ML20211D711
| ML20211D711 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 08/09/1999 |
| From: | Maine Yankee |
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| NUDOCS 9908270121 | |
| Download: ML20211D711 (200) | |
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{{#Wiki_filter:O MAINE YANKEE DEFUELED SAFETY ANALYSIS REPORT (DSAR), REV.17 INSERTION INSTPUCTIONS NOTE PLEASE TAKE NOTE THAT PAGE NUMBERING HAS CHANGED IN VARIOUS SECTIONS DO TO PAGE DELETIONS I IRT OF EFFECTIVE PAGES Remove List of Effective Pages 1-5, Rev.16 and Replace with Pages 1-5, Rev.17 TABI E OF CONTENTS Remove Pages iv and v, Rev.16 and Replace with Pages iv and v, Rev.17 SECTION 1.0 O n - ve Pane No. Rev. No. Pace No. 1==a Rev. No. 1-i 14 1-i 17 1-iii 14 1-iii 17 1-3 16 1-3 17 1-5 16 1-5 17 1-6 14 1-6 17 1-7 16 1-7 17 1-8 16 1-8 17 1-9 16 1-9 17 1-13 14 1-13 17 1-15 14 1-15 17 1-16 14 1-16 17 1-17 14 1-17 17 Figure 1.3-2 14 DELETED l-22 14 1-21 17 1-23 15 1-22 15 i O 1 9908270121 99000939 DR ADOCK O
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O MAINE YANKEE DEFUELED SAFETY ANALYSIS REPORT (DSAR), REV.17 INSERTION INSTRUCTIONS SECTION 2.0 Remove Insg! Pane No. Rev. No. Pace No. Rev. No. 2-22 14 2-22 17 2-23 14 2-23 17 2-24 14 2-24 17 2-33 14 2-33 17 2-34 14 2-34 17 2-35 14 2-35 17 Figure 2.2-2 - 2-38 (Fig. 2.2-2) 17 Figure 2.2 2-39 (Fig. 2.2-3) 17 Figure 2.2-4 - 2-40 (Fig. 2.2-4) 17 2-47 14 2-47 17 2-51 14 2-51 17 SECTION 3.0 Remove existing Section 3.0 and Insert New Section 3.0. SECTION 4.0 Remove Insg1 Pane No. Rev.No. Pace No. Rev. No. 4-3 16 4-3 17 4-4 14 4-4 17 4-9 16 4-9 17 4-10 14 (Table 4.4.1) 4-10 17 (Table 4.4.1) 4-11 14 4-11 17 4-12 16 4-12 17 4-13 16 4-13 17 4-14 16 4-14 17 4-15 16 4-15 17 4-16 16 4-16 17 (Table 4.6.1) 4-17 16 4-17 17 (Table 4.6.1) 4-18 16 - 4-19 16 - O 2 J
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O ANALYSIS REPORT (DSAR), REV.17 INSERTION INSTRUCTIONS SECTION 5.0 Remove existing Section 5.0 and Insert New Section 5.0. SECTION 6.0 Remove existing Section 6.0 and Insert New Section 6.0. SECTION 7.0 Remove Pages 7-2 and 7-3, Rev.14 and replace with attached Pages 7-2 and 7-3, Rev.17. APPENDIX A Remove Page A-34, Rev.16 and replace with attached Page A-34, Rev.17. O O 3 J
O = ac UST OF EFFECTIVE PAGES PAGE REY REMARKS PAGE REV REMARKS PAGE REY REMARKS TABLE OF CONTENTS SEC11DN 1.0 SECTION 2.0 1 14 1-1 17 Table of Contents 2-1 14 Table of Contents il 14 1-il 14 List of Tables 2-li 14 Table of Contents ill 14 1-lii 17 Ustof Figures 2-lii 14 List of Tables iv 16 1-1 14 2-iv 14 List of Fgures - v 16 12 14 2-1 14 vi 14 1-3 17 2-2 14 vli 14 1-4 16 2-3 14 vill 14 1-5 17 2-4 14 ix 14 1-6 17 2-5 14 Tab:a 2.1.1 17 17 2-6 14 Table 2.1.2 1-8 17 2-7 14 Figure 2.1 1 1-9 17 2-8 14 Figure 2.12 1-10 14 Table 1.3.1 2-9 14 Figure 2.1-3 1-11 14 Table 1.3.1 2-10 14 Figure 2.1-4 1-12 14 Table 1.3.1 2-11 14 Fgure 2.1-5 1-13 17 Table 1.3.1 2-12 14 Figure 2.1-6 1-14 14 Table 1.3.1 2 13 14 Figure 2.1-7 1 15 17 Table 1.3.1 2-14 14 Fgure 2.18 1-16 17 Table 1.3.1 2-15 14 1-17 17 Table 1.3.1 2 16 14 1-18 14 Figure 1.31 2 17 14 O 1-19 1 20 14 14 Figure 1.3-3 Figure 1.3-4 2-18 2 19 14 14 1-21 17 2-20 14 1-22 15 2-21 14 2-22 17 2-23 17 2-24 17 2-25 14 2-26 14 Table 2.2.1 2-27 14 Table 2.2.2 2-28 14 Table 2.2.3 2-29 14 Table 2.2.4 2-30 14 Table 2.2.5 2 31 14 Table 2.2.6 2-32 14 Table 2.2.7 2-33 17 Table 2.2.8 2 34 17 Table 2.2.9 2-35 17 Table 2.2.10 2-36 14 Table 2.2.10 2-37 14 Figure 2.2-1 2-38 17 Figure 2.2-2 2-39 17 Figure 2.2-3 2-40 17 Figure 2.2-4 2-41 14 Figure 2.2-5 2-42 14 Figure 2.2-6 2-43 14 Figure 2.2-7 2-44 14 2-45 14 2-46 14
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MYAPC LIST OF EFFECTIVE PAGES PAGE REV REMARKS PAGE REV REMARKS PAGE REV REMARKS 3-44 17 2-47 17 SECTION 3.0 3-45 16 2-48 14 3-46 17 2 49 14 3-1 17 Table of Contents 3-47 17 2-50 14 3-il 17 Table of Contents 3-48 16 2-51 17 3-lii 17 Table of Contents 3-49 14 2 52 14 3-iv 17 Ust of Tables 3-50 16 2-53 14 Table 2.3.1 3-v 17 Ust of Figures 3-51 16 2-54 14 Table 2.3.2 3-vi 17 Ust of Figures 3-52 16 2-55 14 Figure 2.3-1 3-1 14 3 53 17 2-56 14 Figure 2.3-2 3-2 17 3-54 17 2-57 14 Figure 2.3-3 3-3 14 3 55 17 2-58 14 Figure 2.3-4 3-4 14 3 56 14 2-59 14 Figure 2.3-5 3-5 14 3-57 16 240 14 Figure 2.3-6 34 16 3 58 14 241 14 Figure 2.3-7 3-7 17 3-59 16 2-62 14 Figure 2.34 3-8 16 340 17 243 14 3-9 16 3-61 17 244 14 3-10 14 342 17 245 14 Figure 2.4-1 3 11 17 343 17 246 14 3-12 17 344 17 2-67 14 3 13 17 345 17 248 14 3 14 17 346 17 O 249 2-70 14 14 Figure 2.5-1 Figure 2.5-2 3 15 3-16 17 14 347 348 14 14 Figure 3.3-1 Figure 3.3.2 2-71 14 Figure 2.5-3 3 17 14 349 14 Figure 3.3-3 2-72 14 Figure 2.5-4 3-18 17 3-70 14 Figure 3.3-4 2-73 14 Figure 2.5-5 3 19 16 3-71 14 Figure 3.3-5 3-20 17 3-72 14 Figure 3.3-6 3-21 14 3-73 17 3-22 14 3-74 17 3-23 17 3-75 14 3-24 14 Table 3.1.1 3-76 17 3-25 14 Figure 3.1-1 3-77 17 3-26 14 Figure 3.12 3 78 17 3-27 17 3-79 16 3 28 15 340 16 3-29 17 341 17 3-30 17 342 17 3-31 17 343 17 3-32 17 3-84 16 Figure 3.3-9 3-33 17 Table 3.2-1 3-85 17 3-34 14 3-86 17 3 35 16 Figure 3.2-1 3-87 17 3-36 16 Figure 3.2-2 3-88 17 3-37 14 Figure 3.2-12 3-89 17 3 38 14 Figure 3.2-13 3-90 17 3 39 17 3.gj 37
- II 3-92 17 3-4I II 3-93 17 3-42 14 3 94 37 M3 16 3-95 17 l
DSAR il Rev.17 J
I Q MYAPC l.lST OF EFFECTIVE PAGES PAGE REV REMARKS PAGE REV REMARKS PAGE REV REMARKS 3-06 17 SECTION 4.0 SECTMMI 5.0 3-97 14 Figure 3.3-21 3-06 14 Figure 3.3-22 4-1 14 Table of Contents 5-1 14 Table of Contenis 3-00 16 4-il 16 Table of Contents 5 - 11 14 List of Tables 3 100 17 4 111 14 List of Teblos 5-ill 17 List of Fgures 3-101 17 Figure 3.3-23 4-iv 14 List of Figures 5-1 15 3-102 17 41 14 5-2 17 3 103 17 4-2 14 5-3 17 3 104 17 43 17 54 14 3 105 17 Figure 3.3-24 4-4 17 5-5 17 3 106 17 4-5 14 56 14 3 107 17 4-6 14 57 17 3 106 17 4-7 14 5-8 17 Table 5.2.1 3 109 17 4-8 14 5-9 15
' 3 110 17 4-9 17 5-10 17 3-111 17 4-10 17 Table 4.4.1 5-11 17 3 112 17 4 11 17 5-12 14 Table 5.3.1 3-113 14 4 12 17 5-13 14 Table 5.3.2 3-114 14 4-13 17 5 14 14 Table 5.3.3 4-14 17 5-15 15 4-15 17 5-16 14 4 16 17 Table 4.6.1 5-17 15 4-17 17 Table 4.6.1 5 18 14 O 5-19 5-20 14 14 5-21 16 5-22 14 5-23 16 5-24 16 5-25 14 Table 5.5.1 5-26 17 Table 5.5.2 5-27 17 Table 5.5.2 5-28 17 Table 5.5.2 5-29 17 Table 5.5.2 5-30 14 Table 5.5.3 5-31 14 Figure 5.5-1 5-32 14 Fgure 5.5-2 5-33 14 Figure 5.5-3 5-34 14 Figure 5.5-4 5-35 17 5-36 17 Table 5.6.2 5-37 14 5-38 14 5-39 14 SA-1 14 SA-2 14 5A-3 14 5A-4 14 Table S.A.1 SB-1 14 5B-2 14 .
58-3 14 Table 5.B.1 O DSAR iil Rev.17 i
1 r MYAPC UST OF EFFECTIVE PAGES PAGE REV REMARKS PAGE REV REMARKS PAGE REV REMARKS SECTION A.0 SECTION 7.0 APENDIX A 6-1 14 Table of Contents 7-l 14 Table of Contents M 14 Table of Contents 6-il 14 Ust of Trbles 7-il 14 Ust of Tables A-li 14 Ust of Tables 6-lil 14 Ustof Figures 7-lil 14 Ust of Figures A-lii 14 Ustof Figures 61 14 7-1 14 A-1 14 6-2 17 7-2 17 A-2 14 6-3 17 7-3 17 A-3 14 6-4 17 7-4 14 A-4 14 6-5 17 7-5 14 A-5 14 Table A.1.1 64 17 74 14 A4 14 Table A.1.2 6-7 17 Figure 6.1-1 7-7 14 A-7 14 Table A.1.3 64 17 7-8 14 A4 14 Table A.1.4 6-9 17 7-9 16 A-9 14 Table A.1.5 6-10 15 7-10 14 A-10 14 Table A.1.6 6 11 17 7-11 14 A-11 14 Table A.1.7 6 12 14 7-12 14 A 12 14 Table A.1.8 6-13 14 7-13 14 A 13 14 Table A.1.9 7-14 14 A-14 14 Table A.1.10 7 15 14 A-15 14 Table A.1.11 7 16 14 A 16 14 Table A.1.12 7-17 14 A-17 14 Table A.1.13 A-18 14 Table A.1.14 A-19 14 Table A.1.15 i A-20 14 Table A.1.16 A-21 14 Table A.1.17 A-22 14 Figure A.1-1 A-23 14 Figure A.1-2 A-24 14 Figure A.13 A-25 14 Figure A.1-4 A-26 14 Figure A.15 A-27 14 Figure A.14 A-28 14 Figure A.1-7 A-29 14 Figure A.1-8 A-30 14 Figure A.1-9 A-31 14 Figure A.1 10 A-32 14 Figure A.1 11 A-33 14 Figure A.1-12 A-34 17 A-35 14 A-36 14 Table A.2.1 A-37 14 Table A.2.2 A-38 14 Table A.2.2 4 A-39 14 Table A.2.2 A-40 14 Table A.2.2 A-41 14 Table A.2.2 A-42 14 Table A.2.2 A-43 14 Table A.2.2 i A-44 14 Table A.2.2 l A-45 14 Table A.2.3 , A-46 14 Table A.2.3 ! O DSAR iv Rev.17 i
O MYAPC LIST OF EFFECTIVE PAGES PAGE REV REMARKS A-47 14 Table A.2.3 A-48 14 Table A.2.2 A 49 14 Table A.2.3 A-50 14 Table A.2.3 A 51 14 Table A.2.3 A 52 14 Table A.2.3 A 53 14 Table A.2.4 A-54 14 Table A.2.5 A-55 14 Table A.2.6 A-56 14 Table A.2.7 A 57 14 Figure A.2-1 A-58 14 Figure A.2-2 A 59 14 Figure A.2-3 A-60 14 Figure A.2-4 A-61 14 Figure A.2-5 A 62 14 Figure A.2-6 A43 14 Figure A.2-7 A 64 14 Figure A.2-8 A45 14 Figure A.2-g MI6 14 Figure A.210 A47 14 Figure A.2-11 1 O l l DSAR v Rev.17
o MYAPC k Defueled Safety Analysis Report TABL.E OF CONTENTS (continued) 3.0 FACILITY DESIGN AND OPERATION (continued) 3.3 Systems 3.3.1 Fuel Storage 3.3.1.1 Design Basis 3.3.1.2 System Description 3.3.1.3 Design Evaluation 3.3.1.4 System Operation 3.3.1.5 Monitoring and instrumentation 3.3.2 Fuel Handling System 3.3.2.1 Design Basis 3.3.2.2 System Description 3.3.2.3 Design Evaluation 3.3.2.4 System Operation 3.3.2.5 Monitoring and Instrumentation 3.3.3 Spent Fuel Pool (SF) Decay Heat Removal (DHR) Removal System l 3.3.3.1 Design Basis 3.3.3.2 System Description 3.3.3.3 Design Evaluation g- 3.3.3.4 System Operation i 3.3.3.5 Monitoring and instrumentation 3.3.4 Ventilation Systems l 3.3.4.1 Fuel Building Ventilation System 3.3.4.2 Control Room Ventilation System 3.3.4.3 Auxiliary Ventilation Systems Rev.17 DSAR iv
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MYAPC Defueled Safety Analysis Report TABLE OF CONTENTS (continued) 3.0 FACILITY DESIGN AND OPERATION (continued) 3.3.5 Auxiliary Systems l 3.3.5.1 Compressed Air 3.3.5.2 Boric Acid Makeup 3.3.5.3 Primary Water System 3.3.5.4 Primary Vent and Drain System 3.3.5.5 Radiological Waste Processing System 3.3.5.6 Fire Protection System 3.3.5.7 Meteorological Instrumentation 3.3.6 Electrical System l 3.3.6.1 Station Offsite Power l 3.3.6.2 Station Onsite Power l 3.3.6.3 Spent Fuel Pool Island and Balance of Plant Electrical l Distribution System l 3.3.6.4 Plant Conputer and Programmatic Logic Controller (PLC) l 3.4 Control of Heaw I n=ds O c Rev.17 j DMR v
MYAPC SECTION 1J INTRODUCTION AND SUREGARY TABLE OF CONTENTS Secdon 2Ela Emot 1.1 Introduction . 1-1 1.2 Gansmi Plant Descriadon 1-2 1.2.1 ' Design Crtteris 1.2.2 Fuel M.&4 System 1.2.3 Fuel Storage System 1.2.4 RadicioghalWasts Trumament System 1.2.5 Radiological Waste Storage and Disposal System 1.3 Faculty Design Overview 16 1.3.1 f Plant SNm and Populadon 1.3.2 Structures 1.3.3 ChemicalTresernent 1.3.4 Process Instruments 1.3.5 Shielding 1.3.6 Electncal Equipment 1.3.7 System Flow Diagrams 1.4 ldanHRahon of Agents and Contrh 1-21 .
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1.5 MatanalIncomorated tw Reinance 1-22
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1 Rev.17 DSAR 14
1 I O -- SECmON 1J LJST OF PMBLDtES Pigura No. 13 1.3-1 SRs Plan 1.3-2 Deleted ] 1.3-3 Standard Symbols trFlow Diagrams 1.3 4 Flow Diagram Symbols 9 .. Rev.17 O DSAR 1-iii
MYAPC reguistions. The systems and eew nt. of the faciBty are designed to enable the facility to withstand the traditionely defined extemal fbrons that may be imposed by natural phenomena, wNhout loss of the capabigty to protect the public. The industrial Securtly ares is ancioned within ] a security innoe with at aa:mes controlled through a guardhouse.
