Specifications bases for both units 5?5!- that the Technical SDeCifications limits are Drovidsd to ensure that the concentration of radioactive materials released in liee!d w==te effluents from the site to unrestricted areas will be less than the conce..i..iion levels specified in 10CFR20. Aenendix B. Table 11. Allowino rolesses w;tii less than two circulatina water DumDs in oDeration does not _comDromise the ability to meet the 10CFR20 limits.
. *> - N/d/ ReNae Partridae 1/8/97 Certified Reviewers Signad5re Printed Name Date Reviewers certification expiration date: 10/27/99 Assistance provideo by:
Printed Name Scope of Assistance Date PSC review by: 1
~
A Date: \\G Y l l l 1
. . .... = m.
FORM TITLE: FORM NO. REV. 10CFRSS.89 DETERMINATION 1000.131A 3 PC 1 Page 1 of 5 Document No. 2104.014 Rev/ Change No. 30/PC-3 - Title LRW AND BMS OPERATIONS Brief description of proposed change: Incorporate infonnation identified durina field walkdowns. Will the proposed Activity: 1. Require a change to the Operating License including: Technical Specifications (excluding the bases)? YesO No@ Operating License? YesO No@ Confirmatory Orders? YesO No@ 2. Result in information in the following SAR documents (includilig drawings and text) being (a) no longer true cr accurate, or (b) violate a requirement stated in the document: SAR (multi-volume set for each unit)?
, Yes@ nod Core Operating Limits Report YesO No@
Fire Hazards Analysis? YesO No@ Bases of the Technical Specifications? YesO No@ Technical Renuirements Manual? YesO No@ NRC Safety Evaluation Reports? YesO No@
- 3. Involve a test or experiment not described in the SAR?
(See Attachment 2 for guidance) YesO No@ 4. Result in a potentialimpact to the environment? (Complete the EnvironmentalImpact Determination of this form.) YesO No@ 5. Result in the need for a Radiological Safety Evaluation per section 6.1.5? YesO No@ 6. Result in any potential impact to the equipment or facilities utilized for Ventilated Storage Cask activities per Section 6.1.6? YesO No@ 7. Involve a change under 10CFR50.54 for the following SAR documents per Section 6.1.7: QAMO? YesO No@ E Pian? YesO No@
ARKANaAS NUCLEAR ONE FORM TITLE: FORM NO. REV. 10CFR80.80 DETERMINATION 1000.131A 3 PC.1 Page 2 of 5 Document No. 2104.014-Rev> Change No. 30/PC-3 Basis for Determination (QMa*lons 1. 2 & 31: The proposed changes are a direct result of system field walkdowns. is removed from the document in this change (SAR Con is ADDED to the document in this change, and 3) Changed description of location for v the conflict with the SAR, an Evaluation and LDCR have been nerformed and are attached for revie . O Proposed change does not require 10 CFR 50.59 Evaluation per Attachment 1, item #
. (if checked, note appropriate item #, send LDCR to Licensing).
Search Scope: List sections reviewed in the Licensing Basis Documents specified in Question 1,2 a was done on LRS, "all" may be entered under "Section" with the keyword (s) used in par Attach and distribute a completed LDCR per Section 6.1.2 if L Document Section LRS: Unit 2 50.59 2LRW-1004; 2LRW-1005; 2LRW-1006 MANUAL SECTIONS: Unit 2 SAR Section 11, Section 15 FIGURES: Unit 2 SAR All Section 11 Figure 2. p Jay Gary Wellwood Carfffjed IWvieWer1iSigtfature 5-26-98 Printed Name Date Rsviewers certification expiration date: 12-22-99 1 Assistance provided by: Printed Name Scope of Assistance Date Srrch Scope Review Acceptability (NA, if performed by Technical Review per 1000.006) l CIrtified Reviewers Signature Printed Name Date
FORM TITLE: \ FORM NO. REV. 10CPR$0.80 DETWtMINATioN 10ee.131A 3 Page2off ENVIRONMENTAL IMPACT DETERMINATION (UNIT 1 and UNIT 2) ' ' Document No. 2104,014 Rev> Change No. 30/PC-3
-isComplete required. SeetheSectionfollowing Determination.
6.1.4 for additional guidance. If the answer to any checklist item is "Yes", an E
; Will the Activity being evaluated:
k Yes No. O E
~ Disturb land that is beyond that initially disturbed during construction (i.e., n buildings, creation or removal of ponds, or other terrestrial impact)? See Un 2.5-17. This applies only to areas outside the protected area.
O '@ increase thermal discharges to lake or atmosphere? O E increase tower? concentration of chemicals to cooling lake or atmosphere through disc O E increase tower? quantity of chemicals to cooling lake or atmosphere through discharge can O E Modify the design or operation of cooling tower which will change drift characteristics O E Install any new transmission lines leading offsite? O @ Change the design or operation of the intake or discharge structures? O O Discharges any chemicals new or differe'nt from that previously discharged? j O @ Potentially water or groundcause water? a
- spill or unevaluated discharge which may effect neighboring s i 0 0 4 involve surface burying water or water?
or ground placement of any solid wastes in the site area which may ef O E involve incineration or disposal of any potentially hazardous materials on the ANO O ~@- _ Result in a change to nonradiological effluents or licensed reactor power level?
.O @
Potentially ANO site. change the type or increase the amount of non-radiological air emissi L l
~
FORM TITLE. ARKANSAS NUct.EJut ONE FORM NO. REV. 14CFRSS.89 EVALUATION 1000.13 0 3 Page 4 of 5 10CFR50.59 Eval. No. FFM 40% - (Assigned by PSC) Document No. 2104.014 Rev1 Change No. 30/PC-3 Title LRW AND BMS OPERATIONS ATTACHED. EACH QUESTION MUST BE AN CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES to all questions is "No," then the proposed change does . r n 1. Will the probability of an accident previously evaluated in the SAR be increased? Yes O No @ 1 System plays a significant role in. Whether it be in th analyzed forin the SAR depends upon these systems. As such, since no ac 1 systems, the probability that an accident will occur remains unchanged. Containment" makes no reference as to the inclusion . 2. Will the consequences of an accident previously evaluated in the SAR be increased? . Yes O No @ Since none of the accident scenarios analyzed for in the SAR areondirectly or d mis-operation of these systems, the subsequent effects on the amount unchanged. of of 3.' Will the probability increased? of a malfunction of equipment important to safety be Yes O No & The removal of the identified component from the controlling documents in designated or identified as important to safety. Since no equipment important e
. likelihood that a malfunction of said safety equipment also remains' unchallenged. ,
4. Will be the consequences increased? of a malfunction of equipment important to safety Yes O No @ the likelihood of that safety equipment failure is uncha malfunction of safety related equipment will also remain unchanged. p 8 . t i
rom TrrLE. FORM NO. REv. SecPnseas eVALuafloN , 1000.?318 3 Pageiof1 Document No. 2196tig Rev/ Change No. 30/PC 3
' 5. .
Will the possibility of an accident of a different type than any previously evaluated in the SAR be created? Yes O No @ The systems that this document govems oe used to treat radioactive liquid waste prior to release. Additionally, these systems are used to transport liquid radioactive waste fmm one holding device to another, while awaiting treatment or release. The removal of the identified component from this document (that does not exist in the field) will have no credible effect upon the systems operation and designe function. Based on this, the likelihood that another accident not already analyzed for is not credible. 6. Will the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR be created? Yes O No @ The proper operation of these systems have no impact upon any equipment classified as important to safety. Further, even based upon physical location of the system component in question, no equipme importent to safety is located in the vicinity to be impacted by the proposed changes. Since no eq important to safety is served by the systems, and no equipment deemed important to safety is in the n vicinity to be impacted by, the likelihood that a malfunction will occur remains unchanged. 7. Will the margin of safety as defined in the bases for any technical
- specification be reduced?
Yes O No @ Aithough not specifically a margin of safety, the Unit 2 Tech Specs require that all liquid rad treated prior to release (T.S. 3.11.1.3), this is based on P.2t exceeding projected dose limits from to the effluents. Since the 25% value mentioned in the bases could be viewed as a margin of ri could be loosely interpreted as a margin of safety), the following discussion is required. The deletion of this component from the controlling document in no way impacts the proper ope these systems, and as such, does not impact its ability to continue the treatment of liquid radioactiv Since the ability to treat the liquid radioactive waste is not impacted, it can be reasoned that the . continue to be treated as in the past, thus not invalidating the requirement. Since the requirement impacted, the values specified will remain unchanged or challenged. Since the values are u 25% margi'i specified in the bases will also remain unchanged or unchallenged. , M Jay Gary Wellwood Certiff Reviewer's' Signature 5-26-98 Printed Name Date Reviewefs certification expiration date: 12 22-99 Assistance provided by: Printed Name Scope of Assistance Date
. PSC review by:
Date: b S N b y -
Retum Tc: CHARLA CHAPMAN Dtte: 11/25/9 50.59 Ssf:ty Evaluition Subecmmittze Action It:m Form Responsible Individual (s): JAY WELLWOOD Action Item No.: FFN 98 072 DANA MILLAR Date Assigned: 11/30/98 Subcommittee
Contact:
CLAY REED Date Due: 05/30/99
Reference:
OP 2104.014 Rev. 30 PC-3 LRW and BMS Operations Action Questions 3,5 & 7 considered the deletion of the valve but did not consider the addition. Paper changes need to be Required: viewed from a physical change perspective when preparing 50.59 cvaluations. If you need assistance please contact Clay Reed. The original revised 50.59 evaluation should be sent back to the PSC for review with a copy forwarded to Charla Chapman GSB-3W. i l Completed By: Date: Approved Disapproved Request For AdditionalInformation: Revised Due Date: l Reviewed By: umany u.va.ua.wse ama uam.a Date:
-o .-.,..o ~ m .
FORM TITLE:
- FORM NO. REV.
10CFRSS.89 EVALUATION 1000,131a 3 Page 4 ofi 10CFR50.59 Eval. No. FFN 072-1 (Assigned by PSC)
- Document No. 2104.014 RevlChange No. 30/PC-3 Title LRW AND BMS OPERATIONS l
l l A WRITTEN RESPONSE PROVIDING THE BASIS FOR THE ANSWER TO EACH QUESTION MUST B ATTACHED. EACH QUESTION MUST BE ANSWERED SEPARATELY. A SIMPLE STATEMENT OF CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES GUIDANCE FOR RESPONSE. l if the answer to any question on this form is "Yes," then an unreviewed safety question is involved. If the answer to all questions is "No," then the proposed change does not involve an unreviewed safety question.
- 1. Will the probability of an accident previously evaluated in the SAR be increased?
Yes O No @ The removal of the identified component (2LRW-1006 - vent valve currently shown to be installed on a section of piping identified by 2HCD-176-1%) from the system has no impact upon the proper operation of the systems it would support. The component, which has been missing for an indeterminate amount of time, obviously has had no impact to date. Currently there is no accident in the SAR which either the Liquid Rad Waste or the Boron Management System play a role in. Whether it be in the initiator, or the mitigation phase, no accident analyzed for in the SAR depends upon either of these systems. As such, since no accident is dependent upon the proper operation of these systems, the probability that an accident will occur continues to remain unchanged. Even in the most remote classification, the accident analysis designated as "Small Spills or Leaks of Radioactive Material Outside Containment" makes no reference as to the inclusion of either of these systems for credit or impact. continued success in operation of these systems with no adverse consequences (i.e. spills, loss of inventory) further indicates system integrity is preserved even with the absence of the identified component.
- 2. Will the consequences of an accident previously evaluated in the SAR be increased?
Yes O No @ Since none of the accident scenarios analyzed for in the SAR aru dependerit upon the operation or mis-operation of either of these systems (Liquid Rad Waste / Boron Management System), the subsequent effects on the amount of off-site dose release in the event of an accident will remain unchanged by the proposed changes.
- 3. ' Will the probability of a malfunction of equipment important to safety be increased? l Yes O No @
The removal of the identified component from the system in no way has any impact to any equipment 1 designated or identified as important to safety directly or otherwise. Since no equipment important to safety is impacted, the likelihood that a malfunction of safety equipment will occur as a result of removing the identified component from the system also remains unchanged.
- 4. Will the consequences of a malfunction of equipment important to safety be increased?
Yes O No @ The removal of the identified component from the system will not have any impact on any equipment ' identified as important to safety. Since there is no safety-related equipment being affected by the removal of the identifico component, the impact on any postulated off-site dose consequences as a result of safety-related equipment failure will also remain unchanged.
ARKANSAS NUCLEAR ONE FORM TITLE: - FORM NO. REV.
, 10CFRSO.50 EVALUATION 1000,1318 {
3 l Page i of1 5. Will the possibility of an accident of a different type than any previously evaluated in the SAR be created? Yes O No @ The identified systems impacted by the removal of the identified component are used to transport liquid radioactive waste from one holding device to another while awaiting treatment or release, and to facilitate I the release of the treated effluents. Since the identified component has been missing for an indeterminate amount of time, and the systems have operated satisfactorily with its' absence, it can be stated with confidence that the removal of the identified component from these systems will have no effect upon the systems operation and designated function. Given the fact that the absence of this component has had no { 1 effect upon its' parent systems nor any equipment that is considered to be safety-related, the likelihood that another accident to occur not already analyzed foris not considered to be possible. 6. Will the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR be created? Yes O No @ The proper operation of these systems have no impact upon any equipment classified as important to safety. Further, even based upon physicai location of the system component in question, no equipmen , important to safety is locatad within the vicinity that the identified component would have been installed. j Additionally, since the component has been missing for an indeterminate amount of time, and there has been no impact on any safety related equipment to date, the likelihood that a malfunction will occur as a result of the components' removal remains unchanged. 7. Will the margin of safety as defined in the bases for any technical specification be reduced? Yes O No @ Although not specifically a margin of safety, the Unit 2 Tech Specs require that all liquid radwaste be treated prior to release (T.S. 3.11.1.3), this is based on no,t exceeding projected dose limits from exposur to the effluents. Since the 25% value mentioned in the bases could be viewed as a margin of risk (which could be loosely interpreted as a margin of safety), the following discussion is required. The removal of the identified component (which has been missing for an indeterminate amount of tim from the system in no way impacts the proper operation of these systems. In fact, the component has been missing for an indeterminate amount of time, and there have been no adverse consequences associated ! with its' absence and the parent systems' ability to continue the treatment of liquid radioactive waste. Based on this, it should be reasoned that the continued absence of the identified component will have no adverse consequences on the performance of the parent systems' capability to continue the treatment of Uquid Rad Waste. Since the ability to treat the liquid radioactive waste is not impacted, it can be reasoned that the effluent will continue to be treated as in the past, thus not invalidating the requirement. Since the requirement is not impacted, the values specified will remain unchanged. Since the values are unchanged, the 25% margin specified in the bases will also remain unchanged. M_ _ Jay Gary Wellwood Cpffified Re9fewers' Signature 12/31/98 Printed Name Date Reviewers certification expiration date: 12 22-99 Assistance provid:d by: Printed Name Scope of Assistance Date PSC review by: kW Date: ) o
T ARKANSAS NUCLEAR ONE Pace 3 F.ORM TITLE: FORM NO. REV. 10CFR50.59 DETERMINATION 1000.131 A 3 PC-1,2 Document No. SDID-2-97-0425 Rev/ Change No. O Title Post RAS LPSI Pump SAR Clarifications Brief description of proposed change: The LDCR will resolve U2 SAR Discrepancy 2-97-0425 conceming post RAS LPSI Pump Operation. The SAR provides information about system operation and flow limitations post RAS, which while true, are not part of the ANO2 Design Basis, are not reflected in any Operating Procedures and do not need to be included in the Des Basis. This LDCR will remove ihree statements conceming LPSI pump operation in post RAS operation. The ANO2 Design Basis assumes that both 2P-60 LPSI pumps will be secured when the RAS occurs and will not be required to operate after the RAS to mitigate core damage or protect the public. Will the proposed Activity:
- 1. Require a change to the Operating License including:
Technical Specifications (excluding the bases)? YesO Ne@ Operating License? YesO No@ Confirmatory Orders? YesO No@ 2. Result in information in the following SAR documents (including drawings and text) being (a) no longer true or accurate, or (b) violate a requirement stated in the document: l SAR (multi-volume set for each unit)? Yes@ nod Core Operating Limits Report? YesO No@ Fire Hazards Analysis? YesO No@ Bases of the Technical Specifications? YesO No@ Technical Requirements Manual? YesO No@ NRC Safety Evaluation Reports? YesO Nc@
- 3. Involve a test or experiment not described in the SAR?
(See Attachment 2 for guidance) YesO No@
- 4. Result in a potential impact to the environment? (Complete Environmental impact Determination of this form.)
YesO No@
- 5. Result in the need for a Radiological Safety Evaluation per section 6.1.57 YesO No@
- 6. Result in any potential impact to the equipment or facilities utilized for Ventilated Storage Cask activities per Section 6.1.67 YesO No@
- 7. Involve a change under 10CFR50.54 for the following SAR documents per Section 6.1.77 QAMO?
YesO No@ E Plan? YesO No@
ARKANSAS NUCLEAR ONE FORM TITLE: Page 4 FORM NO. REV. 10CFR60.59 DETERMINATION 1000.131 A 3 pC-1,2 Document No. SDID-2 97-0425 RevlChange No. O Basis for Determination (Questions 1,2, & 3): Deleting the three statements conceming post RAS LPSI pump operation does not requ Operating License and does not constitute a test or experiment. Three statements abo will be deleted from the U2 SAR and an Evaluation is attached. No other changes are of the SAR. The three statements being deleted from the U2 SAR are: SAR Section 6.3.2.20.5: Alternatively the LPSIpump suction maybe alignedto the contain . avaliableof mode operations, administrative control of LPSIpump flow to less than 3,500 gpm will en NPSH. l SAR Figure 7.3-7, Section D.2.8 Use ofLPSI Pumps During Recirculation. The Low Pre manually restarted to obtain increasedrecirculation tiow. initiation of containment sump recirculation.SAR Figure 7.310, Section 2. Should O Proposed change does n":t require 10CFR50.59 Evaluation per Attachment 1, item # (If checked, note appropriate item #, send LDCR to Licensing).
~
Search Scope: List sections reviewed in the Licensing Basis Documents specified in questions 1,2 and 3. If sea performed on LRS, the LRS search index should be entered under "Section" with the search parentheses. Controlled hard copies of the documents shill be reviewed (LRS is not verif text, not figures or drawings). Attach and distribute a completed LDCR per Section 6.1.2 required. Document Section LRS: All (LPSI /-10,10/ RAS), ("LPSI pump"), (" containment sump *), (containment sump /- MANUAL SECTIONS: U2 SAR Section 6.3.2.20.5 FIGURES: U2 SAR Figure 7.3-7 and Figure 7.3-10, U2 SAR Figure 6.3-12 [- " Certified 3kiviewers Signature Tim Woodson Printed Name IUf? Date Reviewers certification expiration date: 11-19-99 Assistance provided by: i Printed Name Scope of Assistance Date Search Scope Review Acceptability (NA, if performed by Technical Reviewer per 1000.006) [A % %7 cp /for' w A 3 lt d 1,7 Certified Reviewers Signature Printed Name Date
ARKANSAS NUCLEAR ONE Page 5 FORM TITLE: FORM NO. REV. 10CFR50.59 DETERMINATION 1000.131A J l ENVIRONMENTAL IMPACT DETERMINATION (UNIT 1 and UNIT 2) i Document No. SDlD-2-97-0425 Rev1 Change No. O Complete the following Determination. If the answer to any item below is "Yes", an Environmental Evaluation is required. See Section 6.1.4 for additional guidance. Will the Activity being evaluated: Yes No O E Disturb land that is beyond that initially disturbed during construction (i.e., new construction of buildings, creation or removal of ponds, or other terrestrial impact)? See Unit 2 SAR Figure 2.5-17. This applies only to areas outside the protected area. O E increase thermal discharges to lake or atmosphere? O @ increase concentration of chemicals to cooling lake or atmosphere thruugh discharge canal or tower? O 3 Increase quantity of chemicals to cooling lake or atmosphere through discharge canal or tower? O @ Modify the design or operation of cooling tower which will change drift characteristics? O E install any new transmission lines leading offsite? O @ Change the design or operation of the intake or discharge structures? O O Discharges any chemicals new or different from that previously discharged? O @ Potentially cause a spill or unevaluated discharge which may effect neighboring soils, surface water or ground water? O @ involve burying or placement of any solid wastes in the site area which may effect runoff, surface water or ground water? O E involve incineration or disposal of any potentially hazardous materials on the ANO site? O @ Result in a change to nonradiological effluents or licensed reactor power level? O @ Potentially change the type or increase the amount of non-radiological air emissions from the ANO site.
ARKANSAS NUCLEAR ONE FORM TITLE: Page G FORM NO. REV. 10CFR50.59 SAFETY EVALUATION 1000.131B 3 PC-2 Document No. SDID-2 97-0425 Rev> Change No. 0 10CFR50.59 Eval. No. F rN4 I M Title Post RAS LPSI Pump SAR Clarifications (Assigned by PSC) ATTACHED. EACH QUESTION MUST BE ANSWERE CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES GUIDANC . to all questions is "No," then the proposed change does not invo . 1. Will the probability of an accident previously evaluated in the SAR be increased? YesO No @ 2. Will the consequences of an accident previously evaluated in the SAR be increased? YesO No @ 3. Will the probability of a malfunction of equipment important to safety be increased? YesO No @ 4. Will the consequences of a malfunction of equipment important to safety be increased? YesO No @ 5. Will the possibility of an accident of a different type than any previously evaluated in the SAR be created? YesO No @ 6. Will the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR be created? Yes O No @ 7. Will the margin of safety as defined in the basis for any technical specification be reduced? Yes O No @
. tv Certi[!(d Reviewer's Signature Tim Woodson bh' M Printed Name Date RGviewer's certification expiration date:
11 19 99 Assistance provided by: 4 Printed Name Scope of Assistance Date PSC review by:
]\CF -
Date: M SI
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ARKANSAS NUCLEAR ONE Ptge 7 F,ORM TITLE: FORM NO. REV. 10CFR50.59 SAFETY EVALUATION 1000.1318 3 PC-2
- 1. Will the probability of an accident previously evaluated in the SAR be increased?
No. Deleting information from the SAR about post RAS LPSI pump hydraulic limitations and use beyond their design basis does not alter any of the accident scenarios in the SAR. All of the analyzed accidents will remain in their previously analyzed category for frequency of occurrence. There are no probability changes caused by this deletion. 2. Will the consequences of an accident previously evaluated in the SAR be increased? No. The LPSI system acts to inject large amounts of borated water into the core following a SIAS signal when the RCS pressure drops below the pump's shutoff head. As part of the ANO2 design basis, the LPSI pumps are secured by the Recirculation Actuation Signal (RAS) and are not required to operate for long term core cooling or mitigation of core damage. Aligning the LPSI pumps to the Containment Spray System after a RAS would require an Operator entering the ECCS pump rooms and being exposed to radiation that is beyond the ANO2 design basis. Deleting the SAR references to post RAS use does not alter the LPSI design bases, therefore there is no added potential for increasing an offsite dose release. 3. Will the probability of a malfunction of equipment important to safety be increased? No. Although it is possible to use the LPSI pumps post RAS, there are no design requirements to be able to operate the pumps. Deleting the references in the SAR does not place any additional requirements on any of the ECCS or Containment Spray components, or any other safety related components. The LPSI pumps are not required to operate for long term core cooling.
- 4. Will the consequences of a malfunction of equipment important to safety be increased?
No. The current ANO2 design basis and offsite dose calculations all recognize and account for the LPSI pumps being secured once long term recirculation occurs. Deleting the post RAS references in the U2 SAR does not affect the offsite dose calculations. 5. Will the possibility of an accident of a different type than any previously evaluated in the SAR be created? . No. The ECCS system has several specific functions for preventing and mitigating core damage in response ; to a SlAS signal including long term recirculation. None of the postulated failures for the ECCS system require restarting a LPSI pump after starting long term recirculation. Not using the LPSI pumps after a RAS is , part of the ANO2 design, and not using them is bounded by all of the analyzed accidents in the U2 SAR. i 6. Will the possibility of' amalfunction of equipment important to safety of a different type than any previously evaluated in the SAR be created? No. The ANO2 design basis accounts for the LPSI pumps being secured for long term core cooling. The design of the ECCS system did not feel there would be sufficient need to restart the LPSI pumps post RAS. The design deemed that the probability of malfunction (s) of other safety related equipment was too low to ever require the restart of a LPSI pump.
- 7. Will the margin of safety as defined in the basis for any technical specification be reduced?
No.' The technical specification basis does not define any safety limit that is dependent upon successful post RAS LPSI pump operation. Deleting the SAR references to the post RAS LPSI pump operation has no affect on any safety analysis or margins of safety defined in the basis or any other LBD.
y ' 1 ARKANSAS NUCLEAR ONE Page 7 F,ORM TITLE: FORM NO. REV. 10CFR50.59 SAFETY EVALUATION 1000.131B 3 PC-2
- 1. Will the probability of an accident previously evaluated in the SAR be increased?
No. De!ating information from the SAR about post RAS LPSI pump hydraulic limitations and use beyond their design basis does not alter any of the accident scenarios in the SAR. All of the analyzed accidents wiil remain in their previously analyzed category for frequency of occurrence. There are no probability changes caused by this deletion. 2. Will the consequences of an accident previously evaluated in the SAR be increased? No. The LPSI system acts to inject large amounts of borated water into the core following a SIAS signal when the RCS pressure drops below the pump's shutoff head. As part of the ANO2 design basis, the LPSI pumps are secured by the Recirculation Actuation Signal (RAS) and are not required to operate for long term core cooling or mitigation of core damage. Aligning the LPSI pumps to the Containment Spray System after a RAS would require an Operator entering the ECCS pump rooms and being exposed to radiation that is beyond the ANO2 design basis. Deleting the SAR references to post RAS use does not alter the LPSI design bases, therefore there is no added potential for increasing an offsite dose release. 3. Will the probability of a malfunction of equipment important to safety be increased? No. Although it is possible to use the LPSI pumps post RAS, there are no design requirements to be able to operate the pumps. DW3 ting the references in the SAR does not place any additional requirements on any of the ECCS or Containment Spray components, or any other safety related components. The LPSI pumps are not required to operate for long term core cooling. 4. Will the consequences of a malfunction of equipment important to safety be increased? No. The current ANO2 design basis and offsite dose calculations all recognize and account for the LPSI pumps being secured once long term recirculation occurs. Deleting the post RAS references in the U2 SAR does not affect the offsite dose calculations. , S. Will the possibility of an accident of a different type than any previously evaluated in the SAR be created? . No. The ECCS system has several specific functions for preventing and mitigating core damage in response to a SIAS signalincluding long term recirculation. None of the postulated failures for the ECCS system require restarting a LPSI pump after starting long term recirculation. Not using the LPSI pumps after a RAS is part of the ANO2 design, and not using them is bounded by di of the analyzed accidents in the U2 SAR. 6. Will the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR be created? No. The ANO2 design basis accounts for the LPSI pumps being secured for long term core cooling. The design of the ECCS system did not feel there would be sufficient need to restart the LPSI pumps post RAS. The design deerr.ed that the probability of malfunction (s) of other safety related equipment was too low to ever require the restart of a LPSI pump. 7 Will the margin of safety as defined in the basis for any technical specification be reduced? No. The technical specification basis does not define any safety limit that is dependent upon successful post RAS LPSI pump operation. Deleting the SAR references t , the post RAS LPSI pump operation has no affect on er.y safety analysis or margins of safety defined in the basis or any other LBD. 4
FORM N REtf. 14CFRSS.59 DETERMNum Page 1 of,4 Document No. ODCM Rev1 Change No. Rev.13 PC-2 Title Offsite Dose Oe!ca'-*3en Manual Brief description of proposed change: ODCM Section 2.1. "Radiame+ive Liouid Effluent Monitor Setooint" Steps #3 and M were revised to include the cess!NWa of reduced Unit 1 cirealet!na water cump flow due to circulatina water numa thiei;;;ns and/or cire"8a*Ina water bay ccaf'sw.iion. Add;i;onally, a minimum circulation w_Lr flow rate of 100.000 sss. was !!ee edeed to Sten M. Arssadix 1 Limi Anoendix 2. Limitation L3.3.1 =rs ravised to carrect a sravious ora oo.atical mie+=ka included in Revision 13 of the ODCM. Willthe proposed Activity: 1. Require a change to the Operating License including: Technical Specifications (excluding the bases)? YesO Nox Operating License? YesO Nox Confirmatory Orders? YesO Nox 2. Result in information in the following SAR documents (including drawings and text) being (a) no longer true or accurate, or (b) violate a requirement stated in the document: SAR (multi volume set for each unit)? Yesx nod Core Operating Limits Report YesO Nox Fire Hazards Analysis? YesO Nox Bases of the Technical Specifications? YesO Nox Technical Requirements Manual? YesO Nox NRC Safety Evaluation Reports? YesO Nox 3. Involve a test or experiment not described in the SAR? (See Attachment 2 for guidance) YesO Nox 4. Result in a potential impcci to the environment? (Complete the EnvironmentalImpact Determination of this fnrm.) YesO Nox 5. Result in the need for a Radiological Safety Evaluation per section 6.1.57 YesO Nox 6. Result in any poternialimpact to the equipment or facilities utilized for Ventilated Storage Cask activities per Section 6.1.67 YesO Nox 7. Involve a change u..:!sr 10CFR50.54 for the following SAR documents per Section 6.1.7: QAMO? . YesO Nox E-Plan? YesO Nox
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ARKANSAS NUCLEAR ONE F@RM TITLE: " FORM NO. REV. 1eCFR$0.80 CRTERMINATION 1900.131A 3 PC.1J Page 2 off Document No. ODCM RevjChange No. Rev.13 PC,-2
)
Basis for Determin=*!en IQa==*! ens 1. 2 & 31:
- Please see page 4.
O Proposed change does not require 10 CFR 50.59 Evaluation per Attachment 1. Item. (if# checked, note appropriate item #, send LDCR to Licensing). 1 Search Scope: List sections reviewed in the Licensing Basis Documents specified in Question 1,2 and 3. If a searcl q performed on LRS, the LRS search index should be entered under "Section" with the search s parentheses. Controlled hard copies of the documents shall be reviewed (LRS is not verified anl t text, not figures or drawings). Attach and distribute a completed LDCR per Section 6.1.2 If LB required. Document
' Section LRS:
50.59 Common All (Circ., Circulating w/5 water, liquid w/10 release *, liquid w/10 effluent *, liquid w/10 waste) l MANUAL SECTIONS: U1 TS and bases 3.25.1, 6.8.5,6.9,6.14 U1 SAR 2.8.1,1.4.51,1.4.56,11.0, all chapter 11 tables U2 TS and bases 3/4.11.1,6.9.3,6.9.10 U2SAR 2.4.12, 3.1.6,11.0, all chapter 11 tables QA Manual 16.4 FIGURES: N/A- N/A
)
b - Grea Stephenson dedifie evieweh Signature 1/7/99 Printed Name Date Reviewers certification expiration date: 1/5/00 Assistance previded by: Printed Name Scope of Assistance Date l Search ScyoReview Acceptability (NA, if performed by Technical Review per 1000.006) Af/F i'%- _ k-4 Certified Reviewers dignature 4 esc AA l/9/fi Printed Name Date
f FORM TITLE: ' ! 3ORM NO. REV. 10CFR50.58 DETERMINATION 1000.131A 3 Page 3 of 4 l ENVIRONMENTAL IMPACT DETERMINATION (UNIT 1 and UNIT 2) Document No. QQQd Rev1 Change No. Rev.13 PC-2 Complete the following Determination. If the answer to any checklist item is "Yes", an Environmental Evaluation I is required. See Section 6.1.4 for additional guidance. I
'Will the Activity being evaluated: l Yes No O x Disturb land that is beyond that initially disturbed during construction (i.e., new construction of buildings, creation or removal of ponds, or other terrestrial impact)? See Unit 2 SAR Figure 2.5-17. This applies only to areas outside the protected area.-
O x increase thermal discharges to lake or atmosphere? O x increase concentration of chemicals to cooling lake or atmos;here through discharge canal or tower? O x increase quantity of chem 4:als to cooling lake or atmosphere through discharge canal or tower? O x Modify the design or operation of cooling tower which will change drift characteristics? O x install any new transmission lines leading offsite? O x Change the design or operation of the intake or discharge structures? O x Discharges any chemicals new or different from that previously discharged? O x Potentially caIuse a spill or unevaluated discharge which may effect neighboring soils, sur water or ground water? O x
. involve burying or placement of any solid wastes in the site area which may effect runoff, surface water or ground water?
O x involve incineration or disposal of any potentially hazardous materials on the ANO site? O x Result in a change to nonradiological effluents or licensed reactor power level? O x Potentially change the type or increase the amount of non-radiological air emissions from the ANO site. l l l i I L j
asouussas Nuct. EAR ONe FORM Tm.E; FORM No. REV. 10CPR$0.80 REVEW CONTINUATION PAGE 1a00.131C 3
, Page f of f Document No. 92G3 Rev1 Change No. Rev.13 PC-2 10CFR50.59 Review Continuation Pace Discussion-The Unit 1 and Unit 2 liquid effluent release limits for the dose commitment to a member of the pu provided in the ODCM. The ODCM 3mits provide assurance that effluent releases will result in concentrati below the limits of 100FR20 for dose and concentration. Currently, ANO has set administrative 1% of the ODCM dose limits. Prior to the discharge of any radioactive liquid release tank, a calculatio performed to determine the MPC% at site boundary, dose, dose rate, and radmonitor setpoints associa the release. Calculation variables include tank activity, tank volume, release flow rate, circ water rate, and quarterly dilution volume. The decision to release radioactive liquid is based on the results of th calculations. '
The change to the ODCM is to allow the release of radioactive liquid from the site with no less gpm 100,000 thari of circulating water pump dilution flow. Normally, two or more circulating water pumps are in operation. However, there are certain occasions when it is necessary to throttle the condenser waterbo 3630, CV 3626, CV-3622, and CV 3618). With these valves throttled, the circulating water pump f below the normal igt,500 gpm. An evaluation was performed to determine what affect reduce flow would have on liquid release doses, dose rates, and MPC percentages when releasing hig minimum dilution rate of 100,000 gpm (supplied by U1 SYE) and activity from a previously activity) were entered into the Effluent Monitoring system (EMS) Vax computer program to det doses and MPC percentages. Both the doses and MPC% were only minimally affected. Based on evaluation, release of radioactive liquid can be performed with no less than 100,000 gpm of dilution flow. Chemistry Procedure 1604.017, ' Analysis of Uquid Waste", requires added reviews be performed by C Supervisors or designees prior to the release of liquid waste during periods of reduced dilution flow. Appendix 1 Limitation L3.3.1 and Appendix 2 Limitation L3.3.1 are also being revised to correc error that was, introduced during Revision 13. The words " prior to" are being replaced with "by* to b with what is found in U1 TS Section 6.12.2.5 and Unit 2 TS Section 6.9.4. The error occu Radiological Effluent Technical Specifications into the ODCM. Bagis for Determination-Question 1: Will the proposed activity require a change to the Operating 1.icense? An'swer: This ODCM change involves allowing the release of radioactive liquid from the 100,000 gpm of circulating water pump dilution flow. The Operating License does not address circulating water pumps or specKy a discharge flowrate required to release liquid from the site. Therefore, addition no oforthis longertrue requirement accurate. does not result in any information contained within the Ope E
FORM TITLE- '
~FCAM NO. REV.
10CPR$0.89 REVEw CON 7WuATION PAGE te00.131c 3 Questi:n 2: Will the prop: sed activity result in the CAR dgcuments (COLR, SAR, QAMO, EP, FHA, TS Bases, SER, & TRM) (including drawings ard text) being (a) no longer true or accurate, or (b) vio requirement stated in the document? Answer: This ODCM change involves allowing the release of radioactive liquid from the site with no l 100.000 ppm of circulating water pump dilution flow. The Unit 1 SAR 11.1.2.4 mentions that ' liquid w is diluted in the discharge canal to concentrations far below those stipulated in 10CFR20." Section 11.1.3 describes the design basis for the liquid effluent system. The Unit 1 SAR does not provide specif for circulating water pump operation during normal operation of the plant. The U2 SAR 11.2.6.4 also describes the' design oasis for the liquid effluent system. Section 11.2.8 stat "The amount of flow available for diluting the waste is a function of the number of unit 1 circu;ati in operation. The maximum dilution flow will be 766,000 pm with four circulating water pumps in op the minimum flow will be - 383,000 gpm with two circulating water pumps in operation, Section 11.2.8 nee be changed to address limited occasions when the unit 1 circulating water pumps are throttled and lower flows are experienced. No further changes are necessary to the Unit 2 SAR. Question 3: Will the proposed activity involve a test or experiment not described in the SAR? Answer: This procedure change does not involve a test or experiment that is not described in the SAR.
)
I l FORM MLE: AMAEAs NuC4. EAR ONE ~; FORM NO. REV. 10CPR80.89 EVAL UADoN 1000.1318 3 Pc.2 Page1of,2 10CFR50.59 Eval. No. fM49 O (AssiC9ed by PSC) Document No. QREM Rev> Change No. Bay,13 PC-2 Title _ r Offsite Dose *=!caa8-*8en Manual A WRITTEN RESPONSE PROVIDING THE BASIS FOR THE ANSWER ATTACHED. EACH QUESTION MUST BE ANSWERED SEPARATELY. A CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES GUI If the answer to any question on this form is "Yes," then an unreviewed safety questio to all questions is "No;" then the proposed change does not involve an unreviewed safety qu . 1. Will the probability of an accident previously evaluated in the SAR be increased? Yes No x Reducina does not have any thee d!! den i.ct on flow rate the accidents in thewhea SAE!. sensirains a msaltered and knl previously ev=!a TheseI6re. the prob =b!!itV of an accident I _d in tlw *** is not incr-end. 2. Will the consequences of an accident prev ously evaluated in the SAR be incesased? Yes No x
*dsesReducina the d!!M!cn flow rate when oerformina a mcaitsred and k not have any affect on the accidents in the SSR. Therefpr,35the dose consecuen ;
accident previousiv ov=!"t3 in the SAR is not incras==d. l 3. Will the probability of a malfunction of equipment important to safety be increased? Yes O No x This chance involves !!:-wins less than 100.000 releases of r=d!e=a+!ve linuid with a minimu:ti dilMien rate o safety; s ... The chanos dces not d;iect the csei Gon of any eauisment important to throtgina cire water flow has oreviously h.;a evaluei d. thsiefore, this ss,i, does not ' increase the siebEMGiv of a malfunction of enuism.at imessi nt to safety . 4. Will the consequences of a malfunction of equipment important to safety be increased? Yes No x This chance involves a!!-:;;ias
'a== thea 100.000 ---- -- releases of radioective !!ee!d with a minimum dilu+!en rate of n The chance does not ;; rect the es isGsa of any eaa49..ent imositant to safetv.._..I ee^ there;es.. this ch
.' : -::ss.1 to safety. ase dses not incr==== the c'-sEe esasesuences of a malfunction 5.
. Willthe possibility of an accident of a different type than any previously evaluated in the SAR be created?
Yes O No x 3 chance involves n!!- wlas reL agg,pf mdioactive !!ee!d with a minimum dilutisa rate o than 100.000 ===="-- E ':::: are ? e;i.c si;: ;siaarv e=8culm*! ens slic;;d
-Ie= ==88 -
o_........ aufllcies.Jv low'
* -a rates for =-:here.. This ch=nse dGes not create the sessibility of an ac-:!nnt of dNiettet t.se than any ere;isusiv eva!"-^rd in the SAR.
AnnAmbab huus.a:An one: FORM TITLE: FORM NO. REV. 14CFR50.59 kVALUATBoN 1000.1318 3 PC 2 6. Will the possibility of a malfunction of equipment important to safety of a riifferent type than any prev 8'asty evaluated in the SAR be created? Yes O No x This chance involves anc;;;r.s r '::::: of rse!E= dive !!ee!d with a minimum dilution rate of no isss than 100.000 cs.... The char.as dses not G;..ct the sssie;;sa of any eau lsa. sat important to s.;.ev* there; sis. this chs;;se ec+E not create the arabsb!!!!v of a malfunction of eauipment imr,sstsat to safety of a d;^;.res;; tvss than any s..visusiv sys!ead in the SAR, i 1 1 7. Will the margin of safety as defined in the basis for any technical 1 specification be reduced? ; Yes Nox The Technical Soecifiediens and SAR's of sith.i unit do not csatsin any maroin reistsd to reducino the dilution f:Gw. The ODCM ::sa;=Csas are srsv; fed to er.eure that the cGaces.;i iion of ; radioactive mster;als 7 ' :sse in !!ee!d - === ...n .. .-.. the site to unisei.;ctsd areas will be l i less than the concsi.t.-i;Ga ::v;:; ?==cNied in 16CFR20. Maadix B. Tabh II. Allowir>n releases witii dilution f:s;;;; no isss than 100.000 ssia dess not csis.srsiaise the ability to sr.;;t 10CFR20 limith 1 Groo Stephenson 1/7/99 Certified Efeviewers liignature - Printed Name Date Reviewers certification expiratiori date: 1/5/00 Assistance provided by: Printed Name Scope of Assistance Date PSC review by: 4 Date: l D 4
- n. u . n . .u n. . . . .
FORM TITLE: FORM NO. REV. 10CFR50.59 DETERMINATION 1000.131A 3 PC-1 Page lof d D3cument No SDID 2-98-0153 RevjChange No. N/A Title Unit 2 SAR Sections 7.7.1.1.4 and 10.4.7.2. Revise Description of the Feedwater Control System Brief description of proposed change: Revise statement that only the feedwater control valve is used in the low ower mode. Willthe proposed Activity:
- 1. Require a change to the Operating License including:
Technical Specifications (excluding the bases)? YesO No@ Operating License? YesO No@ Confirmatory Orders? YesO No@
- 2. Result in information in the following SAR documents (including drawings and text) being (a) no longer true or accurate, or (b) violate a requirement stated in the document:
SAR (multi-volume set for each unit)? Yes@ nod l Core Operating Limits Report YesO No@ Fire H&rards Analysis? YesO No@ Bases of the Technical Specifications? YesO No@ Technical Requirements Manual? YesO No@ NRC Safety Evaluation Reports? YesO No@
- 3. Involve a test or experiment not described in the SAR?
YesO No@ (See Attachment 2 for guidance)
- 4. Result in a potential impact to the environment? (Complete the Environmental Impact Determination of this form.) YesO No@
- 5. Result in the need for a Radiological Safety Evaluation per section 6.1.57 YesO No@
- 6. Result in any potential impact to the equipment or facilities utilized for Ventilated Storage Cask activities per Section 6.1.6? YesO No@
- 7. Involve a change under 10CFR50.54 for the following SAR documents per Section 6.1.7:
QAMO? YesO No@ E-Plan? YesO No@ Basis for Determination fQuestions 1. 2 & 31: See continuation sheet. O Proposed change does not require 10 CFR 50.59 Evaluation per Attachment 1, item # (if checked, note appropriate item #, send LDCR to Licensing.)
m ARKANSAS NUCLEAR ONE FORM TITLE: FORM NO. REV. 10CFRSO.59 DETERMINATION 1000.131A 3 PC 1 Page 2 of,4 Document No. SDID 2-98-0153 RevlChange No. ItA Search Scope: List sections reviewed in the Licensing Basis Documents specified in Question 1,2 ud 3. If a keyword search was done on LRS, "all" may be entered under "Section" with the keyword (s) used in parentheses. Controlled hard copies of the documents shall be reviewed (LRS is not verified and searches only text, not figures or drawings). Attach and distribute a completed LDCR per Section 6.1.2 If LBD changes are required. Docum.2!11 Section LRS: 50.59 - Unit 2 All(feedwater control) MANUAL SECTIONS: Unit 2 SAR 7.7.1.1.4 and 10.4.7.2 FIGURES: Unit 2 SAR none SECTIONS REQUIRING CHANGE: Unit 2 SAR 7.7.1.1.4 and 10.4.7.2 ue & Ceitifie' d~ Reviewer's Signature linces (%) Printed Name
,lil9r
~ d Dhte Reviewer's certification expiration date: Ob Assistance provided by:
Printed Name scope of Assistance Date Julie D. Jacks 50.59 Research 5-19-98 Pat Riedmueller Technical inout Z-2-H Sea iSco e 'ew Acceptability (NA, if performed by Technical Review per 1000.006) 1 L. A ve.a
, n- . Juua v ,/,ler Ceitified Review 6rik Signature '
Prirded Name ' ' ' ' Date { l i
\
n m um.wm vue FORM TITLE:
. FORM NO. REV.
10CFRSO.80 oETERMINATioN 1000.131A 3 PC-1 Page 3 of_4 ENVIRONMENTAL IMPACT DETERMINATION (UNIT 1 and UNIT 2) Document No. SDlD 2-98-0163 Rev1 Change No. .NyA Complete the following Determinaticn. If the answer to any checklist item is "Yes", an Environm is required. See Section 6.1.4 for additional guidance. Willthe Activity being evaluated: Yes No O E Disturb land that is beyond that initially disturbed during construction (i.e., new construc buildings, creation or removal of ponds, or other terrestrial impact)? See Unit 2 SAR 2.5-17. This applies only to areas outside the protected area. O 2 increase thermal discharges to lake or atmosphere? O E increase concentration of chemicals to cooling lake or atmosphere through discharge c tower? O @ increase quantity of chemicals to cooling lake or atmosphere through discharge canal or tower? O E Modify the design or operation of cooling tower which will change drift chartcteristics? O 3- Install any new transmission lines leading offsite? O @ Change the design or operation of the intake or discharge structures? O E Discharges any chemicals new or different from that previously discharged? O E Potentially cause a spill or unevaluated discharge which may effect neighboring soils, sur water or ground water? O 2 involve burying or placement of any solid wastes in the site area which may effect runo
. surface water orground water?
O O involve incineration or disposal of any potentially hazardous materials on the ANO site? O E Result in a change to nonradiological effluents or licensed recctor power level? O E Potentially change the type or increase the amount of non-radiological air emissions from the ANO site. l \ 1
)
t.
p ARKANSAS NUCLEAR ONE FORM TITLE: FORM NO. REV. 10CFRSO.69 REVIEW CONTINUATON PAGE 1000,131C 3 Page 4 of,4 Document No. SDID 2-984163 Rev> Change No. N/A 10CFR60.59 Review Continuation Paae Basis for Determination. QueMions 1. 2. and 3:
- 1. Technical Specifications (excluding the Bases), Operating License, and Confirmatory Orders:
- 2. " Safety Analysis Report" - SAR (multi-volume set), COLR, FHA, TS Bases, TRM, NRC SERs:
in the Unit 2 SAR, Sections 10.4.7.2 and 7.7.1.1.4 describe the Feedwater Control System (FW
. description states that in the low power mode, feedwater flow rate is controlled by only the fe control valve, and in the high power mode, feedwater is controlled using feedwater pump speed an bypass contml valves. This is an over-simplification that is somewhat misiesdiag.
The FWCS valve and pump programming logic is not directly a function of the mode in which th operating. The FWCS shifts to the high power mode at ~20% increasing (as determined by shifts to low power mode at -15% decreasing. At these power levels (Iow power mode), th ccatrol better valve and description the would bemain feedwater as follows: regulating valve move as required to maintain steam generat The two steam generators are operated in parallel with each generator's Feedwater Control System (FWCS) maintaining its downcomer water level within acceptable limits. The FWCS for each steam generator consists of a low power mode and a high power mode, in the low power mode, the FWCS operates asis a single element control system based on steam generator level.
- S;;.
the ::cf. 2 th: f::f;;:!:: 5t nt: by ;::"!: ' .;; .!y th: 1::fff:: by;- : de!
!n the high power mode, the FWCS operates asis a 4three-element control system based on steam generator level, feedwater flow, and main steam flow. The FWC8 program logic develops a flow demand which that pm !t: 21. emet controls of-the main and bypass control valves and feedwater pump speed to maintain steam generator levels at the desired
.setpoint.
This revised paragraph should replace the identical paragraphs in Section 7.7.1.1.4 and 10.4.7.2. The other documents listed above do not discuss the FWCS to this level of detail and thus clarification.
- 3. Test or Experiment not described in the SAR:
1 This change does not involve any tests or experiments on any components, systems, or group i
m ~ ow e n unc ) FORM TITLE:
. FORM NO. REV.