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in the permanently defbeled condition, two power sourass are available to provide for spent fuel coolrWmehaup, fire protodion, security and emergency preparedness commurucations functions.
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1.2.2 Fuel HondHng System Fuel handing and storage facmies are provided fbr the safe handing, storage and shipment of both new and spent fuel and are demoned to preclude accidental M. The fuel buliding ; houses a new-fuel un;c. dig area, a new-fuel storage room, a spent fuel pool and the necessary cranes requesd for the handing of the fuel assemblies. Fuel shes not be moved within the Containment Building. The spent fuel pool support systems are also located in the fuel buliding. ] The spent fuel movabis piesorm and hoist is a traveing bridge whidt spans the spent fuel pool and moves on rais over useable spent fuel storage locations. The hoist hook supports ts4L4 tools for grapping and movmg fuel assemblies at a safe depth below the operational water level. The design of the fuel handung system precludes the exposure of operating personnel to abnormal radianon fields as the resut of opmunonel transients in the spent fuel poot ukewise, the demon and operation of the fuel handng system prevents the accidental exposure of a spent fuel assembly to a pc.Tl.m above or near the water surface. New or unrradiated fuel assembles have been removed from the Meirs Yankee site and thendbre do not requre a tw,JL4 system. 1.2.3 Fuel Storage System The spent fuel pool,37 feet wide by 41 feet long by 38 feet deep, is located in the south end of the fuel buuding adjacent to the reactor containtnent. The pool is constructed of romforced concrete with a waI and floor lirung of 1/4-inch thick stainless steel The walls of the spent fuel pool are approximately 6 feet thick. The toor of the spent fuel pool rests on bedrock. The fuel transfer tube, which is 36 inches in diameter, connects the containment refuenng cavity with the fuel pool. The Rev.17 DSAR 1-3 j
MYAPC The fuel pocd coohng system alsoiins3 puruknean loop w ;.s5 of a pump and two mars which may be openned independency of the fumi pool comeng system. Flow tam the discharge of the fuel pool purgication pump is directed thmugh the fusi pool prester and/or postner, while the retum ] is through the fusi pool comang retum sne. The puracation pump can aino be used for sidmmme operation or to escuists pool water through the hast .4 w. A submerged, self-contained dominersamer is instased in the spent fusi pool. Cooang water make-up to the spent fusi pool is avegabie from a variety of sources. These sources include the Pdmary Water Storage Tank through the Pnmary Water system, and, as emergency sourens, the fire water system and potable weenrtom the town of Wiscusset water supply system. soron make-up is avaanble kom the self cordained, portable, SFPI baron addulon tank, minor and pump located in the fusi building. 1.2.4 Radiological Waste Treatment System The origmal design of the R=*-*e Waste Treatment System provided for the discharge of ] g radioactMty to the environment such that decharge was mrummed at al emos in accordance wNh ] Q the requrements of 10CFR 20. The system design provided for the conecdon, storing, procemeng, ) i ..
.u.g, and disposal of solid, squid, and gaseous radioscove wastes kom the plant. The ]
principle design objective of this system was b ensure the genomi pubic was and continues to be ] protected from exposure to radioactive wastes in accordance with 10CFR 20. With the plant in a ] permanent defueled condition, the Radioeceve Waste Treatment System and interfacmg systems ] are bemg abandoned and modified to support decommesioning actMties. Those systems bemg ] abandoned and/or modified are the Baron Recovery, Gaseous, Solid Waste, and interfacing ]
% of the Pnmary Vent exi Dram System. The nonnal sources of radioecdve waste in the ]
delbeied condition includes these wastes generated from decontamination and dwi...; w ;cs ] actMbes and resuiual radioaceve precipitants/w .. ii. resulting from the storage of spent fuel ] in the Spent Fuel Pool. Modifications to the Radioecove Waste Treatment System will prcmde for ] the safe and limited me=*vi, transfer, and temponwy retention and W of radioactive waste. ] 1.2.5 Rad' A-34 Waste Storage and Deposal System Liquid waste may also be sent in liquid form to an approved vendor for treatment and disposal. ]
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. -- c ;'dj, liquid wasin may be di.c v.c direcoy from site in accordanca wah the off-site Dose ]
Calculation Manual. ] Rev.17 DSAR 1-5
WYAPC Depleted domineralizar resins from the treatment Wliquid weste may be stored in temporary hoid- ] up tanks or pumped as a aluny diracey into a ligh integrity Container (HIC). The HIC may be de- ] watered on site using an ,--d Proomms Control Procaidure and sent to burial The hic may ] also be sent to an approved vendor k oSelle procmasing and subesquent burial ] 1 Expended (Ilquid) fBlar cartridges are diMastered and pieced into HIC's or combined wHh the ] normal dry active weste (DAW) streams depending on adivity levels. FBlers placed in HIC's may ] be dowvatored on site or sent to approved oSeite vendors k volume reduction and =%=rit ] bunal ] 1 Dry active waste (DAW) is packaged in approved shipping containers and sent to burial or to an ] approved vendor for further procesang and volume reduction. ] I in au cases the weste motortel is shipped in accordance with at mira c=N= USDOT and USMtC ] reguistions for.the shipment of radioactive material Programs are in piece to monitor and control ] the proosesing and shipment W radiondive vestes. ] O 1.3 Pacility Danign overview Kay design data for plant systems and mmponents ruisvent to the deducied condhion of the Maine Yankee piant are Ested in Table 1.3.1. Data relevant to the plant whbe operating are retained in this table for historical purposes only. 1.3.1 Plant Site and Population The Maine Yankee plant is located on the west shore of the Back River.w.MT ;;y 3.9 miles south of the conter of Wiscasset, Mane. This location is shown in Figure 2.1-3. The minimum avrem radius for the she is slighty greater then 2.000 feet. The outer boundary l of the Low Paa'W Zone, as defined in the 10 CFR 100 regulations, is 6 miles from the plant. Within a 5 mBe radius of the plant site, the resident aar=% denety is estimated to average 72 persons per square mile (1990). The nearest popuisson groupng weiin 5 miles is situated around Rev.17 l O DSAR 14
MYAPC the Town of Wiscesant,3.9 mass ledE d the site. The town W Wiscusset has a parW W about 3,340 people (1990). The surrounding towns of "'.-=.4., Boothbay, Woolesch, and ' Westport leiend, whogy or partie8y within a 5 mes radius, have a popuistion of 8874 persons (1990). The city of Lewiston,24 mies WNW tom the plant site, is the nearest dty wth a popuistion in excess of 25,000. The site characteriodes are discumend in detes in Secdon 2. : 1.3.2 Structures i The makr structures on the sus are the reactor containment, primary aux 81ery bubding, Aael j building, turbine buiding, service huiding and circuledng water pump house. Except for the fuel ]
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bunding, these structures have been abandoned in whole or in part in support of the ] 5 c+: . . ' ' -# process. Referto Seftlan 3 for speelAc detais. ~
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The reactor containment is a steeMned reinforced concrete cygnder with a hemispherical dome and an aseendefy 11st renforood canants hundadon met. it served as a conAnoment barrierduring
, plant operemons and provides adequets radioman shielding for any defueled or c+:=... ' .;.5 v condition.
The turbine bunding housed the turtme generator, the component coo 8ng water heet anchangers ] and pumps, and the two diesel generators. The service building contened the main control room, switchgear rooms, shops and employee ] seceities. The primary auxiimry building housed equpment used for puii&.ani and procesang of water kom ] the reactor coolant system. The fuel building provides space for the storage of spent reactor fuel and also weste *pa=f equipment. The fuel building arrangement is discussed in detail in secnon 3. The arculating water pump house contained the crculeung water and service water pumps. I 1 I The seismic criterie to be used in the design of the structures and equipment in the station are i described in Section 3.1. Rev.17 , 1 l DSAR 1-7 i l l
1 i MYAPC 1.3.3 ChemicalTremenent in .e ,ern.,.ndy d.lud.d - a - - 1.,* a,s, - w. .e - to meinah hamn conenenen in the spentiumi pool. spent fuel pool weser is pertoscany teseed ; to answe the desired quemy. 1.3.4 Pmcess instuments Tempemeurs, pmosure, flow and Equid lami manlearing is provided as requimd to imep the ) operedne personnel insonned of piant ansar fuel seampe condnions and to provide insenadan tem which plant processes can be emiumend arreguished, instument signal transmission ihr the I pientinseuments is elecnic. The piant gessous emuerns are manmored ansar sempied fbr radioacdvity (Primary Vent Stock ] g and Spent Fund Pool). The resues of Spent Fuel Pool monitoring are depisynd and high values are ] ! annundstad. The Spent Puol Pool area radadon montaring sladon is provided to monitor radioactivity at the ] 'd Spent Pusi Pool. l
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1.3.5 Shiniding Shisk5ng is provided so that radiation exposure of personnel does not exceed the recommended Emits .of 10 CFR Part 20. The design of radadon shiniding is dependent both on .e adent of mooses rsquired to a particular incetion and on the sources of radiation adjacent to lhet location. in the delueled condulon, Al. ARA prhcipios apply to assure that personnel have taken the necessary precautions prior to acomming or worteng on contaminnend equipment. The plant is provided with a contal room having P shielding to permit occupancy during as credene accident situations. The rededon shinksng in the plant, in combinadan wnh piant radadon control procedures, ensures that opemeng personnel do not receive radiation exposures in excess of the applicable Imits of 10 CFR Part 20. R should be noted however that, in the defusied condition, control room shieldng is not requimd due to the lack of a signlRoant source term tem any of the design beels anddents. in adddon, control room venttellon is not cmdited in the safety anshees. Rev.17 DSAR 1-8 ' j
r O -- 1.3.6 Eactrical Equipment sudon service tranobrmer (X-16) is connected trough a fused disconnect swlich diracey to the ] 115 KV transmineian Enes. In the event that of-elle poweris intenupted, an on era diesel generabris avaEable fbr standby power. Following a loss of ofene power to the spent fuel pool cooing system, and considering the signincandy diminished decay heat load of spent fbal in the pool, ample Sme is avainbie fbr opermears to inRiste ansmate means of cooang ormaimup ior the spent sual pool prior to subsendel hestup orinventoryloss. annaries are instaned to suppiy any required de power. 1.3.7 System Flow Diagrams Flow diagrams lbr each plant system are incorpomhed in the appropriate secdon of this report The symbols used in thsee disgrams are shown in Figures 1.3-3 and 1.34. 9 Rev.17 I i DSAR 1-9 l o
O -- TABLE L11 (consnued) MABdE YAfAGE DENGN CHAftACTEltISTICS Design Flow (gun) 90M Design Head @) 900 Feedwater Pumps - Electrical - Number 2: 2 ^=;= $ Design Flow (spm) 14,000 Design Head (ft) 203s Feedneter Pump- Steam EMwan-Number 1 Fut CopecRy Design Flow (gpm) 28,000 Design Head (ft) 2200 Cin:uiseng Water Pumps -Number 4 Quarter-Copedly
. Design Flow (som) 10s.500 Design Head (ft) 28 menartami annermeer The inkrmanon in his outioondon is belng retained for historical purposes only.
Design Rating (MVA) 900
, .] Power Foctor .90 Tenan vonegam 22 ManB AuxRiary Systems (e) Chamlemi and Voluma Contral Syminem I The irdarmenon in this = % is being retained for Nstorical purposes only. ]
Normel Latdown Flow Rats (gpm) 80 Maomum Letdown Flow Rote (gpm) 200 Chargng Pumps -Number 3 Fixed-Copedty Demgn Flow (gpm) 150 Design Pressure (poig) - 2850 Auxmary Charging Pump - Number 1 Verloble Speed Demgn Flow (gpm) 10 to 30 Design Pressure (poig) 3M Regenerative Heat Eachonger- Number 1 fun-Copedty Design Heat Transsor(atuair) toox10' O Rev.17 l DSAR 1-13
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TABLE ili (congnued) MADE YA88GE DESIGN CHARACTERISTICS (e) Annemiery Pend avstem The infamedan in this =% is being resined kr histrical purposes only. Steam Generskr Emergency Feed Pumps (P-25A & C) Motorh -Quenety: 2 FuB CapacRy Rodng (gpm) 500 Hood (ft) 2525 Steam Generskr AumEary Feed Pump (P-25B) Turthe-Driven-Quandty: 1 Fuf Capacity Radng (gpm) 500 Head (ft) 2525 (r) .
. . , , , ..:...+..,,
4 The inknmenon in this subsecdon is being rWeined tr histrical purposes only. ] Component Coolhg Pumps-Number 2 Primary 2 Secondary Reeng, End (gpm) - 6000 Head (ft) 190 Heat 4~&.v.s. - Number 2 Primary 2 h Ay vendordets sheetinfo. (atumr) E-4A, Primary 72.5 x 1& E-48, Primary 51.3 x 1r E-5A, Secondary 51.3 x 10' E-58, Secondary 72.5 x 10' (g) anent Fuel Comung Svetam Spent Fuel Pool Storage Capacity (demgn) 1758 assemblies Spent Fuel Pool Storage Capadty (actual) 1432 assemblies 2 consoild. assemb. 2 failed rod cegos Volume (ft') 59,116 Pumps - Number 2 Rodrg, Each (gpm) 772 O Rev.17 DSAR 1-15 1 I
TABLE 1.11 (condnued) MAINE YANKEE DESHIN CHARACTERISTICS Head (ft) 120 Heat Exchanger-Number 1 Rodng (BWhr) 22.3 x 1F Phar-Number 2 Type - Certridge Rodng (gpm) 200 Nominal sha(micmns) 1 ] Dominernazar-Number 1 Roein Type miend bed and sise(r) 2a ] Nominal Flow (gpm) 100 ] Decay heat Removal System (DHR)
]
Pumpe
]
, 2 ]
Nominal Flow (GPM) 1000 andt ] Cooient weemrt ] inhytenegheof ] Design Pressure (PSIG) 300 Oj, Design Temp ('F) 350
]
]
Cooling Units 6 Type
]
Air-water ] Finned Cos ] Fene 18 ] Design Heat Land (Grm99) BTumr 2.88E+6 ] convenninnel Plant AmdRary Svmemma (e) ServienWaterSwatam The infbrmedan in this =% is being retained for e purposes only. ] savice weserPumps-Number 4 FdCopecity Rodng (gpm) 10,000 Head (ft) 66 cansminmane Sarunour. The indbrmation in this = % is being retained for historical purposes only. Type Reirdonad Concrets O Diameter (ft 4x:hes) Height (ft-inches) 135-0 169 6 Rev.17 DSAR 1-16
MYAPC O TAIRE 1.11 (continued) MAINE YAleGE DENGN CHARACTBtlSTICS Liner-MeterialThidmose (Inches) ASTM A516 Grade 80 was sa Dame % Ploor 1/4 Design Pressure (paig) 55 Dealgn Temperature (F) 280 Leak Rete (percent perday) 0.1 O 3 O Rev.17 DSAR 1-17
MYAPC 1.4 + 2."- " =. at *
.:. =.2 c . n a Meine Yankee Atomic Power Company was the scde petitioner h the operating license. The
]
Company was organized by eleven New England udty companies for the purpons of conneucang and operedng the Meme Yankee nucieer generaung piant in Wine ====t Maine. Meine Yankee, as owner and operahr of the sladon, was responedde h the design, ] conneuedon, fabricadon of components, operadon and questy aneurance.
~
Maine Yankee Atomic Power Company is responalbie for the c+ ... - - Aq of the incey, ] The prtndpoi contractor organa:sdons associated with the construcdon and operadon or
-f+: .. E ;.4 of the Maine Yankes plant are: ]
]
Entergy Nudeer, Inc. provides operadng and d+ ... ' ' .;.6 management.
]
The Nudear Services Division of the Yankee Atomm Electnc Company (now Duke LW ::ts and Services) in the performance of ergir : -;.4 studies, design reviews, plant modications, and construction coordination. Combustion Engheering (CE), Inc.(now M as the designer and suppdor of the Nucieer Simem Supply System (NSSS). The NSSS indudas the reactor cooient system, reachr auxEery system components, nudeer and certain proomes instrumentadon, and reactor cantns and protecnon system. The Stone & Webster L@ : M.g Corporation as the designer and supplier of the balance of the plant equipment and structures. Addidonney, Sene a webster constructed the balance of plant with technical advice provided by CE for instacation of the reactorplant components. Stone & Websteris the C+=i.. '
'#_- 4 ]
Operations Contractor (DOC).