10CFR50.49 EVALUATION l 1000.1313 3 l I Page1of 2 10CFR50.5g Eval. No. FFW% MI (Assigned by PSC) Document No. SDID-2484153 Rev> Change No. g Title Unit 2 SAR Sections 7.7.1.1.4 and 10.4.7.2. Revise P=EOisilon of the Feedwater Control System A WRITTEN RESPONSE PROVIDING THE BASIS FOR THE ANSWER TO EACH Q ATTACHED. EACH QUESTION MUST BE ANSWERED SEPARATELY. A SIMPLE S CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES GUIDANCE FOR R If the answer to any question on this form is "Yes," then an unreviewed safety question is involved. If to all questions is "No," then the proposed change does not involve an unreviewed safety question.
- 1. Will the probability of an accident previously evaluated in the SAR be increased?
Yes O No @ The proposed revision to Unit 2 SAR sections 7.7.1.1.4 and 10.4.7.2 provides a clarification to th description of operation of the Feedwater Control System (FWCS). The current SAR descript that only the bypass feedwater regulating valves are modulated by the control system when in the lo power mode of operation. The FWCS valve and feedwater pump programming logic is not a function of whether the system is in the low power or high power mode. Instead, the main and bypass feedwate regulating valves, and main feedwater pumps modulate as required to maintain adequate steam levels in both modes. The proposed SAR revisions have no impact on event initiators. Therefor probability of an accident previously evaluated in the SAR will not be increased.
- 2. Will the consequences of an accident previously evaluated in the SAR be increased?
Yes O No @ The proposed SAR revision does not alter any accident analyses assumptions or responses and as has no impact on the consequences of any analyzed accidents. Therefore, the consequences of an accident previously evaluated in the SAR will not be increased, t
- 3. '
Will the probability of a malfunction of equipment important to safety be increased? Yes O No @ ; The proposed SAR revision provides a clarification to the description of operation of the FWCS. The clarified description of operation of the FWCS does not impact any equipment considered to be imp safety. Therefore, the probability of a malfunction of equipment important to safety will not be increased. 1 4. Will the consequences of a malfunction of equipment important to safety be increased? Yes O No E The proposed SAR revision does not alter any equipment responses of the accident analyses and as suc has no impact on any' safety related equipment. Therefore, the consequences of a malfunction of equipment important to safety will not be increased. 5.
. Will the possibility of an accident of a different type than any previously evaluated in the SAR be created?
Yes O No @
ARKANSAS NUCLEAR ONE FORM TITLE: FORM NO. REV. 10CFRSO.89 EVALUATION 1000.131B 3 Page 2 of_g The proposed SAR revision is bounded by existing accident analyses. Therefor accident of a different type than previously evaluated in the SAR is not created. 6. Will the possibility of a malfunction of Equipment important to safety of a different type than any previously evaluated in the SAR be created? Yes O No @ The proposed SAR revision has no impact on any equipment important to safety an the possibility of a malfunction of any equipment important to safety. malfunction not created. of equipment important to safety of a different type than any previ 7. Will the margin of safety as defined in the bases for any technical specification be reduced? Yes O No @ No margins respect of safety to the proposed arerevision. SAR determined to be involved in the ANO-2 Technical Specifica specification basis will not be reduced. Therefore, the margin of safety as defined in the technical id&>1) C6tiified Reviewers Signature li w k b n J vl1la Printed Name
' ' Date Reviewers certificat!on expiration date: /37 Assistance provided by:
Printed Name Scope of Assistance Pat Riedmueller Date Technical 7-2-98 PSC review by: lAr-- Date: b D
i ARKANSAS NUCLEAR ONE PaQe 1 FORM TITLE: FORM NO. REV. 10CFR50.59 DETERMINATION 1000.131A 3 PC-1 This Document contains 3 Pages. Document No. LDCR ANO-2 SAR Section 6.3 Rev> Change No. Title LBLOCA Analysis Results Correction for CR-ANO-2-1999-0110 Brief tiescription of proposed change: l The calculated peak cladding temperature and maximum cladding oxidation for the limiting large break LOCA, the 0.6 square foot double ended guillotine break at the pump discharge with increased tube plugging and reduced RCS flow, have increased from 2158'F to 2169'F and from 7.2% to 7.5% respectively. ANO-2 SAR Section 6.3 text and tables are modified accordingly. Five SAR figures, representing the resu!ts of this limiting analysis, are also replaced. This change is a correction of an error made by ABB-CE during the original performance of this analysis. This error was documented in CR-ANO-2-1999-0110. Willthe proposed Activity:
- 1. Require a change to the Operating License including:
Technical Specifications (excluding the bases)? YesO No@ Operating License? YesO No@ l Confirmatory Orders? YesO No@ l 2. Result in information in the following SAR documents (incluJmg drawings and text) being (a) no longer true or accurate, or (b) violate a requirement stated in the document 1
. SAR (multi-volume set for each unit)? Yes@ nod Core' Operating Limits Report?
YesO NoS l ] 1 ! Fire Hazards Analysis? YesO NoS j Bases of the Technical Specifications? YesO NoS ! Technical Requirements Manual? YesO NoS NRC Safety Evaluation Reports? i YesO No@ I
- 3. Involve a test or experiment not described in the SAR?
YesO No@ (See Attachment 2 for guidance) l 4. Result in a potential impact to the environment? (Complete Environmental Impact Determination of this form.) YesO No@
- 5. Result in the need for a Radiological Safety Evaluation per section 6.1.57 YesO No@
- 6. Result in any potential impact to the equipment or facilities utilized for Ventilated Storage Cask activities per Section 6.1.6? YesO No@
- 7. Involve a change under 10CFR50.54 for the following SAR documents per Section 6.1.7?
QAMO7 YesO No@ E-Plan? YesO No@
ARKANSAS NUCLEAR ONE FORM TITLE: Pace R FORM No. REV. 10CFR80A9 DETERMINATION 1000.131A 3 PC-1,2 Document No. LDCR ANO-2 SAR Section 6.3 Rev1 Change No. F F@M -035 Basis for Determination (Questions 1, 2,8. 3): The proposed changes represent a correction to the LBLOCA ECCS analyses performed to s plugging and reduced RCS flow. SAR section 6.3 is the only licensing documentation that addre specific LBLOCA ECCS analysis results in any significant level of detail. Consequently, these res given anywhere else and this change affects no otherlicensing document. O Proposed charge does not require 10CFR50.59 Evaluation per Attachment 1, item # (if checked, note appropriate item #, send LDCR to Licensing). Search Scope: List sections reviewed in the Licensing Basis Documents specified in questions 1,2 and 3. If search w performed on LRS, the LRS search index should be entered under "Section" with the search st parentheses. Controlled hard copies of the documents shall be reviewed (LRS is not verified text, not figures or drawings). Attach and distributs a completed LDCR per Section 6.1.2 if LBD c required. Document Section LRS: All ("2158" and "7.2 w/10 oxi" and "72 w/10 oxi") MANUAL SECTIONS: ANO-2 SAR Section 6 FIGURES: SAR 6.3-27a through 6.3-27r b Certified Reviewers Signature ks) Q. feats 5'/Mh1 Printed Name Date Reviewers certification expiration date: 5//l/2000 Assistance provided by: Printed Name scope of Assistance JacQue Lingenfelter Date LRS and SAR search, draft determination 5/10/99 Searc Rev ceptability (NA, if performed by Technical Reviewer per 1000.006) tertified Reviewers Signature Svl U Ov} ElN l99 Printed Neme / Ddte
TT ARKANSAS NUCLEAR O NE Page 3 FORM TITLE: l FORM NO. REV. 10CFR50.59 DETERMINATION 1000.131 A 3 l l ENVIRONMENTAL IMPACT DETERMINATION (UNIT 1 and UNIT 2) l l Document No. LDCR ANO-2 SAR Section 6.3 Rev> Change No. Complete the following Determination. If the answer to any item below is "Yes", an Environmental Evaluation is required. See Section 6.1.4 foradditionalguidance. Willthe Activity being evaluated: Y!Ui N2 O E Disturb land that is beyond that initially disturbed during construction (i.e., new construction of buildings, creation or removal of ponds, or otherterrestrialimpact)? See Unit 2 SAR Figure 2.5-17. This applies only to areas outside the protected area. O E increase thermal discharges to lake or atmosphere? O @ increase concentration of chemhals to cooling lake or atmosphere through discharge canal or l tower? * ! O E Increase quantity of chemicals to cooling lake or atmosphere through discharge canal or tower? O 2 Modify the design or operation of cooling tower which will change drift characteristics? O E install any new transmission lines leading offsite? O S change the design or operation of the intake or discharge structures? O E Discharges any chemicals new or different from that previously discharged? O @ Potentially cause a spill or unevaluated discharge which may effect neighboring soils, surface water or ground water? l ! O E involve burying or placement of any solid wastes in the site area which may effect runoff, surface water or ground water? O 2 involve incineration or disposal of any potentially hazardous materials on the ANO site? O @ Result in a change to nonradiological effluents or licensed reactor power level? i l O 2 Potentially change the type orincrease the amount of non radiological air emissions from the ANO site. l
ARKANSAS NUCLEAR ONE FJRM TITLE: Prgi 1 FORM NO. REV. l 10CFR50.59 SAFETY EVALUATION 1000.131B 3 PC-2 l This Document contains 3 Pages. Document No. LDCR ANO-2 SAR RevlChange No. Section 6.3 10CFR50.59 Eval. No. (Assigned by PSC) Title LBLOCA Analysis Results Correction for CR-ANO-21999-0110 A WRITTEN RESPONSE PROVIDING THE BASIS FOR THE ANSWER TO E ATTACHED. EACH QUESTION MUST BE ANSWERED SEPARATELY. A SIMP CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES GUIDANCE If the answer to any question on this form is "Yes," then an unreviewed safety question is invo to all questions is "No," then the proposed change does not involve an unreviewed safety question. 1. Will the probability of an accident previously evaluated in the SAR be increased? YesO No @ 2. Will the consequences of an accident previously evaluated in the SAR be increased? YesO No @ 3. Will the probability of a malfunction of equipment important to safety be YesO increased? No @ 4. Willthe consequences of a malfunction of equipment important to safety be increased? YesO No @
- 5. )
Willthe possibility of an accident of a different type than any previously YesO evaluated in the SAR be created? No @ ( 6. Willthe possibility of a malfunction of equipment important to safety of a YesO different type than any previously evaluated in the SAR be created? No @ 7. Will the margin of safety as defined in the basis for any technical specification be reduced? YesO No @ AN. h Certified Reviewer's Signature J u :eJ O Fea:ts s/io/rr Printed Name Date Reviewer's certification expiration date: r////2000 Assistance provided by: Printed Name Scope of Assistance Jacque Lingenfelter Date Draft evaluation , S/10/99
/ ___
A ./ PSC review by:
/' Date: 5//zM
( '
/
l l l
ARKANSAS NUCLEAR ONE - Pagi 2 FORM TITLE: FORM NO. REV. 10CFR50.59 REVIEW CONTINUATION PAGE 1000.131C 3 2. Document No. LDCR ANO-2 SAR Section 6.3 Rev/ Change No. , a 10CFRSO.59 Review Continuation Paoe 1. Will the probability of an accident previously evaluated in the SAR be increased? No., n The proposed changes correct errors in the results of the limiting large break LOCA analysis previo , reported in the SAR. The calculated values for peak cladding temperature and maximum dadding oxidation , have increased slightly. These changes are only to the analysis results; no changes have been made to the assumed plant configuration or any other analysis input. There are no new systems, components, substructures, design changes, physical alterations, or operating procedure changes being propo .; this change. The analysis results are in no way related to any accident precursor. The modification of these results will have no impact on the probability of an accident previously evaluated in the SAR. 2. Will the consequences of an accident previously evaluated in the SAR be increased? No The proposed changes correct errors in the results of the limiting large break LOCA ECCS analysis 3 previously reported in the SAR. The calculated value for peak dadding temperature has increased from 2158'F ., to 2169'F and the maximum dadding oxidation has increased from 7.2% to 7.5%. Although these values have ~ increased, they are still within the acceptance criteria established to demonstrate proper operation of the emergency core cooling system (2200*F and 17% respectively). These minor changes do not alter the conclusion that the ECCS performance is acceptable. There are no new systems, components, j substructures, design changes, physical alterations, or operating procedure changes being proposed b these changes. The consequences of accidents previously evaluated in the SAR are unchanged by the proposed changes. 3. Will the probability of a malfunction of equipment important to safety be increased? No i The proposed changes correct errors in the results of the limiting large break LOCA analysis previously ,. reported in the SAR. The calculated values for peak cladding temperature and max %um cladding oxidation 'j have increased slightly. These changes are only to the analysis results; no changes ;.;ve been made to the assumed plant configuration or any other analysis input. There are no new systems, components, substructures, design changes, physical alterations, or operating procedure changes being proposedyby these changes. There are no new or different accident conditions imposed on the ECCS equipment. Consequently, the probability of a malfunction of equipment important to safety will not be increased. .1,y. 3 4. Will the consequences of a malfunction of equipment important to safety be increased? No ,..y The proposed changes correct errors in the results of the limiting large break 1.OCA analysis previously . reported in the SAR. The calculated values for peak cladding temperature and maximum cladding oxidation have hereased slightly. These changes are only to the analysis results; no changes have been made to the an *.ed plant configuration, including operable or operating equipment, or to any other analysis input. There are '{~ no new systems, components, substructures, design changes, physical alterations, or operating . procedure changes being proposed by this change. The acceptable results of the ECCS performance . , . Enalyses, which consider all appropriate single failures (the assumption of no failures produces the most c limiting results for this analysis), demonstrate that the consequences of a malfunction of equipment important to safety will not be increased. 3i m
- 5. Will the possibility of an accident of a different type than any previously evaluated in the SAR be created? No
' The proposed changes correct errors in the results of the limiting large break LOCA analysis previously. reported in the SAR. The calculated values for peak cladding temperature and maximum cladding oxidation .,, i
ARKANSAS NUCLEAR ONE FORM mi.E: Pace 3 I FORM NO. REV. I 10CFR50.59 REVIEW CONTINUATION PAGE 1000.131C 3 l have incriased slightly. These changes are only to the analysis results; no changes have been made to the assumed plant configuration or any other analysis input. No plant modi 6 cations, new components, ' alterations, nor operating conditions are being implemented by this change; therefore no new accide are croataid and no currently non-limiting events are becoming more limiting. 6. Wiffthe possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR be created? No There are no new systems, components, substructures, physical design changes, physical alte nor opieriting procedure changes being proposed by this change. The proposed changes correct erro s in the re' 'ults of the limiting large break LOCA analysis previously reported in the SAR The calculated for peak cladding temperature and maximum cladding oxidation have increased slightly. The to the anafysis results; no changes have been made to the assumed plant configuration or any other input. As there are no physical changes to the plant, the possibility of a malfunction of equipment important t c' to safety of a different type than any previously evaluated in the SAR will not be created. 7. Will the margin of safety as defined in the bases for any technical specification be reduced?
.:: No The proposed changes correct errors in the results of the limiting large break LOCA analysis pr reported iA the SAR. The calculated values for peak cladding temperature and maximum cladding oxidation have increased slightly. These changes are only to the analysis results; no changes have been made to the assumed plant configuration or any other analysis laput. The analysis inputs remain conservative with respect to the Technical Specification and COLR limits. The analysis was performed consistent with approved methodology and the ECCS performance results are still within the established acceptance criteria.
these changes.The margin of safety as defined by the bases for the technical specifications are unaffec 4 e t 1 i o a : i 1 1
p Axuumas NuvLt.m unt: ! FORM TITLE:
. FORM NO. REE 10CFR60.89 DETERMINATION 1000.131A 3 PC.1 l
Page 1 of 3 Document Nn. QAMO Rev3 Change No. 20 l Title Quality Assurance Manual Ooerations Brief description of proposed change: Transfer information in Accendix B to the Units 1 and 2 SARs i Will the proposed Activity: I
- 1. Require a change to the Operating License including:
Technical Specifications (excluding the bases)? YesO No x0 Op'erating License? - YesO No x0 ; Confirmatory Orders? YesO Nox0 2. Result in information in the following SAR documents (including drawings and text) being (a) no longer true or accurate, or (b) violate a requirement stated in the document: SAR (multi-volume set for each unit)? Yesx0 nod Core Operating Limits Report YesO No x0 Fire Hazards Analysis? YesO No x0 Bases of the Technical Specifications? YesO No x0 Technical Requirements Manual? YesO No x0 NRC Safety Evaluation Reports? YesO No x0
- 3. Involve a test or experiment not described in the SAR?
(See Attachment 2 for guidance) YesO No x0
- 4. Result in a potentialimpact to the environment? (Complete 1 i
the EnvironmentalImpact Determination of this form.) ' YesO Ne x0
- 5. Result in the need for a Radiological Safety Evaluation per section 6.1.07
'iesO No x0
- 6. Resu' lt in any potential impact to the equipment or facilities utilized for Ventilated Storage Cask activities per Section 6.1.6?
YesO No x0
- 7. Invo!ve a change under 10CFR50.54 for the following SAR documents per Section 6.1.7:
QAMO? Yesx0 NoO E-Plan? YesO No x0 W
ARKANSAS NUCLEIR ONE FORM TITLE: FORM NO. REV. 10cFR60.88 DETERMINATION 1000.131A 3 PC-1.2 Page J of 3 Document No. QAMO Rev> Change No. 20 Basis for Determination (Questions 1. 2 & 31: applicable sections of the Units 1 and 2 SARs.1. These proposed changes are Assurance Program Manual (QAPM). The ANO QAMO will be repl will not require a change to the Operating License documents. , 2. These proposed changes are to transfer information/ requirements from Appendix B of applicable sections of the Units 1 and 2 CARS. The applicable SAR sections will contain a and not beexisting affected byinformation this revision. will be enhanced by more accurate text from Appendix B. The oJ
- 3. These proposed changes are to transfer administrative information/ requirements from A experimenttonot QAMO the appilcable described sections of the Units 1 and 2 SARs. Therefore, this revision will no in the SAR.
C Proposed change does not require 10 CFR 50.59 Evaluation per Attachment 1, appropriate item #, send 'LDCR to Licensing). f Search Scope: List sections reviewed in the Licensing Basis Documents specified in Question 1,2 and 3. If a s perfonned on LRS, the LRS search index should be entered under 'Section" with the sea parentheses. Controlled hard copies of the documents shall be reviewed (LRS is not ve text, required.not figures or drawings). Attach and distribute a completed LDCR per Section 6.1.2 Document Section LRS: 50.59 - Common 'All" (*QAMO", " Quality Assurance Manual", "QAMO Appendix B",
" Quality Assurance Manual Operation Appendix B',
" Quality Program for Fire Protection" MANUAL SECTIONS:
FHA Unit 1 SAR 9.8.3.1,9.8.3.8 and Unit 2 SAR 9.5.1.5.1,9.5.1.5.8 FIGURES: none u Guy Kevin Floyd CertifipSiewers Signatg 4/7/99 Printed Name Date Reviewers certification expiration date: 12/15/99 Assistance provided by: Printed Name- Scope of Assistance Date Search Scope Review Acceptability (NA, if performed by Technical Review per 1000.006) kNh. Th// 11In,M M . 0 v v2 L AR- $flj Cenified Reviewers 5gjrlid6re Printed Name '
'Date
ARKANSAS NUCLEAR ONE EORM TITLE:
.. FORM NO. REV.
1eCFRSS.88 DETERMINATION teco.131A 3 ENVIRONMENTAL IMPACT DETERMINATION (UNIT 1 and UNIT 2) Document No. QAMO Rev/ Change No. .2Q Complete the following Determination. If the answer to any checklist item is "Yes", an Environme is required. See Section 6.1.4 for additional guidance. Will the Activity being evaluated: 1res M
,2 O 'x0 Disturb land that is beyond that initially disturbed during construction (i.e., new cortstructio buildings, creation or removal of ponds, or other terrestrial imi : 1)? See Unit 2 SAR Figu 2.5-17. This applies only to areas outside the protected area.
O x0 increase thermal discharges to lake or atmosphere? O x0 Increase concentration of chemicals to cooling lake or atmosphere through discharge cana tower? O x0 increase quantity of chemicals to cooling lake or atmosphere through discharge canal or tower? O x0 Modify the design or operation of cooling tower which will change drift characteristics? O x0 instati any new transmission lines leading offsite? O x0 Change the design or operation of the intake or discharge structures? O x0 Discharges any chemicals new or different from that previously discharged? O x0 Potentially cause a spill or unevaluated direharge which may effect neighboring soils, surface water or ground water? O x0 involve burying or place ont of any solid wastes in the site area which may effect runoff, surface water or ground water? O x0 1.lvolve incineration or disper,al of any potentially hazardous materials on the ANO site? O x0 Result in a change to nonradiological effluents or licensed reactor power level? O x0 Potentially change the type or increase the amount of non-radiological air emissions from the ANO site.
F ARKANSAS NUCLEAR ONE FORM TITLE: l FORM MO. REV. ! 10CFR60.69 EVALUATION 1000.1318 3 PC-2 l Pagel of_2 10CFR50.59 Eval. No. FF#M -03e (Assigned by PSC) Document No. QAMO RevJChange No. 20
)
Title Quality Assurance Manual Operations A WRITTEN RESPONSE PROVIDING THE BASIS FOR THE ANSWER TO EACH QUESTION MUST ATTACHED. EACH QUESTION MUST BE ANSWEREJ SEPARATELY. A SIMPLE STATEMENT OF CONCLUSION IS NOT SUFFICIENT, ATTACHMENT 2 PROVIDES GUIDANCE FOR RESPONSE. If the answer to any question on this form is "Yes,"than an unreviewed safety question is involved. If the answer ! to all questions is "No," then the proposed change does not involve an unreviewed safety question. l
- 1. Will the probability of an accident previously evaluated in the SAR be increased?
Yes O No @ l
- 2. Will the consequences of an accident previously evaluated in the SAR be l increased? l Yes O No @
- 3. Will the probability of a malfunction of equipment important to safety be incre tsed?
Yes O No S
- 4. Will the consequences of a malfunction of equipment important to safety be increased?
Yes O No @
- 5. Will the possibility of an accident of a different type than any previously evaluated in the SAR be created?
Yes O No @
- 6. Will the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR be created?
Yes O No @
- 7. Will the margin of safety as defined in the basis for any technical specification be reduced?
Yes O No @ \ 0Y$- fruv vsh oV Y~$~$ l CertifiepevTfwers Signature / / Printed Name/ Date Reviewers certification expiration date: Q/IS /47
/
Assistance provided by: Printed Name Scope of Assistance Date PSC review by- /
/ Date: /
,/
/ /
ARKANSAS NUCLEAR ONE F,ORM TITLE: FORM NO. REV. 10cFR50.59 REVIEW CONTINUATION PAGE 1000.131c 3 Page 2 ofj l Document No. QAMO Rev/ Change No. 20 l 10CFR50.59 Review Continuation Paae Question #1: No accidents identified in the SAR will be affected by this revision. This revision is to transfer administrative information/ requirements presently in Appendix B of tne QAMO to the applicable SAR sections. Therefore, the probability _of an accident previously evaluated in the SAR will not be increased. Question #2: No accidents identified in the SAR will be effected by this revision. This revision is to transfer administrative information/ requirements presently in Appendix B of the QAMO to the applicable SAR sections. Therefore, the consequences of an accident previously evaluated in the SAR will not be increased. Question #3: This revision is administrative and does not effect the function of any equipment important to safety. Therefore, the probability of a malfunction of equipment important to safety will not be increased. Question #4: This revision is administrative and does not effect the function of any equipment important to safety. Therefore, the consequences of a malfunction of equipment important to safety will not be increased. Question #5: This revision will not create a new accident scenario which is outside the bounds of existing accidents evaluated in the SAR. This revision is administrative in transferring information from the QAMO to the applicable SAR sections. Therefore, the possibility of an accident cf a different type than any previously evaluated in the SAR will not be created. Question #6: This revision is administrative and does not effect the function of any equipment imoortant to safety. Therefore, ' the possibility of a malfunction of equipment important to safety of a different ty?e han previously evaluated in the SAR will not be created. Question #7: The basis of the Technical Specifications do not specify a margin of safety with respect to this portion of the fire protection program. Therefore, the margin of safety as defined in the basis for any Technical Specification will not be reduced. i l l
=
ARKANSAS NUCLEAR ONE Papi i FORM TITLE: FORM NO. REV.
- 1 10CFR60,59 DETERMINATION 1000.131A 3 PC-1 Document No. 980366N201 , RevlChange No. O Title 2CV-1024-1 and 2CV-1074-1 Packing Leakoff Line Isolation Brief description of proposed change:
This modification will remove the pacirMg leakoff lines from MFW isolation valves 2CV 1024-1 and 2CV-10741. The lines run from the valves' boniset to an open floor drain. The connection at the valve will be plugged with a 1/2" threaded pipe plug which will be seal welded. A new packing configuration will be installed by maintenance. This type modification has been prepared for otNr Anchor Darling valves at ANO to eliminate waste water concems and possible stem erosion. Will the proposed Activity:
- 1. Require a change to the Operating License including:
Technical Specifications (excluding the bases)? YesO No@ Operating License? YesO No@ Confirmatory Orders? YesO No@
- 2. Result in information in the following SAR documents (including drawings and text) being (a) no longer true or accurate, or (b) violate a requirement stated in the document:
SAR (multi-volume set for each unit)? Yes@ NoO Core Operating Limits Report? YesO No@ Fire Hazards Analysis? YesO No@ Bases of the Technical Specifications? YesO No@ Technical Requlrements Manual? YesO No@ NRC Safety Evaluation Reports? YesO No@ j
- 3. Involve a test or experiment not described in the SAR?
YesO No@ (Gee Attachment 2 for guidance) l
- 4. Result in a potentialimpact to the environment? (Complete Environmental impact Determination of this form.) YesO No@
- 5. Result in the need for a Radiological Safety Evaluation per section 6.1.5? YesO No@
- 6. Result in any potential impact to the equipment or facilities utilized for Ventilated Storage Cask activities per Section 6.1.67 YesO No@
- 7. Involve a change under 10CFR50.54 for the following SAR documents per Section 6.1.77 QAMO?
YesO No@ ! E-Plan? YesO No@ (Y.G 5 3 MI b
I 1 ARKANSAS NUCLEAR ONE l FORM TITLE: Page 2 { FORM NO. REV. 10CFR50.59 DETERMINATifN 1000,131A 3 PC-1 I Document No. gdO366N201 I RevdChange No. 0 i Basis for Determination (Questions 1,2, & 3): ' k Backcround IE'_b __'EV.L & Main Feedwater isolation Valves 2CV-1024-1 and 2CV-1074-1 have packing leakoff l past the lower set of packing to floor drains in the Unit 2 UNPPR. Leakage past the lower
- normally not secured by adjusting the packing v hich creates a humidity and waste water conce
! Additionally, allowing high pressure water / steam to flow past the packing can result in valve vendor approved modification to the valve design ptovided by this change will remove the le valves and seal weld a threaded relug in the port. This modification has been performed { other Anchor at the leakoff port.Darling valves. Ad.fitional packing will be added separate from this design pa ' Question i k A review of the Unit 2 Technical Specifications, Operating License, and Confirmatory Ord l leakoff lines for the MFW isolation valves 2CV-1024-1 and 2CV-1074 Questien 2 l l i A review of all the documents listed in Question 2 was performed using LRS and a HARD SAR. SAR Figure 10.2-3, the P&lD for the Unit 2 Secondary System, will have to be revised deletion of the packing leakoff lines for 2CV-1024-1 and 2CV-1074-1. These packing lea stainless steel tubing that are open to area floor drains. The lines do not have line class d not shown on any drawing except for the system P&lD. ! Questiran 3 The only procedures. test associated with this modification is normal MOV testing in accordance with a I O Proposed change does not require 10CFR50.59 Evaluation per Attac.hment 1, appropriate item #, send LDCR to Licensing). Search Scope: List sections reviewed in the Licensing Basis Documents specified in questions 1,2 and was done on LRS, "all" may be entered under "Section" with the keyword (s) used in pare Attach and distribute a completed LDCR per Section 6.1.2 If L . Document Sectinn ! LRS: All (' packing' leakoff"2CV 1024-1'2CV-1074-1" ) MANUAL SECTIONS: Unit 2 SAR T.O.C. and Section 10 I F RES. Unit 2 SAR Figure 10.2 3
- J.) ) .
Stephen J. Lynn l Certifieg Rev
~
r's Signature Printed Name 6/2-N Date .
t l ARKANSAS NUCLEAR ONE Tage 3 FORM TITLE:
' FORM NO. REV.
'10CFR60.69 DETERMINATION 1000.131 A 3 PC-1 Reviewers certification expiration date: 6/03/99 Assistance provided by: '
Printed Name Scope of Assistance Date Search Scope view Acceptability (NA, if performed by Technical Reviewer per 1000.006) _ thw , Certified Reviewers Signature dieue/w cl 6. L swer- m-F rinted Name Y/74r Date l rm D 'E'. # O 1 l l l l l
l FORM TITLE: ARKANSAS NUCLEAR ONE Page c FORM NO. REV. 10CFR50.59 DETERMINATICN 1000.131 A 3 ENVIRONMENTAL IMPACT DETERMINATION (UNIT 1 and UNIT 2) l l Document No. 980366N201 Rev> Change No. O i Complete required. the following See Section Determination, 6.1.4 for additional guidance. if the ans,wer to any item below is "Yes", a Willthe Activity being evaluated: Yes Ng i O @ Disturb inno that is beyond that initially disturbed during construction (i.eI 2.5-17. This applies only to areas outside the protected ar O E increase thermai discharges to lake or atmosphere? O g increase tower? concentratic , of chemicals to cooling lake or atmosphere through O @ increase tower? quantity of chemicals to cooling lake or atmosphere through discharI O E Modify the design or operation of cooling tower which will change drift characte O E instati any new transmission lines leading offsite? O a change the design or operation of the intake or discharge structures? O @ Discharges any chemicals new or different from that previously discharged? O E Potentially water or groundcause water? a spill or unevaluated discharge which may effect neighbor O E involvewaur surface buryi g or placement or ground water? of any solid wastes in the site area which may e O @
-Involve incineration or disposal of any potentially hazardous materials on the ANO O e Resuit in a change to nonradiologicai effluents or licensed reactor power levei?
O E Potentially ANO site, change the type or increase the amount of non-radiological air emiss r g,: 1 YI ~
erumem huace ucc , FORM TITLE: ( FORM NO. REV. 10CFR50.59 EVALUATION 1000.131B 3 PC 2 Page of rl.Ti S D 10CFR50.59 Eval. No._ F W 6-1E (Assigned by PSC) ! Docurent No. 980366N201 Rev./ Change No. 0 Title 2CV-1024-1 & 2CV-1074-1 Packino Leak Off Line Isolation A WRITTEN RESPONSE PROVIDING THE BASIS FOR THE ANSWER TO EACH QUEST ATTACHED. EACH QUESTION MUST BE ANSWERED SEPARATELY. A SIMPLE STATE CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES GUIDANCE FOR RES If the answer to any question on this form is "Yes," then an unreviewed safety question is invNved. If the answ to all questions is "No," then the proposed change does not involve an unreviewed safety qt;nnon.
- 1. Will the probability of an accident previously evaluated in the SAR be
, increased? i Yes O No @ i The only event which will be affected by this nuclear change is packing leakage which is not a SAR l analyzed accident. Industry and ANO experience indicate that the probability of packing leakage will not be increased and may be decreased. l
- 2. Will the consequences of an accident previously evaluated in the SAR be increased?
i Yes O No @ ( The SAR accident analyses do not identify any offsite dose consequences resulting from valve pa leakage possibly because the pathway for the release of radioactivity is very small. The MFW system also not considered a high activity system. The proposed modification will have no effect on the dose consequences of any accident previously evaluated because the leakage will be treated the same, instead of being routed to the floor drain where gaseous activity would be released into the UNPPR, the le now would be at the stem / packing gland interface. In both cases, the activity is subject to processing auxiliary building ventilation equipment, and so the dose consequences are not increased by this moF. cation. 3. Will the probability of a malfunction of equipment important to safety be increased? Yes O No @ The only equipment important to safety which could be reasonably affected by this modification are valves 2CV 1024-1 and 2CV-1074-1 which isolate MFW to the steam generators. Plugging the leakoff port an reconfiguring the packing will not affect the probability of a malfunction of either of these valves. The result of this modification is that any packing leakage will now be at the packing gland area on the stem rather than the floor drain. Experience has 90wn though that leakage will be minimized. Pressure boundary integrity is maintained on the valve's bont'et by selecting the plug in accordance with system requiremen and seal welding. 4. Will the consequences of a malfunction of equipment important to safety be increased? Yes O No @ The only equipment important to safety, which could be reasonably affected by this modification, are valves 2CV-1024-1 and 2CV 1074-1. The plugging of the stem packing leak-off lines for these valves do not introduce any new postulated failure modes for this equipment that could increase consequences associated with its malfunction. All postulated malfunctions for the valves remain the same and therefore the dose consequences remain unchanged.
ARKANSAS NUCLEAR ONE FORM TITLE: FORM NO. REV.
, 10CFR60.59 EVALUATION 1000.1313 3 PC-2
- 5. Will the possibility of an accident of a different type than any previously evaluated in the SAR be created?
Yes O No @ Plugging the leak-off port on these valves and removing the leak off lines does not change the characteristics of the valves to the extent that new or different types of accidents could be created than those already analyzed.
- 6. Will the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR be created?
Yes O No @ The proposed modification changes only the routing of packing leak-off flow, and does not change the characteristics of the affected valves sufficiently to create a new type of failure not previously analyzed. In addition, the change will not create local conditions sufficiently different to result in the possibility of a new type of malfunction to any other equipment important to safety in the upper north piping penetration room.
- 7. Will the margin of safety as defined in the basis for any technical specification be reduced?
Yes O No @ There are no technical specification bases which contain any margins of safety based upon the routing of valve packing leak-off lines. Performance of this nuclear change will therefore not reduce any margins of safety in the T.S. bases. M - LW Steohen J. Lynn bbN Certified Review Sigr sture Printed Name Date Reviewer's certification expiration date: b'N O Assistance provided by: Printed Name Scope of Assistance Date PSC review by: Date: $ TI $S 4 r p . ... . . . % _ w.s. O-
3 ARKANSAS NUCLEAR oNE FORM TITLE: FORM NO. REV. 10CFR50Je Ll TERMINATION 1908.131A 3 Pc.1 l-OM1 Page 1 of ,,3 l Document No. PC 958007P201 RevJChange No. 1 b N [T M. Title stator Laak Monitorina System Installation Brief description of proposed change: Initial installation of SMar Laak Monitor on U2 Turbine Generator i i Will the proposed Activity:^ i 1. . Require a change to the Operating License including:
- l. Technical Specifications (excluding the bases)?
YesO No@
- Operating License?
YesO No@ \ i Confirmatory Orders? YesO No@ l
- 2. Result in information in the following SAR documents (including drawings and text) being (a) no longer true or accurate, or (b) vic!ste a requirement stated in the document:
SAR (multi-volume set for each unit)? Yes@ NoO Core Operating Limits Report YesO No@ Fire Hazards Analysis? YesO NoS Bases of the Technical Specifications? ' YesO No@- Technical Requirements Manual? YesO No@ NRC Safety Evaluation Reports? YesO No@
; 3. Involve a test or experiment not described in the SAR?
YesO NoS (See Attachment 2 for guidance)
- 4. Result in a potential impact to the environment? (Complete the Environmental Impact Determination of this form.) YesO No@
- 5. . Result in the need for a Radiological Safety Evaluation por section 6.1.57 YesO No@
- 6. Result in any potential impact to the equipment or facilities utilized for Ventilated Storage Cask activities per Section 6.1.67 YesO No@
- 7. Involve a change under 10CFR50.54 for the following SAR documents per Section 6.1.7:
QAMO? YesO No@ E-Plan? YesO No@ Begis for Determination fQuestiorty 1. 2 & 31: See Page 4 of this determination.
' O Proposed change does not require 10 CFR 50.59 Evaluation per Attachment 1, item W . (if checked, note appropriate item #, send LDCR to Licensinc.
FORM TITLE: FORM NO. REV. 10CFR50.CO DETERMINATION l 1000.131A 3 PC.1 b- Page 2 ofj Document No. PC 958007P201 RevJChange No. J fAbI1O Y Search Scope: List sections reviewed in the Licensing Basis Documents specified in Question 1,2 and 3. If a keyword se was done on LRS, "all" may be entered under "Section" with the keyword (s) used in parentheses. Controlled copies of the documents shall be reviewed (LRS is not verified and searches only text, not figures or d Attach and distribute a completed LDCR per Section 6.1.2 if LBD changes are required. Document Section LRS: All (SWC, stator water, stator cooling, generator cooling, IA, instrument Alr) MANUAL SECTIONS: Unit 2 SAR Sections 10.2.2, 9.3.1 l { FIGURES: Unit 2 SAR Figure 3.2-6, Figure 9.3-1 Cert! w- v_/W _ viewersd5fgMiture '
# Gary W. Liffick 11/10/97 Printed Name Date Reviewers certification expiration date: 1/15/98 Assistance provided by:
Printed Name Scope of Assistance N/A Date Search Scope Re Acceptability (NA, if performed by Technical Review per 1000.006) kNA- MdKPf BetielNu) Ruo& 97 Certified Revfewers Signature Printed Name Date
r ARKANSAS NUCLEAR ONE FORM TULE: FORM NO. REV. 10cFRSO.80 DETERMINATION 1000.131A 3 PC 1 L .- DC PAu Page 3 of 4 1 i ENVIRONMENTAL IMPACT DETERMINATION l l (UNIT 1 and UNIT 2) i g7g Document No. _ PC 958007P201 RevlChange No. __1 l QQ{ ' Complete the following Determination. If the answer to any checklist item is "Yes", an Environmental Evaluation I is requi.ed. See Section 6.1.4 for additional guidance, 1 Will the Activity being evaluated: l Yes No l O S Disturb land that is beyond that initially disturt>ed during construction (i.e., new construction of buildings, creation or removal of ponds, or other terrestrial impact)? See Unit 2 SAR Figure 2.5-17. This applies only to areas outside the protected area. O E increase thermal discharges to lake or atmosphere? O S Increase concentration of chemicals to cooling lake or atmosphere through discharge canal or tower? O E increase quantity of chemicals to cooling take or atmosphere through discharge canal or tower? O E Modify the design or operation of cooling tower which will change drift characteristics? l 0 S install any new transmission lines leading offsite? O E Change the design or operation of the intake or discharge structures? O S Discharges any chemicals new or different from that previously discharged? O O Potentially cause a spill or unevaluated discharge which may effect neighboring soils, surface water or ground water? O E involve burying or placement of any solid wastes in the site area which may effect runoff, surface water or ground water? O O involve incineration or disposal of any potentially hazardous materials on the ANO site? O E Result in a change to nonradiological effluents or licensed reactor power level? i l O @ Potentially change the type or increase the amount of non-radiological air emissions from the ANO site. l l 1 L __
ARKANSAS NUCLEAR ONE FORM TITLE: FORM NO. REV. 10CFR50.89 REVIEW CONTINUATION PAGE - 1000.131C 3 QQML Page 4 01J Document No. PC 958007P201 i Rev1 Change No. j. 10CFR50.59 Review Continuation Paae gyo The Stator Water Cooling System is shown on U2 SAR Fig. 3.2-6 (P&lD M-2208{ instrument Air System is shown on U2 SAR Fig. 9.3-1 (P&lD M-2218 Sh. 2). This PC ' installs a stator leak monitoring system integral to the stator water cooling system. It utilizes Instrument Air for cooling and as injection (purge) air. l This PC installs pipe, tubing and several valves. The two SAR Figures and P&lD will ne to be revised to show the new piping and valves. No other changes to the LBDs will be necessary. The text of the SAR will not require revision. f
- 1. The TS, OL, and Confirmatory Orders were reviewed and no sections require revision The review included tables and figures; it was determined that this PC does not affect these documents.
- 2. This PC will require revision of the U2 SAR (Fig. 3.2-6 and Fig. 9.3-1). This PC will have no impact on the Core Operating Limits Report. This PC will not require changes to th FHA, Bases of the Technical Specifications, the Technical Requirements Manual, or th NRC Safety Evaluation Reports.
- 3. This PC does not change the function of any system and does not involve any test o experiment.
- 4. The environmental checklist was reviewed and this PC has no environm
- 5. This PC does not involve processing radioactive material or impact monitored effluent release points.
I
- 6. There is no potential impact to Ventilated Storage Cask equipment or procedures.
- 7. This PC does not involve a change under 10CFR50.54 for the QAMO or the E-plan.
[ ARKANSAS NUCLEAR ONE FORM TITLE: FORM NO. REV. 10CFR60.89 EVALUATION 1000.1313 3 Page 1 of_2
- d. [M 10CFR50.59 Eval. No. H. omPS (Assigned by PSC) g7 Document No. PC 958007P201 Rev1 Change No. _1 gQf Title Stator Leak Monitorina System Installation A WRITTEN RESPONSE PROVIDING THE BASIS FOR THE ANSWER TO EACH QUESTION MUS ATTACHED. EACH QUESTION MUST BE ANSWERED SEPARATELY. A SIMPLE STATEMENT OF CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES GUIDANCE FOR RESPONSE.
If the answer to any question on this form is "Yes," then an unreviewed safety question is involved. If the answer to all questions is "No," then the proposed change does not involve an unreviewed safety question.
- 1. Will the probability of an accident previously evaluated in the SAR be increased?
Yes O No @ Installation of this monitoring system will not change the functions of the Stator Water Cooling System or the Instrument Air System. The new installation is designed to the required codes and standards for these two sy The modifications associated with this Plant Change have no impact on any of the postulated accidents anal the SAR. Additionally, there are no such accidents associated with the Stator Water Cooling System. The o such accident associated with the Instrument Air System is Loss ofInstmment Air. The system functions will remain the same and the probability of a loss ofInstrument Air remains unchanged. The use of a hose (if used) for the initial supply ofInstmment Air will not increase the likelihood of a Loss of Instrument Air. Reference Installation Section 2.f(pages 14 & 14a). The hose is rated for 300 PSI, whic than the operating pressure of the Instrument Air System. Therefore, the likelihood of a failure in these systems will be no greater than that previously evaluated. B this evaluation, the probability of an accident previously evaluated in the SAR will not be increased
- 2. Will the consequences of an accident previously evaluated in the SAR be increased?
Yes O No @ The failure of piping or other components associated with this modification cannot increase the failure consider and consequences presiously evaluated in the SAR. The Stator Leak Monitoring System cannot increase the consequences of a Loss ofinstrument Air event, or any other previously evaluated accidents described in the SAR. Installation or failure of this monitoring system will not increase or affect any dose consequences for the accidents identified in the SAR. This modification cannot increase the consequences of any presiously evaluated event. 3. Will the probability of a malfunction of equipment important to safety be increased? Yes O No @ The Stator Water Cooling System and the Instrument Air System are not essential to safety and the implement of this modification will not change the probability of a malfunction of equipment important to safety. There is no ' equipment important to safety for which the SWC system provides any interface or support function. Insinunent Air provides air for some safety-related components. However, they are all designed to fail safe or they have source of air. Based upon this evaluation, the probability of a malfunction of equipment important to safety will n be increased.
)
FORM TITLE; ARKANSAS NuCt. EAR oNE s.Crnse.= EvacuanoN FORM NO. REU. [ k ms s (c. y58 e 07P2. o t 9M.C. 2 S AtV.1. D cR1- Page _2_, of._2
\
4. Will the consequences of a malfunction of equipment important to safety be increased?
)
i Yes O No @ 1 The consequences of a ='6ejon of the Stator Water Cooling Syseem and the In by this modification. No changes are being made that would increase the consequence systems Additionally, no other consequences of a =In eion of equipment important to safety are being created. 1 5. Will the possibility of an accident of a different type than any previously evaluated in the SAR be created? Yes O No @ The madificarians being performed on the Stator Water Cooling System and the Inst create any new accident scenarios. The design of the Stator Leak Monitormg System is consi already evaluated and this moddication will not alter S for the new configuratson are the same as that for the old configuration, the p and bounding. 6. Will the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR be created? Yes O No @ The function of the Stator Water Cooling System and the Instrument Air System a Change. The SWC system does not involve any safety related equipment or perfo no new safety-related air loads being added to the Instrument Air System, lastrument Air safety-related components. However, they are all designed to fail safe or they have a Based upon this evaluation, the possibility of malfunction of equipment important to sa previously evaluated is not created by this change. 7. Will the margin of safety as defined in the bases for any technical specification be reduced? i Yes O No @ )
'Ihere are no Technical Specification safety limits or bases defined for the Stator Wa Instrument Air System. Therefore, the margin of safety as defined in the TS Basesl will not be reduced. '
A- > Gary W. Liffick "pRevlemis&)l5 nature " Printed Name 11/10/97 Date Reviewer's certification expiration date: 1/15/93 Assistance provided by: Printed Name Scope of Assistance N/A Date PSC review by: 3-- Date: \b .
c n.m. ~ .... s.nn u.~ 1 FORM TITLE: l FORM NO. REV.
, 10CFR50.59 DETERMINATION 1000.131A 3 Pc.1
! Page .1. of.6 Document No. 958033P202 Rev/ Change No. 0 - 958033P202 Title ANO-2 Replace Hydrazine Analyzer PAGE 3 REV 0 Brief description of proposed change: Replace 2AITS-4009 and 2AITS-4014 and associated equipment. Willthe proposed Activity: c
- 1. Require a change to the Operating License including: I Technical Specifications (excluding the bases)?
YesO No@ Operating License? l YesO No@ Confirmatory Orders? YesO No@ 2. Result in information in the following SAR documents (including drawings and text) being (a) no longer true or accurate, or (b) violate a requirement stated in the document: : SAR (multi-volume set for each unit)? Yes@ nod Core Operating Limits Report YesO No@ Fire Hazards Analysis? YesO No@ i Bases of the Technical Specifications? YesO No@ Technical Requirements Manual? YesO No@ NRC Safety Evaluation Reports? YesO No@
- 3. Involve a test or experiment not described in the SAR?
(See Attachment 2 for guidance) YesO No@
- 4. Result in a potential impact to the environment? (Complete the EnvironmentalImpact Determination of this form.)
YesO No@
- 5. Result in the need for a Radiological Safety Evaluation per section 6.1.57 YesO No@
- 6. Result in any potentialimpact to the equipment or facilities utilized for Venulated Storage Cask activities per Section 6.1.6?
YesO No@
- 7. Involve a change under 10CFR50.54 for the following SAR documer.ts per Section 6.1.7:
QAMO? YesO No@ E-Plan? YesO No@ Casis for Determination (Questions 1. 2 & 3): SEE ATTACHED SHEET w-O Proposed change does not require 10 CFR 50.59 Evaluation per Attachment 1, item # . (if checked, note appropriate item W. send LDCR to Licensing). l t
ARKANSAS NUCLEAR ONE FORM TITLE: FORM NO. REV.
. 10cFR$0.80 DETERMINATION 1000.131A 3 Pc.1 Pace -
Document No. 958033P202 Rev./ Change No. A 9 58 033 P 20 2 ~2 o M SE 4.. H y 0 Search Scope: List sections reviewed in the Licensing Basis Documents specified in Question 1,2 and 3. If a ke was done on LRS, *all" may be entered under "Section" with the keyword (s) used in parentheses. copies of the documents shall be reviewed (LRS is not verified and searches only text, not fig Attach and distribute a completed LDCR per Section 6.1.2 If LBD changes are required. Document Section LRS: ANO-2 Tech. Spec.