]
The wesenghouse sectnc Ccw As. as the anginal supplier and ==cer of the turtnnes and elecencei generaer. The low pressure turtunes were rapieced dunns Maine Yankee operations with equipment from Asee Brown-Bovart Combustion LS _ _..q, Wes8nghouse, and EXXON Nudear (now Siemens Power C+.,a..) were suppliers of nucieer fbei at vanous time during the operaeng INeems of Meme Yankee. Rev.17 DSAR 1-21 i
MYAPC 1.5 Material Inenrnarated av Reference Cartain program documents and h topcal reports or analyses have been 6m 1 into the DSAR by reference and are Ested in each secdon as approprints. This documentuden may include information developed by Meine Yanhos, as wet as Yankee Atomic, ABS-CE, weednghouse, Stone and webster, and cIher orgonhadons. Some documentadon that is 6 m - _1 by reference continues to be updmind to assure that the information used is the latest available. These documents indude the fotowing:
- 1. Queuty Assurance Prograrn
- 2. LT-v .iPlan
- 3. Security Plan
- 4. Fire Protection rW .. )
- 5. Off Site Does Ndatir=1 Manuel ]
- 6. Process Contml Program ] {
- 7. Post Shudown Dommmesioning Acevides Report I
]
- a. Technical & J Each of these programs and plans may be modNied as necessary in acccsdance wnh the
~ ~ v ~ r,.ed.,s.od.<
q l o i DSAR 1-22 Rev.15
MYAPC 2.2.7.2 Picy-.. Scope To determine whether or not there is a signitcant eNoct by the plant on the sunounding environment, a two mone sample erdar*m method is employed for rnost media. The zones have been designated as fotows: I Zone 1- The ares wnhin appmuimately a 54 nile radius from the plant site. This area is considered under the inAuence of the plant. Zone 11 - The area outside =-;--M, " '; a 5-mile radius from the plant site. This ares provides background data for the environmental surveys. Since data from each of the two zones are correlated, there is simultaneous monitoring in both areas to establish and incNitate a statistical analysis of the survey data. This allows Maine Yankee to dlNorentiate between plant releases and other abnormal trends in environmental radioactMty due to indout from almospheric .h weapons tests or other sources of radiondMty. Dinact radiation monitoring locations are grouped into an inner ring and an outer emergency response nng, as wed as a control group. 2.2.7.3 Picy-i. Sample Media The program monitors four pathway categories. They are the direct radiation, airbome, waterbome, and ingestion pathways. Each of these categones is mondored by the collection of one or more sample media, which are listed below, and are described in more detail in this section. Additional
]
sampling media may at times supplement those included in the basic pathway categories. Airborne Pathway Waterbome Pathway Air Partculate S-Twiing Estuary Water Sampling Ground Water Sampling l Shoreline Sediment Sampling ingestion Pathway Direct Radiation Pathway Milk Sampling TLD Monitoring Fish and inver1obrate Sampling I Rev.17 =- DSAR 2-22 1
F l MYAPC
- 1. Airtsorne Mondoring Air irivisitu irs stations are established at a total of five locations, four of which are Zone I locations. Airbome particulates are collected by passing air through a 47 l mm fibergises fBer. Air sempier pumps operate continuously and a dry gas motor l is iricispu 11 into tie samp8ng stream to measure the total amount of air sampled in a given interval. -
The air particulate fIters are collected biweeldy and analyzed for gross beta l radioactivity. siwesidy air pareculate meers from each location are wiw: d l quarterty and analyzed for gamma emitting radionuclides. l
- 2. Waterinome Pathways .
A composite sampier at the plant outfall area collects an aliquot of estuary water at least every two hours. On a monthly bass, this ci,,i,,c,e;1ed sample, as well as a grab sample fham the control location, are analyzed 'T genut _ .o g radionuclides. The samples are composited at the erdf,c.mi i laboratory for quarterly tritium analyses. The collection of fresh or ground water samples is not required since no source of water used for disii; dig or irrigation purposes is in an area where the hydraulic gradient or recharge propernes are suitable for contamination. However, samples may be periodically co8ected to provide additional information regarding potential contamination of ground water media. l Sediment cores are maar4ad at both the former and the current discharge areas I on a semiannual schedule. Each sample is analyzed for gamma-emitting radioactivity.
- 3. Ingestion Pathways Milk samples are maar4ad monthly from two locations. The samples are analyzed l for gamma-emitting radionuclides. Samples of food crops from indicator and l control locations are needed only when milk sampling is not being performed since both assess radioactivity levels in the same pathway. At least two commercially or recreationally important marine biological specimens (such as fish, mussels, crabs, and lobsters) are collected at the Long Ledge (discharge) area as well as Rev.17 DSAR 2-23
MYAPC at a contml location on a semiannual or seasonal schedule. Au samples are analyzed for gs.r.i ai.;;; .v radionuclides.
- 4. Direct Radladon Pathway Direct gamme radiadon measurements are obtained at as seasons, which consist of inner ring, outer ring, and control locations. Thermoluminescent Dosimeters (TLDs) are employed to record the integrated gamma radiation exposures over a quarterly period.
2.2.7.4 Emergency Survealance The radiological environmental monitoring program surveitances required when the Emergency Plan has been activated are addressed in the Emeqpncy Plan. 2.2.7.5 Provi T Evaluation and Reports A report on the radiological environmental monitoring program is submitted annusHy to the USNRC. j The report contains a summary, interpretations and an analysis of trends for the results of the radiological environmental surveillance activities for the report period. Included are comparisons with operational contrais and previous environmental surveitance reports, plus a description of the radiological environmental program and maps of au sampling locations. An assessment of the observed impacts of the station on the environment is also included. 2.2.7.6 Land Use Census A census of the relevant land use activities surrounding Maine Yankee is performed annually to ensure that the optimum milk sampling locations are being mondored (or food crop locations if milk sampling is not being done). Specifically, the location of the nearest milk animal, the nearest garden of greater than 50 square meters, and the nearest residence in each of the 16
.T ^ sr4ical sectors within five miles of the plant is identified. Dose calculations are done to determme the optimum milk or food sampling locations.
2.2.7.7 Sample Locations The sampling and monitoring points for the measurements involved in this surveillance program are presented in Table 2.2.10, and locations are shown on Figures 2.2-2 through 2.2-7. DSAR 2-24
MYAPC TABLE 2.2.8 ENVIRONMENTAL MONITORING PROGRAM - PROFILE MEDIA Numberof Sample Type Sampling Frequency Required Analyses Sample Locations Air Particulate Weeldy Gross-beta 5 Quarterly Compoede Gamma WWy Ground Water
- Quarterly H-3, 2 l gamma Wwy Estuary Water Monthly Composite Gamma spectroscopy 2 Quarterly Composite H-3 Dinoct Radiation Quarterly integrated gamma 38 dose O
- Groundwater samples shall be taken when this source is tapped for drinking or IUg' uen purposes in areas where hydraulic gradient or recharge properties are suitable for contamination.
Rev.17 DSAR 2-33
y MYAPC TABLE 2.2.9 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM - FUNCTIONAL MEDIA Nutr.ber of Sampling Sample Sample Type Frequency Required Analysis Locations Milk Monthly Gamma ww 2 l , Food Crop
- Monthly Gamma spectrasmpy 2 l (3 types of broad leaf vegetation)
Sediment Semiannually Gamma spectroscopy 2 Fish and invertebrates Semiannually Gamma ww 2 (2 saw.wMy or orin Season recreationalimportant l species) Performed only if milk sempling is not done. Rev.17 DSAR 2-34
MYAPC TABLE 2.2.10 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM LOCATIONS E=a usrecuan Station Designation Location From Plant From Plant Dan) Air Particulate and Charcoal Filter AP-11 Montsweag Brook 2.7 NW l AP-13 Bailey Farm (ESL) 0.7 NE l AP-14 Mason Steam Sts#on 4.8 NNE l AP-18 Westport Firehouse 1.8 S l AP-29 Dresden 9J W 20.1 N l r.stuary water WE-12 Plant Outfall 0.3 SW hE-20 Kennebec River 9.5 SW arounct water WG-13 Bailey Farm (ESL) 0.7 NE j WG-24 Morse Well 9.9 W O ' - ' - FH/MU/CA/HA-11 Long Ledge Area 0.9 S FH/MU/CA/HA-24 Sheepscot River 11.1 S 5::: .;.; SE-18 Old Outfall Area 0.6 S "SE-18 Foxbird Island 0.6 S MIIR l TM-18 Chewonki Farm 1.9 WSW TM-25 Hanson Farm 18.3 W Direct Radiation TL-1 Old Ferry Road 0.9 N TL-2 'Old Feny Road 0.8 NNE TL-3 Bailey House 0.7 NE TL-4 Westport Island , Rt.144 1.3 ENE TL-5 MY information Center 0.2 E TL-6 Rt.144 and Greenleaf Rd 1.0 E TL-7 Westport Island, Rt.144 0.9 ESE TL-8 MY Screenhouse 0.2 SE TL-9 Westport Island, Rt.144 0.8 SE
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o to e I utsprTras O FIL*RE 2.2-4 Environmen si Radiologies! Sampling Loes-ions Cu:sido 12 Kilome:ers from }!aine Yankee DSAR Rev. 17 2-40
MYAPC 2.3.3 Probable Maximum Flood The Service Water / Circulating Water Pump House and the Fuel Oil Storage Facility have been l abandoned, but roman in piece. The discussion concemog probable maximum flood and neve runup l le retained for historical ruierenos. l 2.3.3.1 Maximum Water Surface Elevation An 'swestigation was made to predict the probable maximum flood level which could occur at the site of the Mame Yankee Atomic Power Station on the Sheepscot River estuary when the probable maximum hunicane is taken as me design basis meteorological event. The investigation is based upon the parameters of the probable maximum hunicane as defined by U.S. Weather Bureau Report HUR 7-97, Interim Report- Meteorological Charactenstics of Probable Maximum Hunicane, Atlande and Gulf Coasts of the United States (Reference 1). This ir ;::"-; " i shows that the maximum wate; levels at the Maine Yankee Power Station due to the probable maamum hurricane are predicted to L a at Elevation 19.9 feet and Elevation 21.4 feet on the plant site and screen wet structure, respectivsly. These levels are based upon the simultaneous occurrence of the maximum storm surge, maximum predicted astronomical tide, an initial rise in mean sea level, estuarian ampilScotion, the probable maximum flood in the Sheepscot watershed, maximum waves in Montsweeg Bay and existence of a channel restriction at the former Cowseagan Narrows C=t::::y. Removal of the causeway and replacement with a bridge increases the degree of conservation of this original work because the causeway acted as a dam to the hunicane surge and caused " water pile-up" at its face. This dam effect is induded in the maximum water elevation mentioned above. As a result of the causeway removal, the surge resulting from the probable maximum humcane can now travel wen up the estuary, thereby lowering the maximum water levels at the plant site. Thus, the maximum water levels presented above are higher than what could occur during a probable maximum flood caused by a probable maximum hunicane. ! Safety measures have been implemented in the design of the plant regarding this design basis i flooding. The screen welis protected up to Elevation 22 feet 0 inches, while the floor grade of the ; I pnncipal power station structures is Elevation 21 feet 0 inches. There will be no significant risk of flood at the site sece the mmimum shore i- , ,i Elevation of 20 feet 0 inches and site grade which vanes from Elevation 20 to 21 feet should preclude water from entering. Rev.17 DSAR 2-47
MYAPC pumps. The design extreme low river water levelis minus 7 feet at the service water pumps. This level has been exponenced occasionally during the plant opw.iing history. The recommended mrumum submergence of the service water pump suction bell to prevent vortex formation and flow interruption is 4 feet. The elevation of the service water pump suction bed mouth is minus 14 feet, 4 inches. Asauming a maximum level drop of 0.5 feet acmss the travehng water screens and the design extreme low river water level of rmnus 7 feet, the mmimum service water pump suction bell submergence would be 6 foot,10 inches which exceeds trurumum recommended submergence with a margin of about 3 feet. 2.3.4 ics Loading, Oil SpiH and Debns Blockage The intake structure is protected by a concrete curten wall which extends across the face of the intake structure and from Elevation 21 feet 0 inches down to -7 feet 0 inches as shown in Figure 5.4-
- 6. The curtain walls provide protection aganst floating objects causing damage to the water intakes.
Submerged objects could damage the intake trash racks or restrict flow though one or more of the intake channels. it is highly improbable that submerged objects would exist in the Back River which could O simultaneously block all intake channels. Ice wRI not form at the intake channels to a depth to l obstruct intake flow. Frazile ice will not form to block the intake flow. The Central Maine Power Company Mason Station has three oil storage tanks located at Birch Point in Wiscasset. One tank has the capacity of 100,000 barrels of oil; the other two have capacities of 132,000 barrels each, malang a total oil strage of 364,000 barrels at the site. It is most unikely that any oil contaned in these tanks could reach the water and have an influence on the safe operation of Maine Yankee. AH of the tanks are enclosed by dikes which would contain any oil coming from any tank within the dike ares. Manually-operated valves in the bottom of the dikes for the purpose of draining off water are maintained in the closed position. Should a tanker spill oil during delivery, thers is the possibility that oil could advance toward the area of Mene Yankee. However, the Maine Yankee intake structure provides a curtain waH which projects 7 feet below the mean sea level. This would prevent oil c7 surface-bome debris from entering the service water system. Therefore, it is extremely unlikely that an oil spill of any kind will cause an operational problem at Mene Yankee. DSAR 2-51 Rev.17
O MYAPC SECTION 3.0 FACIUTY DESIGN AND OPERATION 1 TABLE OF CONTENTS Saclion Illla Eaga 3.1 Danign.QBada... ... . .. ..... . .. . ..... ....... 3-1 3.1.1 Conformance wNh 10CFR 50 Appendix A General Design Creens 3.1.2 Classi6 cation of Structures, Systems, and Components 3.1.2.1 SSCs important to the Defbeicd Condition Of.2.2 Wind, Missile, and Tomado Loadings 3.1.2.3 Water Level (Flood) Design 3.1.2.4 Seismic Design 3.2 structures . ........ 3 27
, ]
3.2.1 Fuel Building
,p 3.2.1.1 General .
!V 3.2.1.2
~
Fuel Unioeding Area 3.2.1.3 New Fuel Storage 3.2.1.4 Spent Fuel Pool 1 3.2.1.5 FuelStorage Racks l 3.2.2 Storage BuAdings 3.2.2.1 Underground RCA Storage Bunker 3.2.2.2 Radiation Controled Area (RCA) Storage Building 3.2.2.3 , LSA Storage Building 3.2.2.4 Warehouse 3.2.2.5 Low Level Waste Storage Building 3.2.3 Serwce Bulk 5ng 3.2.3.1 Control Room Area 3.2.4 Turbine BuBding 3.2.5 - Primary AuxNory Building 3.2.6 Service Waterintake Structure 3.2.7 Fire Pump House 3.2.8 Masonry Wags O DSAR 3-i Rev.17
MYAPC SECTION 3.0 FACIUTY DESIGN AND OPERATION TABLE OF CONTENTS SAldkM1 M 2ast 3.3 Svstems 3.3.1 Fuel Storage. . ...... ....... . 3-39
]
3.3.1.1 Design Basis 3.3.1.2 System Description 3.3.1.3 Design Evaluation 3.3.1.4 System Operation 3.3.1.5 Monitoring and instrumentation 3.3.2 Fuel Handing System... 3-73
]
3.3.2.1 Design Basis 3.3.2.2 Svstem Description 3.3.2.3 Design Evaluation 3.3.2.4 System Operation
]
3.3.2.5 Ir= par *wi and Testing ] ) 3.3.3 Spent Fuel Pool (SF) Decay Heat Removal (DHR) Removal System.............................................................................3-78 ] 3.3.3.1 Design Basis 3.3.3.2 System Description 3.3.3.3 Design Evaluation 3.3.3.4 System Operation 3.3.3.5 Monitoring and Instrumentation 1 1 Rev.17 DSAR 3-il
MYAPC SECTION 3.0 FACALITY DESIGN AND OPERATION TA8LE OF CONTENTS 3 i Sadion IRla Eaga 3.3.4 ventsation Systems. . . ... ...... . .. ...... ..... 345 )
,. 3.3.4.1 Fuel Building Ventilation System
]
3.3.4.2 AuxNiery Ventilation Systems
]
3.3.4.2.1 Primary Auxdiary Bu5 ding Ventilation ] 3.3.4.2.2 Containment Buliding Vendinistion System ] 3.3.4.2.3 Containment Spray Building Ventuation System J 3.3.4.2.4 RCA Budding VentBstion System ] 3.3.4.2.5 Front Omce/ Control Room Building VentHation System ] 3.3.4.2.6 Administrabon Buudng (WART Building) ] , 3 3.3.5 Auxdiary Systems...... ..... .......... .. ................. 3-90
]
3.3.5.1 Boric Acid Make .
]
3.3.5.2 Primary Water System ] 3.3.5.3 Primary Vent And Drain System ] 3.3.5.4 Radioactive Waste PrM g System ] 3.3.5.5 Fire Protechon System ] 3.3.5.6 Metecse;egicisiinstmmentation ] 3.3.6 Electrical Systems....... .... ... ................................................. 3-100 ] 3.3.6.1 Offsite Power 3.3.6.2 Station Onsite Power ] 3.3.6.3 Spent Fuel PoolIsland and Balance of Plant Electrical ] Distribution System ] 3.3.6.4 Programmatic Loge Controller (PLC) ] 3.4 Contml of Heavy Loads ...... 3-111
....... . .. ..................... ..................... . ]
Rev.17 DSAR 3-iii
O -- SECTION 3.0 FACluTY DESIGN AND OPERATION USTOFTABLES M Illia q > 3.1.1 Earthquaka Damping Factons 3.2.1 MasonaryWalls
]
O l I O DSAR g Rev.17
O SECTION 3.0 FACluTY DESIGN AND OPERATION UST OF FIGURES Figum No. ))la - 3.1-1 Response Spectra for 0.05g Maximum Ground Accelemtion 3.1-2 Response Spectra fbr 0.10g Maximum Ground Acceleration 3.2-1 Fuel Building Anangement
; 3.2-2 Fuel Building Arrangement 3.2-3 Deleted l 3.2-4
)
- Deleted i ]
3.2-5 Deleted
]
3.2 4 Deleted
]
3.2-7 Deleted
]
3.2-8 Deleted
]
3.2-9 Deleted
]
3.2-10 Deleted
]
-l 3.2-11 Deleted
]
3.2.12 Fire Pump House 3.2-13 FuelTransferTube 3.3-1 Typical Fuel Assembly 3.3-2 Typical Fuel Spacer Grid 3.3-3 Contml Element Assembly 3.3-4 Contml Element Assembly 3.3-5 Spent Fuel Pool Assembly Placement Lirmtations 3.3-6 High Density Spent Fuel Pool Layout For The Two Region Pool 3.3-7 Deleted ] 3.3 3 Deleted ] 3.3-9 Fuel Pool Cooling Piping 3.3-10 Deleted ] 1 3.3-11 Deleted ] 3.3-12 Deleted ] 3.3-13 Deleted ] Rev.17 DSAR 3-v
O - SECTION 3.0 FACluTY DESIGN AND OPERATION UST OF FIGURES F10um No. -I]lla 3.3-14 Deleted
]
3.3-15 Deleted
]
3.3-18 Deleted
]
3.3-17 Deleted
]
3.3-18 Deleted
]
3.3-19 Deleted
]
3.3-20 Deleted
]
3.3-21 Fire Protection System 3.3-22 Fire Protection System 3.3-23 ' - One Une Diagram - Transmission and Utility lii-iwnnections with MY 3.3-24 One Line Diagram - Auxiliary Power System l 3.3-25 Deleted
]
g% Rev.17 DSAR 3-vi
O uwe SECTION 3.0 FACILITY DESIGN AND OPERATION Section 3.0 discusses the design and operation of the structures, systemt, and components required to sainly secre fuel. It also h- supporting systems such as venesson and auxemy systems used to safely store fuel or support decontamination and a=-:+. . =- v.;.w activities. 3.1 Danige Cr$mrla 3.1.1 Conformance with 10CFR 50 Appendix A General Design Criteria in July of 1967, the Atomic Energy Commission issued the proposed general design criteria for nuclear power plants. These 70 criteria were issued for comment by the industry but had not yet been adopted as a reguistorvoquirement. Nevertheless, as the following dier==ian shows, the Maine Yankee plant has been designed and constructed in accordance with the intent of these criteria. In the following paragraphs, each criterion is stated and its conformance indicated. On September 18,1992, the USNRC confirmed that plants with construction pernts issued prior to O May 21,1971 were evaluated on a plant-specific basis, and beciu"u; 4 the cunent General Design Criteria of Appendix A to 10CFR 50 would provide little or no safety benefit. (Reference 1, SECY-92-223) On August 7,1997, Maine Yankee certilled in accordance with 10CFR 50.82 that the company had permanently ceased power speietice and that all fuel was removed from the reactor vessel (Reference 2, MN-97-89). With the dcx:keting of these certifications, the Maine Yankee license no j longer allowed operation of the reactor or placing of fuel in the reactor vessel. The AEC 1967 Design Critena listed below are relevant to the permanently defueled plant condition. ANALOGOUS 1971 CRITERIA NUMBER CRITERIA Group 1 -Overall Plant Requirements Quality Standards 1 1 Performance Standards 2 2 Fue Protection 3 3 Shanng of Systems 4 5 Record Requirement 5 1 DSAR 3-1
g MYAPC Group 111 - Nuclear and Radianon Controls Control Room 11 19 I instrumentation and Control Systems 12 13 Monitoring Radioactive Reiseses 17 64 Monitoring Fuel and Waste Storage 16 63 Group vill- Fues and Waean stormee systern Pnnention of Fuel Storage Criticality 66 62 Fuel and Waste Storage Decay Heat 67 61 , Fuel and Waste Storage Radiation Shielding 66 61 2 Protecdon Against RadioactMty Release from Spent Fueland Waste Storage 69- - Group IX- Plant Emuents Control of Release of Radioactivity to the Envuonment 70 60 Group I-OveraH Plant Requirements Criterion 1 - Quality Standards Those systems and ceinpor,eid. of reactor facilities which are essential to the prevention of accidents which could afisct the public health and safety, or to mitigation of their consequences, shall be identified and then designed, fabricated and erected to quauty standards that reflect the Irroh of the safety function to be performed. Where generaly recognized codes or standards on design, matartais, fabrication, and inspection are used, they shen be identined. Where adherence to such codes or standards does not sumco to assure a quality product in keeping with the safety function, they shall be supplemented or modified as necessary. Quality assurance programs, test procedures, and inspection acceptance levels to be used shall be identified. A showing of sumciency and applicabmy of codes, standards, questy ensurance programs, test procedures, and inspection acceptance levels used is required. (This Cntonon is directly armlogous to Cnterxy11 of 10CFR 50, Apper.dk A,1971, except that the 1971 version addresses records retention. Maine Yankee record retention requirements are addressed in the Quality Assurance Program. Design, fabrication, and construction records are ] addressed by Criterion 5 of this secdon.)
)
Rev.17 DSAR 3-2
WAPC Rasoonse: Those systems and components which are essential to the prevention of accidents which could a5ect the public hesith and safety, or to mitigation at their consequences, have been designed, fabricated, and wected e questy standards that renect the importance of the safety funedon to be performed. Generally recognized codes and standards on design, materials, fabricadon, and inspection have been used. These wwe supplemented to redoct cunent pracdcas. The t: s - of the systems and components to which this criterton applies include the codes and other standan$s met by these systems. The quality assurance program is submitted to the regulator for review and approval of any proposed revisions which would result in a reduction of previous commitments.
References:
Sections 3 and 4 Cittadon 2 - Performance standards Those systems and components of reactor facilities which are essential to the g_c,Li of accidents which could asset the public health and safety, or to nW@bi of their consequences, shaN be designed, fabricated, and erected to performance standards that will enable the facGity to withstand, without loss of the capabdity to protect the public, the additional forces that might be imposed by natural phenomena such as earthquakes, tomadoes, flooding condibons, winds, ice, and other local site e5ects. The design bases so established shan reflect: (a) appropriate consideration of the most severe of these natural phenomena that have been recorded for the site and the sunounding area, and (b) an appropriate margin for withstanding forces greater than those recorded to reflect uncertasilles about the historical data and their suitability as a basis for design. (This Cntanon is directly analogous to Criterion 2 of 10CFR 50, Appendix A,1971, except that the 1971 version addresses consideration of appropnate combinations of the effects of normal and accident conditions with the e5ects of the natural phenomena and the importance of the safety functions to be performed.) Reasonse: Systems and components which are essential to the prevention of accidents which could a5ect the public health and safety, or to .ie;w.bi of their consequences, are designed, fabricated and erected to perfonnance standards that enable the facNity to withstand, without loss of the capability Rev.14 DSAR 3-3 l
MYAPC to protect the public, the ad Silonal forces that trught be imposed by natural phenomena such as earthquakes, tomadoes, Sooding conditions, winds, ice, and other local site effects. The design
' bases so established redoct appropriate consideration for the most severe natural phenomena that have been recorded for the site and sumzunding area, and appropriate margin Ibr withstanding forces greater than those recorded to ro6ect uncertainties about the historical data and their suitability as a basis fordesign.
On April 17,1979, an eartiquake of approximate magrutude 4 occuned about 10 lolometers west of the plant. On January 9,1982, an earthquake of approximate magnilude 5.75 occuned in Central New Brunswick. Ad=ary=rilly, the Licenseejoined with the Regulator in a program to assess the seismic ruggedness of the Maine Yankee plant. This program is eferred to as the Seismic Design Margins PrevieT. (SDMP). This program is fully described in NUREG/CR426, dated Man:h 1987. On March 26,1987, the Reguistor issued a Salisty Evaluation Report whidi concluded that ad issues assooated with the seisnue design were considered resolved (Ref. 3). i
References:
Sections 2 and 3. Critarion 3 - Fire Pro 8=<*wi The reactor facility shal be designed (1) to minimize the rovbeb"'ty of events such as fires and avplaarvis, and (2) to minimize the potential effects of such events to safety. Ncs.s,T4,ustible and fire resistant materials shad be used whenever practical throughout the facility, particularty in areas containing entical portions of the facility such as containment, control room, and cxamponents of er.girxM safety features. (This Cnterion is directly analogous to Cnterion 3 of 10CFR 50, Appendix A,1971, except that the 1971 version addresses fire protection and fire fighting considerations.) Resoonse: As described in Section 3, the materials and layout used in the station design have been chosen to minimize the possibility and to mitigate the effects of fire. Sufficient fire protection systems and equipment have been provided to mininuze the adverse effects of fire on structures, systems, and components i.Tif,citerd to safety taking into account the dect,r.r,; kat,v plant condition and actMties. Rev.14 DSAR 3-4
MYAPC Critarion 5 - Record Ramamment Records of the design, fabricadon, and construedan of essendal components of the plant shaN be maintained by the reactor operator or under its control throughout the Nfe of the reactor. (This Critorion, in combination wnh Criterion 1 of this secdon, is analogous wth Crlierlon 1 of 10CFR
' 50, Appendix A,1971.)
Raw' 4 Records of design, fabricadon, and construction of components are being maintained for the duration of the license. Desipi calculations are in the possession of Stone & Webster and Yankee Atomic. AN other required design, fabrication, and construction information is in the E-:::::4: of Maine Yankee.
References:
Section 6 and the Quality Assurance Program GROUP 111 - NUCLEAR AND RADIATION CONTROLS k Crttarion 11 - Control Room The facMity shaN be provided with a control room fkom which actions to maintain safe operational status of the plant can be controNed. Adequate radiation protecdon shaN be provided to permit access, even under accident conditions, to equipment in the control room or other areas as
- necessary to shut down and maintain safe control of the facdity without radiation exposures of personnel in ave === of 10CFR 20 limits. It shaN be possible to shut the reactor down and maintain it in a safe condition if access to the control room is lost due to fire or other cause.
Ramoonna: This crtterion is met only to the extent that radiation exposures of personnel in excess of 10CFR 20 Nmits cannot occur based on the available source term from any credible accident. Sufficient ume is avaliable to allow operators to restore cooling or makeup to the spent fuel pool and maintain { exposures well below the Emits of 10CFR 20. Controis for spent fuel pool cooling, makeup and purtlication are located near en equipment and are not in the control room.
References:
Sections 3,5, and 7 Rev.14 DSAR 3-5
_y l MYAPC j Critarian 12 - Instrumar*% and &wdral h- ' . Instrumentation and aantrois shal be provided as required to monitor and maintain variables within prescribed operating ranges. namoonsa.- Instrumentation is provided as required to monitor and maintain significant variables. Controls are provided for the purpose of maintaining these variables within the limits prescribed for safe opwation. In the permanently defueled condition, the principal variables to be monitored include
; suoi pool level, and temperature. Baron concentranon is determined via sampung.
References:
Sections 2,3, and 4 Critarion 17 - Monitoring Radianceve Relaamas Means shen be provided for monitoring the contenment atmosphere, the facWty emuent discharge palhs, and the facWty environs fori=*=*vity that could be released fkom normel opwadons, kom anucipated transients, and from accident conditions. (This Criterion is analogous to Criterion 64 of 10CFR50, App. A,1971, except that the 1971 version addresses spam cent.;,4 g components for recirculation of loss ofa:oolant accident fluids.) e 898000EE The means for monitorinC radiation levels in the spent fuel pool and concentrations on site is ] ps7 sided. The sensitMty and range of this equipment is adequate for operating and anMp#M transients in the permanently defueled condition. Effluent discharge paths where a potential release of radioactive material exists are monitored. j Monitoring equipment has sufficient sensitivity and range for operating and anticipated transients ; in the pwmanently defueled condition. Instrumentation is provided for monitoring radiation levels in the spent fuel pool ares. ]
]
Rev.16 DSAR 3-6
')
O "" An erwonmental surveellance program has been established. The facility's environmental program prowdes effective iinnektu s of radioactive material released from the plant.
References:
Section 4. Crtherton 18 - Moniensino Fumi and W" Atarna. MonW and alarm instrumentation shall be provided for fuel and waste storage and handling armes for conditions that mght contributs to loss of continuity in decay heat removal and to radiation exposures. Resoonse: The spent fuel is monitored by an area detector with audible and visual alarms activated at the detector location and in the control room. For waste handling operations or decontamination operations, portable detectors with audible alarms may be used to monitor radiation exposure. , The spent fuel pool is equipped with high and low liquid level alarms, and a high temoerature alarm which M indicate loss of continuity in decay heat removal capacity of the fuel pool cooling system.
References:
Section 3 and 4 GROUP Vill - FUEL AND WASTE STORAGE SYSTEM Critorion 68 - Pr;;;.T.ca of Fuel Sterna. Critir mmy Criticality in new and spent fuel storage shall be prevented by physical systems or pror*=aar. Such means as geometrically safe configurations shall be emphasized over procedural controls. Ramponsa: Criticality is prevented by geometrically safe configurations. ] Rev.17 DSAR 37
O Irradiated spent fuel is stored under water in a restforced concrete pool, lined with stainless stool. Fuel assemblies are spaced and the racks are so fabncated that enticality is precluded. Although the waterin the pool is generally borated, neither soluble baron nor control rods are required to keep even unwradiated fbol assemblies subcritical. The baron concentration required as a result of the analyzed "mispieced assembly" incident is h w in section 3.3 and 5.2.
Reference:
Section 3.3. Critorion 87 - Fuel and Waste Storage Decav Heat Reliable decay heat removal systems shall be designed to prevent damage to the fuel in storage facilities that could result in radioactivity release to plant operating areas or the public environs. Rasoonse: The spent fuel pool is designed to maintain the water level safely above the spent fuel assemblies at all times. The fuel pool outlet pipe, which serves as the suction pipe to the fuel pool cooling pumps, is siphon 94 to prevent draindown of the pool. The retum piping is located such that siphoning by the cooling system is limited to no less than 10 feet above the active fuel. Additionally, ] a branch syphon break has been instadled to increase this limit to no less than 19' above the active ] fuel. Pool water level may be restored through diverse (offsite and onsite) supplies of fresh or domineralized water. Emergency makeup water and cooling is also available from the fire pond.
Reference:
Section 3. Cntenon 68 - Fuel and Waste Storage Radiatx)n Shielding Shielding for radiation protection shall be prtmded in the design of spent fuel and waste storage facilities as required to meet the requirements of 10CFR 20. (Also refer to Cnterion 67 of this DSAR section.) Rev.16 DSAR 3-8
MYAPC Rannonna: The radiation shielding (water) of the spent fuel pool is designed to pmvide adequate personnel protection. The weses storage and proosesing focaties are shielded as required to protect personnel som exposures in excess of reguistory requirements. In addnion to shielding design,
;..A. nx, of em Radiation Protecdon Plan requirements assures that doses to personnel performing work in these areas are maintained ALARA.
Reference:
Section 3 and 4.
\
crearian as - r, -/- "--, A-- ; p;aeha r - - :0,;i n.=_,; Fuel and W "- S c,= Containment of fuel and waste storage shal be provided if accidents could lead to release of undue amounts of radioactMty to the public environs. Ramponna:
, Fuel and waste storage areas are designed to preclude the inadvertent release of undue amounts of radioactMty. Considenng the significantly dminished source terms in the permanently defueled condition, there are no credible accidents resulting in doses to the public approaching the 10CFR 100 Emits. AB spent fuel and waste storage systems are w, : r; "d; designed with ample margin
, to prevent the possibility of gross rnochanical fagure which could release significant amounts of radioactMty. Backup systems such as noor and trench drains are provided to collect potential leakages to preclude the release of radioeceve materials to the environment. Personnel are rigorously trained and administra#ve procedures are strictly followed to reduce the potential for human error. In addition, indiological Emits on waste storage systems are established to assure that credible accdont conditions wlE not result in doses to the public which eppicidi 1 rom (whole body) exposure over a 2 hour-limited duradon accident. The coriamy a of a fuel handling incident arepresented in Sections 3 and 5. In this analysis, it is demonstrated that undus amounts of radioactMty are not released to the public.
References:
Sections 3,4,5 and 7. Rev.16 DSAR 3-9
R O GROUP IX- PLANT EFFLUENTS c *._L. 70. r . -.; J d " '- - - - M P21 -.ed; to the "a. .* ._- - .i-The fiscWty design shed include those means m b mainten control over the plant radioactive affluents, whether gessous, Equid, or solid. Appropriate hoidup cepecity shes be provided for retention of geneous, Equid, or solid affluents, particulerfy where unfavorable environrnental condisons can be expected to require operanonal nmitations upon the release of radioeceve emuents to the environment. In aN cases, the design for radbecevity control shen be Justilled (s) on the basis of 10CFR 20 requirements for normal operation and for any transient situadon #1st might reasonably be anW to occur and (b) on the bass of 10CFR 100 dosage level guidednes flor potential reactor accidents of =W low probabiDty of occurrence except ! that reducson of the recommended dosage levels may be required where high populanon densines or very large cities een be arrected by the radioeceve emuents. Resonnes: The pient r=rer=reve weste control systems (whicn include the Iquid, gaseous and solid redweste systems) are designed to umit the off-site radiation exposure during normel operation to levels below limits set forth in 10CFR 20.