' ANO-2 Operatino License E
/.NO-2 Confirmatory Orders E
ANO-2 SAR A A 28M9 E Ef.!an E FHA E ANO-2 Bases of the Tech. Specs. E ANO-2 NRC SERs E M (LRS Keywords are listed on the continuation sheet.) MANUAL SECTIONS: 3.6.4.4.3
.19.JJ MJ 9J 10.4.7 10.4.10 Table 9.3-2 Table 10.3 2 FIGURES:
Fioure 9.3-3 Flours 10.4-6 Rooer B. Rucker 5/18/98 Certified Reviewers Signature Printed Name Date Reviewers certification expiration date: 9/11/99 Assistance provided by: Printed Name Scope of Assistance bilb Date Scope Review Acceptability (NA, if performed by Technical Review per 1000.006) Y '-- Certified Reviewers SignatC 'fc1.Y S. ~L v \/ E> 2 2. 9$ Printed Name / / Die
-. ..,~ . ww en w..-
FORM TITLE: FORM NO. REV. 10CFR50.59 DETERMINATION 1000.131A 3 Pc.1 Page 3 of_6 ENVIRONMENTAL IMPACT DETERMINATION (UNIT 1 and UNIT 2) Document No. 958033P202 RevJChange No. 9 958033P202 P f REV 0 Complete the following Determination, if the answer to any checklist item is "Yes", an Environmental Evalu is required. See Section 6.1.4 for additional guidance. Will the Activity being evaluated: Yes g O @ Disturb land that is beyond that initially disturbed during construction (i.e., new construction o buildings, creation or removal of ponds, or other terrestrialimpact)? See Unit 2 SAR Figj 2.5-17. This applies only to areas outside the protected area. ' O @ Increase thermal discha.ges to lake or atmosphere? O @ increase concentration of chemicals to cooling lake or atmosphere through discharge can tower? O j Increase quantity of chemicals to cooling lake or atmosphere through discharge canal or tower? O @ Modify the design or operation of cooling tower which will change drift characteristics? O @ instati any new transmission lines leading offsite? O @ Change the design or operation of the intake or discharge structures? O @ Discharges any chemicais new or different from that previously discharged? O @ Potentially cause a spill or unevaluated discharge which may effect neighboring soils, surfac water or ground water? O @ involve burying or placement of any solid wastes in the site area which may effect runoff, surface water or ground water? O @ Involve incineration or disposal of any potentially hazardous materials on the ANO site? O E Result in a change to nonradiological effluents or licensed reactor power level? O @ Potentially change the type or increase the amount of non-radiological air emissions from the ANO site. l
ARKANSAS NUCLEAR ONE FORM TITLE: FORM No. RER 10CPR$0.80 REVIEW CONTINUATION PAGE ' 1000.131c 3 Page 4 of 6 Document No. 958033P202 Rev> Change No. 1 958033P202 10CFR50.59 Review Continuation Pace pg f gy g L SEARCli.EE9EE The following is the LRS search word / phrase list:
, 2P'59*
2C*145*l 2C*145*l1 l : 2AITS*4009 2AITS*4014 l 2FIC*4014 2FIC*4009 2PVC*4008 2PVC*4013 2PS*4008 2PS*4013 2SS*37 2SS*24 2AIC'6425 2AIC*6426 2E1A 2E1B l l condensate w/10 hydrazine feedwater w/10 hydrazine ' steam generator
- w/10 hydrazine secondary sample system oxygen scavenging hydrazine sss BASIS FOR DETERMINATION-PC 95-8033 (958033P201) replaced the hydrazine contro"ers 2AIC-6426 and 2AIC-6425.
These controllers control the three hydrazine pumps 2P 59A, B and C. After the installation of the new contro the existing analyzers 2AITS-4009 and 2AITS-4014 was discovered. The new controllers have manual because of the frequent erratic signals provided by the existing analyzers. i ER 958033P202 is replacing hydrazine analyzers 2AITS-4009 and 2AITS-4014. The new y all of the functions of the existing analyzers, --f -fd it.; n; 2" .i 's it.; ;':r' ::r; d--'^
-d!"-' 'a 'M ' :it':
grange agnal- An additional pressure and temperature switch and a filter will be added to each a l loop to aid in protecting the new equipment. The new analyzers are separated into two enclosu sample loop. One of the enclosures contains the sample detection sub-components and the other l contains the electronic / electrical sub-components. The post-modification testing criteria for this modification will be contained within the DeI testing section. testing equipment ' These testing sections provides detailed instructions similar to general approve
/ instrumentation.
1 I l l l i i i
esvnau.=NucLem unc FORM TULE: FORM NO. REV. 1 10CFR50.59 REVIEW CONTINUATION PAGE 1000.131C ' 3 Page 5 of_g ; Document No. 968033P202 RevlChange No. 9, 958033P201 Responses to Determination Questions: PAGE 7 REV 0 Question 1. No changes to the Operating License will be required since this modification is structured to comply with Operating License documents listed in question 1. The Technical Specifications have no specific operability requirements for the equipment being added or changed by this modification. No specific testing require are addressed for the equipment / systems being affected by this modification or equipment / system any acceptance tests for this modification. i l Question 2. ! The only License Based Document that is being impacted by this Plant Change is the SAR. SAR Fi which includes drawing M-2223 SH 1 is being revised by this Plant Change. A Licensing Docum Request is included in the Plant Change, and a 10CFR50.59 Evaluation is attached. No other LBDs were m untrue or inaccurate by this modification, nor did this modification violate any requirement stated in the LBDs. The information/ instructions in this modification are below the level of detail contained Question 3. No testing is required by this modification other than typical Post Modification testing. This modification doe constitute a test or experiment not described in the SAR as defined by Procedure 1000.131. This modification will enly provide normal detailed post-modification testing similar to approved generic ANO procedures. Question 4. This modification will not result in any adverse impacts to the environment. The operation of the plant will no changed in any way which will result in changes to the air, water or soll conditions of the site. l l
ARKANSAS NUCLEAR ONE p' FORM TITLE: FORM NO. REV. 10CFRSO.89 REVIEW CONTINUATION PAGE 1000.131C 3 Page g of 6 Document No. 958033P202 RevdChange No. O Responses to Determination Questions: 958033P202 PAaE 8 REV 0 Question 5. I This change does not involve processing of radioactive material outside of Controlled Access. Question 6. This change does not involve any equipment used in handling Spent Fuel Storage Casks. Question 7. This modification will not make the QAMO or the E-Plan statements to be un equipment / systems being modified are below the level of detail contained within these documents1 e
, - . ~ .._ = u .
FORM TITLE: FORM NO. REV.
)
10CFR$0.50 EVALUATION 1000.131s 3 958033P202 Page 1- of 3-PA0E 9 REV 0 1 CFR50.59 Eval. No. F FlW -Ob (Assigned by PSC) Document No. 968033P202 RevdChange No. O Title ANO-2 Replace Hydrazine Analyzer A WRITTEN RESPONSE PROVID!NG THE BASIS FOR THE ANSWER TO EA ATTACHED. EACH QUESTION MUST BE ANSWERED SEPARATELY. A SIMP CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES GUIDANCE F If the answer to any cuestion on this form is "Yes," then an unreviewed safety question is inv to all questions is "No," then the proposed change does not involve an unreviewed safety question. 1. Will the probability of an accident previously evaluated in the SAR be increased? - Yes O No @ j No credit is taken for the Secondary Sample Sve;.m in the accident analyses. The Seconda 33,mple System does not contain any touismsnt that asiforms a 52"_;v-r=!:0:d function, nor does it cani.in any couloment that is crsened for automatic F*!en. This m-:-d!fication e-:== not interface ggg Safety Related eculoment. nor willit edd any Safety Related souls... sat. The sparsbility o Secondary Samole Sveism instrumsnimilon ensures the cualltv of the m=k=un wsi r. circulati water and the turb as cycle water /stssm. The hverszine analu .. are used to monitor the hydiez;ns concentration in the main tsadw:t:r lines and provide an alsrm unen the occurrence of
........ I conditions. The hydrazine analyzers also provide a sional to the hydrazine controllers which continually monitor and co6i vl the addition of hvdiszine by adlustino the hvdiazine DumD ei wiss. Hydrazine allows the removal of oxvnen fisia the feedwater. which is essential to prevent corrosion of wet =+=al surfaces. Oxvoon estrasion can significaath &G.ct the ossi iina lifetime of the 7;sdwster.
with secondary sys;.is . This osckene will help to imersve the Secondarv Sample Systs . however. the samplino system dass not snsct the Feedwater Sy sample from the feedwater systems continually flows throuch the analyzers and into the olant drains. The operation or failure of the hydrazine components in the secondary asmsle system are not accident initiators to any of the accidents listed in the SAR. nor do thev interface with any couloment not increased. that is an initiator. Therefore, the probability of previousiv evaluated LSD accidents is 2. Will the consequences of an accident previously evaluated in the SAR be increased? Yes O No @ The consecuences of accidents ;liscussed in the LBDs will not chance as a result of this Plant
^modification.
Chance. The souloment and actions issuired to mitlaate each accidsat will be unsnectsd b Since the modification has been desianed with the Dissar electrical isolation / separation and with seismic intsarity. the new desian will not fall in a mode that will sovsrselv affect any safety function. The dose consecuences associstsd with previousiv evaluated accidents accidsnie will not previously be affscted evaluated as awillresult in the SAR of this not increase. modification. Therefore. the consec
JuuCANsAs NUCLEAR ONE FORM TITLE: 1 FORM NO. REV. ' i e 10CFR40.58 EVALUATION 1000.1313 3 958033P202 Page2 of 3 Document No. 958033P202 PAGE 10 R E V 0 Rev/ Change No. 0 3. Will the probability of a malfunction of equipment important to safety be increased? Yes O No @ The coulomont affected by this Plant Chanos does not interface (i.e. control) with soulo considered ;mssiteid to se!.ii. This modification ec+5 not affect any oculoment or cablino t ssifsans any csatist or ir .;"sch functions wn;. safety related svei ms. The hydrazine analy si.Ad. a s;sas to the hve..J. e controllers which are part of the Chemical Addition System is a subsystem and/or of Conden--* innd TeeC;;;;.r. The da=!en confiouration of the added as'Er=d ===>!c-i. fonci.;as d!sec==!9n. n; is in ecce dance w;ii ANO desian standards as described in the
. I i
The current desian standards for interfacino with safetv-related cii isciation and ;&asiailon to prevent oroaccetion of a failure from a non-safety system e c;icult ':E;as of this med'";cmilon is in kersino with these standards. i The intesi;is of the safetv-related circuits and pressure boundsries has been ens
;rs;;ns =lainic ;6siell-disa standards as issuited oer applicable Essroved ANO details Drocedures, e
Additional fire !eed!no, heat lesdisc. b nori and diesel loadino have all been proper e Eddresesd. evaluated and found Eccas'+W. There are no new *e!!Ps rne'iae ;6^.vduced to couloment that is important to safety. Based ofon sishability the above a malfunct;en d!sev==!en. of eeu;ss..ent It safety. important to can be determined that this Plantj 4. Will the consequences of a malfunction of equipment important to safety be increased? i Yes O No @ l The souloment important to safety.and cablino lastaiiagi by this modification does not increase reliance ; As csaciuded in- the response to cuestion 3.0. the ohvsical/ electrical csnf;auietien of this msdii; cation ensures that the probability of a malfunct;on___of soulo issssitant to se!.;i has not been increased. Therefore. the offsite dose consecuences l vi;th a malfunction of souloment important to safety mselfication. is not increased as a result of this S. Will tr e possibility of an accident of a different type than any previously evaluated in the SAR be created? Yes O No @ The function of the aaelomont e77.cted by this modification is not recuired for shutdow unit. I'^;==f-d;;en.^. e aC+-^ive c'm: or maintainino rsactar ces!sst ss=+ere ;ra-Gi;iv. It has been related system -_I th&i the sae61. or ::en-- 'GE;.I'=^;ss of this modification will have no neoative imned on a safetv-The insiellation of this modification will also not chance the way gfpi.iions will i= =eand to an ace'esa". No crad;t is taken in the current accident analys ec^ca=*!c or meaaal action by the hydrazine analyzers or controllers. insielled by this ir.sd;iscation will not crm=** any accident initiators. Failure of the coulomont Aeolicable desion r=EEli;&eal. hevG been consideisd (see response to cuestion 3.0) to ensure that system
;massi&ni to 5:^_;i are not issee dized; therefore. It can be concluded that the possibili accid-nt different from any si;4seelv analyzed in the SAR will not be created.
ARKANSAS NUCLEAR ONE FORM TITLE: FORM NO. REV. I
, 10CFR50.89 EVALUATION 1000,1313 3 Page a of 2 4
Document No. 968033P202 958033P202 RevlChange No. g t 1 PAGE ll REV 0
- 6. Will the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR be created?
Yes O No E l All system desion for souloment important to safety will remain the same. None of the circuits l 1 added. deleftd or modified by this desian will be routed in such a manner to cause prosecation of a l failure irt a* Class 1E circuit. The criteria for electrical separation has been maintained and ' conservative adherence to seismic reouirements has been observed to insure comoliance. Also, this modT.GUon does not electrice!!v/ mechanically interfece with couloment important to safety. Thsiebre the possibility of a malfune+!en of eauipment important to safety that involves an initiator or failure of a different tvoe that any previously evaluated in the SAR has not been created.
. t
- 7. Will the margin of safety as defined in the bases for any technical specificationbe reduced?
Yes O No @ The Techni i Specification bases do not establish a maroin of safety for the hydrazine analyzers l 2AITS4009 and 2AITS4014 or controllers 2AIC4425 and 2AIC4426. The installation of the new I ' analyzers 2HYE4009 & 2HYTS4009 and 2HYE4014 & 2HYTS4014 or the modification / addition of the other devices associated with these instrument loops will not affect or alter the existino Tech. Spec. reouireme.n. nor be included in any new reouirements. Based on the above statements, this modificationcwill not reduce the maroin of safety as defined in the h=ses. l l 1 I e Rooer B. Rucker 5/18/98 CeitifieId Neviewers Signature Printed Name Date Reviewers certification expiration date: 9/11/99 Assistance provided by:
. i -
Printed Ni Scope of Assistance Date C PSC review br - Date: l l i 3
! l
'~
\ . l l 1 1
ARKANSAS NuCl. EAR ONE FORM TITLE: Pane 1 FORM NO. REV. ( 10CFR50.59 DETERMINATION 1000.131A 3 PC-1 PAGE b MV.# 8 This Document contains 3 pages. Document No. ER962029P201 Rev1 Change No. 9, Title ,,, Main Feedwater Recire. Valves 2CV4741/2CV4749 and Controller 2FIC4735/2FIC Brief description of proposed change: This Plant Change will change the Flow Indicator Controllers (FIC),2FIC-0735 and 2FIC-074 2UC-5283 in 2002 to the ANO2 mandard controller, replace the Modutrocic actuators 2CV 0741 a Controllers, and basically moves (spare and add) 2 relays at 2C02 to provide 100% overnde receipt of2 out of 3 Feedwater pressure switch logic from 2PSX 0732 (old 2PSX2 0732) and 2PSX-07 Willthe proposed Activity: 1. Require a change to the operating Ucense including: Technical Specifications (excluding the bases)? YesO NoS operating Ucense? YesO NoS Confirmatory Orders?
, YesO Nc3 2.
l Result in information in the following SAR decuments (including drawings and text) being (a) no longer true or accurate, or (b) violate a requirement stated in the document: SAR (multi-volume set for each unit)? I YesE nod Core Operating Umits Report? YesO NoS Fire Hazards Analysis? YesO NoS Bases of the Technical Specifications? YesO NoS Technical Requirements Manual? YesO No@ NRC Safety Evaluation Reports? YesO NoS , 3. Involve a test or experiment not described in the SAR? i (See Attachment 2 forguidance) YesO NoS 4. Result in a potential impact to the environment? (Complete Environmental Impact Determination of this form.) YesO NoS 5. Result in the need for a Radiological Safety Evaluation per section 6.1.57 YesO NoS
- 6. Result in any potential impact to the equipment or facilities utilized for Ventilated Storage Cask activities per Section 6.1.67 YesO NoS 7.
Involve a change under 10CFR50.54 for the following SAR documents per Section 6.1.77 QAMO? I YesO NoS E-Plan? YesO NoS
ARKANSAS NUCLEAR ONE FORM TITLE: Pace 2 FORM NO. REV. 10CFR50.59 DETERMINATION 1000.131 A 3 PC-1 na mv.* o Document No. ER962029P201 Rev) Change No. O Title _ Main Feedwater Recirc. Valves 2CV-0741/2CV-0749 and Con _ Basis for Determination (Questions 1,2, & 3): 2'IIC 5283 in 2C02 to the ANO2 standard contmiler, and rep Controllers. A search was conducted on the LRS of the following documents: 2LFO,2NSE,2SA , FHACALC, JCOFILES, NUREG, QAMOl7, REGUIDE, REGUIDES, ULD, VSCSAR detail to (M-2204, describe Sh.2),7.3-9 the (M-2402, Sh.type 3 & 4) ofcontroller installed. A review of the drawings affected by this . SAR will not be used by this Plant Change. This Plaat Change will not impact the enviro
- I.2 4 1
'5.S- ti CM 2 ee t)
O Proposed change does not require 10CFR50.59 Evaluation per Attachment 1, item #
'(if checked, note appropriate item #, send LDCR to Licensing).
Search Scope: List sections reviewed in the Licensing Basis Documents specified in questions 1,2 and was done on LRS, "all" may be entered under"Section"with the keyword (s) used in parent Attach and distribute a completed LDCR per Section 6.1.2 If L Document Section Unit 2 Tech. Specs _ Unit 2 Operating License all (FWCS, " Feed Water Rectreulation". all controller and valve tag numbers.
- Tee Confirmatorv Orders all (FWCS, " Feed Water Recirculation", all controller and valve tag numbers. " Fe Unit 2 SAR all (FWCS, " Feed Water Recirculation", all controller and valve tag numbers, " Fe ,
_QAMO all (FWCS," Feed Water Rectreulation" all controller and valva tag numbers," Feed _E-Plan all (FWCS " Feed Water Rectreulation", all controller and valve tag numbers, " Fee FHA all (FWCS," Feed Water Rectreulanon", all controller and valve tar numbers " Feed U2 Tech Spec Bases all (FWCS, " Feed Water Rectreulauon", all controller and valve tag numbers. " Feed [NRC SER's all (FWCS, " Feed Water Recirculation" all controller and valve tag numbers " Feed W Unit 2 SAR all (FWCS " Feed Water Recirculatiaa' , all controller and valve tag numbers, " Fee FiFm 7.3-9#. Figure 9.3 l#. Figure 10.4 2, Sh.2#, Figure 1.2 6. and 3.8-31(LDCR at t o > A ,_ ti. 4r m m . Douglas A. Bruce Cehified feviewers'Sigliature 12/12/97 Pnnted Name Date Reviewers certification expiration date: 02/19/99 l Assistance provided by: l i Printed Name Scope of Assistance Date ! I I l Search Scope Review Acce l bility (NA, if performed by Technical Reviewer per 1000.006) K }J Clrrtified Reviewers SignafGre i k a/ M 8 1~ Printed Name Ww 2~.22-9R Date t
ARKANSAS NUCLEAR ONE FORM TITLE: Pace 3 ' FORM NO. REV. 10CFR50.59 DETERMINATI$N 1000.131A 3 PAGE B _ REV.t O ENVIRONMENTAL IMPACT DETERMINATION (UNIT 1 and UNIT 2) Document No. ER962029P201 Rev./ Change No. O., Title _ Main Feedwater Recire. Valves 2CV-0741/2CV-0749 and Controller 2F _ l i Complete the following Determination, if the answer to any item below is "Yes", an Environmental E required. See Section 6.1.4 for additional guidance. Will the Activity being evaluated: ! Yes N_o O @ ! Disturb land that is beyond that initially disturbed during construction (i.e., new construct l buildings, creation or removal of ponds, or other terrestrialimpact)? See Unit 2 SAR 2.5-17. This applies only to areas outside the protected area. 0 @ increase thermal discharges to lake or atmosphere? l D @ increase tower? concentration of chemicals to cooling lake or atmosphere through discharge O @ increase quantity of chemicals to cooling lake or atmosphere through discharge canal o tower? O @ Modify the design or operation of cooling tower which will change drift characteristics? O @ install any new transmission lines !eading offsite? i O @ Change the design or operation of the intake or discharge structures? O O Discharges any chemicais new or different from that previously discharged? O E Potentially cause a spill or unevaluated discharge which may effect neighboring soils water or ground water? l 0 8 involve burying or placement of any solid wastes in the site area which may effect surface water or ground water? O E Involve incineration or disposal of any potentially hazardous materials on the ANO site? O E Result in a change to nonradiological effluents or licensed reactor power level? O E Potentially change the type or increase the amount of non-radiological air emissions from ANO site.
ARKANSAS NUCLEAR ONE FORM TITLE: Page 1 FORM NO. REV. 10CFR50.89 SAFETY EVALUATION 1000.131B 3 This Document contains 5 Pages. REY.#_ @ tocFR50.59 Eval. No. F F tt) , ai 2- t3 l (Assigned by PSC) Document No. ER962029P201 RevlChange No. 0 Title _ Main Feedwater Recirc. Valves 2CV-0741/2CV-0749 and Cont A WRITTEN RESPONSE PROVIDING THE BASIS FOR THE ANSWER A'ITACHED. EACH QUESTION MUST BE ANSWERED SEPARATELY. A CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES GUID If the answer to any question on this fonn is "You." then an unreviewed safety question to all questions is "No," then the proposed change does not involve an unreviewed safety qu 1. Will the probability of an accident previously evaluated in the SAR be increased? YesO t.u d The accident analysis Unit 2 SAR Chapter 15 does not discuss the Feedwater Recirc system in sufficient detail to warrant (continued on Page 3) 2. Will the consequences of an accident previously evaluated in the SAR be increased? YesO No @ . The mitigation of the consequences of an accment previously evaluated in the SA affected (continued on bypage the3)replacement of the Feedwater Recirculation modutronic control valv 3. Will the probability of a malfunction of equipment important to safety be increased? YesO No @
.This upgrade does not alter the function or capability of any equipment rela to perform its safety related function. (continued on page 3) 4.
Will the consequences of a malfunction of equipment important to safety be increased? YesO No @
' The consequences of a malfunction of equipment important to safety is not co upgrade of the controllers nor the valve operators. (continued on page 4) 5.
Will the possibility of an accident of a different type than any previously evaluated in the SAR be created? YesO No @ The highly unlikely concurrent loss of the Faedwater Racirculation controllers, due to a mode controller failure (cont'd on page 4) 6.
.Will the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR be created?
Yes O No @ The highly unlikely concurrent loss of the Feedwater Recirculation controllers, due to mode controller failure (cont'd on page 4) {
_ _ _ _ - --- - ---- - - ---------- ---------~-- '~
- ~ l ARKANSAS NUCLEAR ONE FORM TITLE: Pace R FORM NO. REV.
10CFR50.59 SAFETY EVALUATION 1000.131B 3 7. Will the margin of safety as defined in the ba;es for anPAGE ID technical specification be reduced? REV.# _ d Yes L.; rvo @ The Feedwater Recirculation valves and their respective Control room (continued on page 5) d dd 4. hwo Dou0las A. Bruce
' C6rtified Reviewer's Signature 12/12/97 Painted Name Date Reviewer's certification expiration date: 02/19/99 Assistance provided by:
Printed Name Scope of Assistance Date PSC review by: Date: \o to 94 e ER 982029P201 r -
ARKANSAS NUCLEAR ONE FORM TITLE: Pag 1 FORM NO. REV. 10CFR80/9 REVIEW CONTINUATION PA2E 1000.131C 3 Document No. PAGE ER962029P201 1i REV.# d
- Rev1 Change No. 0 10CFR50.59 Review Continuation Pace 1.
The probability of an accident previously evaluated in the SAR is not increased? (cont'd revisions. None of the accident initiators evaluated in SAR accident analysis are affected the existing controllers to digital controllers and the motor operated. valves to Pneumatic operated valves The modifications performed on the controllers, and the associated microprocessor failure d to software failure, do not invalidate the failure modes / consequences outlined in the SAR. Activities included in this modification do not create any increase in the frequency of any accident initiato new equipment does not exhibit performance characteristics nor have design features that would , 1 increase the probability of system malfunction or require increased operator burden to suppo operation of the system. The equipment is also compatible with the installed environment. T the probability of accidents previously evaluated in the Unit 2 SAR is not increased. 2. The consequences of an accident previously evaluated in the SAR is not increased? i controllers. The function of the Feedwater Recirculation remains unchanged following this mod Modifying the valve operator from electrical to pneumatic will not increase the consequ accident previously evaluated. The instrument air line that is added is 3/8" tubing and can be is i locally at each valve or at another root valve in the instrumsnt air system. The additional instr consumption required is approximately .7 scfm for each valve, for a total of 1.4 scfm, which is a minimal impact on the instrument air loading. Additionally, equipment common cause software for the controliers is mitigated by the hardware design where consequences of failures are alarm ! and control can be bypassed, either manually at the controller or with a manual handwheel at operators. No increase in the offsite dose rate is increased by this activity. Finally, the increased reliability of the new equipment more than offsets the additional rir.k experienced by common c increased. failure. Therefore, the consequences of accidents previously evaluated in Unit 2 SA software
- 3. The probability of a malfunction of equipment important to safety is not increased?(
l The controllers replaced are seismically mounted, and do not affect the operation of ex Relocated and replaced relays to provide shutdown functions to the feedwater and feedwater Recirculation system are also seismically mounted and remain functionally identical to relays. Malfunction of the newly replaced Feedwater Recirculation valve operators and co has no affect on existing safety equipment nor affect the safe shutdown of the plant. The new a operated valves fail as-is just as the old operators. Additionally, separation requirements f related cables and non-safety related cables are addressed in calc. revision to C#culation 95-E 01. PC. Therefore the probability of a malfunction of equipment important to safety is not in 1
ARKANSAS NUCLEAR ONE FuRM TITLE: Pace 2 FORM NO. REV. 10CFR50.59 REVIEW CONTINUATION PA2E 1000.131C 3 PAGE P1 REV.8 D - Pa0% Document No. ER962029P201 Rev1 Change No. _t
- 4. The consequences of a malfunction of equipment important to safety is not increased?(cont'd)
The new components will perform their original intended function. The new comp does not affect any equipment necessary to ensure that o Additionally, failure of the newly replaced valve operators and associated equip this modification actually enhances operator control. Therefore the consequences o equipment important to safety is not increased.
- 5. The possibility of an accident of a different type than any previously evaluated in th not created? (cont'd) because of a software failure, cannot create any new accident not already ad SAR accident analysis per Unit 2 Chapter 15. Year 2000 compliance issues for the addressed, and the controllers are determined to be Y2K compliant. The only other cred loss of instrument air, which has previously been analyze become the initiators of any new type of accident, nor are any new accident sce the original valve manufacturer. Newly installed controlle cccidens possibility is created by this modification. This activity is bounded by previous
[nalysis evaluated in the SAR. Therefore, the possibility of an accident of any previously evaluated in the SAR is not created. a different ty
- 6. The possibility of a malfunction of equipment important to safety of a differe previously evaluated in the SAR is not created? (cont'd)
Recirculation valve operators and Control Room controllers. This change has n equipment important to safety, nor does this equipment provide any safety related functi possible failure of the controllers, due to a common mode digital controller failure, or cir, is minimized by the design of the controllers, which allows for manual operi the microprocessor, or manual operation of the valve. Additionally, software cha{ cfter installation, and configuration changes to the functions of the new controllers l controlled by plant procedures, thereby eliminating MMIinduced failures. Replacemint ofl
, operator with pneumatic actuators does not create any new failure mode because the va cs before and the valve response to the failure is the same. - A loss of instrument ai! '
increased because the new loading is insignificant. Additionally, the addition of an Ai loss of instrument air for 10 full strokes (5 full strokes pe instrument air and not place an undue hardship on the Feedwater System. Ther frilure scenarios have been evaluated, and are bounded by SAR failure analysis. Therefo i possibility of a malfunction of equipment important to safety of a different type than cvaluated in the SAR is not created.
ARKANSAS NUCLEAR ONE FORM TITLE: FORM NO. REV. 10CFRSS.59 DETERMedATioN 1900.131A 2 PC.2.3 Page 1 ofj Document No. , ER-g63203P201 RevlChange No. O Title; installation of Addcr.e: Level Gauce in Sseat Fuel Pool
! Will the proposed Activity: .1.
Require a change to the Operating Ucense including: Technical Specifications (excluding the bases)? YesO No@
. Operating Ucense?
YesO No@ Confirmatory Orders? YesO No@ 2. Result in information in the following SAR documents (including drawings and text) being ' (a) no longer true or accurate, or (b) violate a requirement stated in the document: Core Operating Umits Report YesO No@ SAR (multi-volume set for each unit)? Yes@ nod QAMO?* YesO No@ E-Plan?' YesO No@ FHA YesO No@ Bases of the Technical Specifications? YesO No@ NRC Safety Evaluation Reports? YesO No@
- 3. Involve a test or experiment rot described in the SAR? -
YesO No@ 4.- Result in a potentialimpact to the environment? (Complete Environmental Impact Checklist of this form.) YesO No@ 5. Result in the need for a Radiological Safety Evaluation per section 6.2.4.A?
'YesO No@
6. Result in any potential impact to the equipment or facilities utilized for Ventilated Storage Cask activities per Section 6.2.4.B? Yes@ NoO Basis for CL ;a:aation: This plant change will install a level gauge in the cask loading pit which is attached to the spent fue safety aM housekeeping concems, this indicator will be used as the primary level indicator for the level. The indicator currently being used will still be available for backup indication as it also provides ala indication in the control room. The addition of this indicator will not require a change to the Operatin - because there is no n ention of the indicator used for level verification. The SAR do except for the addition of the indicator to Figure g.1.1. All references to SFP level do not mention any pa! indicator. This plant change does not involve a test or experiment and it does not need a Radiol ! Evaluation since it will be located in Controlled Access. A 10CFR72.48 review is attached.
- Changes to these documents require an evaluation in accordance with 10CFR50.54.
See Section 6.2.1.B. 4
[ ' FORM TITLEi ~ 10CFR80.88 DETERMINATION FORM NO. REV. 1000.131A 2 PC.2.3 Page 2 of 3 Document No. ER-963203P201 RevdChange No. O
References:
List sechons reviewed in the Licensing Basis Documents, specified ,
. If a in q in parentheses. Controlled hard copies of the d searches such as LRS are not controlled and search text only, not figur completed LDCR if LBD changes are required.
Document U2 Tech Specs Section All foool w/5 levell . Doeratina License i Confirmatory Orders All foool w/5 levell 1 SAR All(oool w/5 level) QAAdg, All (oool w/5 level). Section 9.1. Floure 9.1.1 g,-Pjlan. All foool w/5 level) FHA All foool W/5 level) l TS Bases All foool w/5 level) NRC SERs All foool w/5 level) , All (cool w/5 level) I h -r - Certified Reviewer's Signature Dan Churchman 09/18/96 Printed Name Date Reviewer's certification expiration date: 08/22/98 t l Assistance provided by: l Printed Name Scope of Assistance ~ Date 1 l
p ARMNSAS NUCLEAR ONE FORM TITLE: FORM NO. REV. 1eCFRSO.89 DETERMINATION 1000.131A 2 PC-2.3 Page2of_} ENVIRONMENTAL IMPACT CHECKLIST ! (UNIT 1 and UNIT 2) ! i Documerit No. ER-963203P201 Rev/ Change No. O, Complete the following check list. If the answer to any checklist item is "Yes", an Environmental Evaluation is! i required. See Section 6.2.1.E for additional guidance. Willthe Activity being evaluated: Yes My 1 O @ Disturb land that is beyond that initially disturbed during construction (i.e., new construction of i buildings, creation or removal of ponds, or other terrestrialimpact)? See Unit 2 SAR Figure l 2.5-17. This applies only to areas outside the protected area. O O increase thermal discharges to lake or atmosphere? O .E Increase concentration of chemicals to cooling lake or atmosphere through discharge canal or tower? O E Increase quantity of chemicals to cooling lake or atmosphere through discharge canal or tower? ! O E Modify the design or operation of cooling tower which will change drift characteristics? i O 2 Install any new transmission lines leading offsite? ! O E change the design or operation of the intake or discharge structures? ' O E Discharges any chemicals new or different from that previously discharged? O O Potentially cause a spill or unevalustad discharge w hich may effect neighboring soils, surface I water or ground water? O @ involve burying or placement of any solid wastes in the site area which may effect runoff, surface water or ground water? O @ involve incineration or disposal of any potentially hazardous materials on the ANO site? O E Result in a change to nonradiological effluents or licensed reactor power level? O E Potentially change the type or increase the amount of ron-radiological air emissions from the ANO site.
ARKANSAS NUCLEAR ONE FORM TITLE: FORM NO. REE 10CFR80.59 EVALUATION 1000.131a 2 PageIof 10CFR50.59 Eval. No. N~ b ,, (Assigned by PSC) pgG'n.y Document No. ER-963203P201 RevlChange No. 0 0 Title installation of an AddMional Soent Fuel Pool LevelIndicator en FNivM u A WRITTEN RESPONSE PROVIDING THE BASIS FOR THE ANSWER TO EACH Q ATTACHED. EACH QUESTION MUST BE ANSWERED SEPARATELY. A SIMPLE ST CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES GUIDANCE FOR If the answer to any question on this form is "Yes," then an unreviewed safety question is involved. If t to all questions is "No," then the proposed change does not involve an unreviewed safety question.
- 1. Will the probability of an accident previously evaluated in the SAR be increased?
Yes O No @ The probability of an accident previously evaluated in the SAR will not be increased. This additional level indicator will not affect any of the evaluated accidents involving the spent fuel A seismic analysis has been completed for the addition of the indicator to the seismic structure.
- 2. Will the consequences of an accident previously evaluated in the SAR be increased?
Yes O No @ The consequences of an accident previously evaluated in the SAR will not be increased due to the addition of this level indicator. A seismic evaluation has been completed and the indicator has no affect on the spent fuel pool. 3. Will the probability of a malfur ction of equipment important to safety be increased? - Yes O No @ The level indicator and surrounding equipment are not safety related. Therefore, this addition will not increase the probability of a malfunction of any safety related equipment. 4. Will the consequences of a malfunction of equipment important to safety be increased? Yes O No @ The levelindicator and surrounding equipment are not safety related. Therefore, this addition will not increase the consequences of a malfunction of any safety related equipment. ! 5. Will the possibility of an accident of a different type than any previously evaluated in the SAR be created? Yes O No @ 1 This level indicator is being mounted in the cask loading pit as an additional level indication. The current indication will still be available. The new indication will not be around any safety related i equipment and does not interact with any safety related equipment. Therefore, the possibility of a accident of a different type than any previously evaluated in the SAR does not exist. i 1
___ ____ - - - - - - - - - - ~ ARMANSAs NUCLEAR ONE FORM TITLE:
. FORM NO. REU.
10CFR50Je EVALUATION 1000.131B 2 6. Will the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR be created? Yes O No @ The new indication will not be around any safety related equipment and does not interact with safety related equipment. Therefore, the possibility of a malfunction of safety related equipment of a different type than any previously evaluated in the SAR does not exist. 7. Will the margin of safety as defined in the bases for any technical specification be reduced? Yes O No @ The margin of safety for the spent fuel pool will be unaffected by the addition of this level indicator. The current level indicator will remain along with it's alarm functions for the control room. A seismic evaluation has also been completed to show that this indicator will not affect the seismic capabilities of the spent fuel pool.
> C ~
Dan Churchman 09/18/96 Certified Reviewer's Signature Printed Name Date Reviewer's certification expiration date: 08/22/98 Assistance provided by: Printed Namt Scope of Assistance i Date i PSC review by-
~ Date: \
r M D.
fnf V ~/ Page 1 of 2 72.48 R" Mew of ER 963203P201 Document No. ER 963203-P-201 .
Title:
Installation of Additional Spent Fuel Pool Level Indicators Will the proposed Activity:
- 1. Require a change to the VSC Certificate of Cosqpliance?
Yes No X conditions for System Use (including Bases)? Yes No. X 2. Result in a significant increase in occupational exposure related to VSC use? Yes No_ X 3. Result either in information in the VSC SAR or SER being (a) No longer true or accurate, or (b) violate a requirement stated in the document? Yes
- 4. No X Involve a test or experiment not described in the VSC SAR?
Yes No X Basis and
References:
(Include k'aywords if LRS search was used) The proposed activity will add a SFP level indicator the ANO-2 Spent Fuel Po vicin."y. The independent Spent Fuel Storage installation License Basis Do not address scope design details in the Spent Fuel Pool area. Those design charachte of the VSC LBDs. . The following are detailed answers to the 4 determination questions from above: the VSC Certificate of Compliance or Conditions for . installing a ne levelindicators. This indicator is not address .
- 2. Implementation of the proposed activity will HQI result in a significant increase in occupational exposure related to VSC use. The requested changes willimprove w conditions in the ANO-2 Spent Fuel Pool and will not modify Dry Fuel Storage ope in radiation areas.
- 3. The proposed activity will HQT result in information in the VSC SAR or SE ionger true or accumte OR violate any requirements stated in the VSC SAR or SE .
by installing a new levelindicatorin the vicinity of the Cas addressed in the VSC LBDs. FORM TITLE: FORM No. REV. 10CFR72.48 DETERMINATION 1022.039A 0
Page 2 of 2 72.48 Rcview of ER 963203P201 WWbV
- 4. The proposed activity does HQIintroduce any tests or experiments not already discussed in the VSC SAR. No Dry Fuel Storage operational tasks are added by this the changes are to plant equipment not associated with Dry Fuel Storage.
A complete manual LBD search was performed Scope of the LRS search included the VSC SAR, CSU, and SER 1 LRS Search words were: poolw/10 level levelindicat* pooiw/10 design My, Robert J. Priore
\
'Reviesfer's Signattare 7/3/96 Printed Name Date 9/13/97 10CFR50.59 certificaticn Expiration Date W
(,, - i f I' FORM TITLE: ' FORM No. REV. 10CFR72.48 DETERMINATION 1022.039A 0 l
Dae==wat No. 9ST2/kM20f Rev2 Change-No;- +- O
' Title ALTERNATE POWER SUPPLY FOR 2X1A3 AND 43LA Will the proposed Activity:
- 1. _ Require a change to de Operating Heena* including:
( Technical Specificauens (excluding the bases)? Yes O No @ Operating i1,.==? Yes O No @ Confinsatory Orders? Yes O No @
- 2. Result la information in the following SAR docusseats (including drawings and text) being (a) no longer true or accurate, or (b) violate a requirement stated la the document:
Core Operating Limits Report? Yes O No @ SAR (multi-volume set for each unit)? Yes @ No O QAMO?* Yes O No @ E-Plan?* Yes O No @ FRA Yes O No @ Bases of the Technical Specifications?
- Yes - 0 No @
NRC Safety Evaluatha Reports? Yes O No @ Involve a test or experiment act described in the SAR? Yes O No @
- 4. Result la a potential impact to the envir=====t? (Complete the EavironmentalImpact Checklist of this form.) Yes O No @
- 5. Result la the need for a 8"-tj=: Safety Evaluation per section 6.2.4.a?
Yes O No @ 6.' l Result la any potential impact to the equipment or facilitieJ ' utilized for Ventilated Storage Cask activities per section 6.2.4.b? Yes O No @ Casis for Determina*ia=? This inadificarian will install a double throw switch on the wall betund 2B6. This switch will be fed by the exasung feeder from 2B64 to 43LA and by a new feeder from 2B54. This will allow 43LA to be fed by 2B54 when 2B64 is out of semce lhis will be controlled .C - M'=-e My to only be allowed in Mode 5 or Mode 6. This will prevent installing a Temp Alt. each outage to inanatain rndistiaa monitoring in the Refuel area. The switch configuration prevents any ! passiwi ty of tying the two MCCs together. 1.) Although the Operatmg License, the Tech Specs, and the C+ ~=== y Orders do mention the affected MCCs, these ! dar====*e do not address the level c(detail affected by this Modification. 2.) SAR Figures 8.3-11,8.3 15,8.3 66 and 8.3-17 will have to be revised by this modification. The SAR text. the QAMO, E-PLAN, TECH SPEC BASES, or any NRC SER does not specify or desenbe the power source of 43LA or ^ 2XL43. 3.) This Modificanon will not utahze or involve any tests or expenments. 4.) See pass 3 of this 50.59 Deteriniaatina - 5.) No work will be performed in an RCA and nothing will be done that could affect Radiological conditions 6.) This inadincarian will not alter the operation or impact any equipment or facilities utilized for VSC activities. Changes to these dar====*a require an evaluation la accordance with 10CFR50.54. See Section 6.2.1.B. 7oRnt Trrt.fa PORM NO. REY. 10CFR50.59 DETERMINATION 1000.131A 2 PC-2J
r Prge 11 Document No. 9582WP1bl RevdChange No.-- - - O
References:
List sections reviewed la the Licensing Basis Documents, specified in questions 1,2 and was done os LRS, "all" assy be estered under "Section" with the keyword (s) used in parenthe hard copies of the documents shall be reviewed as computer-based searches such as LRS search text only, not figures or drawings. Attach a completed LDCR if LBD changes are require Document ig31gg U2 TECH SPEC ALL "2RE-8233" U2 SAR ALL "2RE-8540" U2 CA.una License ALL " CONTAINMENT PURGE EXHAUST" U2 NSE ALL " CONTAINMENT PURGE MONITOR" U2 SER ALL 2B54 OAMO ALL 2B64 E-PLAN ALL "MCC54" . 43LA. 2XL43 FHA ALL "MCC64" U2 SAR 8.1.2. 8.1.3. 8.1.4. 8.3.1. 8.4; FIGURE 8.3-11. 6.3-15. 8.3-17: TABLE 8.3-1. U2 TECH SPECS 3/44.8. 3.9 11. 3.3.3.9 C N THOMAS N. MARSHALL .7-24-N Cardfied Reviewer's Signature ~ Printed Name - Date Reviewer's certification expiration date: _ 6/20/98 Assistance provided by: Printed Name ' Scope of Assistance Date FORM TITLE: FORM NO. REY. 10CFR50.59 DETERMINATION 1000.131A 2
Document No. %saMPbf - RevxmanrNo.-- o ENVIRONMENTALIMPACT CHECKLIST (UNIT 1 and UNIT 2) Complete the following checklist. If the answer to any checklist item is "Yes", an Environmental Evaluation is required. Se Section 6.2.1.E for additional guidance. i Will the Activity being evaluated: i
.Ytt. .St .O E Disturb land that is beyond that initially disturbed during construction (Le., new construction of buildings, creation or removal of ponds, or other termstrial impact)? See Unit 2 SAR Figure 2.5-17. This applies only to areas outside the protected area.
O E Increase thermal discharges to lake or atmosphere? O E Increase concentration of chemicals to cooling lake or atmosphere i through discharge canal or tower? O E Increase quantity of chemicals to cooling lake or atmosphere thrwegh discharte canal or tower? O E Modify the design or operation of cooling tower which will change drift characteristics? i O E Install any new transmission lines leading offsite? O E Change the design or operation of the inide or discharge structures? O E Discharges any chemicals new or different from that previously
. discharted?
'I O E Potentially cause a spill or unevaluated discharge which may effect neighboring soils, surface water or ground water?
O E Involve burying or pi-----t of any solid wastes la the site area which may effect runoff, surface water or ground water? O -E' Involve incineration or disposal of any potentially hazardous materials on the ANO site? O E Resule la a change to moeradiological emments or licensed Reactor powerlevel? O E Potentially change the type or lacrease the amount of non-radiological air emissions from the ANO site. FORM TITLD FORM NO. REV. 10CFR50.59 DETERMINATION 1000.131A - 2
ARKANSAS NUCLEAR ONE FORM TITLE: . ~ . . . - . FORM NO. REV. 14CFR80.80 EVALUATION 1000.131a 2 1:'p y. q q. cyp. Page /,1 of 10CFR50.59 Eval. No. fUUYf[/ 4 (Assigned by PSC)
'gg Document No. ER-963219P201 Rev1 Change No. A Title Alternate Power Sunolv for 2XL43 and 43LA A WRITTEN RESPONSE PROVIDING THE BASIS FOR THE ANSWER TO EACH QUE ATTACHED. EACH QUESTION MUST BE ANSWERED SEPARATELY. A SIMPLE STATE CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES GUIDANCE FOR RES If the answer to any quashon on this form is "Yes." then an unreviewed safety question is involved. If the to all questions is "No," then the proposed change does not involve an unreviewed safety question.
- 1. Will the probability of an accident previously evaluated in the SAR be increased?
Yes O No @ This activity does not impact any accident initiator for any accident previousiv evaluated in the SAR. It will not ;ac. 22: the Prabeb!ilty of any accident previousiv evaluated
- 2. Will the consequences of an accident previously evaluated in the SAR be increased?
Yes O No @ This activity will not itPa=d r=d3 don da== csE:::aances for any accident evaluetsd in the SAR. It will h;ve no 0.==N on ;nMi==:E.e the "=== cGE M:ances of any accident. This modificaation cannot cae== any accident evalumhd in the SAR and will not impact any miticatino actions which which are ?2t= i ci='M for in the BAR. Therefore, the consecuenehs of an accident previously evs!e=ted in the SAR will NOT be incm===". 3. Will the probability of a malfunction of equipment important to safety be increased? Yes O No @ A Safetv-related breaker is 9-ad to senarate the Non-Q load from the Safety-related bus to ensure l tr.et the % stet-EMV of a malfunction to Safetv-related eaulornent is not increased.
- 4. 1 Will the consequences of a malfunction of equipment important to safety be increased?
Yes O No @ This activity does not emaad any mitioation maeseres or ecuis.T.ent and will not cause the csE:::::a-:== of the nisifunction of any Safetv-ife;ed =auls.T.ent to increase. 5. Willthe possibilityof an accident of a different type than any previously evaluated in the SAR be created? Yes O No @ The activity. this accidents evaluated in the SAR bound any accident that could be costulated to be caused by
ANnANbAh NUCt.t:AN ONE FORM TITLE: FORM NO. REtf. 10CFRSS.88 EVALUAM 1000.1313 2 ["$' I4 6. Will the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR be created? Yes O No @ The only tvos malfune+!en this activity coeM ===51biv cmate is the isss of one Safetv-related b Jnd thatthan G;;.r.iii is evaluated evaletd inin the the SAR. SAR. It cannot create the nossibility of a malfunction of ecuipment 7. Will the margin of safety as defined in the bases for any technical specification be reduced? Yes O No @ This activity will have no impact on any marnin of =_ ..i specified or implied in the bases for any technical saacification. Thomas N. Marshall 2/27/97 (Certified Review 6fs Signature ' Printed Name Date Reviewer's certification expiration date: 6/20/98 Assistance provided by: Printed Name Scope of Assistance Date PSC review by: L b _ Date: A %
ARKANSAS NUCLEAR ONE FORM TITLE:
. . , , . , _ ,, FORM NO. REV.
~
10CFRec.G EVALUATION 1000.1318 2 I Page___ cf 10CFR50.59 Eval. No.JFi l Mf (Assigned by PSC) Document No. ER-963219P201 RevJChange No. I 0,,,,1,, 1 ( Title Alternate Power Seeniv for 2XL43 and 43LA i l A WRITTEN RESPONSE PROVIDING THE BASIS FOR THE ANSWER TO E ATTACHED. EACH QUESTION MUST BE ANSWERED SEPARATELY. A SIMPL CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES GUIDANCE > If the answer to any question on this form is "Yes," then an unreviewed safety question is involj to all questions is "No," then the proposed change does not involve an unreviewed safety question. 1. Will the probability of an accident previously evalucted in the SAR be increased? Yes O No @ This activity does not imaad any accident initiator for any accident previousiv evaluated in t SAR. This activity cannot e=e== any accident previousiv evaluated in the SAR. Therefore, i increase the Probability of any accident previousiv evaluated. 2. Will the consequences of an accident previously evaluated in the SAR be increased? Yes O No @ This activity will not ima=d r='Hetion dose consecuences for eny accident evaluated in the SAR It will have no imEEd on m-M;==%c the dose consecuences of any accident. This modificatio annot eeeee any Ecc!-Ant ev=!a=*=d in the SAR and will not impact any miticatino actions which are taken credit forin the BAR. Therefore. the consecuences of an accident previousiv evaluated in the SAR will NOT be incr====-i. 3. Will the probability of a malfunction of equipment important to safety be increased? Yes O No @ A Safetv-rmwad breaker is e==d to separate the Non-Q load from the Safetv-related bus. This sa! IggL),p &;ie&dv sc.s Ed freivi one Safetv-related bus. This oDtiCnal feed Can only be used durino Mode sve!!Ebh. 5 or Meds 6 to maintain radiation monitorino capability when the normal feeder is not A '=!!e o of this aae!ET. sat would not result in a plant trio nor would it result in any Safetv-related man!ss at beino out of service. Use of a Safetv-related breaker Probability of a malfunction to Seietv-related eculoment is not increased. i 4. Will the consequences of a malfunction of equipment important to safety be increased? Yes O No @
' This activity does not imosct any mitioation measures or eauioment.