References:
Sections 4. i Rev.14 DSAR 3-10 1
MYAPC 3.1.2 n==aisr=Han of Structures, Systems, and Components The plant structures and proomms systems are deselRed hei to their funcilon and tw degree
- of integrity required to protect to public from uncontroNed reisenes of radioacthe byproducts.
Structural design crNorte indude too classes of buildings:
- 1. Claesl Structures Class I structures were designed in accordance with the " Building Code Requirements fbr Reinforood Concrets," ACI-31863, including increases anowed for stresses produced by earthqueim loads in mmbinadan wNh other appropriate loads. Where steel is utEzed, it is designed in accordance with AISC "54=riernean for the Design, Fabncation and Erecten of Structural Steel for Buildings." A Class I structure wR maintain its integrity during the trypve "M earthquake where the combination of the normal operating loads and the seismic stresses do not esoned 90%
of the yield strength. These strudures are primarily constructed with massive reinforced concrete and the M =Ma loads are not a mejor factor in the design. 1 Class I structures and equipment are designed to remain functional during an operational benis earthquake (ground acceleration 0.05g) and maintain fuel pool integrity during the more severe design basis of the hypotho6 cal eenhquake (ground acceleration 0.10g). In addition, some of the Class I structures are designed so that damage will not result from tornado winds ce missiles. The structures, systems and components which have been designed to Class i seismic requirements are listed below: P.AR (only to the extent it may asoct the Fuel Building or other SFPI SSC's) ] Containment Bundng (reenforced concrete substructure and superstructure) ] Fuel Building (reinforced concrete structure and steel superstructure) ] Fuel assemb8es Rev.17 j i DSAR 3-11
MYAPC Fuel Pool Liner ] Valve FP-21 ] Blind flange on Containment side of Fuel Transfer Tube ] FuelTransferTube ] Spent fuel storage racks ] Fuel buiding - yard crane steel support structure, portion within spent fuel pool building only Fuel building-yard crane (CR-3) ] Fuel handling plationn and hoist (CR-g) ] Flow limiters on the Ener leakage detection system
]
Spent Fuel Pool Cooing Loop Suction Piping (from the pool wall up to and including the ] siphon protection) (
]
- 2. Class 11 Structures Class 11 structures are usus5y designed to the requirements of the Unifomi Building Code. Class li structures were not designed for OBE and DSE. They were designed for dead load plus live load plus wind in accordance with AISC "Specillestion for the Design, Fabrication and Erection of Structural Steel for Buildings" and " Building Code Requirements for Reinforced Concrete", ACl-316 63, Part IV-A, utlitzing working stress design methods.
Where Class I structures are connected directly to the Class 11 structures, the interaction between the structures is taken into account in the design of both in addition, shake space between all adjacent Class I structures is provided forin the design. This rattle space is a minimum of 3 inches which is conservative with respect to the actual seismic requirements to prevent impact between f buildings in the event of a disturbance. DiamanHament of Rai=M Daninned Strtewas. Systems and Cwi uensnts -
)
For SSCs formerly designed to Seismic Class I requirements but not credited for peih. Ag a Seismic Class I function in the defueled conditioit, the following criteria apply prior to pedesiis5 dismantlement operatens:
- a. Declassificaton of ceir@-w.te shall be performed in accordance with appropriate WG::-irs and design procedures and processes.
- b. When declassifying an SSC, a 10CFR 50.53 evaluati.m shall be performed if:
- 1) the safety cloh is described in the FSAR, or Rev.17 DSAR 3-12
MYAPC
- 2) it's failure in a seismic event could affect a Seismic Class I cuirifsseit desc2ed in the FSAR in such a menner as to cause an unanalyzed inodont or an accident with offsite doses avemar%g 230 mrom (whole body) or 260 mrom (organ dose).
- c. When declassifying an SSC, a 10CFR 50.54 evaluation shall be performed if the classification is doecribed in the OQAP.
- d. SSCs shall be designed to asismic Class I requirements if, during a seismic event, ils failure has the potential to drain the fuel pool water level lower than 10 foot above the acave fuel.
3.1.2.1 Structures, Systems and Components ;i Weit to The Defueled Condition (ITDC)
,' Ganaral On August 7,1997, Maine Yankee certified per 10 CFR 50.82 that the company had permanently ceased power operadon and that all irradiated fuel had been permanently removed from the reactor vessel (Reference 2). This is a permanent, norwovocable certification that changed Maine Yankee's licensing basis by no longer allowing fuel in the reactor vessel and no longer allowing power operation.
l The license basis for the malonty of Structures, Systems and Components (SSCs) associated with nuclear safety has been changed. Those SSCs which gok performed a reactor safety function (i.e., SSCs which do not support a spent fuel or radiation pihi safety function) need no longer be maintained under nuclear grade controls. t SSC classification involves a determination that an SSC is, or is not, safety-related'. SSCs l classified as safety-related are treated differently by regulation than other SSCs.* } For a plant undergoing decommissioning, the only SSCs which meet the definition of safety-related* are the fuel transfer tube, and spent fuel cooling loop suction piping (from the pool wall ] Rev.17 ]
- 1. sassey reisted sace == moes remed upon to remain funceanel during and famowing design beste everne e enews: al the inseyay or me reeamr cociant prenewe boundary; b) the apabmay to shut down me reactor and mainmin it m a esse shusemen conemon: and op me capsbany to pnwont or mingene the consequences or accidenne that could seeuk M ponendel cAete espoewee comparable to me spaldsense or 10 cPR 100.
- 2. 10 cPR SO Appendk s neees that"the parenant regubemenes or this appendk apply to ad acevises asemang the seiny-reissedfunemonsor ?ssco.
- 3. The Ibet too parts or the sessyelseed dednWon (reecer coolant pressure boundary, and capetely to achieve and mesmein esse shutdown) do not appsy to a decommmelaning piant, given the scense reestomane or 10 CPR 50.82. The third part of me seep doenman (accident consequences comparable to 10 CFR 100 guidelines) also does not apply. At Maine Yankee. the consequences weh me designeconee beeis evente e to decorrenteoloning me neerty swee onsors or magnitude lower then Part 100 guineanos mW lower then the EPA protoceve accon guide Smit.
DSAR N 3-13
MYAPC to and induding the siphon protection), blind flange on Containment side of fuel transfer tube, ] valve FP-21. This results in two areas ofinterest ]
- 1) Maine Yankee's " nuclear grade" processes are based largely upon quality assurance (10 CFR 50 Appendix B) requirements. Raciassfymg all SSCs as non-safety related could lead to the eEminetton of most rnanagement controis in situations where maintaining rigorous management contrais is intended.
- 2) Mene Yankee recognizes that certain functions remain 'irWt to safety in the defueled condition.
It is necessary to redessify SSCs in order to proceed with decommissioning. Strictly fogowing regulatory requirements in rodassdcation results in elimination of most of the cunent management controls, which is contrary to management's intent. Thus, in order to provide an enhanced engineenng controis above that mandated by regulatory requirements, an artificial classdcation system termed "liipun .t to the Defueled Condition" (ITDC) is introduced. The following concoms are addressed within this clasadcation:
. SSCs which support a fuel safety or radiation protection safety function, and Identification of enhanced management and engineering controls are maintained on SSCs classified as ITDC.
It is noted that SSCs which are not defined as within the ITDC classification, or otherwise designated as safety class, are euminated from the license basis. It is not the intention of implementing the ITDC daandcation to raciassify components previously defined as NNS as ITDC. AdditionaEy, there may be other SSCs to which a level of enhanced quality or eiigir.n..irig ovensight has been applied, but do not meet the intent of the ITDC classification. The following enteria are used to detemne which SSCs are designated as ITDC: Criterion 1. The SSC is essential to the normal operation of the storage, control, or maintenance of the spent nuclear fuel or the monitoring of radioactive effluent. ] Criterion 2. The SSC is essential in preventing postulated accidents or incidents involving the storage, control, or mantanance of the spent nuclear fuel or the monitoring ] of radioactive effluent. ] Rev.17 DSAR 3-14
MYAPC Critorion 3. The SSC hislodcol closedcation is the direct result of an outstanding commitment to the USNRC which remains essential to storage, control, or ] maintenance of the spent nudeer fuel; or the monitodng of rMiame+ive emuent. ] Critorion 4. The SSC sationes a requirement bened in regulations winch remain essential ] to storage, control, or maintenance of the spent nucieer fuel; or the monitoring ] of rednecdive affluent as denned in the Maine Yanlose licensing basis. This ] includes any SSC which is independoney required by the umidng Conditions of Operallon (Secdon 3) of the Technical Speci6 cations.' A positive response to any criterion indicates that an SSC is ITDC. ALA..:. 2-e P% and L:.. " " -e on n== of the SSC M===? *-7 e T t The SSC reclass Heation criteria wlE be used as a basis to change various Maine Yankee processes, provided that the change involves an SSC that is non-(TDC and, provided that plant procedures contain an acceptable method for approving the change. The fouowing idnds of
, " software" changes ===aci=ed with non-ITDC SSCs are allowed:
.j
. SSC classi6 cations e drawings, i . calculations, i
. procedures e iss.r.iwicc.; 4 items and corrective actions
]
extemal industry operating expenonce reports
= comrmtments open work orders (in process at the time the decision was made to decommission the l plant) the application of 10 CFR 50 Appendix B criteria provided it does not represent a reduction in commitment.
Use of these critada does not authores:
- a. Activities cie.a g new hazards or initiators not already recognized as part of the current license basis (e.g., decontaminston or dece in ; icning of major components defined in 10 CFR 50.82)
- b. The physical removal / disassembly of existing SSCs, or the installation of new SSCs.
However, it may provide the basis for initiating a hardware change. Rev.17 A The roc metussen ausses tot he appuupdate aguhamy eenge momenem to used kr eNeceng to menOs. DSAR - 3.15
MYAPC
- c. Changes to Technical $+T-:+% requirements applicable to the current mode of operation.
- d. Changes to reguindons, license conditions, rules, and permits untH such time that relief is granted from the regulating authority. However, it may provide the basis for requesting relief from the reguistions, license condmons, rules, and permits.
- e. Changes to commitments. Applicadon of the commitment change process is required to change commitments.
- f. Changes to the OQAP. However, it may provide the basis forinitiedng a change to the OQAP.
- g. Changes to the OOCM. However, it may provide the basis for initiating a change to the ODCM.
L h. Changes to the Emergency Plan. However, it may provide the basis for initisung a change to the Emwgency Plan.
- 1. Changes to the Security Plan. However, it may provide the basis for initiating a change to the Security Plan.
J. Changes to the Fire Protection Plan. However,it may provide the basis for initiating a change to the Fse Protection Plan.
- k. Changes to the Radiation Protection Prognun. However, it may provide the basis fbr initiating a change to the Radiagon Protection Program. i Boundaries and laterfaces for ITDC SSCs SSCs identified as ITDC that require "availab5ty", must meet the following criterion:
"A system, subsystem, train, currw-i :, or device is "available" or will have "availabdity" when it is capable of perfomung its specified function (s)."
implicit in this definibon is the assumption that the necessary attendant instrumentation, controls, power sources or equipment, or other auxiliary equipment that are required to support the available SSC, are capable of performing their support function, as necessary. Engineered Raouirements for ITDC SSCs A higher level of quality is maintained for ITDC components to assure that the capability exists to reliably meet the performance expectations and requirements. The controlled list of ITDC components is govemed by plant procedures. ITDC components are not safety-felsted ceirii.,er-in and are not required to satisfy 10CFR50 Appendix B requirements. Although not required by regulation, the following criteria is developed and applied, as appicpr'-te, to ITDC ! SSCs to assure continued reliability: I Rev.14 DSAR 3-16 ;
~ MYAPc
- a. Design Control Measures will be invoked to assure e regulatory requirements, license basis, and design basis informanon is emscay transisted into = pace =* m, drumnes, procedures and instrucdons. These measures shaN include provisions to assure that appropriate quaky standards are specmed and included h design demuments and that deviations som such standards are controsed. Design changes, haluahe said danges we be aw to engineered design control measures commensurnes wah the imponence of the ssc.
- b. Procurement Document control Measures we tw hvoked to assure met appsable reguistory requirements, design basis, and other requirements which are necessary to assure adequate quality are suitably included or reestenced in the documents for procurement of motortei, equipment, services.
- c. Instructions, Procedures, and Dramngs ActMties affecting sses we be prescrited by documented instructions, procedures, or drumnos, of a type approprisen to the circumstances and we be accompushed in accorden=
wah these inseuenons, procedures, and drawings. Inseuenons procedures, and drawings we include appropriate quantitative or qualitanve acceptance criesna for determining that important actMues have been sa"+:x-w accompeshed.
- d. Control of Purchased Material, Equipment, and Services Measures we be invoind to assure that notorial, equipment, and services conform to the procurement documents. These measures shalinclude provisions, as appropriate, for source evaluation and mal =*ws, objective evidence of quauty fumished, inspection at the source, and examination upon delivery.
- e. ir e * -i inspection of actMbes affectog quality wiH be invoked and executed to verify conformance with the documented instructions, procedures, and drawings for accomplishing the activity.
- f. Handling, Storage and Shipping Measures will be invoked to control the h r,C" g, storage, shipping, cleaning and preservation of material and equipment in accordance with work and inspection instructions to prevent damage or deterioration.
Rev.14 DSAR 3-17
O g. Test Control Survedlance testing wNI be established for SSCs to ensure that me SSCs p i . . satisfactorty commensurate with the importance of its intended safety funcilon.
- h. Measuring and Test Equipment Appropriate controis wE be invoked to assure that measuring and test devices used on SSCs are property contrated, celbrated and adjusted at spedAnd periods to maintain' accuracy within necessary limits.
I
- i. Conecdve Acnon Measures wlN be invoked to assure that conditions adverse to qumEty are prompey identified and corrected. In the case of significent conditions adverse a quality, the measures will assure that the cause of the condition is determined and consedve action is taken to preclude repetition.'
3.1.2.2 Wind, Missile, and Tomado Loadings O De Maine Yankee faclety is capable of withstanding the e5ects of severe winds or tomadoes without loss of capabulty of the safety systems to perform their safety functions. Section 2.2.6 discusses wind and tomado data for this region. The design tomado has a rotational velocity of 300 mph, a velocity of advance of 60 mph, and an extemel vacuum of 3 psig developed in 5 seconds. Thus the total effective velocity is 360 mph. Missiles may travel with the tomado equivalent to (1) a utuity pole 35 ft long,14 inches in diameter, weighing 50 lb/cu. ft. (1850 lbs totel), and traveling 150 mph, or, (2) a 1-ton automob8e traveling at 150 mph. The structures housing spent fuel and adjacent structures are designed to resist the combined e5ects of tomado wind load, pressure drop and missile loads to produce the most critical loading condition. The original design code fbr the spent fuel pool and other Class I structures was ACI 31843. The atowable stresses for shear and flexure are defined by the criteria in secdons 1600 l and 1700 of ACI 318-63. This specifically included the appropriate Wy reduction factor and allowable stresses. De Fuel Building is designed for protection against wind and tomado as follows: Reinforced concrete structure Steel superstructure Fuel building - yard crane steel support structure, portion within spent fuel pool building only Rev.16 DSAR 3-18 i
i I uwc The 6 foot thick reinforced concrete walls, which extend from 12' 6* below ground grade to 26' above ground grade, are designed to withstand the eSects of tomado and missiles. A very substantial degree of added tomado and missile protection is aSorded Ly he below grade construction of the fuel pool. The nominal grade level (20' elev) is 5-;-ui., * 'i at the same elevation as the top of the aceve fuel. The steel frammg above me pool is designed for tomado iondings such that it we not fell into the pool and damage fuel assembnes. The loss of water from the storage pool under the eSect of a tomado would result in a maximum ceiculated loss of Ave feet of water, primer #y due to vortexing. Given that the normal level of the spent fuel pool is at the 44 ft. elevanon and approximately 23 ft. above the aceve fuel, this water loss is not signiscent as it is bounded by the siphoning incident. The fire pump structure is also designed to withstand the effects of wind and tomado. The fire pumps, located at ground grade near the fire pivi.cr.wr reservoir are tomado protected and ] screened from the full eSects of the tomado by the dike. This is a very reliable makeup source of water to the pool. 3.1.2.3 Water Level (Flood) Design 3.1.2.3.1 Hurricane An investigation was made to predict the probable maximum flood level which could occur at the site of the Mene Yankee Atomic Power Station on the Shaarww River estuary when the probable maximum hunicane is taken as the design basis ict:+ '-:-fM event. The irr;::";+"-:s is based I upon the parameters of the probable maximum hurricane as deAned by U.S. Weather Bureau Report HUR 7-97, Interim Report - Meteorological Charactenstics of Probable Maamum Hunicane, t Atlantic and Gulf Coasts of the United Statas and discussed in section 2.2.6. This irr;::";+"-:s shows that the maximum water levels at the Maine Yankee Power Station due to the probable maximum hunicane are predicted to be at Elevation 19.9 feet and Sevation 21.4 feet on the plant site and screen weil structure, respecuvely. These levels are based upon the simultaneous occunence of the maximum storm surge, maximum predicted astronomical tide, an inittel rise in mean see level, estuarine ampilfication, the probable maximum flood in the Sheepscot watershed, maximum waves in Montsweeg Ray and existance of a channel restriction at the former Cowseagan Nanows Causeway. Rev.16 DSAR 3-19 {
MYAPC 3.1.2.3.2 Snow Section 2.2.5.1. Table 2.2.5 shows average snowfeN statistics for Portiend which are considered to be representmove of the she. Structural loading and capacity reduadon factors for buk5ngs required in the defusied candluon provide ample margin to accommodate snow loading. 3.1.2.3.3 ice, Glass, and Temperature Extremos ice loading is discussed in Sections 2.2.5 and 2.3.4 of this report. Gleme and ice storms usuefy occur in the months October through Apri with an average frequency of 1 to 3 storms per year. Ice thidmesses of .25 inches or .5 inches wlR likely occur every year whereas .75 inches is Braly to occur at least every three years. Structural loading and Mty reduction factors for tuddings required in the defueled condition provide ample margin to accommodate ice loading. 1 1 1 1 As h M in sedian 2.2.4, the average January temperature is about 227 wNh between 10 and 20 days of sub more temperatures occuning yearly. Temperature data representathe for the she is provided in Table 2.2.2. During exesnded periods of freezing temperatures, it is possible that freeze damage could occur to water filled piping in buildings no longer maintained due to the defueled condition. The primary concem regarding freezing is the affect on the integrity of the pool. Pool water freezing is not possible provided that the aggregate decay heat load of the stored assemblies is cr-W high. Assurance is provided through mudne operator rounds which monitor and log the temperature of the pool water. Adjacent buildings and rooms conleining siendicent water sources which could potentiasy affect the safe storage of fuel are either temperature controbed to preclude freezing, or the water source is appropnately heat traced or drained, as noomssary. This incVas adequale administrative or design contrais for protedhg the integrity of the fusi transfer tube. Rev.17 DSAR 3-20
i -. 1 MYAPC 3.1.2.4 Seismic Design 3.1.2.4.1 Design Basis AE strudures and elements of the plant are designed in accordance with sound ven lE::iq practice and are considered capable of withstanding seismic foross corresponding to a ground accoloradon of at least 0.03g, in adelon to normal loads, without damage or loss of funcuan. - in addison, as structures and components of the plant which are important from the standpoint of nucieer safety and demogs which could aSoct the health and safety of the pubNc, i.e., " Class l" portion, are designed e meet the vosowing crnerta: .