It will allow monitorina
=aS!L... eat to R.T.&in EE= Ebb durino train outanes, and thus, could decrease the consecuences o a malfund"-:-n of 8; Gi-related =ae! ament . The only oculoment involved that is important to safety is ?"M and the plant is analvzad for a malfunction of 2854.
ARKANSAS NUCLEAR oNE FORM TITLE:
. . . . . . . , , , FORM NO, REV.
10CFR$0.59 EVALUATION 1000.131B 2 5. Will the possibility of an accident of a different type than any previously evaluatedin the SAR be created? Yes O No @ The accidents evaluated in the SAR bound any accident that could be postulated, to be caused by this activitv. The only failures that could be caused by this activity is loss of 2B54 or 2864, but not both. The plant is desioned for the loss of either of these MCCs without an accident beino created. 6. Will the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR be created? Yes O' No @ The only tvoe malfunction this activity could possibly Create is the loss of one Safety-related bus and that is evaluated in the SAR. This activity meets industry codes and standards and will not cetate the possibility of a malfunction of eau, loment different than evaluated in the SAR. 7. Will the margin of safety as defined in the bases for any technical specification be reduced? Yes O No @ This activity will not recuire removino any souloment from service when it is reouired to be operable and technical will have no impact on any maroin of safety specified or implied in the bases for any specification. d' i 4 Thomas N. Marshall 01/13/98 Dertified Revievderspignature 7 Printed Name
, Date Reviewer's certification expiration date: 6/20/98 Assistance provided by:
Printed Name Scope of Assistance Date PSC review by: N c PM - Date: l i
AMAANSAs NUCLEAR ONE FORM TITLE:
. FORM NO. REv.
10CFRSO.89 Revision 1000.131p 2 Page 1 of_1, Document No. 96-8016 RevlChange No. 1 10CFR50.59 Eval. No. Revision No. This form is to be used to document Revisions to 10CFR50.59 Evaluations. Evaluation Refer to sectionafter PSC 6.2.6 of review may this procedure. become necessary due to SRC review, changes to th , Reason for revision to 10CFR50.59 Evaluation: DCPR #3 to modification packaos 96 8016 "Modifir etions to 2CV-1460 SW Soueeze Valve" add numbers to components installed at 2CV-1460. ihese new components and tan numbers wil shown on 2SAR floure 9.2-1 (P8.tD M-2210 st.2). These new components were added to t circuit of valve 2CV-1460 by the oriainal packane. but not shown on the drawino. At close modification packeee. It was determined that in order to expedite maintenance on valve 2C components in the air control circuit should have component tao numbers. Both the oriainal 60. determination and evaluation were revised to included verbiace about the new components. Will the proposed revision result in any additional:
- 1) Change to the Operating License?
Yes O No @
- 2) Change to other Licensing Basis Document?
Yes O No @
- 3) Conduct of tect or experiment?
Yes O No @
- 4) Impact to the environment?
Yes O No @ If yes, describe below and take appropriate action as per initial Determination: Indicate revisions to the 10CFR50.59 Evaluation by placing revision number at the top r page of the form (s). Changes should be lined through, initialed, dated and indicated with the For extensive changes, new forms may be used with revision bars in the marDin deno form to front of previous 10CFR50.59 Evaluation. Retum to the PSC for review. . d des Cepfied Chuck Sosnv 7s 7 jewers Signature Printed Name
' 'Date Reviewers certification expiration date: 8-23-97 PSC review
- W93 AY Date: 9-/~/~~ f 7
? .
1 l
NuCM ONE ~
-"'**N FORM TITLE:
FORM NO. REV. 10CMt90.89 DETERMINATION 1000.131A 2 l Document No. PC 96-8016 RevlChange No. 1 Title Modifications to 2CV-1460 SW BemM Valve Will the proposed Activity: l 1. Require a change to the Operating License including: Technical Specifications (excluding the bases)? YesO No@ Operating 1.icense? YesO No@ Confirmatory Orders? YesO No@
- 2.
- Result in information in the following SAR documents (including drawings and text) being (a) no longer true or accurate, or (b) violate a requirement stated in the document
- j i
SAR (multi-volume set for each unit)? Yes@ nod QAMO?* YesO No@ E-Plan?* YesO No@ FHA YesO No@ Bases of the Technical Specifications? YesO No@ NRC Safety Evaluation Reports? i YesO No@
- 3. Involve a test or experiment not described in the SAR7 YesO No@ ,
4.
. Result in a potentialimpact to the environment? (Complete Environmental Impau Checklist of this form.) .
YesO NoS j
- 5. .
Result in the need for a Radiological Safety Evaluation per section 6.2.4.a? YesO No@ 6. Result in the need for a 10CFR72.48 Review per section 6.2.4.b? YesO No@ i Basis for Cr ;re;,,e;0i? ! This PC will modify the pneumatic control circuit of service water valve 2CV-1460 (squeeze v current setup of the pneumatic control circuit, attempts to take manual control of this valve are da g personnel and equipment. This PC will change the failure position of the valve from " fail-open* n.n PC will also add components to allow safe swap over from pneumatic operation to hand
#1 airsupply isoledon valves, adds two new pressure indicators, adds a new pressure regu tag nuinbers to two amisdag pressure indicators). A review of the LBD's shows that 2CV-1460 the 2SAR and the FHA. However, failure position of the valve and pneumatic components o A- detailed in either writeup. The failure position is shown on figure 9.21 (P&lD M-2210 sh.2) and
$ the valve to " fail-as is* oradding the new components does (Q1&2) The other LBD's ,2T/S,2OL,2CO, QAMO, E-plan,2T/S Bases,2NSE & 2SER are not in n revision, since these documents do not get to a level of detail to discuss the valve in questio h - l require revision. However, the 2SAR, figure 9.2-1 will need to be revised to show the valves os and the new composants. (Q3) This PC does not involve a test or experiment not describ This PC does not impact the environment nor does it result in a (Q5) Radiological Safety
' does not impact the VSC program. Since a yes answer has been marked as a response to qu evaluation will need to be performed and is attached to this determination.
- Changes to these documents require an evaluation in accordance with 10CFR50.54.
See Section 6.2.1.B. L . l .
ARKANaAs NUCLEAR ONE FORM TITLE: , FORM NO. REV. 10cFR80J8 DETERMINATION 1000.131A 2 Page1 of .3 Document No. 96-8016 Rev> Change No. 1
References:
List sections reviewed in the Licensing Basis Documents, specified in questions 1,2 an keyword search was done on LRS, "all" may be entered under "Section" with the in parentheses. Controlled hard copies of the documents shall be reviewed as comp searches such as LRS are not controlled and search text only, not figures or drawings. A completed LDCR if LBD changes are required. Document Section 2SAR 2T/S ALL (2CV-1460,2CV1460, outfall, basin level, cooling tower w/10 basin,2LIC squeeze,2HCD-26-30*) 2OL
- 2CO's "
QAMO " E-Plan " 2T/S BASES
- 2NSE
- 2SER
- FHA "
2SAR Figure g.2-1 (change failure position of valve 2CV-1460 from FO to FAl) C- DJ _ Chuck Sesnv 7 o!r7 - Certifiep Reviegs Signature Printed Name I bate Reviewer's certification expiration date: 8-23-97 Assistance provided by: Printed Name Secpe of Assistance None Date
ARKANSAS NUCLEAR ONE ete- p A FORM TITLE: FORM NO. REV. 10CFR80.89 DETERMmeATION 1000.131A 1 PageJ_,of.3 ENVIRONMENTAL IMPACT CHECKLIST (UNIT 1 and UNIT 2) l Document No. PC 96-8016 Rev/ Change No. 1 Complete the following checklist. If the answer to any checklist item is "Yes", an Environmental Eval required. See Section 6.2.1.E for additional guidance. Willthe Activity being evaluated: Yes go O g oisturb land that is beyond that initially disturbed during construction (i.e., new cons buildings, creation or removal of ponds, or other terrestrial impact)? See Unit 2 SA 2.5-17. This applies only to areas outside the protected area. O' s increase thermal discharges to lake or atmosphere? O g increase tower? concentration of chemicals to cooling lake or atmosphere through discharg O g increase quantity of chemicals to cooling lake or atmosphere through discharge canal or tower? O g Modify the design or operation of cooling tower which will change drift characteristics? O g Install any new transmission lines leading offsite? O a change the design or operation of the intake or discharge structures? O g Discharges any chemicals new or different from that previously discharged? O g Potentially water cause or ground a spill water? or unevalurted discharge which may effect neighboring soils, s O g involve burying or placement of any solid wastes in the site area which may effect ru surface water or ground water? O g involve incineration or disposal of any potentially hazardous materials on the ANO site? O g Result in a change to nonradiological effluents or licensed reactor power level? O g Potentially change the type or increase the amount of non-radiological air emissions fro ANO site. I 1 1 i i
~
FORM TrTLE: asuu6usas NUCLEAR oNE '
. FORM NO. REV.
10CPRSO.88 EVALUATION 1800.1313 2 FFrV- - 0 10CFR50.59 Eval. No._ T M "O Document No. PC 96-4016 (Assigned by PSC)
%9 Rev> Change No. 1 Title Modifications to 2CV-1480 SW Se'_P:+ Valve ATTACHED. EACH QUESTION MUST BE ANS CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES GUID to all questions is "No," then the proposed change does not .
1. Will the probability of an accident previously evaluated in the SAR be increased? Yes O No @ This modification will chance the fei!ere E-3sH;0n of the 2CV-1460 Service Water and add ggrggg several 9:+rsilon. The componsids faltere E-3: to the oneumatic Dailion of the a&m'er to improve rcs ove 4;On of this valve is shcwa on 2SAR floure 9.2-1 and this floure will need kg to be revised to chaaGe the fai!ere E-35nion from FO to FAl. 5 F!en_=e 9.21 wilf =5 = newir s-fe-f cc...-s .e..;. to the casi . ;, sG; ties of the =t.: actuator. This valve is inst:ll portion of the Service Weier System which is non-Q and non-a=is-idc. A rev'r.; of the
.in Sis;_Chsater si and15 of the 2SAR 2SAR (15.1.30 Less of Service Water Systern. & 18.f.34 8 T=Na ofa**
;&.ein.T at Air g 9.2-5 (Service Water System Sinale Fe!!ere Analysis) shows no accideai previousiv evalustad by the 2SAR beina initiated by valve 2CV-1460. The . new ccer.,c; 9:-i-2 to the air === 'y *:nes were ::.'n+2 to handle the Ge-E -"o and :;.T &ikidre.Ja5:scica: at '
=t,z 2CV-1460. The&& new ccipeOnents are no more likelv to fall than any coir--:-5&
li ;,dment be increased. Air Sve; . Therefore. the probability of an sccident previous!v eve!eeted in
\
- 2. .
Will the consequences of an accident previously evaluated in the SAR be increased? - ' Yes O No @ 4g A review of the accidents described in Chapter 15 of the 2SAR (15.1.30 Loss of Serv 15.1,34 ' a** N _ ofiss;ic.T st Air Svsism) and 2SAR Table 9.2-5 (Service Water System Sinole Fa Analysis) shows no ardent previousiv evaluated by the 2SAR beino initiated by valve 60== ce6593eences willwillnot not be imDSCted by this Chance. Therefore. the consecuenc previcesiv evaluated in the SAR be incremead. s
ARKANSAS pasar's war ONE na m s a e FORM TITLE: ~T FORM NO. REtf. 10CPR80.49 DETERMINATION 1000.131a 2
- 3. fy2.J3 Will the probability of a malfunction of equipment important to safety be increased?
Yes O No @ The portion of the Service Water System where this valve is located is co LeaLMcd from the safety isidied DGrison of the service w&ist system . Tw the lake retum dia*haiss line from the safety related LOitian of the system ind&sendsaus is&d to a cGmLC loss of the 2CV-1460. of SW retum flowFailure durina aof theshutdown. clant actuetorThis and could s ma ecs;;as to the emsiaancy C=::-;;. C0mberWe'v 40ns to rm*0's service water flo w have already been scGb b:;N of a 2CV-1460 fe!!ure has not been incr==:indEded ineAO valve falls as is. no matter what h r-+ns to the lanument air system the mGdific4Han installed. the only w&v to fail the valve do*ad will be . to de n er (PMEE neric-jis!!v checia for :eeks on the air cdiadaG. Where as before the modificat on. a leak in the ai_r failure will actue!!v be raducs! by inet:Gac this mGeif on. the orah=hi!ity of a malfunction of aaM=asat important to safety will not be incrgggt 4. Will be the consequences increased? of a malfunction of equipment important to safety Yes No @ 2CV-1460 could imaset aae!c-aent imoortant to sa shutdown cos ;aat Enoineerino Report 90-R 2014-01 failure (i.e. valve GGlaa full de==d) by removino rvno th the valve full open and hGidhai it open. After innte"inc this modification it is to the important eculoment full dotad to safety cesilion. Therefore. will not be increased _. hmead on this evaluatio 5. Will the possibility of an accident of a different type than any previously evaluated in the SAR be created? Yes O No @ Failure of the actuator to remain " apis" after loss send of in makeup water to coolina tower hasin. The safety related portions of the Servic remain unaffected. Failure of this valve is not descissd in the 2SAR. Therefore evaluation, the cessibility of an accident of a different tvoe than previousiv eva created.
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1 _ ,n. - ARKANSAS NUCLEAR DNE
. b FORM TITLE:
~***M 10CFR80.88 DETERMINATION FORM NO. REV.
1000.1313 2 6. fA*f. 3.k.$ Will the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR be created? Yes O No @ . safety would be for 2CV-1460 to fall fully closed. If 2C water system can be oDened to the ECP. and coolino flow restored. The cortio . Egtem in which valve 2CV-1460 is located can be isolated from the safety re By installina this modification, it is less likely that valve 2CV-1460 will fail in( based upon this evaluation. in the SAR *he will notpossibility be created. of a malfunction of eculoment im tvoe than previousiv evaluated 7. Will the margin of safety as defined in the bases for any technical specification be reduced? l Yes O No @ which is cart of the Unit 2 Service Water Systema review of the Unit 2 T - defined in the bases for any technical specification will not be reduced. Based uoon this ev ; l i l ) Cl1 8 Certifi#d Revgers Signature Caucu Sesnv 7[o/g7 Printed Name
/ ' Date Reviewers certification expiration date:_ 8-23-97 Assistance provided by:
Printed Name
-Nstag Scope of Assistance Date PSC review by: _ a&M '= U Date:__ 7-/7-77 I
l l l l
FORM TITLE: NUCLEAR ONE
- j. FORM NO.
10CFR80.89 DETERMINATION REV. 1000.131A 2 PC-2,3 k Page l of k Document No. ER973636P201 RevJCha' ige No. O Title Plant Heatina System Chemical Addition Pot installation Will the proposed Activity: 1. Require a change to the Operating License including: Technical Specifications (excluding the bases)? YesO No@ Operating License? YesO No@ Confirmatory Orders? 2. YesO No@ Result in information in the following SAR documents (including drawings and tex (a) no longer true or accurate, or (b) violate a requirement stated in the document: Core Operating Limits Report YesO No@ SAR (multi-volume set for each unit)? , Yes@ nod QAMO?* YesO No@ E-Plan?* YesO No@ FHA YesO No@ i Bases of the Technical Specifications? YesO No@ NRC Safety Evaluation Reports? 3. YesO No@ Involve a test or experiment not described in the SAR?
~4.
YesO No@ Result in a potential impact to the environment? (Complete Environmental impact Checklist of this form.) YesO No@ l 5. Result in the need for a Radiological Safety Evaluation per section 6.2.4.A? i YesO No@ { 6. Result in any potential impact to the equipment or facilities utilized for Ventilated l Storage Cask activities per Section 6.2.4.87 i YesO No@ j Basis for Determine *Jen: The installation of a new chemical add. tion pot will allow for safe addition of chemic , concentration of chemicals added to the system. The me corrosion is not discussed in the U2 SAR.
' revision to the U2 SAR Figure No. 3.2 2 (P&lD M-2220 Sht 1 and 3).
. FHA. T.S. Bases and the U2 SER's do not provide a level of detail to be impacted by ,
Figure, a 50.59 Evaluation will be completed.is no impact to the environment, ra .
- Changes to these documents require an evaluation in accordance with 10CFR50.54.
See Section 6.2.1.8.
ARKANSAS NUCLEAR ONE 17363f f201 M?AE/ FORM TITLE FORM NO REV. 10CFR60.59 DETERMINATION 1000.131A 2 PC-2.3 l Page 2 of Document No. ER973636P201 Rev> Change No. O
References:
List sections reviewed in the Licensing Basis Documents, specified. in If a quest in parentheses. Controlled nard copies of the doc searches sach as LRS are not controlled and search text only, not figures or d completed LDCR if LBD changes are required. Document Section ZYINDEX 50.59 "ALL" with the keywords: (Unit 2 Index)
" ". Heatina. Plant Heatina. Chemical Addition, Inhibitor. Corrosion inhibitor M AMA ,A y Certified Reviewers Signature Bruce Franklin _ _ _ 3/27/97 Printed Name Date Reviewers certification expiration date: 8-24-98 Assistance provided by:
Printed Name Scope of Assistance .N!b Date
- s scv sur / +vi ! /0 UJ C /
ARKANSAS NUCLEAR ONE FORM TITLE: FORM NO. REV. 10CFR60.69 DETERMINATION 1000.131A 2 PC-2.3 l Page 3 of ENVIRONMENTAL IMPACT CHECKLIST (UNIT 1 and UNIT 2) Document No. ER973636P201 RevdChange No. 0,, Complete the following checklist. If the answer to any checklist item is "Yes", an Environm required. See Section 6.2.1.E for additional guidance. Will the Activity being evaluated: Yes No O O Disturb land that is beyond that initially disturbed during construction (i.e., new conI ' 1 buildings creation or removal of ponds, or other terrestrialimpact)? See Unit 2 S 2.5-17. This applies only to areas outside the protected area. t O O Increase thermal discharges to lake or atmosphere? O E l increase tower? concentration of chemicals to cooling lake or atmosphere through disc l l 0 2 Increase tower? quantity of chemicais to cooling lake or atmosphere through discharge ca O O Modify the design or operation of cooling tower which will change drift characteristics? O @ Install any new transmission lines leading offsite? O @ Change the design or operation of the intake or discharge structures? O @ Discharges any chemicals new or different from that previously discharged? I C @ Potentially water cause or ground water?a spill or unevaluated discharge which may effect neighboring s 0 @ l Involve burying or placement of any solid wastes in the site area which may effect runo surface water or ground water? , O O Involve incineration or disposal of any potentially hazardous materials on the ANO site O O Result in a change to nonradiological effluents or licensed reactor power level? O E Potentially change the type orincrease the amount of non-radiological air emissions ANO site.
ARKANSAS NUCLEAR ONE k /2, FORM TITLE; FORM NO. REV. 10CFR40.89 EVALUATION 1980.131B 2 I i Page 4 of 10CFR50.59 Eval. No. fd#l-(Assigned by PSC) ' 0 Document No. ER973836P201 RevJChange No. O Title. Plant Heatino System Chemical Addition Pot installation ATTACHED. EACH QUESTION MUST BE A CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVID . to all questions is "No," then the proposed change doe . 1. Will the probability of an accident previously evaluated in the SAR be increased? Yes O No @ The oram!!"v relocates the et.4=kel of1-:-Tri-:-n an ec-:?-fir.: crevieuelv for addino ct.e.T.icals to theevaluated plant heatino system. in the BA acce==!E-!!ity and seressne; safety concerns. This resolves function and does not chance the operatino characteristics.This chanos does5-not &T;.c
. -- .- _ .. .: ;e . . . __07.; :----_ n:.; ... ~ r^ "- ^'n: ?;:^ -- x;;
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2. W;ll the consequences of an accident previously evaluated in the SAR be increased? Yes O No @ The consecuences of an accident previousiv evaluated in the SAR will notThis be inc chans. dsas not involve any tvoe of accident as the Plant Heat;as Svei m is no is non-risk sionificant. clant. Implementation of this clant chance will not impact the ossrstion of the 3. Will the probability increased? of a malfunction of equipment impoltant to safety be Yes O No @ chanss sc--5 not involve any ee9!s LeGt laissitEnt to plant saf accessibility to the chemical addition pot on the plant hostino system. The reduces the sct:3tial for personnel chemical contaminction. The Plant Mastino safety related system and is not reauired to mitiaate the consecuences condition. , 4. Will be the consequences increased? of a malfunction of equipment important to safety Yes O No @ The consecuences of a malfunction of soulomont important to safety will not . The be inc the plant hMns svsi m recirculation numns. The new s lo The chamical ade"_ son not will connect across the o s system t-c-c^^Z pumns. No eau le.T,ent important to safety is located in this area a comcGGeat. eddition not or the olant heatino system in this area will not affect any s chemical
FORM TITLE: ARKANSAS NUCLEAR ONE FORM NO. REV. 10CFR$0.89 EVALUATION 1000.131B 2 5. Will the possibility of an accident of a different type than any previously evaluated in the SAR be created? Ye? O No @ The possibjkV increased. of an accident of a different tvy than oreviously evaluated in the SA e The Diant heatino system is non-safety related and is not risk sionificant . The plant The only safety function of the system is to provide ant conta heatino system containment isolation.chemical addition pot can not affect the system roouirement no of pro 6. Will the possibility (J c .nifunction of equipment important to safety of a different type th.an any previously evaluated in the SAR be created? Yes O No @ The possibility of a malfunction of eouipment important to safety of a different usiv type t provide containment isolation. Relocatino the chemical ocation add will not impact the system ability in fulfillino the containment isolation function modes are created by the relocation of the chemical addition not. . No new failure 7. Will the margin of safety as defined in the bases for any tech 5ical specification be reduced? Yes O No @ The plant heatino system is not included in the olantced.technical
. The maroin of safety will not affect the operation of the system northe plant.will not be affect . It d4476 2Ar,tt Certified Re.%wer's Signature Bruce Franklin 3/27/97 Printd Name Date Reviewer's certification expiration date: 8-24-98 Assistance provided by:
Printed Name N/A Scope of Assistance Date C review by: M-_ _ _ N Date: M 1 1
anum monova.ce vne FORM TITLE: FORM NO. REV. 10CFR50.59 DETERMINATION 1000.131A 2 PC-2,3 Page 1 of 5 Document No. 973673P201 RevlChange No. 2 pt 973673P201 Title ANO.2 LOCAL ANALOG DP INDICATIONS FOR 2F6A. 2F6B AND 2F6C P AliF 3 REV 0 Will the proposed Activity:
- 1. Require a change to the Operating License including- u Technical Specifications (excluuing the bases)?
YesO No@ Operating License? YesO No@ Confirmatory Orders? YesO No@
- 2. Result in information in the following SAR documents (including drawings and text) being (a) no longer true or accurate, or (b) violate a requirement stated in the document:
Core Operating Limits Report YesO No@ SAR (multi-volume set for each unit)? Yes@ nod QAMO?* YesO No@ E-Plan?* YesO No@ FHA YesO No@ Bases of the Technical Specifications? YesO NoS NRC Safety Evaluation Reports? YesO NoS
- 3. Involve a test or experiment not described in the SAR?
YesO No@
- 4. Result in a potentialirnpact to the environment? (Complete Environmental impact Checklist of this form.)
YesO No@
- 5. Result in the need for a Radiological Safety Evaluation per section 6.2.4.A?
YesO No@
- 6. Result in any potentialimpact to the equipment or facilities utilized for Ventilated Storage Cask activities per Section 6.2.4.87 YesO No@
Basis for Determination: SEE ATTACHED CONTINUATION PAGE Changes to these documents require an evaluation in accordance with 10CFR50.54. See Section 6.2.1.B.
ARKANSAS NUCLEAR ONE FORM TITLE; FORM NO. REV. 10CFR60.89 DETERMINATION 1000.131A 2 PC-2,3 Pc 9'13673P201 age 2 of 5-Document No. 973673P201 P AGE v' REV
, Rev./ Change No, !
References:
List sections reviewed in the Licensing Basis Document's, specified in questions 1,2 and 3. If a keyword search was done on LRS, "all" may be entered under "Section" with the keyword (s) used in parentheses. Controlled hard copies of the documents shall be reviewed as computer-based searches such as LRS are not controlled and search text only, not figures or drawings. Attach a completed LDCR if LBD changes are required. Document Section ANO 2 Tech. Spec. ALL ANO 2 Operatina License f ANO-2 Confirmatory Orders M A j g ANO 2 SAR M A QAMO ALL E-Plan FHA M ANO-2 Bases of the Tech. Specs. M ANO-2 NRC SERs E ! M LRS Keyword Search: , service water w/10 filter service water w/10 strainer service water w/10 Dressure filters w/10 differential pressure strainers w/10 differential pressure l intake structure l service water w/5 oumo
!!! LIE '
The followina taa numbers were also searched for:
)
F6 JF,,6A 2Fjg F 2F6C 2PDT1426 2PDT1432 2PDT1438 2PDIS1426 2PDIS1432 2PDIS1438 2SW1426A 2SW14268 25W1432A 2SW1432B 2SW1438A 2SW1438B Rooer B. Rucker 7/22/97 Certified Reviewer's Signature Printed Name Date Reviewer's certification expiration date: 9/13/97 Assistance provided by: Printed Name Scope of Assistance Date { j
AMnANT A5 NULLtAM UNt: FORM TITLE:
. FORM NO. REV.
10CFR$0.59 DETERMINATION 1000.131A 2 PC-2,3 Page 3 of_5 ENVIRONMENTAL IMPACT CHECKLIST (UNIT 1 and UNIT 2) PC 973673P201 Document No. 973673P201 Rev> Change No. 0 PAGE# C REVO Complete the following checklist. If the answer to any checklist item is "Yes", an Environmental Evaluation is required. See Section 6.2.1.E for additional guidr.nce. 2 Will the Activity being evaluated: Yes No O E Disturb land that is beyond that initially disturbed during construction (i.e., new construction of buildings, creation or removal of ponds, or other terrestrialimpact)? See Unit 2 SAR Figur 2.5-17. This applies only to areas outside the protected area. I O @ Increase thermal discharges to lake or atmosphere? O g increase concentration of chemicals to cooling lake or atmosphere through discharge canal tower? O @ Increase quantity of chemicals to cooling lake or atmosphere through discharge canal or tower? O B Modify the design or operation of cooling tower which will change drift characteristics? O B instati any new transmission lines leading offsite? O a Cnange the design or operation of the intake or discharge structures? O @ Discharges any chemicals new or different from that previously discharged? O O Potentially cause a spill or unevaluated discharge which may effect neighboring soils, surface water or ground water? O @ involve burying or placement of any solid wastes in the site area which may effect runoff. ' surface water or ground water? O @ involve incineration or disposal of any potentially hazardous materials on the ANO site? O @ Result in a change to nonradiological effluents or licensed reactor power level? O E Potential y change the type or increase the amount of non-radiological air emissions from the ANO site. l l I 1 L
ARKANSAS NUCUEAR ONE FORM TITLE: FORM NO. REU. 10CFRSO.58 REvEW CONTINUATION PAGE 1000.131C 2 Page 4 of.5 Document No. PC 973673P201 RevlChange No. P 9 73 W81 10CFR50.59 Review Continuation Pace Basis for Detenninstion This PC is providing a backup differential pressure indication for the Service Water Basket Straine and 2F6C. DCP 93-2007 installed a DP transmitter at each basket strainer, which provides a signal t Dixson indicator and the Plant Computer. The Dixson indicators provide the required high alarm Control Room annunciators for each basket strainer. This DP loop does not require vital power, sin perform a safety function other than pressure boundary. Therefore, when power for this loop is disru no indication or alarms for the basket strainer differential pressure. This PC is installing three Non-Q mechanical differential pressure gauges. These gauges will be parallel with the existing Basket Strainer DP transmitters. This will provide a direct backup indicatio Strainer DP, This indication should only be used during a malfunction of the existing DP indication accuracy of the existing equipment is greater than the mechanical indication. All of the new equipment will be required to be installed as "Q' equipment to meet the pre requirements of the Service Water System: however, the three new DP gauges will be tagged as Non instrument valves on the high and low pressure side from the existing instruments will act as the between the "Q' and "Non-Q" equipment. These valves will be normally closed, and will be ope readings from isthe backup indication mechanical DP gauges. These readings will only be necessary during a surveil required. Response to Determination Questions:
\
Question 1. No changes to the Operating License will be required since this modification is structured to c Operating Ucense documents listed in question 1. The Technical Specifications have no specific operability requirements for the equipment being added by this modification, and none of the existing equipme affected by this modification. Question 2. l l The only License Based Document that is being impacted by this Plant Change is the SAR which represents drawmg M-2210 SH 1 (Piping and instrument Diagram - Service Water revised by this Plant change. A Licensing Document Change Request is included in the Plant 10CFR50.59 Evaluation is attached. No other LBDs were made untrue or inaccurate by this this modification is below the level of detail contained within these documents.
ARKANSAS NUCLEAR ONE FORM TITLE:
- FORM NO. REV.
10CFR60.59 REVIEW CONTINUATION PAGE 1000,131C 2 Page i of_1 Document No. PC 973673P201 Rev/ Change No. 0-PC 973673P201 PAGE 7 REV O Question 3.
~
No testing is required by this Plant Change other than typical Post Modification testing. This Plant Change does not constitute a test or experiment as defined by Procedure 1000.131. This Plart Change will only provide normal post-modification testing as defined in approved ANO procedures. Question 4. This Plant Change will not result in any adverse impacts to the environment. The operation of the plant will not be changed in any way which will result in changes to the air, water or soil conditions of the site, Question 5. This change does not involve processing of radioactive material outside of Controlled Access. Question 6. This change does not involve any equipment used in handling Spent Fuel Storage Casks.
ARKANSAS NUCLEAR ONE FORM TITLE: FORM NO. REV. 10CFR50.59 EVALUATION 1000,131s 2 Page l of_3
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10CFR50.59 Eval. No. FFklO Kr8 (Assigned by PSC) Document No. 973673P201 Rev./ Change No. O Pc 973673Pt0f Title ANO-2 LOCAL ANALOG DP INDICATIONS FOR 2F6A. 2F68 AND 2F6C PAGE B REV 0 A WRITTEN RESPONSE PROVIDING THE BASIS FOR THE ANSWER TO EACH QUESTIO ATTACHED. EACH QUESTION MUST BE ANSWERED SEPARATELY. A SIMPLE STATEMEN CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES GUIDANCE FOR RESPONS If the answer to any question on this form is "Yes,"then an unreviewed safety question is involved. If the answer question.to all questions is "No," then the proposed change does not involve an unreviewed safety (
- 1. l Will the probability of an accident previously evaluated in the SAR be increased?
Yes O No @ No credit is taken for the Service Water Basket Strainer DP indications. This modificati csaisin any eeWement that performs a safetv-reistsd control function, nor does it contain any eauim.;at that is credited for automatic action. This modification does not interface with Safety i Related eouiw-eat that is not isolated, nor will it add any Safety Related eaula.en;. The sssi-Milty of the service water sve;.m will not be affected by this PC. since the assi=issa of the eervice water system ca:V raed.es one tisin of oisino with one pump which inclue== one basket strainer. This PC will oniv orovide additional berkus indication for differential Gr=Esi=e s it;;; serv;ce water U:"M ei..;sers. The sserstion or failure of the new or eM:"as .i. iner DP led;ca?;sn is not an ect;Cai initiator to any of the accidents listed in the SAR. nor es 5 it interface w;;t any ME!w a; that is an initiator. Therefore, the probability of previousiv evale&d LBD accidents is not increased. *
- 2. Will the consequences of an accident previously evaluated in the SAR be increased?
Yes No @ The consecuences of accidents discussed in the LBDs will not chance as a result of this Pla Chsass. The eeulw.en; and sctisns reovired to mitiaate each accident will be un.G.cted by this j nsdirratisa. S; ace the insdification has been desianed with DroDer Dhysical isolation /seDaration ' and w;th seismic 16;.siiiv. the new desion will not fail in a mode that will adverselv .%.ci any safety func^;ss. The slas;& failure analysis in Table 9.2-5 in the SAR accounts for the failure in the seismic catenorv i sr ::Ps boundarv of one trein. Since the two trains are isolated frain eech other. a failure in one trein will not affect the other train. Each train is desioned for 100% capacity for minimal reoutred flows for all Operatino conditions. The dose consecuences assodeMd with DreviousIV evetEI SCCidents will not be affGCted as a result of this modification. Therefore, the consecuences of accidents Dreviously evaluated in the SAR will not increase. l
m m.. ~ ... e n w u. FORM TITLE; FORM NO. REV. 10cFRSO.69 EVALUATION 1000.131B 2 Page 2 of-3 PC 973673P701
, PAGE ? REV O 3.
Will the probability of a malfunction of equipment important to safety be increased? Yes O No @ The soulomont affected by this Plant Chance does not interface (i.e. controll with souloment that is considered important to safety. This modification does not affect any souloment or cablino that performs any control or !nterlock functions with safety or non-safety related systems. The desion cGnfiauration the of the added followino discussion. eouisa-nt is in accordance with ANO desian standards as d The current desian standards for interfacina with safetv-related pipina#@!na includes physical isolation and seeeration to prevent propaastion of a failure from a non-safety system. The desian of this modification is in keepina with those standards. The lateority of the safety-related plainG/tubina has been ensufed by meetina seismic installation standards per eenlicable soproved ANO details and procedures, e Additional fire loadina. and heat loadino have all been properly addressed, evaluated and found acceptable. Batterv and diesel loadina was not affected by this modification. There are no new failure modes introduced to eculoment that is important to safety, pgd on the above discussion. It can be determined that this Plant Chanae will not increase the probability of a malfunction of souipment important to safety. 4. Will the consequences of a malfunction of equipment important to safety be increased? . Yes O No @ The eauioment, tubina and valves installed by this modification does not increase reliance on eouipment important to safety. As concluded in the resDonse to cuestion 3.0. the ohvsical confiouration of this modification ensures that the probability of a malfunction of souloment important to safety has not been increased. Therefore, the offsite dose cc,nseauences associated with a malfunction of equipment important to safety is not increased as a result of this modification. 5. Will the possibility of an accident of a different type than any previously evaluated in the SAR be created? Yes O No @ The function of the couloment affected by this modification is not reouired for shutdown of the unit, mitiaatino radioactive releases or maintainino reactor coolant pressure intearity. It has been demonstrated that the installation of this modification will have no neoative impact on a safetv-related system or component. The installation of this modification will also not chance the way i Operations will respond to an accident. No credit is taken in the current accident analyses for any automatic or manual action by the existino strainer DP indications. Failure of the couloment installed by this modification will not create any accident initiators. Applicable desian j reouirements have been considered (see response to ouestion 3.01 to ensure that systems ! important to safety are not ieopardized; therefore, it can be concluded that the possibility of an accident different from any previously analyzed in the SAR will not be created.
ARKANSAS NUCLEAR ONE FORM TITLE: FORM NO. REv. 10CFR50.58 EVALUATION ; 1000.131B 2 Page 3 of 3 PC 973673Pzoy ;
- 6. PAGE /0 REV O Will the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR be created?
Yes O No @ All system desions for eautoment important to safety will remain the same. None of the tubina added structure seismic by this /desian will component. be routed in such a manner to cause procaoation of a failur i The criteria for physical separation has been maintained and conservative adherence to seismic requirements has been observed to insure compliance. Also, this modification does not electrically / mechanically interface with eauloment import Therefore, the possibliity of a malfunction of eauipment imponant to safety that invo initiator or failure of a different tvoe than any previousiv evaluated in the SAR has no . 7. Will the marDin of safety as defined in the bases for any technical specificatlor' be reduced? s Yes O No @ The Technical Specifications bases do not establish a maroin of safety for Service Wa Strainers,2F6A. 2F6B or 2F6C. or the existina strainer DP indications. The installation of mechanical strainer DP indicators will not affect or alter the existino Tech. Spec. reautreme be included in any new reauirements. Based on the above statements, tais modification reduce the maroin of safety as defined in the bases, d w
- Roaer B. Rucker 7/22/97 Cer6fied Reviewers Signature Printed Name Date Reviewers certification expiration date: 9/t3/97 Assistance provided by:
Printed Name Scope of Assistance N/A Date PSC review by: - Date:%
l AnKANSAS NUCLEAR ONE FORM TITLE: l
- FORM NO. REV.
10CFR50.59 DETERMINATION 1000.131 A 3 PC-1 Document No. 973744P201 Rev/ Change No. O 1 Title Addition of undervoltage trip bypass to trip circuit breakers.
- h.
- description of proposed change:
See Continuation page. PC 9737 44 P 201 Will the proposed Activity: l PA0E 5 RE V 0 j
- 1. Require a change to the Operating License including:
Technical Specifications (excluding the bases)? YesO No@ Operating License? YesO No@ Confirmatory Orders? YesO No@ i 2. Result in information in the following SAR documents (including drawings and text) being l (a) no longer true or accurate, or (b) violate a requirement stated in the do'cument: SAR (multi-volume set for each unit)? Yes@ NoO Core Operating Limits Report? YesO No@ Fire Hazards Analysis? YesO No@ Bases of the Technical Specifications? YesO No@ Technical Requirements Manual? YesO No@ NRC Safety Evaluation Reports? YesO NoS
- 3. Involve a test or experiment not described in the SAR?
(See Attachment 2 for guidance) YesO No@ ,
- 4. Result in a potentialimpact to the environment? (Complete Environmental impact Determination of this form.)
YesO No@
- 5. Result in the need for a Radiological Safety Evaluation per section 6.1.5?
YesO No@
- 6. Result in any potential impact to the equipment or facilities utilized for Ventilated Storage Cask activities per Section 6.1.67 YesO No@
. 7. Involve a change under 10CFR50.54 for the following SAR documents per Section 6.1.77 QAMO7 YesO No@ E-Plan? YesO No@
ARKANSAS NUCLEAR ONE FORM TITLE: FORM NO. REV. 10CFR50.59 DETERMINATION 1000.131 A 3 PC-1 Document No. 973744P201 Rev./ Change No. O Basis for Determination (Questions 1,2, & 3): See continuation page. O Proposed chango does not require 10CFR50.59 Evaluation per Attachment 1, item (if # checked, note appropriate item #, send LDCR to Licensing). Search Scope: List sections reviewed in the Licensing Basis Documents specified in questions 1,2 and 3. If a keyword was done on LRS, "all" may be entered under "Section" with the keyword (s) used in parentheses. Control copies of the documents shall be reviewed (LRS is not verified and searches only text, not figures Attach and distribute a completed LOCR per Section 6.1,2 If LSD changes are required. Document Section LRS: ANO-2 Tech. Specs ALL (" Reactor Tnp Breakers *
- Shunt Trip" " Under Voltage Bypass" ANO-2 OP. License "TCB" " Reactor Trip Switchgear", "Undervoltage coil", " Shunt coil")
ALL ( As above } ANO-2 Conf. Orders ANO-2 SAR ALL ( As above) ANO-2 TS Bases ALL ( As above) ANO 2 NRC SER, QA Manual ALL ( As above ) PC 97374P201 ALL ( As above ) PAGE G RE V 0 MANUAL SECTIONS: ANO 2 Tech. Specs
. Section 3/4.3.1, Sections 4.3.1.1.1,4.3.1.1.2 and 4.3.1.1.3, Table 3.3-1
( and 4.3-1 ANO-2 Tech Specs Bases Sections 3/4.3.1 and 3/4.3.2 ANO-2 SAR Sections 7.2.1.1.9.6, 7.2.1.1.9.7 and 7.2.1.1.9.8, Table 7.2 5 FIGURES: None M . b ibI-4 . Nick Mehta Certified Reviewers Signature 08 25-98 Printed Name Date Reviewers certification expiration date: 05-05-99 Assistance provided by: Printed Name Scope of Assistance Date Search Scope Review Acceptability (NA, if performed by Technical Reviewer per 1000.006) N' _]/OE'L/A/ SHEMORAl Certified Reviewers Signature 9/0Y/97 Printed Name ' Date
ARKANSAS NUCt. EAR ONE FORM TITLE: FORM NO. REV. i 10CFR50.59 DETERMINATION 1000.131 A 3 ENVIRONMENTAL IMPACT DETERMINATION (UNIT 1 and UNIT 2) t
' Document No. 973744P201 RevlChange No. O Complete the following Determination. If the answer to any item below is "Yes", an Environmental Evaluat required. See Section 6.1.4 for additional guidance.
Will the Activity being evaluated: Y.11 Hg O E Disturb land that is beyond that initially dis *urbed during construction (i.e., new construction of buildings, creation or removal of ponds, or other terrestrial impact)? Sea Unit 2 SAR Figur 2.5-17. This applies only to areas outside the protected area. O E increase thermal discharges to lake or atmosphere? O E increase concentration of chemicals to cooling lake or atmosphere through discharge canal o tower? O E increase quantity of chemicals to cooling lake or atmosphere through discharge canal or tower? O E Modify the design or operation of cooling tower which will change drift characteristics? O E Install any new transmission lines leading offsite? O E Change the design or operation of the intake or discharge structures? O E Discharges any che' micals new or different from that previously discharged? O E Potentially cause a spill or unevaluated discharge which may effect neighboring soils, surface water or ground water? O E involve burying or placement of any solid wastes in the site area which may effect runoff, surface water or ground water? I O E involve incineration or disposal of any potentially hazardous materials on the ANO site? O E Result in a change to nonradiological effluents or licensed reactor power level? O E Potentially change the type or increase the amount of non-radiological air e. missions from the ANO site. l PC 973744P 201 PAGE 1 RE V 0 L ,
ARKANSAS NUCLEAR ONE FORM TITLE: FORM NO. REV. 10CFRSO.89 REVIEW CONTINUATION PAGE . 1000.131C 3 Page 1 of 1 Document No. PC 973744P201 Rev/ Change No. 0 1 1 1pCFR50.5p Review Continuation Pane ' This modification will revise the method of independent test verification of the shunt trip coil in the Reac Switchgear circuit. Presently a spring retum to center switch (HS/ TEST) is used to bypass the shunt srip in th
' shunt bypass' position for testing of the UV trip. Placing this switch in the ' shunt trip" position immecately energizes the shunt trip relay and does not test the PPS matrix test and control room manual trip buttor.~. '
This modification will revise the testing of the shunt trip relay by replacing the ' shunt trip" position of "V,S/ T with the new "UV bypass" position. The new 'UV Bypass" position will allow for independent verification of the shunt trip feature using the PPS matrix test and control room manual trip buttons. This modification will maintai the existing ' shunt bypass" feature to allow the independent verification of the undervoltage trip feature from both the PPS matrix test and control room manual trip buttons. Step 1 Bases of Determination:
)
Question 1. NO. l' No change is required to the Unit-2 Technical Specifications. The Technical Specification discusses the reacto protective instrumentation and reactor trip system response time of each reactor trip including operability and surveillance requirements. However, The Technical Specification does not explicitly require testing the trip devices independently and does not provide the level of detail to address this modification. Question 2. Yes. l This modification will have impact on Unit-2 SAR. Section 7.2.1.1.9.8 discusses the method of independen test verification of the shunt and UV trip coils. This section explains about placing the handswitch in the " shunt trip' position to test the shunt trip capability. The verbiage will be revised to reflect the new 'UV bypass" po for testing the shunt trip relay. See revised section 7.2.1.1.9.8 attached with the LDCR Table 7.2-5 will be revised to add failure modes analysis due to added handswitch contacts in the undervoltage and shunt coil circuit. This modification will not result in revision being necessary for the Unit-2 NRC Safety Evaluation Repor QAM, E PLAN, and bases for Technical Specifications. None of these documents provide sufficient detail to address this modification. Question 3. No. No new tests or experiments not described in the SAR are involved by this modification. PC 973744P201 PAGE 8 RE V 0 l
AMAAN#Mb NU6LEAM UNC FORM TITLE: FORM NO. REU. 10CFRSO.89 EVALUATION 1000.131B 3 PC-2 PC 973744P201 Page 1 of 2 PA6E 9 RE V 0 10CFR50.59 Eval. No. M W0 t bM (Assigned by PSC) Document No. PC 973744P201 RevlChange No. O Title Addition of undervoltane trio bvoass to trio circuit breakers. A WRITTEN RESPONSE PROVIDING THE BASIS FOR THE ANSWER TO EACH QUESTION MUST BE ATTACHED. EACH QUESTION MUST BE ANSWERED SEPARATELY. A SIMPLE STATEMENT OF CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES GUIDANCE FOR RESPONSE. If the answer to any question on this form is "Yes," then an unreviewed safety question is involved. If the answer to all questions is "No," then the proposed change does not involve an unreviewed safety question.
- 1. Will the probability of an accident previously evaluated in the SAR De increased?
Yes O No @ This modification will not increase the probability of an accident previously evaluated in the SAR. The SAR accident analysis chapter 15 (sections 15.1.0.4.1 and 15.1.0.4.2) covers the reactor protective system trips and engineering safety features response times. However, this modification will not impact the response time of each shunt and undervoltage relays. This modification does not change any PPS logic or its design basis and the trip system cannot cause an accident. The only change is to add handswitch contacts to the cimuit to allow the independent testing of the shunt trip coll. This modification still maintains the system reliability and will allow the independent verification of shunt and undervoltage trips from both the PPS matrix test and the control room trip buttons.
- 2. Will the consequences of an accident previously evaluated in the SAR be increased?
Yes O No @ This modification will not increase the consequences of an accident previously evaluated in the SAR. Rather, it will improve inservice testability. The TCBs surveillance procedures require the use of handheld jumpers and lifting leads in the PPS or TCBs to verify independent testing of UV and shunt trip coils. IEEE 279-1971 indicates that that jumpers and lifted leads are not the preferred method of testing for protective system reliability. This modification will meet the above requirement and improve the method of surveillance testing. As describe above this modification will not impact the response time of each reactor trip circuit evaluated in the chapter 15 of the SAR. The modified trip circuit will not create a new accident. Therefore, it will not create a new pathway for release of radioactive material, nor it will affect any barriers which mitigate dose to the public or cause an increase of the onsite doses. 3. Wi;l the probability of a malfur:ction of equipment important to safety be increased? Yes O No @ This modification will not increase the probability of a malfunction of equipment important to safety. This modification does not change the design basis of the Reactor Trip Switchgear circuit. This modification will improve the method of testing by adding a undervoltage bypass function to independently verify the shunt trip function. The replacement handswitch is similar to the existing handswitch and the new handswitch will be procured under Class IE requirements. The undervoltage bypass contact will be used to test the shunt coil and if this contact does not reopen, the undervoltage coil may not de-energize for a plant trip. To ensure that the contact does open after shunt coil testing, the testing of the undervoltage coil will be performed last to ensure that the undervoltage bypass contact has in fact reopened. The probability of the handswitch contact falling after the shunt coil testing is highly remote. Therefore, this will not introduce any new failure modes which will impact on equipment important to safety.
ARKANSAS NUCLEAR ONE FORM TITLE: FORM NO. REU. toCFRSS.s0 EVALUADoN j 1000.131a 3 Pc-2 -
- 4. \
Will the consequences of a malfunction of equipment important to safety be increased? Yes O No @ 1 This modification will not increase the consequences of a malfunction of equipment important to safet This modification will affect only the test circuit and will not impact normal operation of the breakej will be performed in a sequence which ensures the test contacts are reset to normal upon completion of test. The test circuit has no function related to mitigating the consequences of a malfunction of equipm importaat to safety. The malfunction of the reactor trip switchgear circuit has already been evaluated in th Unit-2 SAR and found acceptable. This modification does not alter the function of the circuit, but on handswitch contacts to bypass the trip signal during testing. The modification of the breaker test circuit wi;{ not result in increased potential offsite doses from a malfunction of equipment important to safety. 5. Will the possibility of an accident of a different type than any previously evaluated in the SAR be created? Yes O No @ { As stated above, the SAR accident analysis chapter 15 covers the reactor protective systems trips a engineering safety features response times. However, this modification will not impact on response time each reactor trip. This modification will decrease the possibility to create an inadvertent trip by re the requirements of using handheld jumpens and lifted leads during surveillance testing of trip circuit. This trip system cannot cause an accident. Therefore, this modification will not create the possibi accident of a different type than any previously evaluated in the SAR. 6. Will the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR be created? Yes O No @ The proposed modification will not create the possibility of a malfunction of equipment importa of a different type than any previously evaluated in the SAR. This proposed change involves the method; independent testing of shunt trip and undervoltage coil. This modification will add the undervolt function to test the existing shunt trip coil. The failure of undervoltage bypass contact is being addres SAR table 7.2 5. The existing table 7.2 5 has already evaluated the undervoltage trip coil failing to energize (#108, manual pushbutton). This modifications failure mode would be the bypass conta open. The same scenario applies for the bypass contact as for the manual pushbutton This modification package adds this similar scenario to the undervoltage coil failure (#109). This modification will decrease the possibility to create an inadvertent trip by removing the requirements of using handheld jumpers lifted leads during surveillance testing of trip circuit. This modification will not impact on the SAR accide described in chapter 15, section 15.1.0.4.1. 7. Will the margin of safety as defined in the basis for any technical PC 973744P 201 specification be reduced? ggE Io RE M o Nog This modification does not impact on response time of each reactor trip. No margins were found in the bases for the technical specifications to address this modification. Therefore, there is no change in th margin of safety.