- 1. The design earthqueim is based on a ground accelera6an of 0.05g, and this portion of the plant shall be capable of operating through such earthquake.
- 2. The hypothetical earthqueim is based on a ground acceleration of 0.1g, and this portion of the plant shaR be capable of performing its intended safety funcuan under an earthquake.
- 3. A spectrum analysis is used, with appropriate conservative damping factors.
3.1.2.4.2 Design Data 2 While 0.04g was found to be the maximum probable ground acceleration at the base of the structures on the site, it was dedded to round this upward to 0.05g horizontal for design purposes. The values for seismic design of Class I structures and ceirpenes;. am as stated under " Design Bases." in class I strumures and components, stresses due to normal loads plus the design earthquake do not exceed those design values permitted in the applicable codes, while stresses due to normal iceds plus the hypotheecal earthquake do not exceed the yield stress of the a5seted materiais. Earthquake stresses are based on a herh%hi ground acceleration and a vertical ground acceleration of two-thirds of tfm horizontal, with the two acting simultaneously. Design for Class I structures and components used the " response spectrum" approach in the analysis of the dynamic loads imparted by earthquakes. The seismic design is based on the acceleration response spectrum curves shown in Fgure 3.1-1 for the design earthquake and Figure 3.1-2 for the !ry#:":e earthquake. The curves are derwed from the "Housner Spectrum" normalized to 0.05g for the design earthquake and 0.10g for the !ryM:":e earthquake. The design response spectra (Fgures 3.1-1 and 3.1-2) are a spec %.a;va of the level of seismic design accoloration, or dispiscoment, as a function of natural penod of vibration and damping level. Rev.14 DSAR 3-21
MYAPC i The response spectrum anat / sis is applied to al category I structures and wiW,.id. and groups thereof whose responses may be interdependent, considering their natural period and using appropriate damping factors as listed on Table 3.1.1. I Class I structures and components are designed in the foHowing general manner:
- 1. An analysis is made to determine the natural periods of vibration of the structure using equhelent lump mass systems or distributed mens systems as is considered appropriots. In these analyses, periods and modo shapes are determined fbr each lumped tross mode.1hese data then define participation factors for each structure. Where structures are supported on their own Ibundations, foundation d'aplacements are considered in C ^ iisaq natural periods and participation factors. it should be noted, however, that Class I structures at this site are founded on grande gness. A ;..py, foundation yielding will be very small and may in many cases be neglected without introducing significant error.
- 2. The earthquake design acceleration value for the specdic natural period of the structure or sTwwa being considered is detomaned from Figure 3.1-1 using ri-c.pi'^ _ damping factors. The hortaontal component of the ground acceleradon is taken directly and the vertical wTgwa ls taken as two-thirds of the honzontal value. These components are considered as acting simultaneously.
- 3. For certain structures, and especiaNy for vibratory systems of a highly svA nature, such as a piping system, use of the maximum response value (peak of the curve) correspondng to the appropriate damping factor may be elected in peinya q the stness erdie:. of the system.
- 4. A tabulation of typical d-Tip; s factors which are used Ibr various vibratory systems important to nuclear safety is presented in Table 3.1.1. Conservative values are shown for various materials, methods of construction, and location with respect to the ground.
- 5. The design is then checked to verify that stresses are within acceptable limits for the hypothetical earthquake using Fgure 3.1-2.
Refer to section 3.1.2 for a hsting of SSCs required to be designed to Seismic Class I requrements in the defusied condition. SSCs wtuch were previously designed to Class I cnteria but are not Isted in section 3.1.2, are not credited for performing a safety function (Class I ) in the permanendy defueled condition, and therefore, are no longer required to be designed to Class I requirements. Rev.14 DSAR 3-22
MYAPC 3.1.2.4.3 Seismic Design and Qualification
]
As of March 24,1986, new Class I systems, stuctures, and componards wit be designed and queNAnd to the seismic demands as deAnod by a 0.18g NUREG/CR-0098 50th percent 5e Ground Response Spectra (GRS). Analytical quellAcaNon wRl be to the SEP alioamble stress levels and demping values noted in Reference 4 (i.e., for piping, damping = 3% or PVRC, abowable strees = 2.48h, no OBE) up to any interisco wNh exisung stuctures, systems, and mmponents. Seismic adequacy may aino be demonsrated through similarity by comparison to the documented performance of equipment in natural earthquales (Reference 5), or simuisted earthquakes on testing machines. i
~
Sadion 3.1 Rafarances:
- 1. Memorandum; SECY-92-223-Resolution of Deviations identified During The Systematic Evaluation Program; S J. Chilk, Secretary, USNRC to J.M. Taylor, Sept.18,1992.
- 2. MY Letter to the NRC MN-9749 "CarilEcstion of Cassation of Power Operata and Permanent Removal of Fuel fiom the Reactor, dated August 7,1997.
O 3. Letter: " Seismic Design Margins Program," P.M. Sears, USNRC to J.B. Randazza, Maine Yankee Atomic Power Company, March 26,1987; NMY 87-029
- 4. USNRC Letter to MYAPCO, dated Mardi 26,1987.
- 5. USNRC Letter to MYAPCO, dated February 19,1987, (Generic Latter 87-02).
O Rev.17 r DSAR 3-23
MYAPC TABLE 3.1.1 EARTHQUAKE DAMPING FACTORS Percent of Critical Damping
, DESIGN HYPOTNEmCAL EARTHQUAKE EARTHQUAKE Reedor Containment 2.0 5?
Reinforced concrete structure, other than 2.0 5.0 containment, founded on soll orrock l Reinforced concrets supporting structure, not 2.0
' 5.0 founded on soil orrock Steel-framed structures, including supt ,g structures and foundations Bolted or Riveted 3.0 5.0 Weided 1.0 2.0 Reactor vessel, Intsmais and control rod drives , Weided Assemblies 1.0 1.0 Bolted Assemblies 3.0 3.0 Mechanical equipment, including pumps, fans and 2.0 2.0 similaritems
-{ Piping Sys' ems 1.0 2.0 1 1 Rev.14 DSAR 3-24
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MYAPC 3.2 Structures 3.2.1 Fuel Building 3.2.1.1 General 1; 4 The principal function of the fuel bu5 ding is to provide a location for Sie safe storage of new and spent fuel assemtdes. The building houses a new4uel unioedng asas, a nou4 bel storage room, a spent fuel pool and the necessary cranes required for the handing of the Ibel assemblies. The spent fuel pool cooling system heet emmenger, the fuel pool cosihg pumps and the fuel pool
, purflication pump are located on Elevation 21'0". The spent fuel pool support systems are also ]
l located in the fuel buBding. The fuel building arrangement is shown on Figures 3.2-1 and 3.2-2. 3.2.1.2 Fus! Unloading Area New fuel was shipped to the site in two element shipping casks. A Ave 4an overhead crane in the fuel building was used to unioed the shipping cooks. The spent fbal pool purification system fBlers are located in shielded cubicles below the fuel unloading area. Shield simbs are removed from the 4 fuel unloading floor to replace expended filter cartridge elements. 3.2.1.3 New Fuel Storage Area
, The new4uel storage room is designed for storage of 160 fuel assemblies. The fuel room is
, located over the spent fuel pool cooling pumps and heet exchanger. The fuel rack consists of guide sleeves symmetrically located on the floor at Elevation 31 ft.1-1/2 in. and through the coling of the new4uel room at Elevation 44 ft. 6 in. The fuel room floor has a drain opening located over the spent fuel pool cooling equipment cubicle. The floor opening prevents Twths of the new4uel storage area. The spent fuel pool new4uel elevator winch is siso located in the new-fuel storage room. 3.2.1.4 Spent Fuel Pool Cooling of the spent fuel assemblies during the radioactive decay period is a,wdy;ished in a stainless stool lined reinforced concrete pool filled with borated water. Space is provided in the pool to pleos the spent fbol shipping cask. The pool is serviced by means of the yard crane, as wed as a moveable p; iven with hoist. A new-fuel area adjoins the spent fbel pool. The poolis Rev.17 DSAR 3-27
MYAPC designed to safely reset the hypothetical earthquake or tomado, as wed as the applied loads of the water and fuel. The pool has a reinforced concrete floor fiounded on rock and sidewaNs 6 feet thick whid1 adond F from 12 foot 6 inches below ground grade to 26 feet above ground grade. The concrete is reinforced with #11 bars at 12 inch conter to center spaang with a yield strength of 40,000 pai. The concrets has a 28 day minimum compresolve strength of 3,000 pai. The reinforced spent fuel pool was origineNy designed in accordance with ACI-31843 to resist the appropriots dead, Eve, hydrostatic and maamum hypothetical seismic loadings. The structure was reenalyzed, in support of EDCR g2-111, to demonstrate the acceptabinty of installing the new high densdy spent fusi storage racks. As part of the preliminary decommissioning actMties, the structural evaluations have been performed which demonstrate the adequacy of the SFP concrete and liner to withstand the effects of dead, live and hydrostatic forces in conjunction with an elevated pool water temperature of j 212*F. Completa details of this evaluation are contained in References 3.2-1 and 3.2-2. The pool is cuii-;iM lined i with plates of stainless steel which have test channels behind each weld. The test channels are piped to the spent resin pit sump through four-1 inch tog tale pipes, each with a flow limiter at the end of the pipe. In the event of a malfunction of a Hner weld, the leakage through each telltale is limited to less than 2.5 gpm. The liner is designed as a ASME Section lil, Division 2, Paragraph CC-3720, Liner, Table CC-3720-1, Service Category, Membrane. The piste material is ASTM A240, Type 304 stainions semel Liner Anchors are designed to ASME Section lit, Division 2, Paragraph CC-3730 and are constructed of ASTM A-36 steel. The weld rods used to weld the vertical stiffener flanges to the liner wall liner were ASTM E30g (carbon to stainless steel) with a minimum tensile strength of 81,000 psi. The fuel transfer tube was onginally designed as safety class 2; however, since the containment integrity design basis is not applicable in the defueled condition, it has been raciassified as safety , class 3. It consists of a 36-inch 00,3/8 inch thick, ASTM A312 TP304, stainless steel pipe ! instaNed inside a 40-inch OD stainless steel sleeve as shown in detail on Figure 3.2-13. The inner pipe acts as the transfer tube and connects the containment refueling canal with the spent fuel pool and is welded to the fuel pool stainless steel liner. The outer pipe is fitted with bellows expansion Jomts, backed up by a packed slip joint to con.pensate for any differential movement. Structural steel supports a superstructure of protected metal siding which encloses the pool. The steel frammg above the pool is designed for earthquake and tomado to prevent it from falling into O the pool and damaging fuel assemblies. The masonry well at the south end of the fuel building is not designed for certain wind or earthquake loadings, and, therefore, an evaluation of the DSAR 3-28 Rev.15
MYAPC consequences of a wel collapse was performed. The analysis demonstrated adequate spent fuel pool cooling capabRity and structural rack integrity. 3.2.1.5 Fuel Storage Rads The new and spent fuel pool structures including fuel rocks are designed to withstand the anticipated earthquake loedngs as Class I structures in accordance with the guidance of Regulatory Guide 1.29. Analyses show that the rocks wlE perform theirintended function under both asismic and load drop loedngs in accxirdance with Reguistory Guide 1.124 and NUREG4800. 1 The design ensures that during the event, rack-to rack and red-t>was interadion is +;-;5ri considered. Structural material used in the rack design is ASME Section ll, SA 240, Type 304 seminions stool. The design considered themel loads induced by an operudng temperature of 154 F. Subsequeney an evaluation was performed which documented the acceptability of the racks at a temperature of 212*F. The ANSYS version 4.4A program was used for au computer aided mechanical analysis. The design considered impact loads from a 2500 lb. (submerged weight) fuel element & M from ] 18 inches above a module, a fuel element hangup during removal, and the load induced if an assembly hit the top of a rack while moving at the maximum horizontal velocity of the crane. ]
- W W reenalysis considered impact loads from a 2000 lb. (submerged weight) fuel element ]
dropped from 22.5 incties above a rnodule. Sutu;&.J y and a ennlahin geometry are maintained ] and damage to the stored fuel is minimized. The racks consist ofindvidual storage cous joined into a rack module. The rads are a single tier, reenneer array of free standng modules, not anchored to the pool wans, noor or adjoining racks. Each rack module is pnwided with adjustable support feet. Each fuel rack is a folded metal plate assembly of 14 gage metal, approximately 180 inches high,117 inches wide and 128 inches deep. The folded metal piele assembly is welded to a baseplate, which is supported by adjustable supported feet. Region I contains 5 racks, spaced on a minimum of 10.5 inch centers. Region 11 contains 21 racks spaced on a minimum of 9 inch centers. Spent fuel storage rJas may be moved only in accordance with written procedures which ensures that no rack modules are moved over fuel assemblies. Rev.17 DSAR 3-29
MYAPC 3.2.2 Storage Buildings 3.2.2.1 Undergound RCA Storage Bunker The undaryound RCA storage bunker, also referred to as the high red bunker, is located wNhin
#m protected area about 120 feet northwest of containment. The bunker is a reinforced concrets structure and is partisNy buried below yard grade. The bunker is 27.5 feet by 16 foot and approulmately 12 imot high. The top of the bunhar is about 5 feet higher then the sunnundng yard grade. The bunker is dMded into Ave intemel compartments each separated by 18 ind thick concrete walls. The exterior bunker walls vary from 12 to 18 inches in thickness. The roof of the bunker consists of six removable 18 inch thick concrete roof plugs.