- *" I Nick Mehta 08/31/98 Oertified Reviewers Signature I Printed Name Date l
Reviewers certification expiration date: 05-05-99 Assistance provided by: Printed Name Scope of Assistance Date PSC review by: Date: 0 % N
FORM TITLE: FORM NO. REv.
, 10CFR50.89 DETERMINATION 1000.131A 3 Pc 1 PAGE 3 REV.O Page1of,g Document No. NCP 973786N201 Rev1 Change No. A Title Unit 2 Moisture Seoarator Reheater Reolacement Brief description of proposed change: Replace the HP and LP tube bundles in the U-2 MSR's.
Will the proposed Activity:
- 1. Require a change to the Operating Ucense including:
. Technical Specifications (excluding the bases)?
YesO No@ Operating Ucense? YesO No@ Confirmatory Orders? YesO No@
- 2. Result in information in the following SAR documents (including drawings and text) being (a) no longer true or accurate, or (b) violate a requirement stated in the document:
SAR (multi-volume set for each unit)? Yes@ nod Core Operating Umits Report YesO No@ Fire Hazards Analysis? YesO No@ Bases of the Technical Specifications? YesO No@ Technical Requirements Manual? YesO No@ NRC Safety Evaluation Reports? YesO NoS
- 3. Involve a test or experiment not described in the SAR?
YesO No@ (See Attachment 2 for guidance)
- 4. Result in a potential impact to the environment? (Complete the Environmental impact Determination of this form.) YesO No@
- 5. Result in the need for a Radiological Safety Evaluation per section 6.1.57 Yes@ nod
- 6. Result in any potential impact to the equipment or facilities utilized for Ventilated Storage Cask activities per Section 6.1.67 YesO No@
- 7. Involve a change under 10CFR50.54 for the following SAR documents per Section 6.1.7:
QAMO? YesO No@ E-Plan? YesO No@ Basis for Determination (Questions 1,2 & 31: See attached continuation O Proposed change does not require 10 CFR 50.59 Evaluation per Attachment 1. Item # _, (if checked, note appropriate item #, send LDCR to Ucensing). t
ARKANSAS NUCLEAR ONE FORM TITLE: FORM NO. REV. 10CFR50.88 DETERMINATION 1000.131A 3 PC.1 PAGE Y - REV.O Page 2 of,g Document No. NCP 973786N201 RevlChange No. .Q Search Scope: List sections reviewed in the Licensing Basis Documents specified in Question 1,2 and 3. If was done on LRS, "all" may be entered under "Section" with the keyword (s) used in parenthes Attach and distribute a completed LDCR per Section 6.1.2 if LB Document Section LRS: All Documents contained under LRS 50.59 - Unit 2 All (Scav*, Copper, MSR, 2E12*, Separator, RSLLV, RSHLV, High Load, Low Load, 2FO*, 2EBD*, 2GBD*, M*2203, Excess Steam, Turbine Building. MANUAL SECTIONS: Unit 2 SAR, Section 10 Unit 2 SAR, Section 1.2.2.5 FIGURES: Unit 2 SAR Figure 10.2-2 Unit 2 SAR Figure 10.2-5 (Sheet 1) Unit 2 SAR Figure 7.3-9
!1// Jack Orlicek
/ Certified Reviewers Signature 8/29/98 Printed Name Date Reviewers certification expiration date: 11/28/99 Assistance provided by: Printed Name Scope of Assistance Date Sea Scope Review Acceptabliity (NA, if performed by Technical Review per 1000.006) z# / Douatas Edoell Certified &evie#s Signature 9/26/98 Printed Name Date i i l i l l
FORM TITLE: FORM NO. REV. 10CFR50.88 DETERMINATION 1900.131A 3 PCo1 Page } of_g ENVIRONMENTAL IMPACT DETERMINATION (UNIT 1 and UNIT 2) Document No. 973786N201 Rov 1 Change No. A PAGE S REV o Complete the following Determination. If the answer to any checklist item is "Yes", an Environmental Evaluation is required. See Section 6.1.4 for additional guidance. Will the Activity being evaluated: Yes No O E Disturb land that is beyond that initially disturbed during construction (i.e., new construction of buildings, creation or removal of ponds, or other terrestrial impact)? See Unit 2 SAR Figure 2.5-17. This applies only to areas outside the protected area. O E increase thermal discharges to lake or atmosphere? O E increase concentration of chemica:s to cooling lake or atmosphere through discharge canal or tower? O E Increase quantity of chemicals to cooling lake or atmosphere through discharge canal or tower? O E Modify the design or operation of cooling tower which will change drift characteristics? l j O E install any new transmission lines leading offsite? O E Change the design or operation of the intake or discharge structures? O E Discharges any chemicals new or different from that previously discharged? O E Potentially cause a sp!ll or unevaluated discharge which may effect neighboring soils, surface water or ground water? O E Involve burying or placement of any solid wastes in the site area which may effect runoff, surface water or ground water? O E Involve incineration or disposal of any potentially hazardous materials on the ANO site? O @ Result in a change to nonradiological effluents or licensed reactor power level? O E Potentially change the type or increase the amount of non-radiological air emissions from the ANO site. l L ..
ARKANSAS NUCLEAR ONE FORM TITLE: FORM No. REV. 10CF!160.58 REVIEW CONTINUATION PAGE 1000.131C 3 Page 4 of g Document No. 973786N201 Rev/ Change No. 2 PAGE d REV.O 10CFR50.59 Review Continuation Pace NCP 973786N201 will replace the existing MSFi tube bundles containing 90-10 CuNi tubes with a tube material that is more compatible with the replacement ANO-2 steam generators. This modification will also optimize the efficiency of the replacement tube bundles for the planned 107.5% power uprate of ANO-2. contract / purchase order (NAS 00687/973375) was awarded to Thermal Engineering Intemational, Inc. 1 (fonnerly Senior Engineering) to provide the design, fabrication and delivery of replacement tube bund ANO-2 according to ANO Specification M-2559 Technical Specification for Replacemerf. Tube Bundlesj 2 Moisture Separator Reheaters. This design, change installs shop fabricated replacement tube bundle constructed with tubes made of SA-268, Grade TP439 (UNS Designation S43035) ferritic stainless steel material and designed for 107.5% power uprate conditions. The replacement tube bundle design provided by TEl include several design enhancements based o experience with tube bundle replacements and the availability of proprietary OEM information ANO-2 MSR vessel and tube bundle configuration. These design enhancements include: a chan material, a reduction in excess (scavenging steam) requirements, increased tube surface area and a ve tube bundle orientation. The lower heat transfer coefficient of Type 439 Stainless Steel is offse , increased surface area and vertical tube orientation of the tube bundles. The scope of configuration changes includes the following: (see referenced ER's for further detail) 1. Replacement of the existing high pressure (HP) and low pressure (LP) tube bundles for both MSR's1 B (including piping intemal to the MSR vessel) { 2. Deletion of four (4) flow orifices to be replaced with four (4) valves for scavenging steam control I ER9737861205.
- 3. {
Adjustment ER9737861205. ofinlet steam valve controllogic to the HP bundle for the new tube bundle design per 4. Installation of additionallocalinstrumentation to monitor temperatures and pressures of the HP and LP bundles per ER9737861205. Basis for Determination
- 1. This 50.59 determination addresses the MSR tube bundle replacement as outlined in NCP 973786 modification will not impact the Technical Specifications, operating License, or Confirmatory Orders.
- 2. SAR Figure 10.2-5 sheet 1, Piping and Instrumentation Drawing Reheat Steam and SAR Figure require revision. No other LBD information will be caused to be untrue orinaccurate.
- 3. All MSR related systems will function the same as prior to the modification. NCP 973786N201 w any new or revised test or experiment that will put the plant in a different operating mode than cove current plant procedures. The resulting feedwater temperature as shown in the heat balance prepa system conditions after replacement tube bundle installation has been reviewed by NED. This tem has been confirmed to be within the bounds of the current accident analysis.
- 4. Radiological Safety Evaluation attached.
l
n.n.~...-..~ FORM TITLE: FORM NO. REV.
, 10CFR50.59 EVALUATION 1000.1313 3 Pc-2 Page5of.s 10CFR50.59 Eval. No NIYN N l (Assigned by PSC)
Document No. 973786N201 RevlChange No. g PAGE I 8tEV. O Title ANO-2 MSR Tube Bundle Replacement A WRITTEN RESPONSE PROVIDlNG THE BASIS FOR THE ANSWER TO EACH QUESTION MU ATTACHED. EACH QUESTION MUST BE ANSWERED SFPARATELY. A SIMPLE STATEMENT OF CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES GUIDANCE FOR RESPONSE If the answer to any question on this form is "Yes," then an unreviewed safety question is involved. If the answer to all questions is "No," then the proposed change does not involve an unreviewed safety question.
- 1. Will the probability of an accident previously evaluated in the SAR be increased?
Yes O No E This m'odification replaces the existing 90-10 Cu-Ni tube bundles with shop fabricated tube bundles with Type 439 ferritic stainless steel tubes. The replacement tube material is more compatible with the replacement steam generators. The replacement MSR tube bundles are functionally equivalent to the original design. This modification does not change the function or failure modes of any components, systems, or structures. All previously evaluated accidents listed in SAR section 15 were reviewed for applicability. No additional failure mechanisms not considered by the safety analyses are introduced by this modification. This modification will not increase the probability of previously evaluated accidents
- 2. Will the consequences of an accident previously evaluated in the SAR be increased?
Yes O No E This modification will not affect the ability to mitigate the consequences of any previously evaluated accident. This modification does not change the function or failure modes of any components, systems, or structures. This modification is not directly associated with any systems or components related to accident initiation or mitigation as described in chapter 15 of the SAR. There will be no affect on the dose consequences of any previously evaluated accident in the SAR. The dose consequences and assumptions used in the SAR will not be changed. Therefore, there is no change to the consequences of any previo evaluated accident in the SAR as a result of this modification. 3. Will the probability of a malfunction of equipment important to safety be increased? Yes O No E The MSR tube bundles are not safety related and their replacement does not effect safety-related components or systems. This modification does not change the functinn or failure modes of any components, systems, or structures. The replacement tube bundles will improve plant efficiency, eliminate copper in the secondary system in preparation for steam generator replacement, and be designed for the future 107.5% power uprate. This modification will not increase the probability of a malfunction of equipment important to safety.
ARKANSAS NUCLEAR ONE FORM TITLE: FORM NO. REV. 10CFR50 SS EVALUATION 1000.1318 3 Pc-2 Page 6 of 6 PAGl! f "J!V. 0 4. Will the consequences of a malfunction of equipment important to safety be increased? Yes O No @ The replacement components of the MSR tube bundles is not safety related and does not effect safety or systems. This modification does not change the function or failure modes of any components, systems, or structures. It does not change, degrade, or prevent any action described ir. a accident discussed in the SAR. The consequences of a malfunction of equipment importa not be increased as a result of this modification. I 5. Will the possibility of an accident of a different type than any previously evaluated in the SAR be created? Yes O No E This modification does not change the function or failure modes of any components, systems, It will not create any new failure or accident condition. The possibility of an accident of a d any previously evaluated in the SAR will not be created. 6. Will the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR be created? Yes O No @ This modification involves replacement of equipment in the secondary system that is not safe has no impact on equipment important to safety, it does not change the function or failu components, systems, or structures. The possibility of a malfunction of equipment important to s different type than any previously evaluated in the SAR will not be created as a result of this modi l 7. Will the margin of safety as defined in the basis for any technical specification be reduced? Yes O No S There are no technical specification limits or basis associated with the moisture separator / reheat bundles. This modification does not affect the margin of safety as defined in the bases for any Specification, bases of the LCO's or surveillance requirements.
/
Jack Orlicek 8/29/98 fertified Reviewers Signature Printed Name Date Reviewers certification expiration date: 11/28/99 Assistance provided by: Printed Name Scope of Assistance Date t PSC review by: Date: \\ % D ER 973788N201
[' ARKANSAS NUCLEAR ONE FORM TITLE: FORM NO. REtf. 10CFR60.49 DETERMINATION 1000.131A 3 PC.1 Page / of f Docurr.ent No. NC 973905N201 RevlChange No. 9 Title REMOVE MAIN TURBINE SETBACK CIRCUIT Brief description of proposed change: This NC will permanently remove the turbine setback circuit. Will the proposed Activity: NC 973 905N201
- 1. Require a change to the Operating License including:
PAGE 5 RE V 0 Technical Specifications (excluding the bases)? YesO No@ Operating License? YesO No@ Confirmatory Orders? YesO NoS
- 2. Result in information in the following SAR documents (including drawings and text) being (a) no longer true or accurate, or (b) violate a requirement stated in the document:
SAR (multi-volume set for each unit)? Yes@ nod Core Operating Limits Report YesO No@ Fire Hazards Analysis? YesO No@ Bases of the Technical Specifications? YesO No@ Technical Requirements Manual? YesO No@ NRC Safety Evaluation Reports?' YesO No@
- 3. Involve a test or experiment not described in the SAR7 (See Attachment 2 for guidance)
YesO No@
- 4. Result in a potential impact to the environment? (Complete the Environmentalimpact Determination of this form.)
YesO No@
- 5. Result in the need for a Radiological Safety Evaluation per section 6.1.5?
YesO No@
- 6. Result in any potential impact to the equipment or facilities utilized for Ventilated Storage Cask activities per Section 6.1.67 YesO No@
- 7. Involve a change under 10CFR50.54 for the following SAR documents per Section 6.1.7:
QAMO? YesO No@ E-Plan? YesO No@ I
ARKANSAS NUCLEAR ONE FORM TITLE: FORM NO. REV. 10CFR60.69 DETERMINATION 1000.131A 3 PC-1,2 49'fs A8 Page_Z.Of'f Document No. W973905P201 RevlChange No. 9 Basis for Determination (QL==*!ons 1. 2 & 31: NC 973 905 N201 ' P A0E (p RE V 0 l l 0 Proposed change does not require 10 CFR 50.59 Evaluation per Attachment 1, item. (if #-checked, note appropriate item #, send LDCR to Licensing). Search Scope: List sections reviewed in the Licensing Basis Documents specified in Question 1,2 and 3. If a search was performed on LRS, the LRS search index should be entered under "Section" with the search statement parentheses. Controlled hard copies of the documents shall be reviewed (LRS is not verified and s text, not figures or drawings). Attach and distribute a completed LDCR per Section 6.1.2 If LSD change required. Document Section LRS: , ALL (Turbine, Setback, Governor Valves, Main Feedwater Pump, Main Turtaine,80%,70%, Runback, Turbine Load Limit) MANUAL SECTIONS: ANO 2 SAR 10.2,10.3,10.4.7,15.1.8,15.1.10 ANO-2 TECH SPECS 3/4.7.1.1, 3/4.7.1.5, 3/4.3 FIGURES: ANO-2 S R Figure 7.3-9 (dwgs M-2402 shts 3&4) Certified Reviewers (fignature STEVE CAPEHART Printed Name Y ~I7'i[ ] Date Reviewers certification expiration date: 5/9/99 Assistance provided by: Printed Name Scope of Assistance Date Search Scope Review Ac ptability (NA, if performed by Technical Review per 1000.006) $ tliw2 M uedAl ertified Reviewers Signature
\liues &J sinkt Printed Name ' ' Date 3 1
i
AMnANSAS NUCLEAR ONE FORM NO. REV. 10CFR60.59 DETERMINATION 1000.131A 3 Page1of Y ENVIRONMENTAL IMPACT DETERMINATION (UNIT 1 and UNIT 2) NC U3 905 001 Document No. NC 97390SN201 Rev1 Change No. 9 PAGE 7 RE V 0 Complete the following Determination. If the answer to any checklist item is "Yes", an Environmen is required. See Section 6.1.4 for additional guidance. Will the Activity being evaluated: Yes No O E Disturb land that is beyond that initially disturbed during construction (i.e., new construction o buildings, creation or removal of ponds, or other terrestrialimpact)? See Unit 2 SAR Figu 2.5-17. This applies only to areas outside the protected area. O E increase thermal discharges to lake or atmosphere? C E Increase concentration of chemicals to cooling lake or atmosphere through discharge canal! l tower? O E increase quantity of chemicals to cooling lake or atmosphere through discharge canal or tower? O E Modify the design or operation of cooling tower which will change drift characteristics? O E install any new transmission lines leading offsite? O E Change the design or operation of the intake or discharge structures? i O E Discharges any chemicals new or different from that previously discharged? l 0 2 Potentially cause a spill or unevaluated discharge which may effect neighboring soils, surface l water or ground water? O E involve burying or placement of any solid wastes in the site area which may effect runoff, surface water or ground water? O E Involve incineration or disposal of any potentially hazardous materials on the ANO site? O E Result in a change to nonradiological effluents or licensed reactor power level? O E Potentially change the type or increase the amount of non radiological air emissions from the ANO site.
ARKANSAS NUCLEAR ONE FORM TITLE: FORM NO. REV. 10cFR60.89 REVIEW CONTINUATION PA?E 1000.131c 3 PageIof k Document No. NC 973905N201 RevlChange No. 9 NC 973 905 N201 10CFR50.59 Review Continuation Paae PAGE 8 RE V 0 This NC will permanently remove the ' Main Turbine Setback to 80% on Loss of One Main Feedwater Pump' circuit. The circuit is currently removed by T-Alt 97-2-004 which was implemented in March 1997. The Main Turbine Load Setback circuit is designed to set the turbine load back to 80% (800 Mw equivalent in the EHC system) upon the loss of one Main Feedwater (MFW) pump. The Setback instantaneously reduces the Control Valve position to the 80% load equivalent position. The apparent purpose of the circuit is to reduce the steam demand to the main turbine which should reduce the bolloff rate in the steam generators. The reduced boiloff rate will assist in keeping the steam generator levels above the low level trip setpoint of 23% SG Narrow Range level thereby preventing a reactor trip. The end result is that the unit feedwater demand is within the flow capacity of one MFW pump. The circuit is currently removed by T-Alt 97-2-004 which was implemented in March 1997. The removal of the automatic action to reduce plant power with this circuit will require operator action to reduce plant power in order to prevent a reactor trip on low steam generator level. This action is found in AOP 2203.027 Rev. 4, Loss of Main Feedwater Pump, floating step 12 which instructs the operator to ' Verify Turbine Setback 720 to 760 MWe'. QUESTION 1 - Operatino License t The Main Turbine Load Setback circuit is not discussed in the level of detail present in the ANO-2 Technical Specifications, Operating License or any Confirmatory Orders. QUESTION 2 - SAR Documents The Main Turbine Setback circuit is shown on Logic Diagrams M-2402 shts 3 and 4. These drawings correspond to SAR Figure 7.3-9. QUESTION 3 - Test or Experiment The modifications made by this NC do not require any post-modification testing.
)
QUESTION 4 - Environmental imoact The modifications made by this NC do not require an Environmental Impact Evaluation per the Environment impact Checklist. QUESTION 5 - Radiolooical Safetv Evaluation The work performed by this NC does not involve the processing of radioactive material and will not affect monitored ventilation or drainage pathways. QUESTION 6 - Ventilated Storace Cask The components associated with the Main Turbine Setback circuit do not interface with the systems associated with the VSC project. QUESTION 7 - QAMO or E-PLAN The Main Tuitine Setback circuit is not referenced or discussed in the QAMO or E PLAN
E } ArmANTA5 NUGLt:AH ONE FORM TITLE: 1 FORM NO. REV. 10CFR60.59 EVALUATION 1000,131B 3 PC-2 i Page ] off NC 973 905 N201 PAGE p ; RE V 0 10CFR50.59 Eval. No. Ff& 98- 427 (Assigned by PSC) Document No. NC 973905N201 RevlChange No. 9 Title REMOVE MAIN TURBINE SETBACK CIRCUIT A WRITTEN RESPONSE PROVIDING THE BASIS FOR THE ANSWER TO EACH QUES ATTACHED. EACH QUESTION MUST BE ANSWERED SEPARATELY. A SIMPLE STATEME CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES GUIDANCE FOR RESPO If the answer to any question on this form is "Yes," then an unreviewed safety question is involved. If the answe to all questions is "No," then the proposed change does not involve an unreviewed safety question.
- 1. Will the probability of an accident previously evaluated in the SAR be increased?
Yes O No @ There is no discussion of the setback circuit contained or credited in the accident analysis. The setback is not an accident initiator and does not relate to causing an accident. The function of interest, i.e. turbine setback, is intended to prevent a reactor trip upon loss of a MFW pump vs. mitigating an accident. This function will be performed by operator action. Therefore, the probability of an accident previously evaluated in the LBD is not increased.
- 2. Will the consequences of an accident previously evaluated in the SAR be increased?
Yes O No @ The setback circuit does not place any additional demands on any equipment important to mitigation of any accidents. The setback circuit is not an accident initiator and does not relate to causing an accident. Therefore removal of the setback circuit does not increase the dose consequences of any of the accidents previously cvaluated in the SAR. 3. Will the probability of a malfunction of equipment important to safety be increased? Yes O No @ The setback circuit affects the operation of the main turbine govemor valves which are non-Q. The setback circuit function for the main turbine govemor valves is not considered an effect on equipment important The setback circuit does not interface electrically or mechanically with equipment important to safety. Tn it is concluded the removal of the setback function will not increase the probability of a malfunction of equip important to safety. 4. Will the consequences of a malfunction of equipment important to safety l be increased? Yes O No @ The setback circuit function for the main turbine govemor valves is not considered an effect on equipment important to safety. Given the setback circuit does not affect the ability of any equipment important to safet mitigate any accident, the dose consequences associated with malfunctions of equipment important to safety not increased. i
ARKANSAS NUCLEAR ONE FORM TITLE: FORM NO. ret /. 10CFR50.89 EVALUATION 1000.131B 3 PC-2 NC 973 905 N201 PAGE /0 RE V 0 5. Will the possibility of an accident of a different type than any previously evaluated in the SAR be created? Yes O No @ The function of the equipment being removed from service by this NC is not required for shutdown of Ii mitigating radioactive releases or maintaining reactor coolant pressure boundary integrity. The function of the setback circuit to automatically reduce plant power upon loss of a MFW pump s being replaced with ope action. The setback circuit function is not considered an accident initiator. Therefore, the possibility of an accident of a different type than any previously evaluated in the SAR is not created. 6. Will the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR be created? Yes O No @ The setback circuit does not interface with or affect any equipment considered important to safety. The removal a different of the setback circuit does not create the possibility of a malfunction of equipment important to type. 7. Will the margin of safety as defined in the basis for any technical specification be reduced? Yes O No @ The setback circuit is not referenced in the ANO-2 Tech Specs. Therefore, there are no Tech Spec bas define margins of safety. , l Certified' Reviewers S(gnature STEVE CAPEHART Printed Name f'M ~ Date i Reviewers certification expiration date: 6/9/99 Assistance provided by: Printed Name Scope of Assistance Date PSC review by: '\1% Date: S D
~~~,.- nu m nune ~
FO'.T.M TITLE: l FORM NO. REV. 10CFR50.88 DETERMINATION i 1000.131A 2 PC-2,3 \ Page 1 of.4 Document No. g73932P202 Rev) Change No. 9 Title PAGE 1 REV.O ' Relocation of Outaae Control Center (OCCl to CA-2 For Unit 2 Outane Will the proposed Activity: l j l 1. Require a change to the Operating License including: l j l Technical Specifications (excluding the bases)? YesO No@ Operating License? I YesO No@ l Confirmatory Orders? l YesO Nc? $ l 2. ! Result in information in the following SAR documents (including drawings and text) being (a) no longer true or accurate, or (b) violate a requirement stated in the document: Core Operating Limits Report YesO NoS SAR (multi-volume set for each unit)? Yes@ nod QAMO?* ' YesO No@ E Plan?* YesO Nc@ FHA Yes@ nod Bases of the Technical Specifications? YesO No@ NRC Safety Evaluation Reports? YesO No@ 3. Involve a test or experiment not described in the SAR? YesO NoS 4. Result in a potential impact to the environment? (Complete Environmental impact Checklist of this form.) YesO No@ 5. Result in the need for a Radiological Safety Evaluation per section 6.2.4.A? YesO No@ 6. Result in any potential impact to the equipment or facilities utilized for Ventilated Storage Cask activities per Section 6.2.4.87 YesO No@ Basis for Determination: _See Attached Continuation Sheet 4
- Changes to these documents require an evaluation in accordance with 100FR50.54.
See Section 6.2.1.B. t
___ - - ------ ----------------------~-~ FORM TrrLE. ARKANSAS NUCLEAR ONE FORM NO. GElf. 10CFR50.50 DETERMINATION 1000.131A 2 PC 2,3 Page 2 of.4 Document No. 973932P202 RevlChange No. 0,
References:
List sections reviewed in the Licensing Basis Documents, specified . If a in questio in parentheses. Controlled hard copies of the docum searches completed such LDCR aschanges if LBD LRS are arerequired. not controlled and search text only, not figures or d Document Section ANO-2 SAR FHA All* (Fire protection, Fire Barriers, Penetrations)
,ANO 2 Tech Specs All* (Fire protection, Fire Barriers, Penetrations)
ANO E-Plan All* (Fire protection, Fire Barriers, Penetrations) NUREG All* (Fire protection, Fire Barriers, Penetrations) All* (Fire protection, Fire Barriers, Penetrations) M >N/ EY 7\ -To'- Y Cedified Reviewers Signature Herbert R. Rideout WW97 Printed Name Date Reviewers certification expiration date: 2/15/98 Assistance provided by: Printed Name Scope of Assistance Date
.f g Y " g gt
FO ;M TITLE. _._~~m. - FORM NO. f EV. 10CFR60.69 DETERMINATlON 1000.131 A 2 PC.2.3 Page 3 off ENVIRONMENTAL IMPACT CHECKLIST Document No. 973932P202 (UNIT 1 and UNIT 2) [ O RevlChange No. O Complete the following checklist. If the answer to any checklist item is "Yes", an Environmental required. See Section 6.2.1.E for additional guidance. Will the Activity being evaluated: Yes g O O Disturb land that is beyond that initially disturbed during construction (i.e.. new construc buildings, creation or removal of ponds, or other terrestrialimpact)? See Unit 2 SAR 2.5-17. This applies only to areas outside the protected area. O O increase thermal discharges to take or atmosphere? O 2 increase tower? concentration of chemicals to cooling lake or atmosphere through discharge O @ increase tower? quantity of chemicals to cooling take or atmosphere through discharge canal or O S Modify the design or operation of cooling tower which will change drift characteristics? O O Install any new transmission lines leading offsite? O @ Change the design or operation of the intake or discharge structures? O O Discharges any enemicats new or different from that previously discharged? O @ Potentially water or groundcause water? a spill or unevaluated discharge which may effect neighboring soils . O E involvewater surface burying or placement or ground water? of any solid wastes in the site area which may effect runof , O @ involve incineration or disposal of any potentially hazardous materials on the ANO site? O O Result in a change to nonradiological effluents or licensed reactor power level? O @ Potentially ANO site. change the type or increase the amount of non-radiological air emissions i
ARKANSAS NUCLEAR ONE FORM TITLE. FORM NO. REV. 10CFRSO.89 REVIEW CONTINUATION PAGE 1000.131C 2 Page 4, of 4 Document No. 973932P202 Rev> Change No. 0 10CFRC0_.59 Review Continuation Pace The proposed modification is to relocate tho_Outaos Control Center (OCC) to the exi offices located on El 386' of the auxiliary buildino. specifically Room 2126 and 2146. Th renovation will remove the exis;;..s walls for Room 2145 to allow more room for the olenn . Conditions for this area do not irra=d any of the documents listed under item e1 above th the information in the listed documents to be untrue or cause error. However, ne due to f in documents listad underitem 2 a SAR and FHA chance r nawill GARbe Fla.1.2 reouired 3 and Pre- to b i Fire Plan Dr:das FZ-2063 to r,sse;ent the oronosed confiauration. The imonet . e on thes that an evaluation of the msdincation be addressed. This evaluation is attached cedure as ner p reauirements. PM ___ gg Q .
,,. .~
... -n w. .r.
FORM TITLE: FORM NO. REV. 10CFRSO.59 EVALUATION 1000.1315 2 4 Page j. off PAGE REV.O 10CFR50.59 Eval. No. FF u- 9-)-Of o9 (Assigned by PSC) Document No. 973932P202 Rev1 Change No. g Title Relocation of Otm Control Center (OCCi to CA-2 For Unit 2 Ottre A WRITTEN RESPONSE PROVIDING THE BASIS FOR THE ANSWER TO EACH ATTACHED. EACH QUESTION MUST BE ANSWERED SEPARATELY. A SIMPLE ST CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES GUIDANCE FOR If the answer to any question on this form is "Yes," then an unreviewed safety question is involved. If t to all questions is "No," then the proposed change does not involve an unreviewed safety question.
- 1. Will the probability of an accident previously evaluated in the SAR be increased?
Yes O No @ No new or avia*!no =cenarios will be i.nr+%' by the reasc f:-as in this area. No -:t-- "":-s +@ that include this mun as part of an sc-:!d: ;; condition. In eeneral the renovations beins r -'e will-not inci.ee. nor f::r :: any exisi;as Lic-t+t-iEh as it relates to cssditions stated in any i RD.
- 2. Will the consequences of an accident previously evaluated in the SAR be increased?
Yes O No @ No previousiv evaluated accident will be impacted by the renovations to this area. No new conditions will be initiated. Renovations will not include any safev sionificant activW= nor will any limits be exceeded that may already exist in this area. 3. Will the probability of a malfunction of equipment important to safety be increased? Yes O No @ No increased probeb!!ity of a malfunction of eouisment imssiisnt to s.:.;i will exist. Svsnm ossi.;;60 parameters are not imes-ted and existino desion M::: recuise.nea;. are still -- -:7.i ined. No couloment is impacted by this mod that will be prevented from fulfillino a desion functica. 4. Will the consequences of a malfunction of equipment important to safety be increased? Yes O No @ No eautoment imoortant to safety is impacted by this chance. Area beino renovated is non-safety sionificant and has no components beino installed or removed that fall within this cEEGeis.
- 5. -
Will the possibility of an accident of a different type than any previously evaluated in the SAR be created? Yes O No @ No new accident are created by this renovation. No safety sionificant components are beino impacted that will chance this condition.
.........,-w.y6 FORM TITLE:
10CPR80.80 EVALUATION FOAM NO. REV. 1000.131B 2 PC 973WP,70) PAGE 8 REV.O Page2of2 6. Will the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR be created? Yes O No @ No new possibilities of souleT.snt malfunctions will be oenerated by this renovat Nosafety sianificant components are beina imoscted that will increase or chance; conditions. 4 7. Will the margin of safety as defined in the bases for any technical specification be reduced? Yes O No @ No maroins inouts. are imeeded. The renovation to the area has no impact to maroin
, 1 -
c/ A / Herbert R. Rideout Ciedifisti Revieweir's Signature 4/4/97 Printed Name Date Reviewer's certification expiration date:_ 2/15/98 Assistance provided by: Printed Name Scope of Assistance Date PSC review by: A
--_b DateAY
U394'1 poi Doc:: ment No. PC" G abua RevdChar.ge Nr. - 0 Title ALTERNATE RELIABLE ANO-2 POWER FOR VSF-9 PC 973567P301 Will the proposed Activity: PAGE R+ REV D 2./
- 1. Require a change to the Operating License including:
Technical Specifications (escluding the bases)? Yes No G Operating License? Yes O No @ Confirmatory Orders? Yes O No @ 2. Result in information in the following SAR documents (including drawings and text) being (a) no longer true or accurate or (b) violate a requirement stated in the document: Core Operating Limits Report? Yes O No @ SAR imulti-volume set for each unit)? Yes @ No O QAM0?a Yes No O E-Plan?* Yes O No D FHA Yes O No @ Basen of the Technical Specifications? Yes O No E NRC Safet) Evaluation Reports? Yes O No @
.l. Imohe a test or esperiment not described in the SAR?
Yes No G J. Result in a potential impact to the environment? (Complete the EndnoirnentalImpact Checklist,of this form.) Yes No @
- 5. Result in 6. 1eed for a Radiological Safety Evaluation per section 6.2.4.a?
o Yes O No @ Hesult in any parential impact to the equipment or facilities utilised for Ventilated Storage Cask activities per section 6.2.4.b? Yes O No D it.m Inr lietermination:
- s. - n.ninnen: 10: It.o.a l'or ik:tenmnation Chan::es to these documents require an evaluation in accordance with 10CFR50.54.
. Nee .Nectinn 6.2.1.B.
I t Hul IIIl.I.: FORM No. REY. 10CFR50.59 DETERMINATION 1000.131 A 2 PC-2.3
P:ge _ 22 Document Nr. PC 973980P301 RevdChange No. O R;ferences: List sections reviewed in the Licensing Basis Documents, specified in ques: ions 1,2 was done on LRS, "all" may be entered under "Section" with the keyword (s) used in hard copies of the documents shall be reviewed as computer-based searches such search text only, not figures or drawings. Attach a completed LDCR if LBD changes are r Document Eggtig,n, ANO-1 TECH SPECS ANO-lSAR ALL (B55. 2B64. VSF-9. 2VSF-9. Control T.com isolation. CREV. CREVS) Chapter 8. Chapter 9 - Text and Firures ANO-lSER OAMO I E-PL A N FHA l i ANO-2 TECH SPECS f ANO-2 SAR ANO-2 SER Chapter 8. Chapter 9 - Text and Fieures
- FIGURES 8.3-66 and *8.3-15
{ DHr o a. AAuw </{r /9, Certilled Reviewer's Signature Printed Name 4)at(e Resien er'. certification expiration date: f/7/ff
/ /
higarce prmided by: 4 Printeel Name Scope of Assistance Date Ubii 1. nile Initial scarch and first draft. I i PC 973967P301 P AGE ** RE V O l 22 I ( nul i I I l.l': FORM NO. REY. lHCFR50.59 DETERMINATION 1000.131 A 2
Pcge 23 Doe: ment Ns _ PC 973980P301 Rev./ Change N 0 ENVIRONMENTAL IMPACT CHECKLIST (UNIT I and UNIT 2) Complete Section 6.2.1.Ethe following for additional checklist. guidance. If the answer to any checklist item is "Yes", an Environmental Evalu . Will the Activity being evaluated: PC 973967P301 Yes No PAGE y REYO O E Disturb land that is beyond that initially disturbed during constrwetion (i.e., new construction of buildings, creation or
- removal of ponds, or otner terrestrial impact)? See Unit 2 SAR Figure 2.5-17. This applies only to areas outside the protected area.
] x increase thermal discharges to lake or atmosphere?
2 E increase concentration of chemicals to cooling lake or atmosphere through discharge canal or tower? O @ Increase quantity of chemicals to cooling lake or atmosphere through discharge canal or tower? e O @ Modify the design or operation of cooling tower n hich will change drift characteristics? [ @ install any new transmission lines leading offsite? [ @ Change the design or ope $ation of the intake or discharge structures? [ @ Discharges any chemicals new or different from that previously discharged? [ Q Potentially cause a spill or unevaluated discharge which may effect neighboring soils. Surface water or ground water? [ Q involve burying or placement of any solid wastes in the site area s n hich may effect runoff, surface water or ground water? [ @ involve incit.cration or disposal of any potentially hazardous materials on the ANO site? [ @ Result in a change to nonradiological effluents or licensed Reactor poner level? [ @ Potentially change the type or increase the amount of non-radiological air eminions from the ANO site. I 4 Hol iIi1.1.'; tosas No. REv. IUCFR50.59 DETERMINATION 1000.131 A 2
Prge 24 Docrment Nr. _ PC 973980P301 RevdChange No. 0 10CFR50.59 Review Continuation Pare Ib I Nbb7I30l Dnerintion of Changg PAGE e REV 0 11us I'lant Change mil provide a permanent mechanism to supply ANO-1 Control Room Emergency Fan Filte VSF-9 :ad one tram of the Control Room isolation circuits (Cl41 and C141B) with alternate reliable powe VSF 9 und 2VSF 9 are rerpecuvely the ANO l and ANO-2 Control Room Emergency Fan Filter umts y start on a Control Room isolation signal. The EsEm1 power configuration for the fan motors of VSF-9 and 2VSF-9 are ALTERNATE ALTERNATE NORMAL EMERGENCY NORMAL EOUIPMENT POWER POWER EMERGENCY VSF-9 POWER POWER B5641 DG1/DG2 to B56 N/A N/A 2VSF-9 2B54-C3 2DG1 to 2B54 B5666 DGI/DG2 to 356 durmg penods of time when the opposite unit cannot maintain .
- ec n ca ms op Spectticauon 3N 1 and the ANO-2 Technical Specification 3.7.6.1 both require two independent r onditioning and C Av lihranon o stem.4 be operable. If only one system is operable then the other system shall s or both unit be res shall begm shutdoun iCold shutdown withm 36 hours for ANO-l and withm 30 hours for . ANO-2) The w here bmh trams are out of service. is to install a Temporary Modification which feeds VSF 9 bre -
I empaan Modideauons required the installation of temporary conduit. cable and rr wiring of ANO-1 eqmp .
. Also. one train of the Control luum isolanon uremts are ternporanly powered from umt 2 power via the tnstallation of e farger Ilus I'lant Change pennanently mstalls a more sophisticated version of the Temporary Modification peration to w
inndh snap pmer to VSF-9 from the " Normal" to the "Altema n u nh an custme inechamcalh mterlocked -break before make" starter. ne line side of the -second green-sta thesel baded Mrr 2HM cubicle DI. Additionally the control scheme of B5553 y is modified b th
- i. unetloded u nh the starter coils to electncally isolate the non-selected starter. This new hand
.noenth cie,ted pmer source for H5553 (either BS5 bus power or 2B64DI), A larger 300VA CPT is ins adihnonal sapasm neated to pmer Cl4I and Cl4iB which are the common and green-tratn portion of the a on circuits.
I o,i. .onnsienon oi un I'lant Change the power configurnuon for VSF 9 and 2VSF-9 will be as follows: i ' ALTERNATT. ALTERNATE I I NORMAL EMERGENCY NORMAL !
! UIUPMENT POWER EMERGENCY ;
POWER POWER POWER i VSF-9 B5553 1 DGl/DG2 to B55 2B64DI 2DG2 to 2B64DI
' VSF 9 2B54-C3 2DG1 to 2B54 B5666 DGl/DG2 to B5666 Ountion i Ram fne Determination 6
W han::c Io the non-bases portton of the ANO-l or ANO-2 Techmcal Specifications areent required level of as they i .ici.nl so addie the pec 6cs of breaker t)pe. his moditicauon will not require a change to the ANO-l t onunn.non i irders a, these documents do not contam the level of detail required toange. address this Plant Ch Ountion 2 Bash for Determination hhance omie hu YSiso u the ANO-l SAR documents are reqmred as thev do not contam a sutlicient level of or the Control Room isolation ctremts. No changes to the text of the ANO-2 SAR documents a suinuent lesel at detail to address the specifics of cross 4:es to ANO-l for the power, source for VSF-9 the s a on circuits or oihei sh.ued t ontrol Room EmergencyllVAC components. ANO-2 SAR Figures .
.- - R315(E-2015
- s. sh 4)and R 3 mil require eeuwon as a iesnh ot onduit mstallauon and the use of a MCC breaker 2h64DI by this design change.
P:ge 25 Doc: ment Na _ PC 973980P301 RevdChange No. O Ouestion 3 Basis for Deter-instion The testing
. clumpe required by this change consists of normal wiring checks. No tests or experunents not desenbed in the SAR Ouestion J Basis for Determination PC 973567P301 there are no environmental concerns aszciated vith this change. See attached checklist. PAGE EHP' REY 0 16~
Ouestivm 5 & (. Basis for Determination Tlu 6hanFe does not involve any radioactive material nor does it involve any equipment or procedures assoc catL s An RSE is not required A 10CFR72.48 review is not required. 1 i e l l l
Page 26 1)ocument No. PC 973980P301 Rev./ Change No. O Title ~ ALTERNATE RELIABLE ANO-2 POWER FOR VSF+9 A WRITTEN RESPONSE PROVIDING THE BASIS FOR TiiE ANSWER TO EACII QUESTION ATTACHED. EACH QUESTION MUST DE ANSWERED SEPARATELY. A SIMPLE STATEMENT CONCLUSION IS NOT SU7FICIENT. ATTACIIMENT 2 PROVIDES GUIDANCE FOR RESPONSE If the answer to any sp.esann on this forr ;s "Yes," then an unreviewed safety question is involved. If th answer to all questions is "No," then the PC 973967P301 proposed change does ..ot involve an une criewed safety question. PAGE R9 REV 0
- 1. 2- 6 Will the prubability of an accident priously evaluated in the SAR be increased?
Yes O No @
- 2. Will the conscquences of an accident previously evaluated in the SAR be increased?
Yes O No @
- 3. Will the probability of a malfunction of equipment important to safety be increased?
Yes O No @ 4. Will the consequences of a malfunction of equipment important to safety be increased? i Yes No B
- 5. Will the possibility of an accident of a different type than any previously evaluated in the SAR be created?
Yes No @ 6. Will the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR be created? Yes O No @ l 7. Will the margin of safety as defined in the bases for any
- te mical specification be reduced?
Yes O No @ w DHi e n. OHmAJ Ceriili61 Reticher's Signature 4'f//f97 Printed Name / Ddte Reviewer's certification capiration date: i 7 d/f
/ /
1 Assistance provided by: Printed Name Scope of Assistance Date CHRIS LITTLE INITI AL SEARCrl & DRAFT PSC review by: Date: $ 0
Pcge 27 Documnt No. PC 973980P301 RevdChange No. O 10CFR50.59 Review Continuation Page PC 973G67PS01 PAGE Rf REV 0 i 3-7 The Control Room Emergency Ventilation and Filtration system is designed to mitigate t maintaining the Control room Etwironment smtable for habitation. This system cannot initiate an a ANO l or ANO-2 LBDs. This change does enhance operation of the Control Room Emergency Vent system by providing an alternate method of providing reliable power to shared components (VSF-9, the probability of any accident previously evaluated in the SAR is not increased by this design chan ,
- 2. The Control Room Emergency Ventilation and Filtration system provides barners to onsite ( .
changes made by this design change do not affect the ability of the Control Room Emerge isolate the Control Room and maintain Control Room habitability. Since the design of Control Ro and Fihration system will be mamtained by this change the consequences of any accidem previo ANO-2 SARs will not be increased. i The circuns added/ modified by this design chang are electncally and physically isolated from the redun
.iremis/ components of the Control Room Emergemy Ventilation and Filtration system. The modification pe poner and control circtuts and the common (Cl411 and green train (Cl41B) portion of the Emergency C Grcmts does not introduce any mechanism by ';<hich a single failure could cause the Emergency miem to fail. As a result of this Plant Cnange the nonnal source for 2VSF-9 is 2B54-C3 (red tram) and l poner source for VSF-9 is 2B64 D1 (green tram). MCCs 2BS4 and 2B64 are redundant, independent l The nonnal ANO-l power source and the ahernate ANO-2 power source for VSF-9 are not redu
.onsidered independent as both as both meet ANO design criteria for safety related ciremts. A fauh on e ieopardue both rebmdant trains for either ANO-l or ANO-2. Installation of this design change does n
.mahchance dessen sis desenbed in the LBD. The probability of a malfunction of equipment important to safety l 4
i No moddicanons are inade to the Control Room Emergency Ventilation and Filtration system e whic ambts ofil" e stem to perform its safety function. No new failure modes are introduced No equipment ns are malfunc
.mroduced uluch are not aircady bounded by,smgle failure analysis. Therefore, the consequence
.mnonam m safen) uill not be increased. l I
c i ommi Room Emercenev Ventilation and Filtrauon system is designed to mitigate the conse nm.nmne the Control room Environment smtable for habitanon. This system cannot minate an accident 3 W m.hMthe orFonirc: ANO-2 LBDs The addition of an alternate reliable power supply to VSF-9 does not introduce y a! Room Emergenes Ventilauon and Filtration system could become an accident initiator
<.. idem of a different type than previously evaluated in the SAR will be created as a resuh of this desi
. modiikanons are made to Control Room Emergenes Ventilation and Filtration system eqmpmI i
.a.u the .duh:3 of this s3 stctu to perfonn its safety funcuon. No new equipment malfuncuons or failur m h are not alread) bounded 3 b smgic failure anal > sis. Therefore. the possibility of eqmpmel
.m ines mmh es alnated in the SAR will not be created.
~N
\1.nems of safet) as defined in the ANO-l or ANO-2 Tech Spec Bases associated with the Contr nut.amn and Filtranon system will be reduced as a result of this design change. The installauon of an aher 3
i 1.iu-lEs power suppl niected s3 sicm to ANO-1/ANO-2 shared components (VSF 9. Cl41. Cl418) will maintam the desig \
00M M TilLE: runa sw. ns. e . 10CFRSS.50 oETERMINATioN 1000.131A 3 Pc-1 Page1of N Document No. PC 974062P201 RevdChange No. ,q Title 2E/P4410 AND 2E/P4811 SIGNAL DRIVERS Brief description of proposed change: INSTALL SIGNAL DRIVERS FOR 2E/P4810 AND 2E/P4811 Will the proposed Activity: l PC 974062 P201
- 1. Require a change to the Operating License including:
PA6E 35 RE V 0 Technical Specifications .(excluding the bases)? YesO No@ l Operating License? YesO No@. Confirmatory Orders? YesO No@
- 2. Result in information in the following SAR documents (including drawings and text) being (a) no longer true or accurate, or (b) violate a requirement stated in the document:
SAR (multi-volume set for each unit)? Yes@ NoO Core Operating Limits Report
- YesO No@
Fire Hazards Analysis? YesO No@ Bases of the Technical Specifications? YesO No@ Technical Requi: aments Manual? YesO No@ NRC Safety Evaluation Reports? YesO No@
- 3. Involve a test or experiment not desenbed in the SAR? YesO No@
(See Attachinent 2 forguidance) 4 Result in a potential impact to the environment? (Complete > the Environmentalimpact Determination of this form.) YesO No@
- 5. Result in the need for a Radiological Safety Evaluation per section 6.1.57 YesO No@
- 6. Result in any potential impact to the equipment or facilities utilized for Ventilated Storage Cask activities per Section 6.1.67 YesO No@
- 7. Involve a change under 10CFR50.54 for the following SAR documents per Section 6.1.7:
QAMO7- YesO No@ E-Plan? YesO No@
m- . - . . . . _ . . . . FORM TITLE: FORM NO. REV. 10CFRSO.88 DETERMINATION 1000.131A 3 PC 1,2 Pagelof 4
. Document No. PC 974062P201 Rev> Change No. A PC 974062 P201 Basis for Determination A= Mons 1. 2 & 31:
PAGE F6 RE V 0 0 Proposed change does not require 10 CFR 50.59 Evaluation per Attachment 1, item # . (If checked, note appropriate item #, send LDCR to Licensing). Search Scope: List sections reviewed in the Licensing Basis Documents specified in Ouestion 1,2 and 3. If a seatch was performed on LRS, the LRS search index should be entered under "Section' with the search statement (s) used in parentheses. Controlled hard copies of the documents shall be reviewed (LRS is not verified and searches only text, not figures or drawings). Attach and distribute a completed LDCR per Section 4.1.2 If LBD changes are required. Document Eggggn LRS: ALL (2 HIC-4812A, 2E/P.4810, 2E/P-4811, 2CV.4810, 2CV 4811, M-2231. BACKPRESSURE CONTROL, LETDOWN BACKPRESSURE, CVCS) MANUAL SECTIONS: SAR 9.3.4.2.1, 9.3.4.2.2, 9.3.4.2.3, 't.4.1.8.2 FIGURES: SAR FIGURE 9.3 4 STEVE CAPEHART Certified Reviewer (Signature Printed Name Date Reviewers certification expiration date: 5/9/99 Assatance provided by: Printed Name Scope of Assistance Date Se rch Scope Review Acceptability (NA. if performed by Technical Review per 1000.006) 00,lldd Pad a CasslaJ WMr Certified Reviewers Signature ' Printed Name Date
FORM TITLE: FORM NO. REU. 10CFR80.80 oETERMINATION 1000.131A 3 Page b of N ENVIRONMENTAL IMPACT DETERMINATION (UNIT 1 and UNIT 2) PC 974062 P 201 Document No. PC 974062P201 Rev/ Change No. A PA6E 37 RE V 0 Complete the following Determination if the answer to any checklist item is "Yes", an Environmental Evaluation is required. See Section 6.1.4 for additional guidance. Willthe Activity being evaluated: 1res N J O E Disturb land that is beyond that initially disturt>ed during construction (i.e., new construction of buildings, creation or removal of ponds, or other terrestrial impact)? See Unit 2 SAR Figure 2.5-17. This applies only to areas outside the protected area. O @ Increase thermal discharges to lake or atmosphere? O O Increase concentration of chemicals to cooling lake or atmosphere through discharge canal or tower? O E Increase quantity of chemicals to cooling lake or atmosphere through discharge canal or tower? O E Modify the design or operation of cooling tower which will change drift characteristics? O E Install any new transmission lines leading offsite? i l O @ Change the design or operation of the intake or discharge structures? j 1 O S Discharges any chemicals new or different from that previously discharged? l O E Potentially cause a spill or unevaluated discharge which may effect neighboring soils, surface i water or ground water? O @ involve burying or placement of any solid wastes in the site area which may effect runoff, surface wateror ground water? 1 O E involve incineration or disposal of any potentially hazardous materials on the ANS site? O E Result in a change to nonradiological effluents or licensed reactor power level? O 'E Potentially change the type or increase the amount of non-radiological air emissions from the ANO site.
ARKANSAS NUCLEAR ONE FORM TITLE: FORM NO. REV. 10CFR$0.58 REVIEW CONTINUATION PAGE 1000.131C 3 PageIof N Document No. PC 974062P201 Rev> Change No. 9 PC 974062 P201 10CFR50.59 Review Continuation Pace PADE 38 RE V 0 This PC will install new valve positioners and associated signal drivers for RCS letdown backpressure valves 2CV-4810 and 2CV-4811. The installation of the new valve positioners was justified per ER 974062P201. This ! PC builds on the ER justification for the positioners and provides justification for the signal drivers. The ope and failure modes for valves 2CV-4810 and 2CV-4811 are unchanged by this PC. l QUESTION 1 - Operatino License The RCS letdown backpressure valves are not discussed in the level of detail present in the ANO-2 Technical i Specifications, Operating License or any Confirmatory Orders. j QUESTION 2 - SAR Documents The signal drivers being added by this PC will be shown on P&lD M-2231 sht 1 (CVCS). This drawing corresponds to SAR Figure 9.3-4. i QUESTION 3 - Test or Exoeriment The post modification testing performed by this PC is within ANO procedures. QUESTION 4 - Environmental impact The modifications made by this PC do not require an Environmental impact Evaluation per the Environmental Impact Checklist. { QUESTION 5 - Radiolooical Safety Eve!ustion The wont performed by this PC will not affect monitored ventilation or drainage pathways. The radioactive material associated with this PC will be the old valve positioners. QUESTION 6 - Ventilated Storace Cask The components associated with the RCS letdown backpressure valves do not interface with the systems associated with the VSC project. QUESTION 7- QAMO or E-PLAN The RCS letdown backpressure valves are not referenced in the QAMO or E PLAN. l l i J
9 AmuwsAs NUCLEAR ONE FOf MTITLE:
+ FORM NO. RElf.
10CFRse.69 EVALUATION 1000.131a 3 PC 2 PCTN0(f2920l p.,, ,_ o, N 10CFR50.59 Eval. No. TT @ D M h (Assigned by PSC) Document No. PC 974062P201 RevlChange No. 9 Title 2E/P-4810 AND 2E/P-4811 SIGNAL DRIVERS A WRITTEN RESPONSE PROVIDING THE BASIS FOR THE ANSWER TO EACH QUESTION MUST BE ATTACHED. EACH QUESTION MUST BE ANSWERED SEPARATELY. A SIMPLE STATEMENT OF CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES GUIDANCE FOR RESPONSE. If the answer to any question on this form is "Yes," then an unreviewed safety question is involved, if the answer to all questions is "No," then the proposed change does not involve an unreviewed safety question.
- 1. Will the probability of an accident previously evaluated in the SAR be increased?
Yes O No @ There is no discussion of the letdo@n backpressure valve configuration discussed in the accident analysis. The operation of the valves (air to open, fail closed) is not being changed per this PC. The components being added by this PC (i.e. signal drivers and air filters) do not interface with any equipment that is associated with SAR accidents. .
- 2. Will the consequences of an accident previously evaluated in the SAR be increased?
Yes O No @ The signal drivers and air filters being added by this PC do not place any additional demands on any equipment important to mitigation of any accidents. The signal drivers and air filters do not interface with any equipment used to mitigate the consequences of any accidents previously evaluated in the SAR.
- 3. Will the probability of a ma: function of equipment important to safety be increased?
Yes O No @ The signal drivers, air filters, valve positioners and valve operators are classified as NON-Q. The valve body is classified as S-R1. The operation and failure modes of the letdown backpressure valves is not affected by this PC. The signal drivers are powered from instrument AC panels 2Y1 brkr 2 and 2Y2 brkr 2 which are classified Q and as such are considered equipment important to safety. The panel breakers supply power to a variety of non-Q loads in 2009 via shorting terminal blocks. The justification for providing power to non-Q components from a Q supply panel is that the panel breaker serves as the Q - Non Q (Class 1E - Non Class 1E) boundary break. In addition to the breaker, a 20 amp fuse is installed in the power feed to 2C09. The only consideration related to ' malfunction of equipment' is adding electrical loads beyond the capacity of the associated breakers. The additional breaker loading has been considered and determined to be within the load margin of the panel breaker. Therefore, the probability of a malfunction of equipment important to safety is not increased.
- 4. Will the consequences of a malfunction of equipment important to safety be increased?
Yes O No @ The signal drivers interface with equipment important to safety (Electrical Panels 2Y1 and 2Y2 - See discussion in Question 3). Given the signal drivers do not affect the ability of panels 2Y1 and 2Y2 to supply power to any other equipment important to safety required to mitigate an accident, the dose consequences associated with malfunctions of equipment important to safety are not increased.
g ARKANSAS NUCLEAR ONE FORM TrfLE: FORM NO. REE 10CFR80.80 EVALUATION 1000.1318 3 PC-2 PcA74062Rzai 5. Will the possibility of an accident of a different type than any previously fwf 40 W6 evaluated in the SAR be created? Yes O No @ The function of the equipment being modified by this PC is not required for shutdown of the radioactive releases or maintaining reactor coolant pressure boundary integrity. The operation of the letdown backpressure valves is not considered an accident initiator. 6. Will the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR be created? Yes O No @ The signal drivers interface with equipment important to safety (Electrical Panels 2Y1 and 2Y2 addition of the electrical loads does not create any new failure modes for the panels. The signal drive not accident initiators. Therefore, the installation of the signal drivers will not create the possibili accident of a different type than any previously evaluated in the SAR.
- 7. Will the margin of safety as defined in the basis for any technical specification be reduced?
Yes O No @ The letdown backpressure valves are not referenced in the ANO-2 Tech Specs.
/ ~
STEVE CAPEHART CertWied Reviewefs signature Printed Name Date Reviewer's certification expiration date:- 5/9/99 Assistance provided by: Printed Name Scope of Assistance Date ! PSC review by: Date: $ N
c ARKANSAS NUCLEAR ONE FORM TITLE: FORM NO. REV. 1eCFR$0.SS DETERMINATION 1ecc.131A 3 Pc.1 Page ,l of 3 Document No. ER974326P201 Rev/ Change No. g Title External Limit Switches Removal for 2CV-5859-2 and 2CV-6882-2 Brief description of proposed change: Removal of abandoned external limit switches. Will the proposed Activity:
- 1. Require a change to the Operating License including:
Technical Specifications (excluding the bases)?
' Operating License? YesO No@
Confirmatory Orders? YesO No@ YesO No@ 2. Result in information in the following SAR documents (including drawings and text) being (a) no longer true or accurate, or (b) violate a requirement stated in the document: SAR (multi-volume set for each unit)? Yes@ nod Core Operating Limits Report Fire Hazards Analysis? YesO NoS Bases of the Technical Specifications? YesO No@ Technical Requirements Manual? YesO No@ NRC Safety Evaluation Reports? YesO NoS YesO No@
- 3. Involve a test or experiment not described in the SAR?
(See Attachment 2 for guidance) YesO No2 4, Result in a potential impact to the environment? (Complete the Environmental Impact Determination of this form.) YesO No@
- 5. Result in the need for a Radiological Safety Evaluation per section 6.1.57 YesO NoS
- 6. Result in any potential impact to the equipment or facilities utilized for Ventilated Storage Cask activities per Section 6.1.67 YesO NoS
- 7. Involve a change under 10CFR50.54 for the following SAR documents per Section 6.1.7:
QAMO? E-Plan? YesO 'NoS YesO No@ Basis for Determination (Questions 1. 2 & 31: This change will not require any changes to the operating liscense. Section 9.3.2.3 of the Unit 2 SAR will need to be changed to reflect the deletion of the external limit switch position indication on the sampling panel (2C-116) and SAR Figure 9.3.2 will need to be changed to show the deletion of 225-58598-2 and 2Z5-5852B-2. No other SAR documents will be affected. This modification will'not constitute a test or experiment, will not have any impact on the environment, nor pose any postulated impact on the Ventilated Storage cask cetivities.This determination indicates that a 50.59 Evaluation will be needed due to the
-impact on the Unit 2 SAR.
O Proposed change does not require 10 CFR 50.59 Evaluation per Attachment 1, item # ,(if checked, note appropriate item #, send LDCR to Licensing). PC E R914 3 2B P201 8 PAGE REV 0
f FORM TITLE ARKANSAS NUCLEAR ONE FORM NO. REV. 10CFR$0.50 DETERMINATION l 1000.131A 3 PC-1 l Page 2 of 3 Document No. ER974326P201 t Rev> Change No. 0 Search Scope: . l List sections reviewed in the Licensing Basis Documents specified in Question 1 was done on LRS, "all" may be entered under "Section" with the keyword (s) used copies of the documents shall be reviewed (LRS is not verified and searches o Attach and distribute a completed LDCR per Section 6.1.2 if LBD changes are required . Document LRS: Section U2 50.59 DOC. All LRS Search The Following word searches were made: "2CV-6859-2,2CV58 Sampling System, Secondary Sampling, External Limit Switch,2C 118, Sampling Panel" MANUAL SECTIONS: U2 SAR Section 9.3.2 Process Sampling Systems FIGURES: 9.3.2 g M-2237 Sh 1 Rev 55 e j i Certified Reviewers Signature Steve Chandler 10 13-97 Printed Name Date Reviewers certification expiration date: 1-6-99 Assistance provided by: Printed Name Scope of Assistance Date Search Scope Review Acceptability (NA, if performed by Technical Review per 1000.00 YY h L a A hln e< k ll Certified Reviewefs Signature Printed Name
. lelNVf
' Date l
PC ER914 326 P201 PAGE 9 REV 0
ARKANSAS NUCLEAR ONE FORM TITLE:
. FORM NO. REtf.
10CFR80Je DETERMINATION 1000.131A 3 Pc 1 Page 3. ofj ENVIRONMENTAL IMPACT DETERMINATION (UNIT 1 and UNIT 2) Document No. ER974326P201 Rev> Change No. ,0 Complete the following Determination. If the answer to any checklist item is "Yes", an Environmenta is required. See Section 6.1.4 for additional guidance. Willthe Activity being evaluated: Yes p O e Disturb land that is beyond that initially disturbed during construction (i.e., new constructio buildings, creation or removal of ponds, or other terrestrial impact)? See Unit 2 SAR 2.5-17. This applies only to areas outside the protected area. O @ Increase thermal discharges to take or atmosphere? O e increase concentration of chemicals to cooling lake or atmosphere through discharge ca tower? O g increase quantity of chemicals to cooling lake or atmosphere through discharge canal or tower? O O Modify the design or operation of cooling tower which will change drift characteristics? O @ install any new transmission lines leading offsite? O @ Change the design or operation of the intake or discharge structures? O e Discharges any chemicals new or different from that previously discharged? l O O Potentially cause a spill or unevaluated discharge which may effect neighboring soils, s water or ground water? O @ Involve burying or placement of any solid wastes in the site area which may effect run surface water or ground water? O E Involv3 incineration or disposal of any potentially hazardous materials on the ANO site? O E Result in a change to nonradiological effluents or licensed reactor power level? O E Potentially change the type or increase the amount of non-radiological air emissions from t ANO site. I i PC E R 914 3 26 P201 PAGE /0 REV 0
ARKANSAS NUCLEAR ONE FORM TITLE: FORM NO. REV. 10CFRSO.59 EVALUATION 1000.1313 3 Page loff 10CFR50.59 Eval. No. i 7 Dc/F % (Assigned by PSC) Document No. 974326P201 RevdChange No. 0_ Title External -limit switches removal from 2CV-5859-2 and 2CV-5852-2 A WR11 TEN RESPONSE PROVIDING THE BASIS FOR THE ANSWER TO EACH QU ATTACHED. EACH QUESTION MUST BE ANSWERED SEPARATELY. A SIMPLE STATE { CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES GUIDANCE FOR RES If the answer to any question on this form is "Yes," then an unreviewed safety question is involved. If the a to all questions is "No," then the proposed change does not involve an unreviewed safety question.
- 1. Will the probability of an accident previously evaluated in the SAR be increased?
Yes O No E The process sampling systems are not essential for safe plant shutdown and serve no emergency function during operation. This modification does not alter any other valves components position to be lostoperation at panel but will cause indication of the motor operated 2C-116. the control room for these valves position indication. Indication will remain at 2C-16 in All valve automatic control is from information onlythe control room and the indication provided to 2C-116 is for indication. the modification package and will notSecondary indication is the only issue affected by an accident will occur. change in any way the probabilities that Removal of the external limit switches will not introduce any new accident initiators or cause any scenarios previously evaluated conditions. in the SAR to provide less mitigative capabilities under accident 2. Will the consequences of an accident previously evaluated in the SAR be increased? Yes O No @ The consequences of previously evaluated accidents will not be affected by this modification. Loss of indication for the steam generator samples at 2C-116 will not increase or decrease the consequences of an accident because this panel is not monitored except when sampling is in progress. There is no increase in the consequences of an accident previously evaluated because this system is not analyzed for mitigative actions in the event of an accident. During an accident condition the valve positions will be monitored and controlled from the 2C-16 control panel in the Unit 2 control room. There are no increased consequences from an evaluated accident as a result of this modification. 3. Will the probability of a malfunction of equipment important to safety be increased? Yes O No S The external limit swit'Bes being deleted for 2CV-5859-2 and 2CV-5852-2 provide secondary position i. ucation to chemistry panel 2C-116 for the convenience of the chemists. Removal of these position indication switches will not alter the functionality of the equipment being monitored because primary position indication internal limit switches.is provided to the control room by the Limitorque actuator The switches being removed are all external to the valve and valve actuator and have been non functional for over 7 years. The valve and valve actuator are the components that they will be unaffected in their ability to functionare important toproviding) safety and position indication) by the modification as proposed. (including There will be no change to the probabilities important to safety. of malfunction of the equipment (valve or actuator) '
?c y 9/Y12C P 20/
'O ## Aod
i ARKANSAS NUCLEAR ONE FORM TITLE: FORM NO. REV. 10CFR40.89 EVALUATloN 1000.131a 3 ; Page2of_2
- 4.
Will the consequences of a malfunction of equipment important to safety be increased? . Yes O No @ There . is no functional loss to the equipment important to safety that is affected by this modification. .Since no functional loss occurs with the
. of modification their are no additional consequences postulat'ed by the installation this, mod package.
5. Will the possibility of an accident of a different type than any previously evaluate (in the SAR be created? Yes O No E Operation of.the Process Sampling Systems without indication at panel 2C-116*is ' not outside the bounds of any accident scenario that has previously evaluated. been Any new postulated accidents evaluated with the same criterik would not take credit for this indication to mitigate the. accident ' therefore the possibility.of an accident of a different type than any previously evaluated in the LBD will not be created. 6. Will the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR be created? Yes O No @ The limit switches to be removed provide secondary position indication for valves Because important of this anyto changes safety but tothey themselves do not have any safety functions. the external limit switches cannot cause a malfunction of equipment important to safety of a type that is different from that which has'been previously evaluated. 7. Will the margin of safety as defined in the bases for any technical specification be reduced? Yes O No S There are no technical specifications regarding the function of the \
. external position indication switches'which provide indication to 2C-116. Since there are no tech specs related to these lindt switches, there is also no bases associated margins as with same defined in switches the basisandfor no corresponding margins of safety. Safety changed as a result of this modification. technical specifications will not be
.<,1
'T /
g - Steve Chandler CertifieMi Revieners Signature 10-13-97 Printed Name Date
- Reviewers certification expiration date: /- f, - 9 9 Assistance provided by:
Printed Name Scope of Assistance Date PSC review by: h ( (. Date: O3D ! l PC ER9M 32B P201 PAGE /2 REV 0 {
ARKANSAS NUCLEAR ONE PifiF 2 Pao)d r i, FORM TITLE: "' FORM NO. MEV. 10CFR50.59 DETERMINATION 1000.131A 3 PC-1 This Document contains 3 Pages. N:" ment No. 974343P201 Rev1 Change No. 0 Title 2PSV-0706 Flange Addition Brief description of proposed change: This Plant Change will add two flange pairs to the one inch diameter piping containing EFW suction relief valve 2PSV-0706. This will allow 2PSV-0706 to be easily removed for maintenance by unbolting the two flange pairs. There w change to the function or operation of the system as a result of this Plant Change. The planned modifica system have been qualified for the seismic category I and other applicable loads in accordance with the ASME Se Class 3 Code. The flanges being added are ASME Section III, Class 3 which is consistent with the existing p Will the proposed Activity:
- 1. Require a change to the Operating License including:
Technical Specifications (excluding the bases)? YesO No@ Operating License? YesO No@ Confirmatory Orders? YesO No@ 2. Result in information in the following SAR documents (including drawings and text) being (a) no longer true or accurate, or (b) violate a requirement stated in the document: SAR (multi-volume set for each unit)? Yes@ nod Core Operating Limits Report? YesO No@ ; Fire Hazards Analysis? I YesO No@ l Bases of the Technical Specifications? YesO No@ ; NRC Safety Evaluation Reports? YesO No@ {
- 3. Involve a test or experiment not described in the SAR?
(See Attachment 2 for guidance) YesO No@
- 4. Result in a potentialimpact to the environment? (Complete Environmental Impact Determination of this form.)
YesO No@
- 5. Result in the need for a Radiological Safety Evaluation per section 6.1.57 YesO No@
- 6. Result in any potential impact to the equipment or facilities utilized for Ventilated Storage Cask activities per Section 6.1.7?
YesO No@
- 7. Involve a change under 10CFR50.54 for the following SAR documents per Section 6.1.77 QAMO?
YesO No@ E-Plan? YesO No@ M
ARKANSAS NUCLEAR ONE Pace 2 FORM TITLE: FORM NO. REV. 10CFR50.59 DETERMINATION 1000.131A 3 PC-1 Document No. 974343P201 l Rev./ Change No. O { Basis for Determination (Questions 1,2, & 3): 97 4 34 3P2 C RE
- 1. %e flanges added by this Plant Change are passive piping components and do not change the function operation of the system. There is nothing in the Operating License that will be required to be cha result of this Plant Change.
- 2. Unit 2 SAR Figure 10.4-2 (P&lD M-2204, sht. 4) will be revised to reflect the additional flanges ad Change. Dere is nothmg in this Plant Change other than the SAR figure mentioned above statement or information contamed in the SAR documents to become untrue or inaccurate,
- 3. There are no tests or experiments as described in the SAR involved with this Plant Change.
Proposed change does not require 10CFR50.59 Evaluation per Attachment 1, Item # checked, note appropriate item #, send LDCR to Licensing). (If Search Scope: List sections reviewed in the Licensing Basis Documents specified in questions 1,2 and 3. search was done on LRS, "all" may be entered under "Section" with the keyword (s) used in pa Controlled hard copies of the documents shall be reviewed (LRS is not verified and searche figures or drawings). Attach and distribute a completed LDCR per Section 6.1.2 if LBD c required. Document Section , LRS: All ("2PSV-0706", flange w/50 emergency feedwater, emergency feedwater w/10 relief) MANUAL SECTIONS: Unit 2 SAR Section 10.4.9 FIGURES: 10.2.3.10 4.2
*/#4./gud-- Keith Butler 2/20/98 Certified Reviewer's Signature Printed Name Date Reviewer's certification expiration date: 11/21/98 Assistance provided by; Printed Name Scope of Assistance I none Date Search Scope Review Acceptability (NA, if performed by Technical Reviewer per 1000.006)
Y $ lvY>H h:>cV'Sc d 55- W Certifie4 Reviewer's Signature Printed Name Date
l 1 ' ARKANSAS NUCLEAR ONE Pao) 3 } FORM TITLE: i FORM NO. REV. 10CFR50.59 DETERMINATION 1000.131 A 3 l ENVIRONMENTAL IMP.\CT DETERMINATION (UNIT 1 and UNIT 2) 97 4 34 3P2 0 ' ' PAGE 5 RE, Document No. 974343P201 Rev1 Change No. O Complete the following Determination. If the answer to any item below is "Yes", an Environmental Evaluation is required. See Section 6.1.4 for additional guidance. Will the Activity being evaluated: Yes No
.O E Disturb land that is beyond that initially disturbed during construction (i.e., new construction of buildings, creation or removal of ponds, or other terrestrial impact)? See Unit 2 SAR Figur 2.5-17. This applies only to areas outside the protected area.
O S Increase thermal discharges to lake or atmosphere'? { i @ Increase concentration of chemicals to cooling lake or atmosphere through discharge canal o i l tower? O O Increase quantity of chemicals to cooling lake or atmosphere through discharge canal or tower? O E Modify the design or operation of cooling tower which will change drift characteristics? O Install any new transmission lines leading offsite? O Change the design or operation of the intake or discharge structures? O Discharges any chemicals new or different from that previously discharged? O E Potentially cause a spill or unevaluated discharge which may effect neighboring soils, surface
- water or ground water? j l 0 .
Involve burying or placement of any solid wastes in the site area which may effect runoff, I surface water or ground water? O Involve incineration or disposal of any potentially hazardous materials on the ANO site? O O Result in a change to nonradiological effluents or licensed reactor power level? Potentially change the type or increase the amount of non-radiological air emissions from the ANO site. j i
ARKANSAS NUCLEAR ONE Pagt 1 FORM TlTLE: FORM NO. REV. 10CFR50.59 SAFETY EVALUATION 1000.131B 3 This Document contains 1 Page. Document No. 974343P201 RevdChange No.. 0 10CFR50.59 Eval. No. PMf-Ols3 Title 2PSV-0706 (Assigned by PSC) A WRITTEN RESPONSE PROVIDING THE BASIS FOR THE ANSWER TO EACH ATTACHED. EACH QUESTION MUST BE ANSWERED SEPARATELY. A SIMPLE S CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES GUIDANCE FOR If the answer to any question on this form is "Yes," then an unreviewed safety question is involved. If t to all questions is "No," then the proposed change does not involve an unreviewed safety question. 1. Will the probability of an accident previously evaluated in the SAR be increased? YesO No @ This modification is adding two flange pairs to the one inch diameter piping containing EFW suction re 2PSV-0706. The flanges being added are passive piping components and do not change the function or operation of the system. The piping system will remain qualified in accordance with the ASME Sect Code for the seismic Category I and other applicable loads. This modification does not affect any of any of the events evaluated in the SAR since there is no affect on the function or operation o 2. Will the consequences of an accident previously evaluated in the SAR be increased? YesO No @ This modification does nothing to change the function or operation of the system. This chang of the mitigating functions associated with any of the accidents evaluated in the SAR. The flange are qualified in accordance with the ASME Section lit, Class 3 Code in accordance with the 2HBC-85 line requirements. There is nothing being done by this modification that could affect offsite doses. 3. Will the probability of a malfunction of equipment important to safety be increased? YesO No @ The flanges being added by this modification package will not change the function or operation of The flanges are constructed in accordance with the ASME Section lit, Class 3 Code of the mater for the applicable line class 2HBC-85 in Specification ANO-M-2555. The piping system is qualified in a; with the ASME Section lil, Class 3 Code for the addition of the flanges. Since the system function anl
. will not change, and the flanges meet all the design requirements of the system, this modification whi m I the probability of a malfunction of equipment important to safety to be increased.
4.- Will the consequences of a malfunction of equipment important-
' to safety be increased?
YesO No @ ' The flanges being added by this Plant Change are passive piping components that do not change . operation of any equipment important to safety. The addition of the flanges will not change the method of or have any affect on the consequences of a malfunction of any equipment important to safety. 5. Will the possibility of an accident of a different type than any previously evaluated in the SAR be created? YesO No @ The flanges being added by this Plant Change are passive piping components that do r'ot change th operation of any system. The addition of the flanges meets the design requirements for the system. The of the flanges does not cause any condition that is different than the existing system such that an accide different type than previously evaluated in the SAR could be created. 97 4 34 3 P2 01 PAGE 6 RE V 0
ARKANSAS NUCLEAR ONE. Page 2 FORM TITLE: FORM NO. REV. 10CFR50.59 SAFETY EVALUATION 1000.131B 3
- 6. Will the pos:,1bility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR be created?
l YesO No @ This Plant Change does not constitute a functional change in any system as evaluated in the SAR being added are passive piping components which do not have any interaction with any equipment imp safety in any way that is different than the existing configuration. There are no new postulated failure mode any equipment important to safety as a result of this modification. 7. Will the margin of safety as defined in the bases for any technical specification be reduced? ) l YesO No @ ' There are no margins of safety as defined in the bases for any technical specification that are to the addition of flanges as installed by this Plant Change. M/ #kIh M. Keith Butler 2-23-98 Certified Reviewer's Signature i Printed Name Date l l Reviewer's certification expiration date: 11/21/98 l Assistance provided by: Printed Name Scope of Assistance NA Date
* ,ff0 h, /
/ I PSC review by:'
/
7 ' ' Date: f f l l 97 4 34 3 P2 01 PAGE 7 .RE V 0
r: , w pgn l KANSAS NUCLEAR ONE Page i FORM TITLE: FORM NO. REV. 10CFR50,59 DETERMINATION 1000.131A 3 PC-1 This Document contains 3 Pages. Document No. 974346P201 Rev./ Change No. _0 Title 2PSV4o97 Flange Addition Brief description ofproposed change: This Plant Change will add a flange pair on each side of Containment Spray relief valve 2PSV-5697. T 5697 to be easily removed for maintenance. There will be no change to the function or operation of t this Plant Change. The planned modifications to the piping system have been qualified for the se pair on the non-Q,2HCD-22-1",2PSV-5697 discharge piping. applicable loads Will the proposed Activity: 1. Require a change to the Operating License including: Technical Specifications (excluding the bases)? YesO No@ Operating License? YesO No@ Confirmatory Orders? YesO No@ 2. Result in information in the following SAR documents (including drawings and text) being (a) no longer true or accurate, or (b) violate a requirement stated in the document: SAR (multi-volume set for each unit)? Yes@ nod Core Operating Limits Report? YesO No@ Fire Hazards Analysis? YesO No@ Bases of the Technical Specifications? YesO No@ NRC Safety Evaluation Reports? YesO No@ 3. Involve a test or experiment not described in the SAPJ (See Attachment 2 for guidance) YesO No@ 4 Result in a p atentialimpact to the environment? (Complete Environmental impact Detercoination of this form.) YesO No@ 5. Result in the ne ed for a Radiologi l Safety Evaluation per section 6.1.57 YesO No@ 6. Result in any potential impact to the equipment or facilities utilized for Ventilated Storage Cask activities per Section 6.1.77 YesO No@ 7. Involve a change under 10CFR50.54 for the following SAR documents per Section 6.1.77 QAMO? E Plan? YesO No@ 974 34 6 P2 01 YesO NoS PAGE 3 REV 0
_,974346P201 ARKANSAS NUCLEAR ONE FORM TITLEt Page 2 FORM NO, REV. 10CFR50,59 DETERMINATION 1000.131A 3 PC-1 i Document No. 974346P201 Rev./ Change No. _ 0 Basis for Determination (Questions 1,2, & 3): 974 34 6 P2 01
- 1. The flanges added by this Plant Change are passive PAGE piping components 4' and d REV 0 result of this Plant Change. operation of the system. There is nothing in the Op 2.
Unit 2 SAR Figure 6.2-17 (P&ID M-2236, sht.1) will be revised to reflect the ad Change. an exception to using Section 6.2.2.2.1 weldedjoints in the system. of the Unit 2 SAR will require revision to note that th 3. There are no tests or experiments as described in the SAR im olved with this Plant C l Proposed change does not require 10CFR50.59 Evaluation per Attachment (If 1, Item checked, note appropriate item #, send LDCR to Licensing). Search Scope: List sections reviewed in the Licensing Basis Documents specified in q , search was done on LRS, "all" may be entered under "Section" with theI; keywo figures or drawings). Attach and distribute a com required. e i Document Section LRS: All (" containment spray", " relief valve", flange *, ESF w/50 leak *, ECCS w/50 leak *) MANUAL SECTIONS: Unit 2 SAR Sections 6.2.2.15.1.13. Table 151.13-5 FIGURES: 6.2-17 M-dddM_ Keith Butler Certified Reviewer's Signature 6/18/98 Printed Name Date Reviewer's certification expiration date: _11/21/98 Assistance provided by: Printed Name Scope ofAssistance none Date Search Scope Review Acceptability (NA, if performed by Technical Revi . (( % hms eA/ Cettifiedneviewer's Signature 7-MP Printed Name Date .
974346P201 ARKANSAS NUCt. EAR ONE Pace 3 FORM TITLE: FORM NO. REV. 10CFR50,59 DETERMINATION 1000.131A 3 ! l ENVIRONMENTAL IMPACT DETERMINATION (UNIT 1 and UNIT 2) A H 6 P2 01 PAGE 5 REV 0 Document No. 974346P201 Rev/ Change No. O Complete the following Detemunation. If the answer to any item below is "Yes", an Environmental Evalu required. See Section 6.1.4 for additional guidance. l Will the Activity being essluated: i Yes ,N,_g O @ i Disturb land that is beyond that initially disturbed during construction (i.e., new construction' buildings, creation or removal of ponds, or other terrestnal impact)? See Unit 2 SAR 2.5-17. This applies only to areas outside the protected area.
@ Increase thermal discharges to lake or atmosphere?'
Increase concentration of chemicals to cooling lake or atmosphere through dischargei tower? 0 @ Increase quantity of chemicals to cooling lake or atmosphere through discharge canal or
@ Modify the design or operation of cooling tower which will change drift characteristics?
I
@ Install any new transmission lines leading offsite?
@ Change the design or operation of the intake or discharge structures?
O @ Discharges any chemicals new or different from that previously discharged?
@ Potentially cause a spill or unevaluated discharge which may effect neighboring soils, water or ground water?
@ Involve burying or placement of any solid wastes in the site area which may effect ru surface water orground water?
O @ Invohr incineration or disposal of any potentially hazardous materials on the ANO site? O @ Result in a change to nonradiological effluents or licensed reactor power level? O @ Potentially change the type or increase the amount of non-radiological air emissions from the ANO site.
974346P201 ARKANSAS NUCLEAR ONE FORM TITLE: Page 1 FORM NO. REV. 10CFR60.59 SAFETY EVALUATION 1000.131B 3 This Document contains 3 Pages. Document No. 974346P201 Rev.lChange No. 0 10CFR50.59 Eval. No. Fi44M-lC Title 2PSV-5697 Flange Addition (Assigned by PSC) introduction This Plant Change will add a flange pair on each side of Containment Spray relief va . do not change the function or operation of the system. Un revised to reflect the additional flanges added by this Plant Change. , . piping joints are welded except the containment spray pum This section leakage willin be as calculated changed Calculation by this Plant Change. The number of flanges is a facto 97-R-2002-01. shown in SAR Table 15.1.13 5 and affects the dose calculationsSection 15.1.13 is the Large Break LOCA Accident. as discus flanges. Following are excerpts from the response:ER 9743461201 has b on the pressure relief valve 2PSV-5697 willbe minimal." The response als cun'entlyin revision to incorporate these and other changes. Pending the results of this calcu Table 15.1.13-5 willbe revised accordingly.' Based on this response, it is concluded that this Plant Change will have negligible affect on consequences of a LOCA as discussed in SAR section 15.1.13.4. include the SAR change to paragraph 6.2 2.2.1 and the change to SAR , . . y Figu necessary changes to SAR Table 15.1.13-5 or to the dose rate calculations as discussed in will be initiated by NED as a result of revision to Calculation 97-R-2002-01. . performed by NED for the resulting SAR changes. An additional 50.59 Review will be . Nuclear the Engineering above discussion. Design is included as a required review for this Plant Change to ATTACHED. EACH QUESTION MUST BE ANSWE CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES GUIDA . If the answer to any question on this form is "Yes,' then an unreviewed safety question i to all questions is *No," then the proposed change does not involve an unreviewed safety que . 1. Will the probability of an accident previously evaluated in the SAR be increased 7 YesO No @ { The flanges being added are passive components and do not change the function or The piping system will remain qualified in accordance with the ASME Section ill, Class 2 C . C;tegory ev:luated I and in the SAR.other applicable loads. This modification does not affect any of the initiato PL 97439;,p2.ol Paa l' Rzv. D
974346P201 ARKANSAS NUCLEAR ONE
)
FORM TITLE: Paoe R
- FORM NO. REV.
10CFR50.59 SAFETY EVALUATION 1000.131B 3 2. Will the consequences of an accident previously evaluated in the SAR be increased? Yes O No @ ; This modification will add a flange pair which introduces a potential leak path in the Conta The number of flanges is a factor in determining the ECCS leakage as determined in Calculat The calculated leakage from Calculation 97 R 2002 Ot is shown in SAR Table 15.1.13-5 and aff calculations as discussed in SAR section 15.1.13.4. Section 15.1.13 discusses the Large Breakj ER 9743461201 has been issued from Nuclear Engineering Design which concludes that flanges on the pressure relief valve 2PSV-5697 will be minimal. Based on this response, it is co)j dose consequences of a LOCA as discussed in SAR section 15.1.13.4 will not be increased. I 3. Willthe probability of a malfunction of equipment Important to safety be increased? . YesO No @ - The flanges being added by this modification package w!!! not change the function or opera s The flanges on the safety related suction side of the valve are constructed in accordance wit lit, Class 2 Code of the material as specified for the applicable line class 2GCB-74 in Specifica The piping system is qualified in accordance with the ASME Section ill, Class 2 Code for the ad flanges. Since the system function and operation will not change, and the flanges meet all th important to safety to be increased. requirements of the system, this modification will not 4. Will the consequences of a malfunction of equipment important to safety be increased? YesO No @ The flanges being added by this Plant Change are passive piping components that do not cha operation of any equipment important to safety. The addition of the flanges will not change the meth or have any affect on the consequences of a malfunction of any equipment important to safety. 5. Will the possibility of an accident of a different type than i any previously evaluated in the SAR be created? l YesO No @ l The flanges being added by this Plant Change are passive piping components that do not ch operation of any system. The addition of the flanges meets the design requirements for the system of the flanges does not cause any condition that is different than the existing system such that an different type than previously evaluated in the SAR could be created. 6. Will the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR be created? YesO No @ This Plant Change does not constitute a functional change in any system as evaluated being added are passive piping components which do not have any interaction with any eq safety in any way that is different than the existing configuration. There are no new postulated fail any equipment important to safety as a result of this modification. 7. Will the margin of safety as deffned in the bases for any technical specification be reduced? Yes O No @ There are no margins of safety as defined in the bases for any technical specification th to the addition of flanges as installed by this Plant Change. P C 9 7 'f3 % P20s Pm 7 REVO
974346P201 FO RM TITLE: ARKANSAS NUCLEAR ONE Pace 3 FORM NO. REV. 10CFR60.59 CAFETY EVALUATION 1000.1313 3 N Nd bk Certified Reviewer's Signature M. Keith Butler 72398 Printed Name Date Reviewer's certification expiration date: 11/21/98 Assistance provided by: Printed Name NA Scope of Assistance Date PSC review by: Date: _ NbI i 1 i l l I
AnuunAb NUGLtAH ONt; FORM TITLE: FORM NO. REV. 10CFR50J0 DETERMINATION 1000.131A 2 PC.2,3 Page 1 of 3 Document No. 974369P201 RevfChange No. ,,,0 Title 2P3A and 2P3B discharoe valves limit switch contact modification. Will the proposed Activity:
- 1. Require a change to the Operating License including:
Technical Specifications (excluding the bases)? YesO No@ Operating License? YesO No@ Confirmatory Orders? YesO No@ 2. Result in information in the following SAR documents (including drawings and text) being (a) no longer true or accurate, or (b) violate a requirement stated in the document: Core Operating Limits Report YesO No@ SAR (multi-volume set for each unit)? Yes@ nod QAMO?* YesO No@ E-Plan?* YesO No@ FHA
. YesO No@
Bases of the Technical Specifications? YesO No@ ! NRC Safety Evaluation Reports? YesO No@
- 3. Involve a test or experiment not described in the SAR?
YesO No@
- 4. Result in a potential impact to the environment? (Complete Environmental impact Checklist of this form.)
YesO No@
- 5. Result in the need for a Radiological Safety Evaluation per section 6.2.4.A?
YesO No@
- 6. Result in any potential impact to the equipment or facilities utilized for Ventilated Storage Cask activities per Section 6.2.4.87 YesO No@
Basis for Determination: See continuous Page. , Changes to these documents require an evaluation in accordance with 10CFR50.54. See Section 6.2.1.B.
mese ARKANSAS NUCLEAR ONE FORM TITLE: FORM NO. RElf. 10CFR60.89 DETERMINATION 1000.131A 2 PC 2.3 Page 2 of,3 Document No. 974369P201 Rev./ Change No. 0
References:
List sections reviewed in the Licensing Basis Documents, specified in questions 1,2 and 3. If a keyword search was done on LRS, "all" may be entered under "Section" with the keyword (s) u in parentheses. Controlled hard copies of the documents sha!! be reviewed as computer based searches such as LRS are not controlled and search text only, not figures or drawings. Attach a completed LDCR if LBD changes are required. Document Section ANO-2 Technical Specification ALL (Cire. Water System, Condenser) ANO-2 SAR ALL (Cire. Water System, Condenser) ANO-2 SER ALL (Cire. Water System, Condenser) FHA ALL (Cire. Water System, Condenser) QAMO and E-Plan ALL (Cire. Water System, Condenser) Unit-2 TS Bases ALL (Cire. Water System, Condenser) l cm . w a, a Nick Mehta 09-10-97 Certified Reviewers Signature Printed Name Date Reviewers certification expiration date: 05-05-99 Assistance provided by: Printed Name Scope of Assistance Date i P.C 9 743 69 P 201 PAGE 16 REVO
o, . ,..,. ..===. w,n. FORM TITLE: FORM No. REV. 10CFR60.69 DETERMINATION 1000. 31A 2 Pc4.3 Page 3 of 3 ENVIRONMENTAL IMPACT CHECKLIST (UNIT 1 and UNIT 2) Document No. 974369P201 RevlChange No. O Complete the following checklist. If the answer to any checklist item is "Yes", an Environmental Evaluation is required. See Section 6.2.1.E for additional guidance. Willthe Activity being evaluated: Yes g O @ Disturb land that is beyond that initially disturbed during construction (i.e., new construction of buildings, creation or removal of ponds, or other terrestrial impact)? See Unit 2 SAR Figure 2.517. This applies only to areas outside the protected area. O a increase thermal dischames to lake or atmosphere? O B increase concentration of chemicals to cooling lake or atmosphere through discharge canal o tower? O g increase quantity of chemicals to cooling lake or atmosphere through discharge canal or tower? O O Modify the design or operation of cooling tower which will change drift characteristics? O E Install any new transmission lines leading offsite? O B Change the design or operation of the intake or discharge structures? O E Discharges any chemicals new or different from that previously discharged? O B Potentially cause a spill or unevaluated discharge which may effect neighboring soils, surface water or ground water? O E Involve burying or placement of any solid wastes in the site area which may effect runoff, surface water or ground water? O E involve incineration or disposal of any potentially hazardous materials on the ANO site? O E Result in a chant,., so r onradiological effluents or licensed reactor power level? O E Potentially change the type or increase the amount of non-radiological air emissions from the ANO site, PC 974369P201 PAGE 17 REV O
ARKANSAS NUCLEAR ONE FORM TITLE: FORM NO. REV. 10CPRSS.59 REVIEW cotmNuATioN PAGE 1900,131C 2 Page,lof_2 Document No. 9'14369P201 Rev> Change No. ! pg g gg 10CFR50.59 Review Continuation Pace LCP 95-6004 was implemented to revise the logic for the circulating water system pump discharge valvas during 2R11 refueling outage. The change involved one of the trip funl circulating pumps. The change was that with both pumps running and both discharge closure of one discharge valve will trip its associated pump. This change was im intamallimit switch LS6 from the discharge valve into the trip circuit of the associated provides the positive proof of discharge valve closure which generates a respective pum CR-2 97-0150 described that the valve 2CV-1224 limit switch contact LS8 opens before l contact LS6 closes. This prevents 2P3A from stopping or tripping per the above discus LS6 and LS8 are located at the same rotor. This modification will move the pump trip inte limit switch LS6 to LS9 and set LS9 to close at 97% to 99% of close travel or before LS LS9 and LS11 are on the same rotor, the function of LS11 will be moved to LS13. Hi-Lo interlock setpoint during valve stroke from 25% to 17%. The limit switch contacts being changed by this modification will have no affect on th circuit of the valves. However, to prevent excessive runout and reverse pump rotatio and accidental trip-out of a pump, the discharge valve is currently programined to clos to 25% open (approx. 9 seconds) and then complete closing the remaining 25% at s 21 seconds) for a total closing time of 30 seconds. With the proposed change, the proces close at fast speed to 17% (approx.10.3 seconds) and then complete closing the remain slow speed (approx.14.4 seconds) for a total closing time of approximately 25 seconds. T closing time will help ensure that pressure is maintained in the system header, this allow vacuum to be maintained. STEP 1. BASES FOR DETERMINATION Question 1. NO ; No change is required to the Unit-2 Technical SpeJfication (TS). Circulating Water pum discharge valves are not directly addressed, and Technical Specifications do not provide detail to address this modification. l No Unit-2 Confirmatory Orders or Operating License were found to address this modification. Question 2. YES. The modification will have an impact on Unit-2 SAR. This Plant Change revise the functional description in existing logic for the circulating water pumps and its discharge valves logic. ; changes will have an effect on SAR Figure 10.4-4. The modification will not result in revision being necessary for the Unit-2 NRC Safety EvalI ' Reports, QAMO, E-Plan and bases for technical specifications. None of these documeI sufficient detail to address this modification. I
ARKANSAS NUCLEAR ONE FORM TITLE: FORM NO. REV. 10CFRSO.58 ftEviEW CONTINUATION PAGE 1000.131C 2 I Page 2 ofj l Document No. 974369P201 RevlChange No. 1 _10CFR50.59 Review Continuous Pace Question 3. No. l i The proposed modification changes the pump stop and discharge valve close controls to enha system reliability. It does not change the function orinterface to other systems. Therefore, this modification affect doesof not safe operation involve the plant. a test or experiment not described in the Unit-2 SAR tha I Question 4. NO. I This modification will not change the physical area or have any impact upon the environment as evaluated on page 3 of this review. ) Question 5. No. This Plant Change will not affect the requirements of section 6.2.4.a of procedure 1000.131. Question 6. No. This Plant Change will not affect the requirements of section 6.2.4.b of procedure 1000.131. t P.C 9 74369 P201 l P. AGE 19 REV 0 1 1
NUCLEAR ONE FORM TITLE: FORM NO. REV. 10CMtse.se EVALUATION 1000.1313 2 PC 974369P201. Page ,1 off PAGE 2o REV 0 10CFR50.59 Eval. No.__ F F & 0-Ni (Assigned by PSC) Document No. 974369P201 RevlChange No. A Title 2P3A and 2P3B @ charae valves limii switch medification. ATTACHED. EACH QUESTION MUST BE AN CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES . If the answer to any question on this form is "Yes," then an unreviewed safety question is to all questions is "No,".then the proposed change does not involve an unreviewed . If the answer safe 1. Will the probability of an accident previously evaluated in the SAR be increased? Yes O No @ The SAR accident analysis chapter 15 covers the accidents that may be related to the circulating water system. Section 15.1.7 addresses a turbine trip due to loss of condenser vacuum and section 15.1.28 address the loss of condenser due to failure of the circulat water system. The proposed modification is built around enhancing system reliability. This modification will actually reduce the stroke time of the valve closure from approximately 30 seconds to approximately 25 seconds. The actual stroke time is changed in the closed direction only. This faster closing time will help ensure that pressure is maintained in the system header, thus allowing condenser vacuum to be of a valve failure or a failure of the system to maintain con based upon this evaluation the probability of an accident of previously evaluated in the St.R will not be increased. 2. Will the consequences of an accident previously evaluated in the SAR be increased? Yes O No @ The consequences of the loss of condenser vacuum accident listed in chapter 15.1.7, turbine trip due to loss of condenser vacuum and section 15.1.28, a loss of condenser vacuum due to failure of the circulating water system are not altered by the proposed 'l modification on the control system of the pump and its discharge valve. The proposed modification as previously described enhances the system reliability and will not alter a assumptions previously made in the evaluation of chapter 15 accidents. In addition, the change in the stroke time of the discharge valve closure does not create a new pathwa for release of radioactive material, nor it will affect any barriers which mitigate dose to the public or an increase of the onsite doses. Therefore, the consequences of an accident previously evaluated in the SAR are not increased. 3. Will the probability of a malfunction of equipment important to safety be increased? Yes O No @ l'n SAR chapter 15, loss of condenser vacuum was evaluated. The proposed covers the circulating water pumps and their discharge valves logic and controls. The circulating water system is the primary heat sink for the condenser, and does nut serve or connect to any safety related equipment. Important to safety related equipment could not be impacted by the proposed change, nor could it directly impact another piece of equipment which is not important to safety equipment. This proposed change will not
FORM TITLE: ARKANE'AS NUCLEAR ONE
%0CFR60.89 EVALUATION FORM NO. REV.
1000.1313 2 replac'e tfie important.to existing safety will not becomponent increased. from the system. Therefore, the probability of equ t P. AGE '2.1 REV0 4. Will be the consequences increased? of a malfunction of equipment important to safety Yes O No @ As stated la answers to the above questions, the circulating water system is used for steam corypensation in the condenser. The electric power source for the circulating wa diesel generator. The proposed change to the stroke time closed direction coes not alter the valve or system function. The modification does not ' create new malfunctions of equipment important to safet in an increased radiological release dose consequence. y and therefore could not result 5. Will the possibility of an accident of a different type than any previously evaluated'in the SAR be created? Yes O No @ The propoIed change covers the circulating water pumps and their discharge va Loss of circulating water system is bounded by the accident analysis in SAR chapter 1 as described in above answers. This modification change the valve closing stroke time time will help ensure that pressure is maintained in the s condenser; vacuum to be maintained. The proposed change enhances system reliab and direct (y or indirectly would not create new situations different from those considered by previous analysis. Therefore, the proposed modification will not create an accident of a different}ype than previously evaluated in the SAR. 6. Will the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR be created?
~
Yes O No @ The proposed modification covers control and logic and does not add any new equipment to the circulating water system. As previously stated, the circulating water system is solely used for condenser service and is not backed by emergency diesel and therefore the possibility of a malfunction of equipment im different type than previously evaluated in the SAR will not be created.