A floor drain system directs any Equids enmar*=d in the bunker to a sump. Liquids camachad in the sump may be pumped to the spent resin pit sump in the nearby RCA storage building through an ] underground pipe or temporary hose. The spent resin pit sump discharges to the Liquid Waste ] Holdup Tank (TK-109) where any liquids would be collected for processing. ] I The bunker provides temporary storage for radioactive wastes before they are moved to the low ] level weste storage building or shipped for processing and diapnaal The outside yard craneis ] avalleblo to move weste containers in and out of the bunker. Weste may be stored in the bunker ] to aNow for some decay before being placed in the low level waste storage building or untR ] anengements are made for permanent off-site disposal. For purposes of decommissioning the high rad bunker has been abandoned. However, the ] C+x-i.,, ' :'Mg Operations Contractor may use the facinty for short term waste storage. Its ]
<:: M/-x will be retained until the faciRty is no longer in use. ] l 3.2.2.2 Radiation Controlled Area (RCA) Storage Building The RCA storage building, adjacent to the fuel building houses the Duratek waste guco ;rs skid, )
a decontamination ares and a waste solidification area. The waste solidification area is no longer ] operable. The Duratok skid has been modified to support processing of wasta liquid from ] decommissoning activities. Tank 109 (TK-109) has been converted from resin hold up to liquid ] waste hold up. TK-109 is used to feed the Duratek processing system. The decontamination area ] provides an area to support fuel pool cask decontamination as well as general equipment ] decontamination. The area is tied into the remaining plant ventilation system and is kept under a ] alight negative pressure. ] Rev.17 DSAR 3-30 l
MYAPC 3.2.2.3 Low SpeaSc Activity SSA) Storage Buking This building houses the L3A compeenr and sened as the siorage building fbr LSA contamers to ] isoep them too of damage or deterksellon. It extends tom the south well of the RCA buBding to the containment, but is not strucasusy attached to either. The buliding is designed norwmeiner asisty. The LSA compactor compresses LSA meterial it is vented into the RCA fBlered vendedon system. The LSA sump pump dischegos to the spent resin pit sump in the RCA Buildhg. 3.2.2.4 warehouse This buBding is located outside the RCA. The building is used to store replacement components and eqispment. 3.2.2.5 Low LevelWaste Skunge Building The low level weste storage building is located on the plant site, outside of the protected area. The building is 154 feet by 64 feet by . . Mr ~/ 25 feet high tom ground to roof Hne with floor elevation at 26 feet. The outer wets of the building consist of steel siding with one foot thick concrete shinid weis inside he sidng 2 a height of 16 feet. A trudt boy is included to atow ecomes , ihr shipping weste containers. The handing provides for interim on mes storage of low level weste and fbr storage of contaminated equipment. 3.2.3 Service Building The Service Building has been abandoned in its entirety by Maine Yankee and no longer serves ] any functioni' ,We to the safe semage of fuel. ] 3.2.3.1 Control Room Area ] The main control room, located in fue Gate House, contains the ce.Lvi. and instrumentation ] necessary to monitor and control various areas and equipment required for safe storage of spent ] fuel The main control room is designed to be avaisbie at an times. The decayed source term ] in the defueled condition, in conjunction with the location and design of the control area, provides ] 1 sufficient protection to ensure that control room personnel will not be subjected to doses which ] would exceed 10CFR20 limits. Equipment in this area has been designed to minimize the ] ! possibNity of a condition which could imod to possible inocessabilty or evacuation. In the event that ] this ares becomes inaccessible, the cont.uls for spent fuel pool cooling and makeup, water ] treatment and weste WW-'. are hxated at control stations remote from the main control room. ] Rev.1's DSAR 3-31
3.2.4 Turbine Building The turbine building housed the secondary plant components and systems. Though t
]
buMng is not a Class I seismic structure, it is designed forwind forces which are greate seismic forces as determined from a combined solemic anedyas of the service bund room and the turtaine hen. The turbine hmE wlE remain intact during the DBE. The Turbine had is considered partesy abandoned wlBi certain restricsons. This ] stru suppost on the north and east exterior weis for SFPI control cabling. Demolition activities are
]
atowed whis ensuring interimons witt aveRoble SSCs (tant afflos hetway structure, admin
]
buildmg structure, fire suppression systems, and designated orange marked equipment o
]
north and east interior and exteriorwells) are not interfered with.
]
3.2.5 Primary AuxiRary Building (PAB) All exterior and interior concrete walls and siebs are designed to safely resist the hypoth earthquake. 1ho PAB has been partinEy abandoned wEh certain restrictions. This assessment of the PA
]
alow for modifice6ans and limited demoution activities to this structure as long as 1) containm
]
closure can be established to most the requirements of the DSAR, 2) the integdty and seism
]
stabitty of the Spent Fuel Pool and support equipment is not challenged and 3) af effluents are vi a monitored and =; -;ssd reisese path. ]
]
3.2.6 Service Waterintake Structure The Service Water intake Structure has been abandoned and serves no function] relating safe storage of spent fuel.
]
3.2.7 Fire Pump Building ] The fire pump house is shown on Figure 3.2-12. It is located near the water storage pond an houses the equipment servicing the Equid portion of the fire protection system. The building two fire pumps, a pressure mantonance pump and a hydn>pneummec tank. A diesel engine driv one pump while the other is motor-driven. The building houses the diesel fuel tank, the battedes and control board required for the desel operation. Rev.17 DSAR 3-32
MYAPC 3.2.8 Masonry Walls Masonry wees in eW vicinity at seisty veseend equipment have been evaluated e detem*w wheiher they we withstand as pa=h N design loads (seismic, tomado, hunicane, ansched equipment, em.). Table 3.2.1 uses the wens evetusend and notes those which are AAy quenSed and those whim ate soeumed e fan, but such secure we not result in um damage e equipment. auseen ao-11 categmy osanimons: category 3: wass whose couapes we not asset sorsty-reisted equipment. This Table has been revised b address only those wens that are required in the pwmeneney ] denseind condman. Table 3.2.1 ] nOOu narnmenCs wAu. eausrunwr BUE.DeIG WAU. 10 LOCATION DRAWe00 LOCATION PROTECTED CATEGORY Fuel FB 441 Spent Fuel 11850fA 12A South Weg Spent Fuel 3 Pool Spent Fuel PudRcedon Elev 44'-6" Bukeng and Cooing Retum Unos and spent fuel store 0* rocks. Rev.17 DSAR 3-33
MYAPC Secdon 12 - Rafarancma:
- 1. YNSD Latter to R. Fraser from D.L Magnarelli/W.E. Henries, " Review of AES Analysis of SFP for Elevated Temperatures," dated Nov. 25,1997.
- 2. YNSD Ndanart No. MYC-2001, " Analysis of SFP Structure for Elevated Pool Water
. Temperature " dated November 25,1997.
1 1 O 4 Rev.14 DSAR 3-34
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- FIRE PUMP HOUSE Maine Yankee.-
MYAPC 3.3 Exatama 4 3.3.1 Fuel Storage 3.3.1.1 Fuel Storage Design Basis Danign critarim - (
)
The fuel storage design critoria are as follows:
- 1. Crtlicotty in new and spent fuel storage is prevented by physical design instures or processes.
A geometrically safe conAguration is emphasized over procedural contrais. 2.' Appropriate shielding is provided to most the requirements of 10CFR 20.
- 3. The fuel buBding is continuously monitored by ares spec 5c detectors. Auede and visual alarms are activated at the detectorlocations and in the Control Room for radiation levels in exoses o predetermined ilmits.
- 4. Spent fuel storage systems are:
- a. designed to prevent or mnigste accidents which could lead to the release of signincent amounts of radioecevity affaceng the putdie heenh and sassey,
- b. designed, fabricated and erected to wNhstand, the addWonal forces that might be imposed by natural phenomena.
The spent fuel pool cooling, malm-up, and puis ;;or, system design criteria is as follows:
- 1. Prevent damage to spent fuel which could cause radianceve release to plant operating areas er the putWie environs.
- 2. Maintain adequate level of pool water to provide shielcAng for radiation pivi dk, .
- 3. Maintain clarity and purity of the borated water in the spent fuel pool.
- 4. The spent fuel pool is equipped with a high and low Equid level alarm, and a high temperature alarm which will indicate loss of continuity in decay heat removal.
Damion assis l The fuel storage and handling design basis is as follows- i
- a. Store 1754 complete spent fusi assemblies. In addition, other inadiated components such as ] l CEAs are stored in the fuel assemblies. The total maximum decay heat lood to be stored in the l
poolis s5.96 x 10' BTU /hr (BTP AS8 9-2 as of 10/19g7). This assures that the calcuisted time- ] to boil and boil-off rates are arlary
- to provide ample time for operators to take remedial schons in the event of fluiirs of forced cooling incident. '
Rev.17 ' DSAR 3-39
MYAPC
- b. The design basis for preventing wiMy is that, considering the possible vanadons, there is 95% probabdity at a 95% confidence level that the effective multiplication factor (K.) of the fuel assembly array win be less than 0.95 for flooded conditions without credit for barated water.
- c. A minimum boron concentration of 670 ppm was assumed in the analysis for to tw=N ]
mispiaced fuel asserr.'Wy. A higher administradve Hmit is required by technical specincadons. The minimum concentration was determined to be required to maintain a K,less than .95 for the most resc8ve am:ident condulon. This concentradon is required whenever fuel assembEss are stored in the pool and a spent fuel pool assembly piacoment vertAcation has not been performed since the last movement of fuel assemblies in the apent fuel pool. The analysis considered various rpierw=#Em conAgurations including a fuel assembly leying hortaontapy on , the top of the spent fuel racks, a fuel assembly pieced adjacent to or on the outside comer of the fuel racks and a fuel assembly placed in the wrong rock.
- d. Maintain fuel cladding integrity in the event forced cooling is lost and cooling occurs by SFP boiling (212*F) at the water surface. The cooling water make-up rate aw= de zw_2
{ losses. Makeup suppses are aveambiein the event of an adonded ions of offsete powerincident.
- e. The new and spent fuel pool structures, inckxNng fuel racks, are designed to wNhstand #m ar*ir-e=d earthquake loadings as Class I struenses. The spent fuel pool is Ened wth stainless steel to ensure against loss of water.
- f. The spent fuel rr::ks are designed to Seismic Class I requirements, capable of sustaining a
, temperature of 212*F, and able to withstand the dropping of a 2500 lb. (submerged weight) ]
assembly from a height of 18' above the top of the racks without incumng damage which could result in w Mi . Af subsequent evaluation has also shown that the racks are able to withstand ] the &ww;rs of a 2000 lb. (submerged weight) fuel assembly from a height of 22.5 inches above ] the top of the racks without incumng damage whid1 muld result in u Mif or spent fuel cooling ] concoms. ]
- g. The spent fuel pool concrete structure is a Seismic Class I structure and capable of sustaining a service temperature of 212*F. The steel framing above the pool is a Seismic Class I structure and is tomado resistant such that it will not fall into the pool and damage fuel assemblies (Note:
The tomado resistant design evaluated fun tomado wind pressures on the bare structural steel frame). l
- h. The does rules at the surface of the SFP from spent fuel assemblies do not exceed 2.5 mromer in fuel building passageways during normal storage (i.e., 50 mrom/hr during fuel handling Rev.17 DSAR 3-40
MYAPC operations). Dose rates at the outside surface of the walls adjacent to the spent fuel area do not exceed the maximum radiation zone level for the arme.
- i. siphon protection is pnwided to preclude inadvertent draining of the SFP to a level lower than
-.- wl.r.- ^ ';10 foot atxwe the top of the active fuel Additional margm (to -. ; u4,7 ^ fi ig')
was pnwided by the instatation of a branch syphon break. J. The minimum spent fuel pool water buk temperature shell be 40*F. ]
- k. spent fbei storage racks may be moved only in accordance with written procedures which enounns that no rack modules are moved over fuel assemblies.
3.3.1.2 Fuel Pool Storage System Description The fuel pool storage and handling system is designed to prevent or trutgate the consequerres of accidents ===ar4=*=4 with the storage of fuel. A (::n/-:- of the fuel and irradiated compo1ents is presented below followed by the structures, systems, and components suppareng the .ufe storage and handling of the spent fuel and irradiated components. A description of the spent ttvA handEng system eqtspment is contained in section 3.3.2. 3.3.1.2.1 Fuel Assembiles (Typical) Figure 3.3-1 shows the details of the fuel assembly. Except for the two conann,4=d=1 assemblies, as assemblies consist of five guide tubes, nine fuel spacer grids, upper and lower and fittings, fuel rods and bumable absortier rods, if required. The number and type of bumable absorber rods in each type of assembly were c@ dependent. The structural frame of the assembly consists of the guide tubes, spacer grids and end fittings. The four outer guide tubes are mechanically attached to the stainless steel and fittings and the spacer grids are welded or mecharucally attached to all five guide tubes. A sleeve is provided in the upper region of the guide tube. The sleeve is made of staniess steel and is chrome plated on the ID. The sleeve is E-:f^'d f located within the guide tubes and cuirg rd. of the upper and fitting. The lower end fitting is a cast stainless steel structure. The lower end fitting contains flow holes for the fuel rods, and positioning holes for the guide tubes. A hold-down device is .'r,ww.pu. J into the upper and fitting and features a hold-down plate which acts on the underside of the fuel alignment plate (Fgure 3.3-1). The hold down piste is loaded by coil springs which are located at each of the guide tube posts. The inconel spnngs are posdioned at the upper and of the assembly so that the spring load combines with the assembly weight in Rev.17 DSAR 3-41
WAPC counteraceng #m upward hydraulic forces. The spnngs are sized and spnng preload selected, sus that a summent not downwed force we be maintaned for as normal and anecipated transient saw and temperatures conditions. The fuel assemt# upper and stung is a cast stainless steel structure. It saves as an anschment for the guide tubes and as the lifting exture. The fuel rod specor grids (Figure 3.3-2) maintan the fuel rod pitch over #m fb5 length of the Ibel rods. Typical grids are fabncated from profammd strips intadocked in an egg crate fashion and welded together. Each fuel rod is supported by springs and arches opposite the springs. Du springs press the rod against the arches to restrict relative motion between the grids and the fuel rods. The springs and arch posinons may be reversed from grid to gdd to provide addmanal restriction to relative motion. The perimeter strips also contain springs and arches in addlion to special features to prevent hang up of grids during a refueling operation. There are two types of rods contained within the typical assembly: fuel rods and discrets humable absorber rods. A bdef description of each type of rod is contained below. Fuel Rod The fuel rods consist of UO pedets, a compression spring and in the ABS-CE and EXXON design spacer discs, all encapsulated within a Zircaloy-4 or ZlRLO tube. The UO pellets are dished at both ends to accommodate the effects of thermal expansion and swelling. The fuel cladding is slightly cold worked Zircaloy-4 or ZlRLO tubing. The diametral gap between the pellet and clad ID is chosen to meet design cnteria regarding clad stresses and strains, and transfer of heat from the pellets. The compresson spring located at the top of the fuel pellet column maintams the column in its proper position during handling and shipping. It also provides support for the clad in the plenum region to prevent local buckling. In the ABS-CE and EXXON designs, an alumina spacer is located at the upper and of the fuel pellet stack to insulate the pienum region and to prevent UO chips from entering the plenum region. In ABS-CE fuel, an aiurmna spacer is located at the bottom of the fuel pellet stack to reduce lower end cap temperature. The pienum above the pellet column prowdes space for axial thermal expansion of the fuel column and for accommodation of fission gas. 1 Rev.14 DSAR 3-42
wApc Diacrate Bumabia Absorbar Roda Dancripdan olearses humable absorber rods are machenicdy similar to fuel rods except that they each contain a column of bumable poison penets insemed of fuel pests. The poison moserial is alumine wnh unilbrmly dispersed baron carbide partidos. The poison material is a 122.7 inch long column
.,,, a ~i contered wthin the 136 to 137 inch long aceve fbal region.
Componenes stored wahin the Ibel assemWy guide tubes in the spent fuel pool are: conkol element assemWy and noukon source assembly. A description of each is provided below. Contml Element Annembly The CEA (shown in Figure 3.3-3) comprised M 9ve inconel tubes 0.948 inches in diameter. Each tube (finger) is seeied by welded end caps. A gas expansion space is provided to Emit maximum 1 tubes stress due to intemel pressure de,M by the release of helium gas and moisture from the baron carbide. The overas length of the control elements assembly is appsoximately 161-5/16 inches. Four Angers are assembled in a square anny around the centrdy located SRh finger. The fingers are joined by an upper and nenng. i .. Most CEAs are fd @ and comprise Ave aethe Angers w.;.;..; .g B.C penets. The fbur outer Angers on d CEAs and al Ave Angers on most CEAs have sdver-indium. cadmium fWed Ups. Some CEAs are designed as part-strength CEAs, uWzing stainless steel potets in place of B.C in h angers. The fus simngth and part-strength CEAs are described in Figure 3.3 4. Neutron Sourca AssamhEas l _Between two and four neutron source assembses were instded in the reactor Ibr each cycle. Each source may include both a startup and sustainer source, orjust a sustaner source. The sources are held in vacant CEA guide tubes by trwens of protrusions which fit into slots cut into the top of the upper end fitting posts. O Rev.16
MYAPC 3.3.1.2.2 Fuel Storage System C::4-C- -i This section identilles the SSCs which comprise the fuel storage system. A d:: A;-C-ri of each is contained below.
. Fuel pool
. Transfertube and looietion valve
. Spent fuel storage racks
. New fuel storage Fuel Pool The fuel pool,37 feet wide by 41 feet long by 38 feet deep, is located in the south end of the fuel bulking adecent to the reactor containment. The pool is constructed of reinfbreed concrete with a well and floor lining of 1/4-inch thick stainless steel. The liner ensures another level of definnse against water leakage through the structure. Welds in the liner are backed up by test channels which are piped to the spent resin pit sump. An area is provided in the fuel pool for inspection of fuel
]
assemblies and space is also provided for a spent fuel cask. The spent fuel pool volume is
=;-E-wte ^:Y 59,116 ft'.
The fuel pool is protected from being drained down to a level =f--MiiV 10N feet above the active fuel assuming a rupture of any pipe normauy connected to the pool. The consequences of the siphoning incident is discussed in section 3.3.1.3. The anti-siphon device on the suction side of the fbol pool coadng pumps is safety class 3, ::': .
- j designed and periodicelly inspected fbr debris. j 4
The design of the pool, in conjunction' with analyses concluding that the MY spent fuel pool concrets ] structure and liner are e-p=hi- of resisting thermai stresses due to prolonged elevated pool water
!' temperatures of 212* F without failure, provides reasonable assuranos that the pool water level will f not be a5ected by credible incidents other than W.pvie;m losses due to normal and loss of forced i c - ang ac a nis.
Transfer Tube and Isointion Valve The fuel transfer tube penetration is provided for fuel movement between the refueling canal in the containment and the spent fuel pool. The transfer tube design prevents draindown through two independent machenical devices; by an isolation valve (FP-21) located in the spent fuel pool and a blind fienge seal at the refueling canal end.
'As noted in Secdon 3. a tranch syphon tweak has been installed which ir creeerm the margin to approximately 19'.