~
7. Will the margin of safety as defined in the bases for any technical specification be reduced? Yes O No @ No margins' were found in the bases for the technical specifications. Therefore, there is no change in the margin of safety, l\A. > eA t ot Certified Reviewer's Signature Nick Mehta 09-10-97 Printed Name Date R; viewer's certification expiration date:_ 05-05-99 Assistance provided by: l Printed Name 1 Scope of Assistance Date i PSC review by: Date: S O
AMKANSAS NUCLEAR ONE Pace i FORM TITLE: FORM NO. REV. 10CFR50.59 DETERMINATION 1000.131 A 3 PC-1
.. 3 l-This Document contains 4 Pages.
Document No. _ER 974820P201 RevdChange No. O Title 2 TIS-5413 SETPOINT CHANGE FOR SFP Hi TEMPERATURE ALARM Brief description of proposed change: This modification involves a change to the Spent Fuel (SFP) High Temperature Alarm setpoint, which is generated by 2 TIS-5413. The alarm, which was previously set at 160'F, is being moved in the conservative direction to alarm at 150*F. Also, in conjunction with the modification, a change is being made to resolve a discrepancy noticed in the ANO-2 SAR. (Continued on Page 2) 1 Will the proposed Activity:
- 1. Require a change to the Operating License including:
Technical Specifications (excluding the bases)? Yes O No @ Operating License? Yes O No 2 Confirmatory Orders? Yes O No @ 2. Result in information in the following SAR documents (including drawings and text) being (a) no longer true or accurate, or (b) violate a requirement stated in the document: 1 SAR (multi-volume set for each unit)? Yes @ No O Core Operating Limits Report? Yes O No @ Fire Hazards Analysis? Yes O No @ Bases of the Technical Specifications? Yes O No @ Technical Requirements Manual? Yes O No @ NRC Safety Evaluation Reports? Yes O No @
- 3. Involve a test or experiment not described in the SAR?
(See Attachment 2 for guidance) Yes O No @ i 4. Result in a potential impact to the environment? (Complete Environmental Impact Determination of this form.) Yes O No @
- 5. Result in the need for a Radiological Safety Evaluation per section 6.1.57 Yes O No @
- 6. Result in any potential impact to the equipment or facilities j utilized for Ventilated Storage Cask activities per Section 6.1.67 Yes O No @
- 7. Involve a change under 10CFR50.54 for the following SAR documents per Section 6.1.77 QAMO?
Yes O No @ l E-Plan? Yes O N6 @
)
ARKANSAS NUCLEAR ONE FORM TITLE: Paog 2 FORM NO. REV. 10CFR80.89 DETERMINATION 1000.131 A 3 PC 1 Document No.- ER 974820P201 p g: i :: , A Rev> Change No. 0 Basis for Determination (Questions 1,2, & 3): reflect the instrument's original 'as built
- range of 50-300*F in lieu of 50 200*F.A chang This is a setpoint change only to a contml room alarm. The alarm is used to alert cont abnormal temperature in the SFP. There are no other physical additions, deletio equipment. The function of the component is alarm and local indication (via local indicato related temperature control functions or equipment interlocks.
The Safety Analysis Report is marked "yes' because a Fuel Pool Temperature ins i incorrectly stated in Table 9.1-4. A Licensing Document Change Request (LDCR) is includ modification package to change the range to 50-300*F. 1 maintaining the pool temperature at or below 150'F for maxi However, there were no references to a value for a high temperature starm.
}
There were no references found in any other documents relative to a SFP high . tempi are no changes required to Operating License documents or other SAR documents. Also, thj not result document. in information in these documents being no longer true or accurate or violatei { O Proposed change does not require 10CFR53.59 Evr.iustion per Attachment 1, item # (if checked, note appropriate item #, send LDCR to Licensing). Search Scope: List sections reviewed in the Licensing Basis Documents specified in questions 1,2 a was done on LRS, "all" may be entered under "Section" with the keyword (s) used in pare Attach and distribute a completed LDCR per Section 6.1.2 if Document
, Section LRS: ANO-2 Technical Specifications ALL (Soent Fuel Pool. Hioh Temperature)
ANO-2 Operating License ALL (Soent Fuel Pool. Hioh Temperature) ANO-2 Confirmatory Orders ALL (Spent Fuel Pool. Hioh Tempareture) ANO-2 SAR QAMO ALL (Spent Fuel Pool. Hioh Temperature) E-Plan ALL (Soent Fuel Pool. Hioh TsmDarature) FHA ALL (Soent Fuel Pool. Hioh Temperature) ALL (Smnt Fuel Pool. Hioh Temperature)
ARKANSAS NUCLEAR ONE
. FORM TITLE: Pace 3 FORM NO. REV. I 10CFR50.59 DETERMINATION 1000.131 A 3 PC-1 Document Section i,.. I 1 SER ALL (Soent Fuel Pool. Hiah Temperature) l ANO-2 Tech Spec. Bases l ALL (Spent Fuel Pool. Hiah TemDerature)
COLR ALL (Spent Fuel Pool. Hiah TemDerature) MANUAL SECTIONS: Section 9.1.3. Table 9.1-4 ' FIGURES:
)
i _N. L3 d W k %0) Stephen L. McKissack 1/14/98 Certihed Reviewers Signature Printed Name Date Reviewer's certification expiration date: 6/10/99 Assistance provided by: Printed Name Scope of Assistance i Date ! l i Search Scope Review Acceptability (NA, if performed by Technical Reviewer per 1000.006) MccJ k Certified Reviewers Signature
.I M b m so A 1 7';hI Printed Name Date I O
l l i I
ARKANSAS NUCLEAR ONE FORM TITLE: PaDe o FORM NO. REV. 10CFR60.59 DETERMINATION 1000.131A 3 ENVIRONMENTAL IMPACT DETERMINATION ' (UNIT 1 and UNIT 2) Document No. ER 974820P201 RevdChange No. O Complete the following Determination. If the answer to any item below is "Yes", an Envi required. See Section 6.1.4 for additional guidance, t ( j Willthe Activity being evaluated: Yll!! .Ng O @ Disturb land that is beyond that initially disturbed during construction (i.e., new buildings, creation or removal of ponds, or other terrestrialimpact)? See Un 2.5-17. This applies only to areas outside the protected area. O. E increase thermal discharges to lake or atmosphere? ! O O
. tower?
increase concentration of chemicals to cooling lake or atmosphere through di O @ increase tower? quantity of chemicals to cooling lake or atmosphere through discharge c O E Modify the design or operation of cooling tower which will change drift characteristic O g instati any new transmission lines leading offsite? O E change the design or operation of the intake or discharge structures? O @ Discharges any chemicals new or different from that previously discharged? O @ Potentially water or groundcause water?a spill or unevaluated discharge which may effect neighboring O @ involvewater surface burying or water? orground placement of any solid wastes in the site area which may eff\ { O E involve incineration or disposal of any potentially hazardous materials on the ANO si O E Result in a change to nonradiological effluents or licensed reactor power level? O @ Potentially ANO site. change the type or increase the amount of non-radiological air emiss i 1 I I 1
ARKANSAS NUCLEAR ONE Page S FORM TITf.E: FORM NO. REV. 10CFR50.59 SAFETY EVALUATION 1000.131B 3 7 d This Document contains 1 Page. Document No. ER 974820P201 Rev1 Change No. _0 10CFR50.59 Eval. No. FNM Cl3 , Title 2 TIS-5413 Setpoint Chance for SFP HI Temperature Alarm (Assigned by PSC) , A WRITTEN RESPONSE PROVIDING THE BASIS FOR THE ANSWER TO EAC ATTACHED. EACH GUESTION MUST BE ANSWERED SEPARATELY. A SIMPLE S CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES GUIDANCE FOR If the answer to any question on this form is "Yes," then an unreviewed safety question is involved to all questions is *No,' then the proposed change does not involve an unreviewed safety question.
- 1. Will the probability of an accident previously evaluated in the SAR be increased?
Yes O No @
- 2. Will the consequences of an accident previously evaluated in the SAR be increased?
YesO No @
- 3. Wilt the probability of a malfunction of equipment important to safety be increased?
YesO No @ 4. Will the consequences of a malfunction of equipment important to safety be increased? Yes O No @
- 5. Will the possibility of an accident of a different type than I any previously evaluated in the SAR be created? I YesO No @ !
6. Will the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR be created? , YesO No @ 7. Will the margin of safety as defined in the bases for any technical specification be reduced? YesO No @ Mob J Pf-OVVma J Stephen L. McKissack 1/14/98
'Cetiified Reviewers Signature Printed Name Date R2 viewers certification expiration date: 6/10/99 Assistance provided by:
Printed Name Scope of Assistance Date PSC review by: A - Date: 19
ARKANSAS NUCLEAR ONE FORM TITLE: Page 2 . FORM NO. REV, 10CFR50,59 REVIEW CONTINUATION PAGE 1000.131C 3 F413E REV.< Document No. ER 974820P201 _ RevjChange No. _0
' 10CFR50.59 Review Continuation Paoe i Responses to questions on Page 1.
- 1. Will the probability of an accident previously evaluated in the SAR be increased?
j 2 TIS-5413. The High Temperature Alarm, which was pre conservative abnormal direction tejnoar=*nfo in mm to alarm SFP at 150*F. This is a setpoint change only to alert control room ol; anniamentIin addition, the modification package changes a discrepancy w) ment ran e in the ANO- ! (JAR JTable 9.1_-4J to ag_ree with the ori2inal _'as installed" inst tin}pnt ranceJHe
~
alarin and loi:ilindication (via localinc ; icatorIonly as there are no reistad temperatura ennjrolfunctione n j df% of_ span) considerations ar_e not crediLtadditional soulomant inter or deem.ed p factor for any accident _ conditions) No existing i accidents irTlfie SAR are initiated by failure of an alarm ~or inipropier operator actions ba the probability of an accident previously evaluated in the SAR will not be increased.
- 2. Will the consequences of an accident previously evaluated in the SAR be increased?
i Je .150*Efin addition, the modification packagt changes ~a 'discr
~
^
~
)
G. GAR, gable g.1-4] to_ aAree the pricinal "ps.irimtaHad" i h the instrurnefit ranDe trittie ANO-2 (nent_rance. is l hardware or physical c tanges to eERing equipment. ion involves no n w temperature condition (including the effects of instrument inaccuracy) will b Temperaturei Alarmf.Qll scale range (30ppQp*Mnd instrumen SAR will not be increased. credited for mitigating an accide61 condition. Thus, the conseque
- 3. Will the probability of a malfunction of equipment important to safety be increased?
Though the setpoint for 2 TIS-5413 is being lowered, Me function of the component, which room alarm and localindication, has not changed, re are no equipment changes, control functions or additional interiocksinvolved with this_m e range of 2 TIS-5413, as stated in the SAR bein Qiharged (o coire'ct3 dis'crepancyiorl accurately rqttle'ct ttagJngtrument's original The a armas built" ran setpoint is being lowered bf10*FOff3,which is weTwithin the d'esign operiting spec sca sons of the component. Therefore, the setpoint change or correcting the range discrepancy will have no effect on probability of equipment failure / malfunction.
- 4. Will the consequences of a n'alfunction of equipment important to safety be increased?
This to sw modiftcation lowers odif (consavative p grq the EFPp, Tsmisimiute Alarm fgu41$p*f dange)?Sn addet cation is correcting discrepancy tq asc_urately the instrument's as but Jhg SA _ There are no equipment etianges or controi~ functions invol lowenne the setpoint, the presence of an abnormal high temperature conditio isA . son. _ v _ --- Onaccuracy)will be alarmed earlier and thus, permit operators to take ainmore high temperature condition, Howawar the effects ti _ ofinstrume3 __ anae' Tin ~rninrrolino Ih' e'
. (accuracy _(1% spanDconsiderations,e SFP High Temperature Alarm,%ull scale e, rgand instrume t.f a malfunction of equipment impo nt to of credited for mitigating an accident condition. Thus, tne consequences safety is not affected.
- 5. Will the possibility of an accident of a different type than previously evaluated in the SAR be c This modification lowers (carimarvative r directiarn the SFP HiotLTemnerminre
~
A (_to 150*E!!iiadditi[on fanje_in the SAR. thefmojiification is co~rectingla discrepancy to accuratefy I ~ ref here are no equiftnint changes or control functions iWyolv~ed with this modification.
ARKANSAS NUCLEAR ONE FORM TITLE: Pace 3 FORM NO. REV. 10CFR50.59 REVIEW CONTINUATION PAGE 1000.131C 3 Changing the setpoint does not present a new accident initiator or failure mode. The setpoint c is within the design operating specifications of the_ component;and will be mart *Astomary calib practices. As a result, the setpoint change ohinstrument range discrepancy)oes not create the probability of a different accident than previously evaluated in the SAR. ~
- 6. Will the possibility of a malfunction of equipment important to safety of a different type than evaluated in the SAR be created?
The SFPpioh TemneratureMarm setoointAnly o,r,gyvides audible s _andf visualindicatigapf_a rat highi conditiomin addition, the modification is correcting a discrepancy to accurately reflect thelnstrument's a range in the SAR. There are no ejq imntrol.fynctions.plinterface_s with other equipme_nt , nyolved with this.mMe setpoint change beinfmade is within the design operating specifications of the component and will be made using customary calibration practices. No functional changes are b which mioht induce a malfunction of mouinment important to safety. Therefore, the setpoint change Giorrecting the instrument _ range i_n thje SA irnportant to safety ~than previously evTa usted in the SAR.'ll not create more of a possibility of a m
~
- 7. Will the margin to safety as defined in the bases of any technical specifications be reduced?
The SFP High Temperature Alarm setpoint is not mentioned in the technical specification bases. SFP High Temperature Alarm setpoint does not affect a margin to safety as defined in the Technical Specifications. PAGE- O _ REV./ ; e 9
ARKANSAS NUCLEAR ONE F M U I- W Hb ' W Page i FORM TITLE:
- FORM NO. REV.
10CFR50.59 DETERMINATION 1000.131 A 3 PC-1 Document No. PC 974899P201 RevdChange No. O Title 2R13 Service Water Pipe Replacement Brief description of proposed change: This plant change provides details relevant to the 2R13 Service Water Piping Replacement Modification. Approximately 533 feet of carbon steel piping has been identified as requiring replacement on the Loop 1/ Loop ll Service Water System. Approximately 8 feet of carbon steel piping will be replaced with stainless steel pipe on Loop ll Service Water supply line to Emergency Feedwater Pump 2P-7A. The criteria used for the areas chosen is based upon historical trending of the pipe wall thickness, system flow testing, and engineering evaluation. Specific locations are indicative of sections susceptible to relatively higher corrosion rates as determined by similar piping replacements previously performed. Will the proposed Activity:
- 1. Require a change to the Operating License including:
Technical Specifications (excluding the bases)? YesO No@ Operating License? YesO No@ Confirmatory Orders? YesO No@ 2. Result in information in the following SAR documents (including drawings and text) being (a) no longer true or accurate, or (b) violate a requirement stated in the document: SAR (multi-volume set for each unit)? Yes@ nod Core Operating Limits Report? YesO No@ Fire Hazards Analysis? YesO No@ Bases of the Technical Specifications? YesO NoS i Technical Requirements Manual? YesO NoS NRC Safety Evaluation Reports? YesO No@
- 3. involve a test or experiment not described la the SAR?
(See Attacilment 2 for guidance) YesO No@
- 4. ' Result in a potential impact to the environment? (Complete Environmental l
impact Determination of this form.) i YesO No@
- 5. Result in the need for a Radicaogical Safety Evaluation per section 6.1.57 YesO No@
- 6. Result in any potential impact to the equipment or facilities I
utilized for Venillated Storage Cask activities per Section 6.1.6? YesO No@ ! 7.- Involve a change ur. der 10CFR50.54 for the following SAR documents per Section 6.1,77 !
'QAMO?
YesO No@ E-Plan? YesO No@
l ARKANSAS NUCLEAR ONE ~ i FORM TITLE: Pace 2
. FORM NO. REV.
10CFR50.59 DETERMINATION 1000.131A 3 PC 1 Document No. PC 974899P201 Rev1 Change No. 0 l Basis for Determination (Questions 1,2, & 3): PC 974899P201 See Continuation Page PAGE 3/ REV 0 O Proposed change does not require 10CFR50.59 Evaluation per Attachment 1, item # (if checked, note appropriate item #, send LDCR to Licensing). Search Scope: List sections reviewed in the Licensing Basis Documents specified in questions 1,2 and 3. If a was done on LRS, "all" may be entered under"Section" with the keyword (s) used in parentheses. copies of the documents shall be reviewed (LRS is not verified and searches only text, no Attach and distribute a completed LDCR per Section 6.1.2 If LBD changes are required. Document Section LRS: ANO-2 Tech. Spec. ALL ANO 2 Operatina License Search words are listed on the ALL ANO-2 Confirmatory Orders continuation sheet for the LRS ALL searches. ANO-2 SAR ALL QAMO ALL E-Plar! ALL FHA ALL ANO-2 Tech. Spec. Bases ALL ANO 2 SER ALL MANUAL SECTIONS: TS 3.8.1.2 i TS 3.9.8.1 TS 3.9.8.2 SAR 9.2 FIGURES: 2SAR4 2.1 UWAhe W Certified Reviewer's Signature O W c$ 5~ T%t \/ Y2N 9k Pnnted Name / Date Reviewer's certification expiration date:
/0// 7 97 Assistance provided by:
Printed Name Scope of Assistance Trevena S. Jackson Research Date 3/23/98 Searc Scope Review Acceptability (NA, if performed by Technical Reviewer per 1000.006)
. $ IIaf _ Lee Printed R. %lsc4e 7l a c.i u Cdrtified Revie%er's Signyure Name
/ D6te
ARKANSAS NUCLEAR ONE Pact 3 FORM TITLE:
- FORM NO. REV. 4 10CFR50.59 DETERMINATION 1000.131 A 3 ENVIRONMENTAL IMPACT DETERMINATION 0 00 (UNIT 1 and UNIT 2) PAGE 32 REV 0 Document No. ,PC 974899P201 Rev1 Change No. O Complete the following Determination. If the answer to any item below is "Yes", an Environmental Evaluation is required. See Section 6.1.4 for additional guidance.
Will the Activity being evaluated: Yes Ng O @ Disturb land that is beyond that initially disturbed during construction (i.e., new construction of buildings, creation or removal of ponds, or other terrestrialimpact)? See Unit 2 SAR Figure 2.5-17. This applies only to areas outside the protected area. O O Increase thermal discharges to lake or atmosphere? O O Increase concentration of chemicals to cooling lake or atmosphere through discharge canal or tower? O 9 increase quantity of chemicals to cooling lake or atmosphere through discharge canal or tower? O @ Modify the design or operation of cooling tower which will change drift characteristics? O E install any new transmission lines leading offsite? O @ Change the design or operation of the intake or discharge structures? O @ Discharges any chemicals new or different from that previously discharged? O B Potentially cause a spill or unevaluated discharge which may effect neighboring soils, surface water or ground water? O @ Involve burying or placement of any solid wastes in the site area which may effect runoff, surface water or ground water? O O involve incineration or disposal of any potentially hazardous materials on the ANO site? O B Result in a change to nonradiological effluents or licensed reactor power level? O @ Potentially change the type or increase the amount of non-radiological air emissions from the ANO site.
ARKANSAS NUCLEAR ONE FcRM TITLE: Paon 4 FORM NO. REV. 10CFR50.59 DETERMINATION 1000.131A 3 Document No. PC 974899P201 Rev./ Change No. o PC 97 48 9 9P2 10CFR50.59 Review Continuation Pace ! l SEARCH SCOPE The fobowing is the LRS search word / phrase list: Service Water Piping w/10 Pip
- or Stainless Code Repair l SA*
2HBC* 2HCC* SWS w/10 Pip
- Responses to Determination Questions I Question 1:
The Tech. Spec., Operating License, and Confirmatory Orders were reviewed and revision. These documents do not get to a level of detail which would be impacted by this modification. The overall operation and function of the Service Water System has not ch Question 2: I A review of the 2SAR documents was conducted to determine whether t these documents being modified. The operation and function of the system will not chan configuration of the system will be the same except for the change from carbon steel to sta This change will affect the 2SAR documents due to a change in the P&lD (SAR Fig. 9.2 line class associated with the modification. Based upon this determination, a 50.59 evaluation required. Question 3: TR modification does not involve a test or expenment not described in the 2SAR. The p rep; aced required essentially to verify like-for-like and the operation of the system will not change. No un system performance. i Question 4: The response to all the questions on the environmental impact checklist is no. Therefore, th proposed change will not result in potential impact to the environment. Question 5: A review of the section 6.1.5 shows no impact by this modification. Therefore, this modification not result in a Radiological Safety Concern.
ARKANSAS NUCLEAR ONE Page S FORM TITLE: FORM NO. REV. 10CFR50.59 DETERMINATION 1000,131 A 3 Question 6: PC 974899P201 PAGE JY REV 0 Operation and configuration of the service water system to the 4 pent fuel pool cooling system remains unchanged. Therefore, there is no impact to the VSC program or associated equipment. Question 7: The steps in this modification will not make the QAMO or the E-Plan statements be untrue or inaccurate. The steps in this modification are below the level of detail within these documents.
FORM TITLE: ARKANSAS NUCLEAR ONE buC h Mt V [J Page 1 FORM NO. REV. 10CFR50.59 SAFETY EVALUATION 1000.131B 3 This Document contains 2 Pages. l Document No. ER 974899P201 RevlChange No. 0 ioCFRso.59 evai. No. m - cccs Title 2kI3 ddrVM6 W9kT '^"'" M 7dDh@WN _ A WRITTEN RESPONSE PROVIDING THE BASIS FOR THE ANSWER ATTACHED. EACH QUESTION MUST BE ANSWERED SEPARATELY. A S CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES GUIDA If the answer to any question on this form is "Yes," then an unreviewed safety question to all questions is "No," then the proposed change does not involve an unreviewed safety que . 1. Will the probability of an accident previously evaluated in the SAR be increased? Yes O No @ Changing 12 feet of Service Water piping from carbon steel to stainless steel will not affec performance of the Service Water system except to make it more reliable. The Loss of Se Water system accident identified in the SAR will not be affected by this piping chang failure of stainless steel piping versus carbon steel piping will be essentially be no different. Also, the probability of stainless steel piping failure is essentially no different increased.steel since stresses are still well below allowable, therefore the probability will n carbon 2. Will the consequences of an accident previously evaluated in the SAR be increased? Yes O No @ Changing 12 feet of Service Water piping from carbon steel to stainless steel'will on the consequences of any accident specifically the Loss of Service Water acc change Water. to stainless steel will still be bounded by the evaluation of any piping failure on Se 3. Willthe probability of a malfunction of equipment important to safety be increased? YesO No @ The probability of equipment failure will not be changed by the replacement of 12 feet of S Water pipe with stainless steel instead of carbon steel. The stainless steel will be more relia i' from a corrosion perspective. Stress allowable forthe new piping will be maintained within requirements. not increase. Based on this the probability of a malfunction equipment important to sa 4. Will the consequences of a malfunction of equipment important to safety be increased? YesO No @ The consequences of matfunction of equipment important to safety will not be increased s the replacement of 12 feet of Service Water pipe with stainless steel will have no new me for malfunction. Since the failure mechanism for the stainless steel piping is the same as the carbonifsteel change a leakpiping was tothe potential for a leak will be the same. Thus, dose consequences will n occur. 5. Will the possibility of an accident of a different type than any previously evaluated in the SAR be created? YesO No @ No new accident scenario can be postulated with the change of Service Water pipe from c steel to stainless steel. The pipe will still function as previously designed and the only fai mechanism could be leakage which is covered under the Loss of Service Water accident. All n piping will meet code requirements and allowable stresses will be below maximums.
ARKANSAS NUCLEAR ONE FAbE 3/o FORM TITLE: NE V [] Pcce 2 FORM NO. REV. 10CFR50.59 SAFETY EVALUATION 1000,131B 3 6. Will the pbssibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR be created? YesO No @ No new equipment malfunction can be postulated with the change of Service pipe from carbo steel tp stainless steel. A leak in the piping would actually be less likely to occur since the new pipe material will have better resistance to corrosion. 7, Will th4 m'prgin of safety as defined in the bases for any technicalipecification be reduced? YesO No @ The technical specification related to Service Water requires that two trains of Service Water operable.iThe bases for this is to ensure adequate cooling is available. The change of Servi Water pipe from carbon steel to stainless steel will not affect thes sbility of the Service Water system to provide adequate cooling. Work will be accomplished in a manner that will maintain required margins of safety.
- &Y / @; $ $ - )., u V 7f2 TV Certified Revjewer's SigrWse Printed Namp '
Date Reviewer's certific$ tion expiration date:
/c// 7 k 9
/ J Assistance providdd by; Printed Name Scope of Assistance Trevena S. Jackson Research Date S/17/98 k - _
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#4 a
9
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' ARKANSAS NUCLEAR ONE Page 1 I- FORM TITLE: FORM NO. REV. ! I 10CFR50.59 DETERMINATION 1000.131A 2 PC-2,3 Document No. 974904P201 Rev1 Change No. O Title ' 2 VET-2 Expansion / Compression Tank Vent Valve Addition Willthe proposed Activity: 4
- 1. Require a change to the Operating License including:
Technical Specifications (excluding the bases)? YesO No@ Operating License? YesO No@ , Confirmatory Orders? YesO No@
- 2. Result in information in the following SAR documents (including drawings and text) being ;
(a) no longer true or accurate, or (b) violate a requirement stated in the document: Core Operating Limits Report i YesO No@ SAR (multi-volume set for each unit)? Yes@ nod QAMO?' YesO No@ E-Plan? YesO No@ FHA YesO No@ Bases of the Technical Specifications? YesO No@ NRC Safety Evaluation Reports? YesO No@ l 3. Involve a test or experiment not described in the SAR? YesO No@
- 4. Result in a potential impact to the environment? (Complete Environmental impact Checklist of this form.)
YesO No@
- 5. Result in the need for a Radiological Safety Evaluation per section 6.2.4.B? YesO No@
- 6. Result in any potential impact to the equipment or facilities utilized for Ventilated Storage Cask activities per Section 6.2.4.87 YesO No@
Basis for Determination: see Page 4 l I. Changes to these documents requirs an evaluation in accordance with 10CFR50.54. See Section 6.2.1.B. ! l ! 1 ! 1 \ l l
ARKANSAS NUCLEAR ONE Paos 2 FORM TITLE: FORM NO. REV. i 10CFR50.59 DETERMINATION 1000.131 A 2 PC-2,3 Document No. 974904P201 RevlChange No. O
References:
List sections reviewed in the Licensing Basis Documents, specified in questions 1,2 and 3. If a , keyword search was done on LRS, "all" may be entered under "Section" with the keyword (s) used
!n parentheses. Controlled hard copies of the documents shall be reviewed as computer-based searches such as LRS are not controlled and search text only, not figures or drawings. Attach a l completed LDCR if LBD changes are required.
Document Section Unit 2 SAR, Unit 2 Tech Specs, All (service air, chilled water, chilled water pumps. 2 VET-2, 2VP-1 A, l Unit 2 SERs, Operating License, 2VP-18) Confirmatory Orders, QAMO, E-Plan, Ucit 2 SAR Figure 3.2-4 l n - $pDben[Avnn 9-3& 97 Sign ture Certifief Revie [ Printed'Namp Date Reviewer's* certification expiration date: [o/ .3 77 !
/
Assistance provided by: l Printed Name Scope of Assistance Date
ARKANSAS NUCLEAR ONE Page 3 FORM TITI.E: FORM NO. REV. 10CFR50.59 DETERMINATION 1000.131 A 2 PC-2,3 ENVIRONMENTAL IMPACT CHECKLIST (UNIT 1 and UNIT 2) Document No. 974904P201 Rev/ Change No. O Complete the following checklist. If the answer to any cher.klist item is "Yes", an Environmental Evaluation is required. See Section 6.2.1.E for additional guidance. Will the Activity being evaluated: - Yes M2 O E Disturb land that is beyond that initially disturbed during construction (i.e., new construction of buildings, creation or removal of ponds, or other terrestrialimpact)? See Unit 2 SAR Figure 2.5-17. This applies only to areas outside the protected area. O E increase thermal discharges to take or atmosphere? O E increase concentration of chemicals to cooling lake or atmosphere through discharge canal or tower? O E increase quantity of chemicals to cooling lake or atmosphere through discharge canal or tower? D E Modify the design or operation of cooling tower which will change drift characteristics? O E Instali any new transmission lines leading offsite? O E Change the design or operation of the intake or discharge structures? O E Discharges any chemicals new or different from that previously discharged? O E Potentially cause a spill or unevaluated discharge which may effect neighboring soils, surface water or ground water? O E involve burying or placement of any solid wastes in the site area which may effect runoff, surface water or ground water? i O E involve incineration or disposal of any potentially hazardous materials on the ANO site? O E Result in a change to nonradiological effluents or licensed reactor power level? O E Potentially change the type or increase the amount of non-radiological air emissions from the l ANO site. l A-
meerrtre even.s.can urne FORM TITLE: FORM NO. REV. 10cFR$0.80 REVN!W CONTWAAAT3oN PAGE 1000.131C 1 i Document No. 974904P201 Rev1 Change No. j. Page 4 10CFR50.59 Review Continuation Pace Description of Channe This modification will add service air valve 2SA-1006 which will be used to vent tank 2 VET water make-up is needed. This tank is used to maMtain a suction head for the main chill water pumps and to provide make-up water to the chill water system by using domestic water pressurized by service air. Domestic water pressure cannot overcome the service air pressure so current practice is to manually lift piessure relief valve 2PSV-3801 so that domestic water can enter the tank. The new vent valve will simplify this vperation and ensure repeatability of the tank's safety relief valve. Question 1. This modification will aot require a change in the Operating License documents. Question 2. Unit 2 SAR Figure 3.2-4 (P&lD M-2222 Sh.1) will need to be revised to show the addition of service air valve 2SA-1006 which will be used to vent tank 2 VET-2. No SAR text is affected by this change. . Question 3. This change will not result in any new or revised test or experiment. Question 4. This modification will not result in any change in radiological release practice and will not result in an additional impact to the environment. See checklist. Question 5. No Radiological Safety Evaluation per section 6.2.4.A is needed since this design change does not involve the processing of radioactive material outside of the Aux. Bldg., Reactor Bldg., or Low Level Radwaste Storage nor creates a new pathway outside of the monitored ventilation or drainage pathways. Question 6. This change does not affect Ventilated Storage Cask equipment or facilities.
c - ARKANSAS NUCLEAR ONE Page 1 FORM TITLE: F&RM ND. REV. 10CFR50.59 SAFETY EVALUATION 1000.131B 2 This Document contains 1 Page. Document No. 974904P201 Rev/ Change No. 0 10CFR50.59 Eval. No. fpts) ff ? -l*$7 (Assigned by PSC) Title 2 VET-2 Expansion / Compression Tank Vent Valve Addition A WRITTEN RESPONSE PROVIDING THE BASIS FOR THE ANSWER TO EACH QUESTION MUST BE ATTACHED. EACH QUESTION MUST BE ANSWERED SEPARATELY A SIMPLE STATEMENT OF CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES GUIDANCE FOR RESPONSE. If the answer to any question on this form is "Yes," then an unreviewed safety question is involved. If the answer to all questions is "No," then the proposed change does not involve an unreviewed safety question.
- 1. Will the probability of an accident previously evaluated in the SAR be increased? Yes O No @
- 2. Will the consequences of an accident previously evaluated in the SAR be increased? Yes O No @
- 3. Will the probability of a malfunction of equipment important to safety be increased?
Yes O No @ l 4. Will the consequences of a malfunction of equipment important i to safety be increased? Yes O No @
- 5. Will the possibility of an accident of a different type than any previously evaluated in the SAR be created?
l Yes O No @
- 6. Will the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR be created? YesO No @
- 7. Will the margin of safety as defined in the bases for any technical specification be reduced?
YesO No @ 0 . Stephen J. Lynn 9/30/97 , CertH ed Rev rs gnature Printed Name Date Reviewers certification expiration date: 6/3/99 Assistance provided by: Printed Name Scope of Assistance Date
/
n /
,_ i) M y/ ,
PSC review by: / /h Date: /8 [ y</ . < < r l
1 ARKANSAS Nucl.UR ONE 7.i 2 FORM TITLE: f . FORM N3D. REV. 10CFR50.59 REVIEW CONTINUATION PAGE 1000.131C 2
' Document No. 974904P201 Rev/ Change No. 0 10CFR50.59 Review Continuation Pace
- 1. Will the probability of an accident previously evaluated in the SAR be increased?
The addition of vent valve 2SA-1006 will not increase the probability of an analyzed accident since there are n accidents evaluated in Chapter 15 of the Unit 2 SAR related to the Service Air System nor the Chilled Water System. These systems are not required for a safe shutdown of the plant nor are the supply air and water from these systems needed for an ESF component to perform its intended safety function.
- 2. Will the consequences of an accident previously evaluated in the SAR be increased?
Vent valve 2SA-1006 will not increase the dose consequences of any analyzed accident. The valve insta on a non-radiological system and in a non-radiological area of the turbine building. No radiological barriers are affected by this change and no new pathways for the release of radiation are created.
- 3. . Will the probability of a malfunction of equipment important to safety be increased?
The installation of this non-Q service air valve will not have any impact on any equipment importan Service air is not used by any safety related component to perform its safety function. The expansi of the Chilled Water system which also has no safety related functions. Chilled water is not used b related corrponent to perform its safety function. 4. Will the consequences of a malfunction of equipment important to safety be increased? The installation of this non-Q service air vent valve can in no way affect offsite nor onsite dose consequ to malfunctions of equipment important to safety. Service air and chilled water are non-radiological system are not used for any plant response to an analyzed accident. The dose for personnel responding to accidents ca not be affected by this change, and plant access is not affected,
- 5. Will the possibility of an accident of a different type than i
any previously evaluated in the SAR be created? ' 1 The only accident that could result from this modification is the loss of service air from the 2 VET-2 expans tank caused by leakage of the new valve. Low pressure in the expansion tank could result in the loss of suc head for the Chilled Water Pumps. Since these pumps along with the entire Chilled Water System serve no safety related function, no credible accident can be created by this service air valve addition.
- 6. Wlit the possibility of a malfunction of equipment important l to safety of a different type than any previously evaluated in the SAR be created? .
i The only two systems related to this change are the Service Air and Chilled Water Systems and these s
. not provide air or water for any component's safety related function. i 7.
Will the margin of safety as defined in the bases for any technical specification be reduced? The System. Unit 2 Technical Specifications do not address any margins of safety for the Service Air or the Chilled W
n .e o ARKANSAS NUCLEAR ONE PeGe 1 FORM TITLE: FORM NO. REV. 10CFR80.89 DETERMINATION 1000.131 A 3 PC-1,2 This Document contains 4 Pages. Document No. 974991N201 Rev/ Change No. O 1 Title ANO 2R13 SU/BD FILTRATION SYSTEM TIE-lNS Brief descrip' ion of proposed change: ER 974991 n. quested a modification package to install a steam generator blowdown filtration system. The i filtration system will be needed for plant operation between 2P99 and 2R14 prior to the Unit 2 steam generator ! replacements. Yt.!s Nuclear Change Package 974991P201, will only address the necessary isolation and connection points thai will be made during 2R13 to allow the installation of the filtration system during power operations prior to 2P99. ER response 974991P202 will address the installation of the flitration system and all design and operation issues. Specifically this Nuclear Change Package will do the following: 1); install a 12 inch manually operated carbon steel butterfly valve,28D-32, in line 2HBD-750-12" to provide isolation between the condensate pump discharge and the start-up and blowdown demineralizers. This valve will be normally open, and will not change the operation or function of the system. I
- 2) install three 4-inch manually operated gate valves and piping for ise!ation and connection points for a start up i and blowdown filter skid that will be added later by ER 974991P202. The three gate valves will be used for inlet and outlet isolation and a bypass. The in-line valve,2DB-36, will be normally open, and therefore the function and operation of the system will not be changed by this mod package.
- 3) Install new piping supports 2HBD-754-H31 and 2HBD 754-H32.
Willthe proposed Activity:
- 1. Require a change to the Operating License including: '
Technical Specifications (excluding the bases)? YesO No@ l Operating License? YesO NoS Confirmatory Orders? YesO No@
- 2. Result in information in the following SAR documents (including drawings and text) being (a) no longer true or accurate, or (b) violate a requirement stated in the document:
SAR (multi-volume set for each unit)? Yes@ NoO l Core Operating Limits Report? YesO No@ l: Fire Hazards Analysis? YesO No@ Bases of the Technical Specifications? YesO NoS Technical Requirements Manual?
~ '$*'""'r >
O
~ ~ ' ' ' "
YesO No@ NRC Sa'ety Evaluation Reports? YesO No@
- 3. Involve a test or experiment not described in the SAR?
YesO No@ ! (See Attachment 2 for guidance)
- 4. Result in a potential impact to the environment? (Complete Environmental Impact Determination of this form.) YesO No@
- 5. Result in the need for a Radiological Safety Evaluation per section 6.1.57 YesO No@
- 6. Result in any potential impact to the equipment or facilities utilized for Ventilated Storage Cask activities per Section 6.1.6? YesO NoS
- 7. Involve a change under 10CFR50.54 for the following SAR documents
ARKANSAS NUCLEAR ONE Paos 2 --- FORM TITLE: ~ FORM NO. REV. 10CFR60.59 DETERMINATION 1000.131 A 3 PC-1,2 per Section 6.1.7? QAMO? YesO No C E-Plan? YesO No@ i I 1 1 t t a, g, W .. ny .- D%
'---TM
ARKANSAS NUCLEAR ONE Page 3 FORM TITLE: F8RM NO. REV. 10CFR50v89 DETERMINATION 1000.131A 3 PC-1,2 Document No. 974991N201 Rev1 Change No. O Basis for Determination (Questions 1,2, & 3):
- 1) The changes made by this Plant Change will not change the operation or function of the Start-Up and Blowdown Domineralizer system or the Steam Generator Blowdown or Condensate systems. There is nothing in the Operating License that will be required to be changed as a result of this Plant Change.
- 2) Unit 2 SAR Figure 10.4-7 will require revision to show the branch piping and flanges added by this PC.
- 3) There are no special tests or experiments involved with this Plant Change.
O Proposed change does not require 10CFR50.59 Evaluation per Attachment 1, item # (if checked, note appropriate item #, send LDCR to Licensing). Search Scope: List sections reviewed in the Licensing Basis Documents specified in questions 1,2 and 3. If search was performed on LRS, the LRS search index should be entered under "Section" with the search statement (s) used in parentheses. Controlled hard copies of the documents shall be reviewed (LRS is not verified and searches only text, not figures ordrawings). Attach and distribute a completed LDCR per Section 6.1.2 If LBD changes are required. Document Section LRS: start *uo. blowdown. demineraliz*. SGBS. SGBD MANUAL SECTIONS: 10.4.6.10.4.8.10.4.10 FIGURES: 1Q4-2 M. Keith Butler 10/15/98 Certified Reviewers Signature Printed Name Date Reviewers certification expiration date: 11/21/98 Assistance provided by: Printed Name Scope of Assistance Date i none-l Sea h Sc Review Acceptability (NA, if performed by Technical Reviewer per 1000.006)
! AW John Harvey /d/da Y CdttifiedPeviewertSignature Printed Name '
, Cfate l
1
, . 5' . I ., .a. [
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l ARKANSAS NUCLEAR ONE Page 4 FORM is:L.Es i FURM NO. REV. l 10CFR50.59 DETERMINATION 1000.131A 3 l ENVIRONMENTAL IMPACT DETERMINATION i (UNIT 1 and UNIT 2) Document No. 974991N201 RevlChange No. O Complete the following Determination. If the answer to any item below is "Yes", an Environmental Evaluatio required. See Section 6.1.4 foradditionalguidance. 1 Willthe Adivity being evaluated: 1 Y.ns No l O B Disturb land that is beyond that initially disturbed during construction 0.e., new construction of buildings, creation or removal of ponds, or otherterrestrialimpact)? See Unit 2 SAR Figure 2.5-17. This applies only to areas outside the protected area. O @ increase thermal discharges to lake or atmosphere? O E increase concentration of chemicals to cooling lake or atmosphere through discharge canal or tower? l O E increase quantity of chemicals to cooling lake or atmosphere through discharge canal or tower? O E Modify the design or operation of cooling tower which will change drift characteristics? ' O E Install any new transmission lines leading offsite? O @ Change the design or operation of the intake or discharge structures? O E Discharges any chemicals new or different from that previously discharged? O @ Potentially cause a spill or unevaluated discharge which may effect neighboring :wn=, t ,rface water orground water? O E involve burying or placement of any solid wastes in the site area which may effect runoff, surface water or ground water? O E involve incineration or disposal of any potentially hazardous materials on the ANO site? O @ Result in a change to nonradiological effluents or licensed reactor power level? i O @ Potentially change the type or increase the amount of non-radiological air emissions from the ANO site.
- ' E.
l
ARKANSAS NUC1. EAR ONE Pepe 1 FORM TITLE: FORM NO. REV. 10CFR50,59 CAFETY EVALUATlON 1000.1318 3 PC-2 This Document contains 2 Pages. Document No. 974991N201 Rev/ Change No. 0 10CFR50.59 Eval. No. FFJ0N Wi (Assigned by PSC) Title ANO-2R13 SU/BD Filtration System Tie-ins A WRITTEN RESPONSE PROVIDlNG THE BASIS FOR THE ANSWER TO EACH QUESTION MUST BE ATTACHED. EACH QUESTION MUST BE ANSWERED SEPARATELY. A SIMPLE STATEMENT OF CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES GUIDANCE FOR RESPONSE. If the answer to any question on this form is "Yes," then an unreviewed safety question is involved. If the answer to all questions is *No," then the pmposed change does not involve an unreviewed safety question.
- 1. Will the probability of an accident previously evaluated in the SAR be increased?
YesO No @ This modification will not change the function or operation of any system. A failure to any of the components added by this modification to the Start-Up and Blowdown Domineralizer system will not affect any initiators of any of the accidents evaluated in the SAR. Therefore, the probability of an accident previously evaluated in the SAR will not be increased.
- 2. Will the consequences of an accident previously evaluated in the SAR be increased?
YesO No @ This modification will not change the fundion or operation of any system or component. As stated in Unit 2 SAR section 10.4.10.3, "The Start-Up and Blowdown Demineralizer system has no potential radioactivity release path to
' the environment". Therefore, this modification will not increase the consequences of an accident previously evaluated in the SAR.
- 3. Will the probability of a malfunction of equipment important to safety be increased?
YesO No @ As stated in SAR section 10.4.10.3, the Start Up and Biowdown Demineralizer system has no safety related function, and failure of any component in the system would not affect the safe shutdown of the plant. This modification will not change the function or operation of the system. Therefore, the probability of a malfunction of equipment important to safety will not be increased by this modification.
- 4. Willthe consequences of a malfunction of equipment important to safety be increased?
YesO No @ As stated in SAR section 10.4.10.3, the Start-Up and Blowdown Domineralizer system has no safety related function, and failure of any component in the system would not affect the safe shutdown of the plant. This modification will not change the function or operation of the system and will do nothing that could affect any equipment important to safety. Therefore, the consequences of a malfunction of equipment important to safety will not be increased by this modification.
- 5. Willthe possibility of an accident of a different type than any previously evaluated in the SAR be created?
YesO No @ This modification package will add piping and valves to a non-safety related system, and the function and operation of the system will not change as a result of the modification. Nothing is being done by this mod package that will create the possibility of an accident of a different type than any previously evaluated in the SAR. d n . % . -. n.:.. L :
l ARKANSAS NUCLEAR ONE i FeRM TITLE: Page 2 FuRM NO. REV. { 10CFR60.59 SAFETY EVALUATION 1000.131B 3 PC-2
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- 6. Will the possibility of a malfunction ot squipment important to safety of a different type than any previour:y evaluated ]
in the SAR be created? YesO No @ As stated in SAR section 10.4.10.3, the Start-Up and Blowdown Demineralizer system has no safety re function, and failure of any component in the system would not affect the safe shutdown of the plant. Thi modification will not change the function or operation of the system. Therefore, the possibility of a malfunction equipment important to safety of a different type than any previously evaluated in the SAR will not be created
- 7. Will the margin of safety as defined in the basis for any technical specification be reduced? ,
YesO No @ There is no margin of safety as defined in the basis for any technical specification associated with thj affected by this modification package. ' Keith Butler 10/15/98 Certified Reviewers Signature Printed Name Date Reviewers certification expiration date: 11/21/98 Assistance provided by: Printed Name Scope of Assistance John Harvey gl/ Date Reviewed for accuracy 10/15/98 PSC review by: Date: 0 $ 4%
' . . ' .~.\_ .O ,._.1r. . _ ._Q_. ._;
anrumenonuws.csm unc FORM TITLE: FORM NO. REV. 10CFR80.59 DETERMINATION 1000.131A 3 PC 1 ) Page 1 of.D Document No. N5054P201 RevdChange No. 9 ' l Title RCS REFUELING LEVEL TUBING MODIFICATION Brief description of proposed change: See Continuation Sheet Willthe proposed Activity:
- 1. Require a change to the Operating License including:
P L 975054P201 Technical Specifications (excluding the bases)? P1GE 3 RE V 0 YesO No@ Operating License? YesO No@ Confirmatory Orders? YesO No@
- 2. Result in information in the following SAR documents (including drawings and text) being (a) no longer true or accurate, or (b) violate a requirement stated in the document: !
SAR (multi-volume set for each unit)? Yes@ nod Core Operating Limits Report - YesO NoS l Fire Hazards. Analysis? ' YesO No@ Bases of the Technical Specifications? YesO No@ Technical Requirements Maiiual? YesO No@ NRC Safety Evaluation Reports? YesO No@
- 3. Involve a test or experiment not described in the SAR?
(See Attachment 2 for guidance) YesO No@
- 4. Result in a potential impact to the environment? (Complete the Environmentalimpact Determination of this form.)
YesO No@
- 5. Result in the need for a Radiological Safety Evaluation per section 6.1.57 YesO No@ j
- 6. Result in any potential impact to the equipment or facilities utilized for Ventilated I Storage Cask activities per Section 6.1.6?
YesO No@
- 7. Involve a change under 10CFR50.54 for the following SAR documents per Section 6.1.7:
QAMO? YesO No@ an? YesO No@ O Proposed change does not require 10 CFR 50.59 Evaluation per Attachment 1, item # ,(if checked, note appropriate item #, send LDCR to Licensing).
ARKANSAS NUCLEAR ONE FORM TITLE: FORM NO. REV. 10CFR40.40 DETERMINATION 1000.131A 3 PC.1.2 Page 2 of.Q Document No. 975054P201 Rev./ Change No. A Basis for Determination fou==+!ons 1. 2 & 31: PL 975054P201 Search Scope: PAGE 4 RE V 0 List sections reviewed in the Licensing Basis Documents specified in Question 1,2 and 3. If a search was performed on LRS, the LRS search index should be entered under "Section* with the search statement (s) parentheses. Controlled hard copies of the documents shall be reviewed (LRS is not verified and search
-required.
text, not figures or drawings). Attach and distribute a completed LDCR per Section 5.1.2 If LBD change Document Section LRS: ANO-2 Tech. Snee. &Q ANO-2 Operatina i leanse
&Q ANO-2 Ceslline :V Orders &g ANO-2 SAR 84 QAMO &R ,
Ef140 ) AM Etf8 AM ANO-2 ma==s of the Tech. Specs.
&Q ANO-2 NRC SERs 84 (LRS Keywords are listed on the continuation sheet.)
MANUAL SECTIONS: TS 3/4.9.81 TS 3/4.9.8.2 TS 3/4.4.1.3 TS 4.0.5 SAR 9.1.3 SAR 9.1.4 SAR 9.3.6 SAR 5.2 SAR 5.1 SAR5.5.10 SAR 5.5.12 SAR 5.6.3 SAR 4.2.1.2.4.7 SAR 3.6.4.2 SAR 3.1.4 SAR 7.7.1.1.9 8AR 7.6.2.5 8AR 3.10.2.2.11 FIGURES: SAR 5.1-3 Rooer B. Rucker 10/4/98 dertified Reviewers Signature Printed Name Date Reviewers certification expiration date: 9/11/99 Assistance provided by: Printed Name N/,8 Scope of Assistance Date Search Scope Review Acceptability (NA, if performed by Technical Review per 1000.006) hW Certified Reviewers Signature Krw /rwmA ten,/f Printed Name Date 4
FORM TITLE: ! FORM NO. REV. I
, 10CFR50.58 DETERMalATION 1000,131A 3 Page 3 of.g ENVIRONMENTAL IMPACT DETERMINATION (UNIT 1 and UNIT 2)
P t 975 051 P201 Document No. 975054P201 RevlChange No. 2 PAGE f RE V 0 I Complete the following Determination. If the answer to any checklist item is "Yes", an Environmental Evaluatio ! is required. See Section 6.1.4 for additional guidance, i Will the Activity being evaluated: Yes No
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l O E Disturb land that is beyond that initially disturbed during construction (i.e., new construction cf bulidings, creation or removal of ponds, or other terrestrial impact)? See Unit 2 SAR Figure 2.5-17. This applies only to areas outside the protected area. O E Increase thermal discharges to lake or atmosphere? i O 2 increase concentration of chemicals to cooling lake or atmosphere through discharge canal or tower? O E increase quantity of chemicals to cooling lake or atmosphere through discharge canal or tower? O S Modify tne design or operation of cooling tower which will change drift characteristics? O G install any new transmiss!on lines leading offsite? O @ Change the design or operation of the intake or discharge structures? ! O E Discharges any chemicals new or different from that previously discharged? O E - Potentially cause a spill or unevaluated c'ischarge which may effect neighboring soils, surface water or ground water? l I 0- 2 Involve burying or placement of any solid wastes in the site area which may effect runoff, surface water or ground water? O E involve incineration or disposal of any potentially hazardous materials on the ANO site? l 0 3 Result in a change to nonradiological effluents or licensed reactor power level? l i O E Potentially change the type or increase the amount of non-radiological air emissions from the ANO site. l e
ARKANSAS NUCLEAR ONE FORM TITLE: FORM NO. REV. 1eCFRSS.80 REVIEW CONTINUATION PAGE - 1000.131C 3 Page 4 of.g Document No.' 975054P201 RevjChange No. 9 PC 975054P201 10CFR50.59 Review Continuation Pace PAGE 6 RE V 0 Brief description of proposed chanoe: This 10CFR50.59 determination / evaluation covers the design, installation and testing of the RCS Refueling Le modification. Since 2R6, several condition reports have been written on the RCS Refueling Level System. The RCS Refu Level System consists of two differential pressure transmitters (2LT-4791 and 2LT-4792) and a Tygon Tube. All three of these level instruments are tied to the top of the pressurizer for the dry reference leg, and all three are connected to the *A* hot leg for the variable leg. 2LT-4791 and the Tygon Tube utilize the same pressurizer tap, but have separate 3/8" tubing runs (inside the D-ring) from the pressurizer. 2LT-4792 utilizes a separate pressurizer tap with its 3/8" tubing run (outside the D-ring). A single 3/8" tubing run from the "A" hot leg tap is connected to the three level components near the instruments. These instruments have drifted low on several occasions, but on at least one occasion all three indications halted during a draindown. This modification will provide tubing / piping corrections to solve these problems. ER 975054P201 will replace all of the reference and variable leg tubing from its source to its component. The tubing size for the reference legs will be increased from 3/8" to 3/4" tubing, and the tubing size for the vari will be increased from 3/8" to 1/2* tubing. A second variable leg will also be added from the pipe connection to the transmitter. 2LT-4791 and 2LT-4792 will each have its own reference and variable leg tubing runs, and the Tygon Tube will tee from the 2LT-4791 reference and variable tubing. The post-modification testing criteria for this modification will be contained within the Design Change Sum testing section. These testing sections provides detailed instructions similar to general approved procedures for testing equipment / instrumentation. . The RCS Refueling Level indications are used to determine RCS level during reduced inventory conditions. These indicators can provide indication when the RCS level is between the bottom of the "A" hot leg and the of the pressurizer. When RCS level is below the bottom of the pressurizer, the normal RCS pressurizer level instruments are out of range low. If the head is removed from the reactor vessel, then the reactor vessel level ' monitoring system is disabled. During certain times of a refueling or maintenance outage, RCS Refueling Lev indication is the only RCS level indication available to 0;a, store The RCS Level Indicators are not part of the Shutdown Operations Protection Plan (SOPP), so Y Lese indier*;rs are unavailable other means will be used to satisfy the RCS Level requirement. RCS Lem is moe c.a. cal during a draindown activity than during stead state level conditions. Certain outage activities
.....cn are based on the availability of the RCS Refueling Level indicators, may be halted, but these indicators are not required for a safety-related function.
SEARCH SCOPE The following is the LRS search word / phrase list: 2Ll*4791 2Ll*4792 2LT*4791 2LT*4792 refueling w/5 level res w/5 level " hot leg" w/5 level tubing supports
' shutdown cooling" tubing details pressurizer w/5 refueling adc tubing requirements res w/5 refueling tubing
F FORM TITLE: I FORM NO. REV. 14CFRSS.58 REVIEW CONTINUATION PAGE 1000.131C 3 i i Page i of.Q Document No. 975054P201 RevJChange No. A Responses to Determination Questions: P C 975 054 P201 PAGE 7 % V 0 \ Question 1. 1 No changes to the Operating License will be required since this modification is structured to comply with the Operating License documents listed in question 1. The Technical Specifications 3/4.9.8.1 and 3/4.9.8.2 have specific operability requirements for the Shutdown Cooling (SDC) System during refueling operations. Technical Specification 3/4.4.1.3 also has specific opera requirements for the SDC System. The RCS Refueling Level indications are used to verify RCS level during certain levels / evolutions during reduced inventory, so the requirements of SDC level can be determined by the RCS Refueling Level Indications. Technical Specification 3/4.9.9 deals with Reactor vessel water level du fue! movement. RCS Refueling Level indications can also be used to verify this RCS level. This modification shall be scheduled, so SDC operability is not questioned, because RCS level indication is not available. The existing indications will not be disturbed until a window when using the RCS Refueling indications for RCS level is not required. This modification does not require any changes to these requirements, and these requirements shall be followed durini all phases of this modification. Question 2. The only License Based Document that is being impscted by this Plant Change is the SAR. SAR Figure 5.13,i drawing M-2230 SH 2 (RCS P&lD) is being revised by this Plant Change. A Licensing Document Change 4 Request is included in the Plant Change, and a 10CFR50.59 Evaluation is attached. No other LBDs were made untrue or inaccurate by this modification, nor did this modification violate any requirement stated in the LBDs. The information/ instructions in this modification are below the level of detail contained within t Table 9.3 25, " Shutdown Cooling System instrumentation Application *, provides a list of instruments for SDC, an the RCS Refueling indicators 2LI-4791 and 2LI-4792 are listed as control room level indications with a high ala , Also, the RCS Refueling Level Tygon Tubing is listed as local level indication. Section 7.7.1.1.9 of the SAR describes the system instrumentation and controls. This section lists "RCS refueling level
- as instrumentation provided to enable the operator to evaluate system performance and detect malfunctions. The SAR also contains general details for valves, the integrity of the RCS pressure boundary, pressurizer design parameters, general design criteria, tubing and piping. The details are the design bases for the current RCS Refueling Level design.
These same design bases were followed for the new design, so these requirernents shall still be true after the modification.
ARKANSAS NUCLEAR ONE FORM TITLE: FORM NO. REV. 1eCFR80.59 REVEW CONTINUATION PAGE . 1000.131C 3 Pagegof,j Document No. 975054P201 RevlChange No. 9 P C 975 051. P 201 Responses to Determination Questions: PA6E F RE V 0 Question 3. No testing is required by this modification other than typical Post-Modification testing. This modifica constitute a test or experiment not described in the SAR as defined by Procedure 1000.131. This modifica only provide normal detailed post-modification testing similar to approved generic ANO procedures. Question 4. This modification will not result in any adverse impacts to the environment. The generation of typica is not considered as an adverse impact to the environment. The operation of the plant will no way which will result in changes to the air, water or soil conditions of the site. Question 5. This change does not involve processing of radioactive material outside of Controlled Access. Question 6. This change does not involve any equipment used in handling Spent Fuel Storage Casks. Question 7. This modification'will not make the QAMO or the E-Plan statements to be untrue equipment / systems being modified are below the level of detail contained within these documents.
FORM TITLE: FORM NO. REV. j
, 10CFR40.59 EVALUATION 1000,131s 3 PC-2 Page1 ofj I P t 9 75 0 51, P 201 -
RE V 0 10CFR50.59 Eval. No. PdMM PAGE 9 (Assigned by PSC) Document No. 975054P201 RevJChange No. 9 Title ANO-2 RCS REFUELING LEVEL TUBING MODIFICATION A WRITTEN RESPONSE PROVIDING THE BASIS FOR THE ANSWER TO EACH QUESTION MUST BE ATTACHED. EACH QUESTION MUST BE ANSWERED SEPARATELY. A SIMPLE STATEMENT OF CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES GUIDANCE FOR RESPONSE. ' If the answer to any question on this form is "Yes," then an unreviewed safety question is involved. If the answer to all questions is "No," then the proposed change does not involve an unreviewed safety question.
- 1. Will the probability of an accident previously evaluated in the SAR be increased?'
Yes O No @ No credit is taken for the RCS Refuelina Level ind! cation in the acciden; analyses. This I med;isceiion dses not contain any eauipment that performs a sefetv-releted conimi function. nor l does it contain anv =ee!ement that is credited for automatic action. This madificetion does not i 101Prface with 5_ .;v-r=8=*=d eee!sment that is not !se8=+=d. This medu c=*!on will add isolation l eauisment which is safetv-reim**d. All new safetv-r='-'ad !!e!= tion eauisment wes desisned and I instelled usina the reouired criteria for the ace!!cetion of the eau lsT.ent. The eserebility of the RCS Refuelina Level indications ensures that adeauste indication and = nins is ave!!sble durina reduced inventorv cssditians for the promet detection of incorrect RCS l eve:. Promst d.iection of low /hiah l eve: will mduce the ea*ential for dam =ee to --*ctv-related eeanment. and is an intesral
;;;.T.ent in RCS level durina refuelina/ maintenance e"*=ce activities. This modification will enhance the availabilltv of the RCS Refuelina Levelindications. Even theueh RCS Refuel ins Level ID.(Mations are used as an ind!cetion of level for the Shutdswa Ces;!ns ISDC) Svei.ni. the operability of tba SDC System is not part of this PC. since RCS Refuelina Leve: does not provide inout into SDC Svet ni controls. The operation or failure of the RCS Refuelina Lave: is not an accident initiator to any of the accidents !!=*ed in the SAR. nor does it in;edece with any eauisment that is an initietor. Therefore. the probability of previousiv eve!s;ed a nr1 acciden's is not '
increased.
- 2. Will the consequences of an accident previously evaluated in the SAR be increased?
Yes O No C The consecuences of accidents discussed in the LBDs will not chance as a result of this I modification. This modification da=s not chance the current RCS Refuelins Level desion. This med;iication is oniv nrovidina an enhancement to the existina desion. The eauisinent and actions { issuised to mitianta each accident will be unaffected by this modification. Since the modification ! has been desianed with proper electrical and mechanical isolation /sensistion. siessure boundary analysis. sis;ns analysis and with seismic in;esiiiv. the new desian will not fall in a mode that will adverselv eTT-ci any safety function. The dose consecuences associr:d with sieviousiv evaluated j accidents will not be affected as a result of this modification. Therefore. the consecuences of I accidenis sievisusly evaluated in the SAR will not incr=ese.
ARKANSAS NUCLEAR ONE FORM TrrLE: FORM NO. REE 10CFRse.s0 EVALUATION 1000.1318 3 PC.2 w P C 975 051. P 201 Page 2 ofj PAGE/0 RE V 0
- 3. Will the probability of a malfunction of equipment important to safety be increased?
Yes No The eauipment affected by this modification does not interface fl.a. conirvl) with eauisment that is cons!dsred imasitent to safety excest for RCS steesure boundary components. The new isolation valves beina added for this med;;;c.iion will Drsvlde the reauired doub!e isolation for one of the RCS Refuelina tranemltt.iissss. (The vii.er leen already has double isolation val = instelled.) The new =l=;. s reaul einents. This L:ns. *nhins and fitiinas viese analyzed for the anolicable RCS pressure cogg ) med;T; = dss does not e;t.ct any eau lsinen; or enhllas that serfsims any cc .iivl or ;n; i;sck func"c-ns with s.;e's or non ;.;< rel;ted sve;.ms. The desian confinuration of the ad '=d eau sment is in accsidsace with ANO desian standards as desciiiied in the foll discussions. The current desian standards for interfacina with s&fe;<-related circuits includes
- .iiscel isoistisa and sessisiion to sievent srassection of a failure frcin a non-safety sve; in. The circuit d==!an of this modification is in Weina with those standards. (
e The current d==!an stanaards for interfacina with sei.iv-related nicias/tubina includes shvsical issist;sa and ;;E-arsison to cr. vest propaastion of a failure frcin a non-safety sve;-in. standards, The cloinaltubina das!cn of this modification is in keepino with those e The intearity of the safetv-related svei.ms/ components has been ensur.d by meetina
;;'ernic instellation sisadsrds ner anolicable approved ANO details and procedures.
Add li;onal fire !esd!na and heat loadina have all been arsserly add,ressed evaluated and found acceptable, e Batterv and diesel !aad!na was not affected by this modification. Fiessure boundarv intenrity was maintain by the selection of material / components that were cualified for the line cia;; that was modified. Pinina med;iscations were des!aned. evaluated, analvred. and approved followina current ANO acc=stsd srectices and stspdards. e Tubina to ninina connections v;;ie desisned evaluated. analyzed and approved foilswina cui..nt ANO accested practices and standards. e There are no new failure modes introduced to eauisment that is important to safety. Based of the above discussion. It can be determined that this modification will not increa probab!!ity of a malfunction of eau;sinent important to safety. 4. Will the consequences of a malfunction of equipment important to safety be increased? Yes O No 5 The oloina. tubina. fittinas and emblina installed by this modification does not increase reliance on sou;sissat imssitsnt to se;eis. As concluded in the response to nuestion 3.0. the shv=! call;h;iiical ccafisuisiion of this modification ensures that the probability of a malfunction of eau;sinent imssitsnt to sei.iv has not b;;n incraesed. Therefore the offsite dose consecuences mssociated with a malfunction of eauipment important to safety is not increased as a result of this modification.
Axevaane suut.;i TAR ONE FORM TITLE: FORM NO. REU.
. 10CFR60.80 EVALUATION 1000.131a 3 PC-2 Pt 975054P201 Page 3 of_;l 5.
PAGEll RE V 0 Will the possibility of an accident of a different type than any previously evaluated in the SAR be created? Yes O No @ The function of the eautoment e;;.cted by this modificstlan is not reouired for shutdown of the unit or m;; inst;ns rad!asctive r:t:::;: coolant ss===ere ;n;.srity. hcw;ver this mcdifice' ion dGes Ef;.ct maintain lns reac32r The lin==t on the RC5 sr::i.as beendary will be within the accepted parameters durina installation and t-st;ns. and efier the in= ="-*!on of this modification the RCS oressure boundarv will not be csmsiis.d. As W--d in the .,.. .dina cuestions. the desian resulreinen' of the RCS w&s ms;n;-lned with no excist; Gas acesid;ns to accesied and accroved ANO ould: lines. ft has been dea,sne;.m.d that the instsil:^isn of this mcd;;;cstion will have no naastive imoact on a s.: ;i-related sys'.m . or ceinssnent. The n;^-Ha'len of this mad lTication will also not chanse the wev Csei-;; ens will resased to an scEl dent. No cr.dit is *maran in the current accident analv;n for any emea-20F or manual actica by the RCS Reic;llns Level. Failure of the soul sir,ent in= 2nad by this modif; =2: On will not ci="= any see! den;. initiators. Anolic=h!e desian resu;rewen;. h:ve haan con =l der.d (see i;;nense to or=Elion 3.01 to ensure that systems imscrtant to cr. e^v are not leessid;med: therefsr.. It can be concluded that the possibility of an accident dlflerent fisin any Dr.viousiv analyzed in the SAR will not be created. 6. Will the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR be created? Yes O No @ All system desians for eauloment important to ssfeiv will remain the same. None of the circuits added circuit. by this des!an will be routed in such a manner to cause oronacation of a failure in a Class 1E None of the field-routed tet-ina will be routed in such manner to cause proceaetion of a failure in a ASME 111. Cl=== 1. 2 or 3 or a Se;ea.;c Class 1 structure /comssnent. The criteria for
- trical separation has been maint.ined and conserssiive adherence to ==!:mic reouirew.ents has been observed to insure ccmslimace. Also. this modif cstion does not ::: i icallv/ mechanically interface with eauipment important to s. isis. Therefore. the possibility of a malfunction of soulomant important to safety that involves an initiator or failure of a different tvoe than previously evaluated in the SAR has not t:::n created.
- 7. Will the margin of safety as defined in the basis for any technical specification be reduced?
Yes O No G The Technical Specifications h==as do not establish a marain of safety for the RCS Refuelino Level indications. The installation of the r.;w sisins. tubina. valves and associated fittinas will not affect or alter the existino Tech. Soec. reauir. men's nor be included in any new reouirements. Based on the absve si.iements. this modification will not reduce the maroin of safety as defined in the bases. Rooer B. Rucker 10/18/98 Certflied Reviewers Signature Printed Name Date Reviewers certification expiration date: 9/11/99 Assistance provided by: Printed Name M6 Scope of Assistance Date PSC review by: W Date: \\ Q.hb
mamm suur.m out: FORM TITLE: FORM NO. REV. 10CFRSO.88 DETERMINATION 1000.131A 3 PC.1 Page L of 9 Document No. 975109P201 RevlChange No. ,j Title Stator Water Coolina Conductivity Ce ector Ren!sceinsat and PS Snubbar installation Brief description of proposed change: This PC makes two chanoes to the Stator Water Coo llas system. The first chance is the ree!eces,,ea; of the exlsi;ns ccaductivity analyzers and d;;;;tlsa c;ll;. The ex;;;;as analyzers are broken and obsolete. so tt.sv are bslaa res!sced witti more msdera eaulsment. l The second chance is the addition of insi cmeat snut-t-ers in the tubina lines feedina i runback oressure switches. The add;i;sa is r.;;ded due to sisssure flue *uetions in the sve;es, which result in sourious ac*a=*!on of these switct.ss. t Will the proposed Activity:
- 1.
Require a change to the Operating License including: i Technical Specifications (excluding the bases)? YesO No@ i Operating License? - YesO No@ Confirmatory Orders? i l - YesO No@ l 2. ! Result in information in the following SAR documents (including drawings and text) being ! (a) no longer true or accurate, or (b) violate a requirement stated in the document: SAR (muP.i-volume set for each unit)? Yes@ NoO l Core Operating Limits Report YesO No@ ) Fire Hazards Analysis? YesO No@ i Bases of the Technical Specifications? YesO No@ Technical Requirements Manual? YesO No@ NRC Safety Evaluation Reports? YesO No@
- 3. Involve a test or experiment not described in the SAR?
(See Attachment 2 for guidance) YesO No@
- 4. Result in a potential impact to the environment? (Complete the Environmentalimpact Determination of this form.)
YesO No@ , 5. Result in the need for a Radiological Safety Evaluation ! per section 6.1.57 YesO No@ 6. Result in any potential impact to the equipment or facilities utilized for Ventilated Storage Cask activities per Section 6.1.67 YesO No@
- 7. Involve a change under 10CFR50.54 for the following SAR documents per Section 6.1.7:
QAMO? YesO No@ E-Plan? YesO No@
FORM TITLE: FORM NO. REV. 10CFR50.s8 DETERMINATION 1000.131A 3 Pc 1.2 {. l Page2 ,of 7' Document No. 975109P201 Rev/ Change No. A Basis for twannin=*!on (Quesuc-ns 1. 2 & Sh
- 1) The Operating License documents do not discuss the stator water conductivity meters or t lines nor is it necessary to include any new information into these documents as a result of this m]
- 2) There is an impact to the the multi volume SAR set. The conductivity cell assemblies contain a val is of a type different than the existing valve. The existing valve is a gate valve, and the new one is valve. This will require a change to Figure 3.2-6 which is the P&lD for the Stator Water Cooling other documents in the SAR do not discuss the subject components in the SWC system.
- 3) The only test to be performed as a result of this modification are routine post modification te proper operation of the equipment. The types of post modification tests to be performed will be leak che proper instrument operation, and proper pressure switch response. These tests will be performed wh system is not required to be operable. Margins to safety will be unaffected. There will also be no other systems ability to prevent or mitigate an accident.
- 4) Per the attached Environmental impact Determination no impact was identified and an environmenta evaluation is not required.
- 5) The proposed modification does not involve any aspect of radiological substances. There a processing issues or new pathways associated with this modification. This system is a closed lo The only direct fluid inte: Tace is with the Condensate Transfer system which provides make-up w SWC. orNo radiological processing is involved nor is there an interface (existing or new) with a system area.
- 6) This modification does not involve the spent fuel, spent fuel pool, or processing of spent fue l
equipment is located in the turbine building in an area remote from the fuel handling areas and does no
- affect the handling process. The VSC equipment and load paths are not affected. The modification does impact any equipment in the Fuel Pool t
, rain bays or roadway areas used by the dry fuel handling process.
- 7) The E-Plan and QAMO do not discuss the subject equipment. Further evaluations are not required. !
O Proposed change does not require 10 CFR 50.59 Evaluation per Attachment 1, item (if checked, note #,,_,_, appropriate item #, send LDCR to Licensing). )
)
Search Scope: i List sections reviewed in the Licensing Basis Documents specified in Question 1,2 and 3. _ If a search wEs performed on LRS, the LRS search index should be entered under "Section" with the search statemen parentheses. Controlled hard copies of the documents shall be reviewed (LRS is not verified and sear text, required.not figures or drawings). Attach and distribute a completed LDCR per Section 6.1.2 If LSD chanl Document SE1120 LRS: U2 50.59 Search ALL (Keywords: SWC, stator, conductivity, instrument snubbers, main generator, beckman, 2CITS9771, 2CITS9772, 2CE9771, 2CE9772, 2PS9777B, 2PS9777C, 2PS9777D, C9771, C9772 MANUAL SECTIONS: SAR multi volume set Section 3 list of figures, Sections 1.2.2.5,10.2 FIGURES: SAR multi volume set Figure 3.2 6 e
ARKANSAS NUCLEAR ONE FORM TITLE: FORM NO. REV. 10CFRSS.89 DETERMINATION 1000.131A 3 PC 1.2 Document No. 975109P201 Rev/ChanQe No. Q Pese?.*f2-dad n eddfA Cleveland Reasoner Cer1ified Reviewers Signature 9-30-98 Printed Name Date
- Reviewers certification expiration date: 11 4-98 Assistance provided by:
Printed Name Scope of Assistance Trevena Jackson Date Research 1-11-91 Search Scope Review Acceptability (NA, if performed by Technical Review per 1000.006) O. de bm . Certified ReviewerTSignature Li 61.s~ w/ 44ErrWr_. f[JM/2P Printed Name Date l 4
F _ , FORM TITI.E: FORM NO. REV. 10CFR60.60 DETERugNATION 1000.131A , I l ENVIRONMENTAL IMPACT DETERMINATION (UNIT 1 and UNIT 2) Document No. 975109P201 Rev/ Change No. A Complete the following Determination. If the answer to any checklist item is "Yes", an Environmentai EvaluatI is required. See Section 6.1.4 for additional guidance. Will the Activity being evaluated: Yes No O 2 Disturb land that is beyond that initially disturbed during construction (i.e., new construction of I l buildings, creation or removal of ponds, or other terrestrial impact)? See Unit 2 SAR Figure ' 2.5-17. This applies only to areas outdde the protected area. O E increase thermal discharges to lake or atmosphere?
- O E Increase concentration of chemicals to cooling lake or atmosphere through discharge canal or l tower?
O @ increase quantity of chemicals to cooling lake or atmosphere through discharge canal or tower? 0 @ Modify the design or operation of cooling tower which will change drift characteristics? l l 0 2 instati any new transmission lines leading offsite? O E change the design or operation of the intake or discharge structures? l O E Discharges any chemicals new or different from that previously discharged? l O 2 Potentially cause a spill or unevaluated discharge which may effect neighboring soils, surface water or ground water? O E involve burying or placement of any solid wastes in the site area which may effect runoff, ! surface wa'ter or ground water? O E involve incineration or disposal of any potentially hazardous materials on the ANO site? O 2 Result in a change to nonradiological effluents or licensed reactor power level? O 2 Potentially change the type or increase the amount of non-radiological air emissions from the ANO site.
ARKANSAS NUCLEAR DNE FORM TITLE: FORM NO. REV. 10CFRSO.89 EVALUATION 1000.131a 3 Pc.2 Page I of 7 10CFR50.59 Eval. No._ (Assigned by PSC) fFMI-lU5 Document No. 9'f5109P201 Rev> Change No. A Title Stator Water CeeHna conductivity Ce ector Res!=-w.est and PS snEt-t+r installation A WRITTEN RESPONSE PROVIDING THE BASIS FOR THE ANSWER TO EAj ATTACHED. EACH QUESTION MUST BE ANSWERED SEPARATELY. A SIMP 1 CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES GUIDANCE FO fl If the answer to any question on this form is "Yes," then an unreviewed safety question is involve to all questions is "No," then the proposed change does not involve an unreviewed safety question. 1. Will the probability of an accident previously evaluated in the SAR be increased? Yes O No @ There are several accident scenarios that have a turbine-generator trip or malfunction as an acci With regard to this modification the issue to consider is whether the modification will result in an inc frequency of the turbine to trip and/or not trip when it should. The conductivity analyzers and associated hardware have no direct trip function nor will they preve from tripping. They give an indication and alarm that may identify a problem that necessitates shutdown, increased. but the instrumentation will not directly result in a plant trip. Therefore turbine trip fr The addition of the instrument snubbers in the tubing to the turbine runback pressure switches re thorough assessment. For this assessment it is assumed that a turbine runback would potentially le trip. A plausible scenario is that the snubbers could become clogged and could mask a low pressure c This condition is highly unlikely since SWC is a very clean system and the snubbers are in a dea subjected to continuous flow. Also three snubbers are used; one for each pressure switch. This was don prevent failure of one snubber from taking out all three switches. The switches use a 2 out of 3 logic. S is a plausible scenario, an assessment is needed to determine if it really represents a increase Note that the pressure switches (2PS-97778,C.D) initiate a turbine runback not a turbine trip. How pressure condition were masked from the switches, overheating of the generator stator would result which cou eventually lead to malfunction of the generator and turbine-generator trip. The eventual trip would b desirable trip. If a low pressure condition existed it would be desirable to trip the turbine. Howeve action of the snubber does not change the probability of the trip demand. The probability of the trip i with the probability of the low pressure condition. A trip is not desired nor will it occur due to thej clogged. There may be undesirable equipment damage because of the response delay, but the c ' that the trip demand frequency is not changed due to the clogged snubbers. So the probability of not increased.
' In the scenario discussed above, the condition is described where the turbine doesn't trip when it shou type of accident that would be most closely associated with the turbine not tripping is Excess Heat Remov to Secondary System Malfunction. One of the scenanos in this accident is rapid uncontrolled openin turbine admission valves. Inability to close the valves on a trip is a comparable scenario although it doe match the explicit description in the SAR. The condition of the turbine not tripping on low SWC pressure relevant to the ability to close the turbine steam admission valves, if closure of the turbine valves is SWC low pressure switch is not the component used to detect / initiate this condition. Therefore turbine the when a low SWC pressure condition exists has no relevance to probability of an accident associa turbine.
The conclusion is that there is not an increased probability of tripping the turbine. Likewise, there isn't an increased probability of the turbine not inpping in an accident scenario. '
FORM TITLE: FORM NO. REV. 10CFR50.59 EVALUATION 1000.131B 3 PC.2 { Po c,
- cj 75108) i>Lo t Itar V o p g gg7 2.' Will the consequences of an accident previously evaluated in the SAR be increased?
Yes O No @ The SWC does not have any function associated with limiting offsite dose in an accident scenario or any other scenario. The function of the system is simply to remove heat from the Main Generator stator windings. There are no radiological boundaries or interfaces with the system. The system does not support any other system that is associated with limiting offsite dose
- 3. Will the probability of a malfunction of equipment important to safety be increased?
Yes O No @ The SWC is associated with the turbine-generator set that is a system that is important to safety. It is important because the turt)ine valves provide MS isolation, Gland steam supports condenser vacuum, and the Generator supplies normal AC power. . The conductivity analyzers provide a pressure boundary for the SWC and provide indication of the SWC conductivity. The design quality and standards of the new equipment are the same as the existing equipment, so the reliability of the equipment is unchanged. Therefore the likelihood of failure of this equipment leading to malfunction of the generator is unchanged. The pressure switch instrument snubbers are installed to decrease the probability of malfunction of other equipment. The reduction in malfunction probability is accomplished by reducing the potential that spurious indications lead to a tutt>ine runback or trip.
- 4. Will the consequences of a malfunction of equipment important to safety be increased?
Yes O No @ The SWC system does not affect the functions of the turbine-generator that are important to safety (maintain vacuum, isolate steam flow). Likewise it is not associated with limiting any consequences associated with failure of turbine-generator. Offsite dose is not impacted by this system either directly or indirectly by leading to malfunction of another system.
- 5. Will the possibility of an accident of a different type than any previously evaluated in the SAR be created? Yes O No @
The SAR already considers a variety of accidents with turbine trips or malfunctions as initiating events. The only plausible accident related issues for SWC is causing a turbine trip. Since a turt>ine trip is considered a relatively frequent event, the accident scenarios have appropriately considered this possibility. Failure cf the SWC would not result in any other type of accident not already considered.
- 6. Will the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR be created? !
Yes O No @ ! It is inherent in the design of the turbine generator that problems in the auxiliary systems could necessitate a turbine trip. Accordingly, the turbine design considers the potential for SWC to fail and lead to a turbine runback / trip. So the system design envelops any impacts of a complete failure of the SWC. ~
- 7. Will the margin of safety as defined in the basis for any technicn' specification be reduced?
Yes O No @ SWC does not provide any margin of safety. There are no TS associate with the turbine auxiliary systems. Likewise there are no presenbe SWC system ilmits or operating parameters that have to be maintained in order to ensure the proper performance of fission barriers. -
FORM TITLE: ARKANSAS NUCLEAR ONE FMUb 8 FORM NO. REV. 10CFR40.59 EVALUATION 1000.1313 3 PC-2 pec.
- 97Fi09Paoi ReV o 3,e757 2 dado Cleveland Reasoner Certified Reviewers Signature 9-30-98 Printed Name Date Reviewers certification expiration date: 11-4-98 Assistance provided by:
Printed Name Scope of Assistance Date PSC review by: Date:_ \G , b $$ l l l 1
ARKAN*AS NUCLEAR ONE Page 1 FORM TITLE: FORM NO. REV. 10CFR50.59 DETERMINATION 1000.131 A 3 PC-1 This Document contains 3 Pages. Document No. PC 980184P201 RevlChange No. O nn nnn4n, Title EFW ETA FEED iu ouvivono01 Brief description of proposed change. F GE /3 REV 0 This Plant Change modifies one valve in the Condensate and Feedwater system and piping in the Steam Generator Feedwater Chemical Feed System in order to feed ETA from the Amine pumps to the suction of the Auxiliar Pump (2P-75) and the suction of the Emergency Feedwater Pumps (2P-7A & 2P-7B). Changes will be require Fig.10.4-2 (P&ID M-2204, Sh.1 & Sh. 4) and SAR Fig.10.4 5 (P&ID M-2240, Sh.1). Will the proposed Activity:
- 1. Require a change to the Operating License including:
Technical Specifications (excluding the bases)? YesO No@ Operating License? YesO No@ Confirmatory Orders? YesO No@ 2. Result in informstion in the.following SAR documents (including drawings and text) being (a) no longer true or accurate, or (b) violate a requirement stated in the document: SAR (multi-volume set for each unit)? Yes@ nod Core Operating Limits Report? YesO No@ Fire Hazards Analysis? YesO No@ l Bases of the Technical. Specifications?
., YesO No@
Technical Requirements Manual? YesO No@ NRC Safety Evaluation Reporis? YesO No@
- 3. Involve a test or experiment not described in the SAR?
(See Attachment 2 for guidance) YesO No@
- 4. Result in a potentialimpact to the environment? (Complete Environmental impact Determination of this form.)
YesO No@
- 5. Result in the need for a Radiological Safety Evaluation per section 6.1.57 YesO No@
- 6. Result in any potentialimpact to the equipment or facilities utilized for Ventilated Storage Cask activities per Section 6.1.67 YesO No@
- 7. Involve a change under 10CFR50.54 for the following SAR documents per Section 6.1.77 QAMO? .
YesO No@ t E Plan? YesO No@
1 ARKANSAS NUCLEAR ONE l FORM TITLE: Pace 2 { FORM NO. REV. 10CFR50.59 DETERMINATION 1000.131 A 3 PC-1 Document No. PC 980184P201 PC 980184P201 Rev./ Change No. O Basis for Determination (Questions 1,2, & 3): PAGE /t' REV 0 This Plant Change modifies one valve in the Condensate and Feedwater system and piping in t Feedwater Chemical Feed System in order to feed ETA from the Amine pumps to the suc Pump (2P 75) and the suction of the Emergency Feedwater Pumps (2P-7A & 2P-7B). Chan Fig.10.4-2 (P&lD M 2204, Sh.1 & Sh 4) and SAR Fig.10.4-5 (P&ID M 2240, Sh.1). No other changes to the LBDs will be necessary. The text of the SAR will not require revision.
- 1. The TS, OL, and Confirmatory Orders weic reviewed and no sections require revision. The figures; it was dwntined that this Plant Change does not affect these documents.
- 2. This Plant Cha .y will require revision of the U2 SAR (Fig.10.4-2 & Fig.10.4 5). This changes to the Core Operating Limits Report, the FHA, Bases of the Technical Specifications, the T Requirements Manual, or the NRC Safety Evaluation Reports.
- 3. This Plant Change does not change the function of any system and does not involve any test
- 4. The emironmental impact determination was completed. This Plant Change does not have
- 5. This Plant Change does not involve processing radioactive material or impact monitored
- 6. There is no potential impact the Ventilated Storage Cask equipment or procedures.
- 7. This Plant Change will not require changes to the QAMO or the E-plan.
O Proposed change does not require 10CFR50.59 Evaluation per Attachmeb 1, Item # item #, send LDCR to Licensing). (If checked. note appropriate Search Scope: I List sections reviewed in the Licensing Basis Documents specified in questions 1,2 and 3. I LRS, "all" may be entered under 'Section" with the keyword (s) used in parentheses. Controli documents shall be reviewed (LRS is not verified and searches only text. not figures or drawinl completed LDCR per Section 6.1.2 If LB,D changes are required. Document Section LRS: All (EFW. Emercency Feedwater. Amine) MANUAL SECTIONS: 1049.15.1.8.151.14.364.513.10.35.1046 FIG S: 104-2.104-5
' s Gary W. LifHek Certif 1 ewer's $4isplifre" 3/30/98 Pnnted Name Date Reviewer's certification expiration date:
1/9/2000 Assistance prmided by: Printed Name Scope of Assistance Tim Woodson Technical Input Date 3/17/98 Searth Scope Review Acceptability (NA,if performed by Technical Reviewer per 1000.006) N Yt (dwnsem Certifie4 Reviewer's Signauste 9.S Pnnted Name Date
ARKANSAS NUCLEAR ONE FORM TITLE: Pace 3
- ~ FORM NO. REV.
10CFR50.59 DETERMINATl!N 1000.131 A 3 i ENVIRONMENTAL IMPACT DETERMINATION PC 980184P2 0I (UNIT 1 and UNIT 2) PAGE /5 REV 0 Document No. PC 980184P201 RevlChange No. O Complete the following Determination. If the answer to any item below is "Yes", an Emironmen See Section 6.1.4 for additional guidance. Will the Activity being evaluated: m . O @ Disturb land that is beyond that initially disturbed during construction (i.e., new construction of buildings, creation or removal of ponds, or other terrestnal impact)? See Unit 2 SAR Figu : This applies only to areas outside the protected area. ' O E Increase thermal discharges to lake or atmosphere? I O E Increase concentration of chemicals to cooling lake or atmosphere through discharge canal o' O @ Increase quantity of chemicals to cooling lake or atmosphere through discharge canal or tower? O E Modify the design or operation of cooling tower which will change drift characteristics? i O @ Install any new transmission lines leading offsite? O @ Change the design or operation of the intake or discharge structures? O E Discharges any chemicals new or different from that presiously discharged? O Potentially cause a spill or unevaluated discharge which may effect neighboring soils, surface w ground water? O E Invohr burying or piacement or any solid wastes in the site area which may effect runofT, surfa water or ground water? O E Involve incineration or disposal of any potentially hazardous materials on the ANO site? O E Result in a change to nonradiological effluents or licensed reactor power level? O @ Potentially change the type or increase the amount of non. radiological air emissions from the AN site.
ARKANSAS' NUCLEAR ONE Pace i FORM TITLE: FORM NO. REV. 10CFR50.89 SAFETY E.VALU. A.Tl2N. . . 1000.131B 3 This Document contains 2 Pages. PAGE /4 REV 0 Document No. 980184P201 Rev/ Change No. 0 10CFR50.59 Eval. No. FRl41054 Title (Assigned by PSC) EFW ETA FEED A WRITTEN RESPONSE PROVIDING THE BASIS FOR THE ANSWER TO EACH QUEST ATTACHED. EACH QUESTION MUST BE ANSWERED SEPARATELY. A SIMPLE STATE CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES GUIDANCE FOR RESPO If the answer to any question on this form is "Yes," then an unreviewed safety question is involved. If the a questions is "No," then the proposed change does not involve an unreviewed safety question. 1. Will the probability of an accident previously evaluated in the SAR be increased? Yes O No @ This change affects only 1/2" and 3/8" chemical feed piping. The new valve and piping compone requirements of the piping system being modified. There are no possible applicable accidents discussed in the SAR parameters. for this chemical feed piping. This Plant Change improves the ability to maintain The modification of this chemical addition piping has : ; impact on the probability of any of the po accidents in the SAR. 2. Will the consequences of an accident previously evaluated in the SARbeincreased? Yes O No @ This chemical feed system does not perform any safety functions for any accident evaluated in the Change affects only the type of chemical that can be introduced into the suction of the EFW failure of piping or components associated with this Plant Change will not increase the failure cons consequences previously evaluated in the SAR. 3. Will the probability of a malfunction of equipment important to safety be increased? Yes O No @ The piping components being added/ modified are not safety related, nor will they affect any saf equipment. All new piping and components associated with this Plant Change meet the specification existing piping systems. This chemical feed system does not perform a safety function, nor can it incre probability of a malfunction of equipment important to safety. 4. Will the consequences of a malfunction of equipmedt important to safety be increased? Yes O No @ The new valve and piping components meet the same requirements of the piping system being m Additionally, important the to safety. failure of any of these components cannot affect the consequences of a m
ARKANSAS NUCLEAR ONE FORM TITLEt P'on 2
. FORM NO. REV. I 10CFR60.59 SAFETY EVALUATION
- 1000.1313 3 r ,
PC 980184P201 P AGE LZ RE V 0 I 5. Will the possibility of an accident of a different type than any previously evaluated in the SAR be created? Yes O No @ l This Plant Change affects only the type of chemical that can be introduced into the suctio pumps. The Emergency Feedwater System t/ill function as before and no new accident s by this Plant Change. The possibility of a new type accident than previously evaluated in created by this Plant Change. - 6. Will the giossibility of a malfunction of equipment important to safety of a different type than any presiously evaluated in the SAR be created? Yes O No @ The piping components being added/ modified are not safety related, nor will they affec equipment. The chemical (Amine) to be fed through this system is compatible with all this Plant Change. No new malfunctions of equipment importa 7. Will the margin of safety as defined in the bases for any technical specification be reduced?
- Yes O No @
The modifications associated with this Plant Change will not affect the margin of safety a any Technical Specifications. The function of the Emergency Feedwater System is not aff Change. Likewise, no Technical Specification margin of safety is reduced. 1
' 84 rtified Revat9 tis Signature Garv W. Liffick Printed Name f
/ f9 6 ate Resiewer's certification egiration date:
1/9/2000 i Assistance prtnided by: ! Primed Name Scope of Assistance Tim Wancienn TechnicalInput Date Mi/,/ff PSC review by: Date: \
AHKANSAS NUCLE AR ONE \ FORM TITLE: Pace 1 J FORM NO. REV. l - 10CFR50.59 DETERMINATION 1000.131 A 3 PC-1 PC 98 0243P201 This Document contains 3 Pages. Document No. PC 980243P201 PAGE // REV 0 Rev/ Change No. O Title Replace Valve 2PS-162 Brief description of proposed change: This Plant Change replaces valve 2PS-162. which is an isolation valve from RCS to the Post Accident Samp (PASS). The existing valve is a 1/2" socket weld gate valve. The replacement valve is a 3/4" socket weld gate va by the same manufacturer. The replacement valve is built to the same standards and has the same pressure and temperature ratings. The only changes required to the SAR are the addition of piping reducers to be shouTi on P&ID M-2237, Sh.1. which is SAR Fig. 9.3 2. Will the proposed Activity:
- 1. Require a change to the Operating License inclu(ng:
Technical Specifications (excluding the bases)? YesO No@ Operating License? YesO No@ Confirmatory Orders? YesO No@ 2. Result in information in the following SAR documents (including drawings and text) being (a) no longer true or accurate, or (b) violate a requirement stated in the document: SAR (multi-volume set for each unit)? Yes@ NoO Core Operating Limits Report? YesO No@ Fire Hazards Analysis? YesO No@ Bases of the Technical Specifications? YesO NoS Technical Requirements Manual? YesO No@ NRC Safety Evaluation Reports? YesO No@
- 3. Involve a test or experiment not described in the SAR?
(See Attachment 2 forguidance) YesO No@
- 4. Result in a potentialimpact to the environment? (Complete Environmental Impact Determination of this form.)
YesO No@
- 5. Result in the need for a Radiological Safety Evaluation per section 6.1.5?
YesO No@ l S. Result in any potentialimpact to the equipment or facilities l l utilized for Ventilated Storage Cask activities per Section 6.1.67 YesO NOS 7 Involve a change under 10CFR50.54 for the followmg SAR documents per Section 6.1.7? QAMO7 YesO NoS E Plan? YesO No@ l l l
ARKANSAS NUCLEAR ONE Page 2 FeRM TITLE: FORM NO. REV. 10CFR50.59 DETERMINATION 1000.131A 3 PC-1 Document No. PC 980243P201 RevlChange No. O PC 980243P201
. Bidis for Determination (Questions 1,2, & 3): P A0 E /A, RE D This Plant Change replaces valve 2PS-162. This valve is shown on U2 SAR Fig. 9.3-2 (P&ID M-??'7, Sh 1). The Figure and P&lD need to be revised to show the piping reducers en each end of the valve. The en: a g valve is a socket weld gate valve in a 1/2" piping line. The replacement valve is a 3/4" socket weld gate valve.
No other changes to the LBDs will be necessary. The text of the SAR will not require revision.
- 1. The TS, OL, and Confirmatory Orders were reviewed and no sections require revision. The resiew included tables figures; it was determined that this Plant Change does not affect these documents.
- 2. This Plant Change will require revision of the U2 SAR (Fig. 9.3-2). This Plant Ch mge will not require cha Core Operating Limits Report, the FHA. Bases of the Technical Specifications, the Technical Requirements Ma the NRC Safety Evaluation Reports.
- 3. This Plant Change does not change the function of any system and does not imulve any test or experiment.
- 4. The environmental irnpact determination was completed. This Plant Change does not have any environmental
- 5. This Plant Change does not involve processing radioactive material or impact monitored effluent relea
- 6. There is no potential impact to the Ventilated Storage Cask equipment or procedures.
- 7. This Plant Change will not require changes to the QAMO or the E-plan.
O Proposed change does not require 10CFR50.59 Evaluation per Attachment 1, item(If#checked, note appropriate item #. send LDCR to Licensing). Search Scope: List sections reviewed in the Licensmg Basis Documents specified in questions 1,2 and 3. If a keyword searc LRS. "all" may be entered under "Section" with the keyword (s) used in parentheses. Controlled hard cop documents shall be reviewed (LRS is not verified and scarches only text. not figures or drawings). A? tach and dis completed LDCR per Section 6.1.2 if LBD changes are required. Document Section LRS-PASS. Post Accident Sarnplinn. 2PS 16J MANUAL SECTIONS: 93.2.9.3224.11.3610 FIGUR S- 9 1-2 AM d/ Gan W. Liffick 3/9/98 Ccided Revieweggnature' Pnnted Nanic Date Reucwcr's cenification expiration date: 1/9/2000 Assistance provided by: Pnnted Name Scope of Assistance Butch Hollowoa Date Technical input 3/9/98 Search Scope Review Acceptability (NA,if performed by Technical Reviewer per 1000.006) bMOW D~r~cM NOU. Ctuc A Tl1lU' Certified Reviewer's Signature Pnnted Name Date .
p ~s . .m~ .~ am, v.~
- ripe o FORM TITLE
FORM NO. REV. 10CFR50.59 DETERMINATION 1000.131 A 3 ENVIRONMENTAL IMPACT DETERMINATION PC 98024 3P201 (UNIT I and UNIT 2) PAGE I3 REV 0 Document No. PC 980243P201 Rev1 Change No. O Complete the following Determination. If the answer to any item below is "Yes", an Environmental Evalua See Section 6.1.4 for additional guidance. Will the Activity being evaluated: Yes No O @ Disturb land that is beyond that initially disturbed during construction (i.e., new construction of buildings creation or removal of ponds, or other terrestnal impact)? See Unit 2 SAR Figure 2.5-17. This applies only to areas outside the protected area. O @ Increase thermal discharges to take or atmosphere? O @ Increase concentration of chemicals to cooling lake or atmosphere through discharge canal or tower? O @ Increase quantuy of chemicals to cooling lake or atmosphere through discharge canal or tower? O @ Modify the design or operation of cooling tower which will change drift characteristics? O @ Install any new transmission lines leading offsite? O O Change the design or operation of the intake or discharge structures? O @ Discharges any ch micals neu or different from that previously discharged? , O O Potentially cause a spill or unevaluated discharge which may effect neighboring soils surface water or l ground water' O. @ invalve burying or placement of any solid wastes in the site area which may effect runoff. surface water or ground water? O @ involve inemeranon or disposai orany potentially hazardous materials on the ANO sitc? O @ Result in a change to nonradiological effluents or licensed reactor pov<,r level? O @ Potentially change the type or increase the amount of non-radiological air emissions from the ANO site.
FORM TITLE: ARKANSAS NUCLEAR ONE Pace FORM NO. REV.
- 10CFR50.59 SAFETY EVALUATION 1000.1313 3 PC 98 024 3VZ01 This Document contains 2 Pages.
Document No. PC 980243P201 PAGE /f ~REV 0 . RevlChange No. 0 10CFR50.59 Eval. No. FFN-YC1 Title Replace Valve 2PS-162 (Assigned by PSC) A WRii 1eN RESPONSE PROVIDING THE BASIS FOR THE ANSWER T ATTACHED. EACH QUESTION MUST BE ANSWERED SEPARATELY. A CONCLUSION IS NOT SUFFICIENT. ATTACHMENT 2 PROVIDES GUIDAN If the answer to any question on this form is "Yes." then an unreviewed safety question i questions is "No," then the proposed change does not invcive an unreviewed safety question. 1. Will the probability of an accident previously evaluated in the SAR be increased? Yes O No @ This change consists of replacing a valve in the PASS System. Only one accident applicable: "Small Spills or Leaks of Raaioactive Material Outside Containment" Since the replacement valve and associated piping reducers meet the same component pressure / temperature req code qualified, the probability ofleakage from the system has not been increased. 2. Will the consequences of an accident previously evaluated in the SAR be increased? Yes O No @ The applicable accident is "Small Spills or Leaks of Radioactive Matenal Outside Containment" The type efTect dispersed into the ground water and the distance it would parameters. and therefore cannot increase the consequences of this accident. 3 Will the probabihty of a malfunction of equipment i imponant to safety be mercased' Yes O No @ ; The replacement valve meets the specifications and codes of the existing valve. The the replacement valve is the same as that of the old valve. The non-Q valve a being rep safety function. nor does it suppon other equipment that performs a safety function. Ba probabihty of a malfunction of equipment imponant to safety will not be increased.
.I Will the consequences of a malfunction of equipment imponant to safety bc increased?
Yes O No @ l The replacement valve and its associated piping reducers mee}}