Rev.17 DSAR 3-44 i i
Spent Fuel Storage Racks The 26 stainless steel, posson curtain storage rocks provide for the storage of a maximum of 1758 fuel assemtdes. The fuel poolds esperated into two regions. Region I contains #ve racks b store up to 228 fuel assemblies, spaced on a minimum of 10.5 inch omnters and is designed for initial enrichments up to 4.5 weight percent U-235. Region il contains 21 rods to store up to 1530 fuel ammemtdes, spaced on a rninimum of 9 indi centers. Figure 3.3 6 shows the high denelly spent fuel
' pool layout for the two region pool. The spent fuel is stored in barated water. Water is necessary to provide shiniding and cooling fbr removal of the decoy host being released by the spent fuel assemtdes. The new and spent fuel pool structures including fuel racks are designed to withstand j the andcipated earthquake loadings as Class I structures. Refer to Figure 3.3-5 for a description of I restrictions for fuel which may be placed in each region.
The rocks are a single tier, rectuineer array of free standing modules and are n'1 A "i designed. The design ensures that during the event, adequate distances between the assemblies is moentained to prevent a wrMy event and to prevent adverse interaction between the pool structure and the reds. Structural meterial used in the rack design is ASME Section 11, SA 240, Type 304 or ASTM A240 Type 304L stainless stool.
' O >wFi -irii i ir==6 reri i: r-ee-..>sei--trii-baron carbide particles in a Type 1100 aluminum matrix. Boral Panels consist of 2 outer sheets of er type 1100 aluminum that clad a sintered plate of boron carbide in a type 1100 aluminum matrix.
New Fuel Storage The new fuel storage area, located at elevation 31 feet in the fbel building, contains faciuties to receive, handle, and store new fuel assemtdes up to 5.5 weight percent U-235. The new fuel is stored dry in racks that have a center-to-center spacing of 20 inches. Although not applicable to the delbeled condition, there are provisions for storing 160 fuel assemblies in the new fuel storage area. New fuel may not be placed in the pool. ( Rev.16 DSAR , 3-45
O 3.3.1.2.4 Fuel Pool Cochng, Makeup and Purf5 cation System r ::g.3 Fuel Pand Hast Enchanger The Ibel pool host enchanger is a cross-flow heet exchanger of the shei end-lube design. The heat exchanger is designed to cool the fuel pool water based on the foRowing condMons: Shen Side Tuba Side 108'F in and 111*F out 116*F in and 113*F out ] 2.88 x 17 BTUhrheet transfer 2.88 x 10' Blumr heat transfer ] h The Ibel pool heat emmenger is designed lbr 150 psig and 225'F on both the shot and tube sides. The tubes are 304 stainions steel, and the shell is of carbon steel. The above values are beood on a he flow of 1700 gpm, a challside flow of 1000 gpm and an outside tempemlure of 87'F and ] the *w heat load as of June 22,1999.
]
Fuel Pool Cooling Pumns There are two fuel pool cooing pumps designed for maximum efficiency at 772 gun per pump at a
- i. . total pump head of--;;-w,- * 'i 80 ft. The pumps may be operated at otherflows and heads. The casing is designed fbr 125 psi at 250*F. The intamal wetted surfaces of the pumps are type 316 stainless steel or an equivalent cast form. The pumps are installed for paraNel operadon. ]
i a Fuel Pool PurtAcation Pumn The fuel pool purt5cetion pump is designed for 250 gpm. The casing is designed for 200 psig at
, 250*F. The intomal wetted surfaces of the pump are type 316 stainless steel or an equivalent cast i
form. The pump is used for fuel pool cooling when the heat load is low, and it can also be used for slamming operations. Sidmmers are provided to couect precipitants on the water surface and pnwont it's build up on the pool well surface. Fuel Pool Filters The fuel pool preAlter and post filter are of the cartridge type and are located in the pt.ifn tion loop to remove the particulate or undissolved solids. The profilter retains 95% of the particles nominely 1 micmn, or larger, at a flow of 200 gpm. The vessel is constructed to Type 304 stainless ] steel and is rated at 200 poig and 250*F. The post fRter retains 95% of the particles nominally 1 ] O Rev.17 ~ DSAR 3-46
O micron, or larger, at a Aow of 200 gpm. The vessel is constmetod to Type 304 stainless steel and is rated at 200 poig and 250*F. Fumi Pool Daminarakar A mixed bed, undensator dominwaher is located in the spent fuel pool to remove dissohad sodds ] i Rom the water by ionic exchange with the resin. The veneel is constucted primerey of Type 304 ) stamines steel and has a design p of 50 poig. The vessel nonnety rests on the fuel pool Hoor ] at the 7.5 ft. elevation and circulates fuel pool water, at a nominal rate of 100 gun, using an intamally ]
; mounted pump and mobr. Fuel pool water is drawn into the vessel by the pump, through two inlet ]
i conneesons, and through the process bed. Water is discharged ham the vessel through a hose ]
, assembly which diskisules the water to a remote area of the fbel pool, ensuring proper pool ]
circulation. The domineralizar is sized to cxantain +;-yuev t.y 28 cubic feet of process media. ] i FuelPool System Vakes As valves in the fuel pool system are austandic stainless stool. The gate and check valves are butt-weided. The diaphrayn and bau valves are socket-weided. Fuel Pool System Phing AM the piping used in the fuel pool system is type 304 stainless steel with welded connections throughout, except for Sanged connections at the pump suction and discharge. Code Raouirements The design of aN components in the fuel pool cooling system complies with the followmg codes and regulations: Fuel Pool Heat Exchanger Tube side ASME Section its Class C Paragraph UW-2(a) of Sedion Vill applies Shell side ASME Section Vill Fuel Pool Cooling Pumps No code Fuel Pool Punfication Pump No code Fuel Pool Filters ASME Section til Class C Paragraph UW-2(a) of Sedian Vill applies Underwater DomineraRasr ASME Section Vill, Not Stamped ] Fuel Pool System Piping ANSI B31.1 Rev.17 DSAR 3-47 l
wApc 3.3.1.3 Design Evaluodon Several considerssons and analyses deternne em safety of fuel pool hendung and storage operadons. These include: Thermal Analysis Radiological Analysis Fuel Aeoembly Structural and Motorial Consideradons Fuel Pool SSC Structural and Meterial Considerations CrlucentyAnalysis The passive design features of the pool prevent signi6 cant loss of pool water and assum that suscient time is provided for operators to identify any incident condition and take the run==y action to restore coo 8ng, and, if requked, provide rnakeup to the spent fuel pool. The design and administradve controls assure that the Ibel ls sofsiy stored and the public and worters are adequately prrearemi Severalincident soonerlos are described below. These include: No pumps running (loss of forced cooing) Loss of host sink Siphoning Fuoi Handling incident Cask Drop Incident
, To determine the significance of these events, each of these incidents are reviewed in terms of the design criterte and design basis requirements. An evaluation is performed for each incident to determine the time available to the operatorto discover the incident and take remedlel actions. The consequences of each event are then evaluated in terms of pool water level and the radological consequences. Structures, systems, and wiiwa, .s' , credited in pnwonting or mitigating im event
, are then classsfied E,cu,J..g to its safety function. Similarly, activities credited to mitigets the condition are then defined. SSCs which are safety class are contained in section 3.1. A h* of the analyses and considerations pertaining to the defueled condition is contained below followed by a description of the incidents and the consequences. 3.3.1.3.1 Analyses and Considerations Thermal Analysis The host load as a function of time is prended in Table 5.5.2. The reduced heat load on 12/21W97 is ceiculated to be 5,420,000 811#hr.which :-;-;-MT ^J/represents a 75% reduction in heet load Rev.16 DSAR 3 48
MYAPC from the previously analyzed fuH core off-lood scenano when the plant was operating. 'me revised residual decay energy release rate was calculated using the assumptions and form datinns of Branch Technical Posidan (BTP) ASB 9-2" Residual Decay Energy fbr Light Water Reactors for Long-Term CooRng" and is consistent wWi the guidance contained in the Standard Review Plan (NUREG 0800), section 9.1.3. It also considers the long term uncertainty fraction for cooling times grooter than 10' seconds and are very conservative. 73rne do Sodf in this analysis, the " time to boil" ovaluelion was performed for three levels (elevations) of water, each at five initial pool temperature contSilons. The approximate elevations shown in the following table: Elevation Height from Height above Description (ft.) pool floor pt.) andive fuel (ft.) 31.5' 23 5 10' This elevation saw,ds to lowest point j on the cooling water retum piping into the
; SFP.
l 40' 32 5 19' This elevation corresponds to the level at I which the cooling water suction piping penetrates the liner. 43' 35 S 22' This elevation corresponds to the water level one foot below the normal pool water (i.e.. the low level instrument setting) The time to boil is also a function of the initial temperature of the pool water. Accordingly, for each elevation, time to boil is enhd=w at inRisi pool water temperatures of 80*F,100*F,120*F,140*F, and 154*F. Time to boil was nah w for each of the three water levels for differing heat loads as a function of time after shutdown and is shown in Table 5.5.2. A spent fuel pool heat iced test, honunofter referred to as the " passive cooling" test, provides more realistic site specific results. The passive cooling test demonstrated that the heat loss due to the effects of conduction, convedion, and n;-:- "-:-7 results in a significantly longer " time-to-bod" and i lower " bod-off' rate it is therefore more representative of pool cooling than the BTP ASB 9-2 calculation results. As a ruis of-thumb, tie f.pi::16.a;4 " passive cooling" results are 5-yc,4T ^y 4 71% of the BTP ASB 9-2 results. An example of the differences are illustrated in the table below. Rev.14 I DSAR 3-49
MYAPC Initial waterlevel = ehrv. 43' BTP ASS 9 2 Results Poselve Cooling Results Final waterlevel = elev. 31' 12/2Sf97 12/2Sf97
, ,, initial PoolTeenp. =80*F 5.42E+4 ETulhr 3.86E44 BTUthr Time to Boil 67.36 hrs. ; 2.8 days 93.5 hrs. ; 3.9 days Bos off Rate 11.84 gan:1.47 ft/dsy 8.3 gpm ; 1.05 ftkisy Days to reach a10 ft. etmo active 8.16 days 11.42 days fuel Radiological Analysis An evaluation was performed to determine the upper limit radiation fields in the fuel building and at surrounding onsste and offsde locations resulting irorn gamma radiation from the spent fuel in the fbel pool. This evaluation was performed as a function of both decay and waterlevel. The results are contained in Table 5.5.3.
The analyses are mnservative for ' die fogowing major reasons:
- 1. The SFP cross-sectional area was assumed to be loaded with disc.ii.ii,.d assemb!ies, each assembly having the same operating history as the worst case assembly currently in the pool.
- 2. No credit was taken for the shielding provided by the assembly and the fuel rack structural material above active fuel.
- 3. The sky shne radiation was based on the assumption that the gamma rays escaping from the fuel pool had a spectrum identical to the uncoluded spectrum originating within the fuel, thus allowmg the radiation to propagate to farther distances.
The basic findings in this analysis are as follows:
- 1. The dose rates on elevation 46', by the edge of the pool, are approximately a factor of 2 lower than those at the over-pool fA.JuiTii.
- 2. A water depth of about 11 feet over the fuel is needed for a dose rate less than 1 mrom/hr on the platform iir i+i'l'i above the pool.
- 3. For 23 feet of water above active fuel, the dose rate on the over-poolF A.Junii is less than .0003 rad /hr. ,
- 4. With 4 feet of water above the active fuel and one year of decay, the sky shine radiation level at the worst case receptor analyzed (20 m from the SFP conter) is =J-isvec':'i 15 mrom/hr:
this sky shine radiation level drops to =J-;-T.;,vc':'i 5 grem/hr at 610 meters (exclusion area ' boundary). Rev.16 s,- OSAR 3 50
WYAPC De aphomng incident assumes that siphoning occurs from the cooling water retum piping inlet and represents the worst case scenano. This inadent results in a worst case pool water level g-; wet r:'i 10 feet above the active fuel'. Radiological doses are illustrated in the table below. ] Water Level at 10 Fest Above the Active Fuel Decay time Sky shine at RalAudon Said on platform Radiatian fleid on 46' . since 124/96 840 meters tiready above pool elevation beesde pool W.AB) l 1 year .0003 rem /hr < 4 mrom/hr < 2 mrom/hr
, 2 years data not < 2 mrumthr < 1 mrom/hr avadable i l
3 years .00007urem/hr < .8 mrum/hr < .4 mrom/hr Radiological consequences are discussed for each inadent resulting in a loss of water level. Consistent with the operating plant requirements, dose rates around the fuel pool can r.c.wriOy be expected to be less than 2.5 mrom/ hour dunng normal spent fuel pool storage conditions. During fbei handling operations, the maximum dose rate to the operators is admmistratively maintained
'O below 50 mrom/ hour. This limit is preserved when analyzing the incidents below. Radiological effects are dist-ad in terms of direct radiation Gold, sky shme, pool surface evaponstion, and in the case of the fuel handling inadent, the decontammation factor for lodine scrubbing. Direct radiation and sky shine are a function of shielding (e.g., water level). The radiation dose rates affected through pool surface evaporation are applicable when the water is significantly heated or boiling. It is virtually a constant which is typically in the range of 1.28 mrem / hour (whole body) and 1.8 )
mrom/ hour (organ-lungs). The decontamination factoris a funcbon of waterlevel above the fuel pin. lodine is not a factor unless there is a breach of the fuel pin (e.g., fuel handling incident). Even then, the short half life of iodine coupled with the fuel decay time minimizes the source term dose affects contributed by radio-lodines. l l 1 i
%s noted in seenon 3, a branch syphon break has been instamed which increases the margin to approxirnstely 19'. ]
Rev.16 DSAR 3 51
MYAPC Fumi P - ar Striw*ral and *' " @ Cc.="r-- ' -s The fuel assemblies are designed to maintan their structural integrity under steady-state and transient operating condidons, as wen as under normal handing, shipping, and refueling loads. The design takes into account dlSerential thermal expansion of fuel rods, thermal bowing of fuel rods and CEA guide tubes, irredledon e5ects, and weer of aR components. Medanical tolerances and cieerances were established on the basis of the functional requirements of the components. AN components, induding welds, are highly resistant to corrosionJn the reactor and fuel storego erwironment. End core was made up of 36,192 Zircaioy 4 or ZlRLO clad rods in 217 assemblies. The rods contain slightly enriched uranium in the form of sintered UO pellets, bumable absorber, or weher-fuisd rods. The principle design structural critoria for the fuel rods is that the predicted permanent strain of the cladding is less than 1.0% during the fuel Nietime. Fuel Pool SSC Striw*wal er.d a'"# Cc.; :.t._:'.-e The fuel pool waN and floor slab is constructed of 6' 0" thick reinforood concrete with a waN and floor Ining of 1/44nch thick stainless steel to ensure against loss of water. The new and spent fuel pool structures including fuel racks are designed to withstand the anticipated earthquake loadings as seismic Class I structures. The spent fuel pool Ener is designed QAR. A structural evaluation was performed to determine the dead load and h%v.t. tic force a5 set on the SFP concrete war, floor stab, and stamiess steel Ener when the pool is subsected to an elevated pool water temperature of 212*F. Even when revising the thermal load factor, ACI allowables were met for a di5erential temperature of 186*F. Furtherqualitative dia-lans showed that even if the ACI asowabies were ex= w (i.e., long term loss of aN SFP cooling during the coldest possible winter temperatures), the water retaining capacity of the pool would be maintained since the pool side concrete remains in compression and the Ener has suf5cient ductNety to maintain inventory integrity. It was concluded that in the unlikely event that a loss of forced flow incident occurred, the concrete walls and base slab have adequate cz.pecity to resist the forces and moments generated by the self weight, water pressure, and themmi e5ects due to the bod-off of the pool. The load factors assumed { for the evaluation are for a service load condition and therefore does not restrict the elevated temperature condition to a, one-time only, Hmited duration incident. The spent fuel racks are designed to Seismic Class I requirements, constructed of ASME Section 11, SA240, Type 304, or ASTM A240 Type 304L stainless steel. The steel alloy is reasonably j corrosion resistant to the oxidizing effects of most electrolytes at low concentrations. The steel is I asp to corrosion in acidic solutions (pH less than 7.0) containing chloride or fluoride anions and can lead to petting of the material. Control of the water impunties are provided by the SFP purtfication domineralizer and filters and a chemical control program to assure that impurities are minimal and win not a5ect the structural integnty of the racks, liner, piping, vessels, exchangers and pumps. - DSAR . 3-52 Rev.16
l MYAPC The neutron absorber poison materialis *Boral"; a carmet componde matsnal made of baron carbide particles in a Type 1100 alummum metrtx. This componds is tughly durable and heat resistant. Boral pensis conost of two outer sheets of type 1100 aluminum that cied a sintered plate of boren carbide in a type 1100 alummum matrix. The type 1100 alummum material imports sufficient corrosion resistance by 'v....;.5 an alummum oxide layer on its surteos when W to n 5 agents. This oxide layer is stable in environments with a pH range frorn 4.5 to 8.5 (which is the operudng range ] of the spent fuel pool). The Boral penois are instaged snug between the outside wad of the storage ] cats and the 304 stainions steel sheath that is welded to the wed. Vent holes at the comers of the stainions steel sheath create a su#icient vent path for any potendel hydrogen produced by a water-aluminum reaction. The neutron absorber capaNmy of Boral is assured by fabrication design and material selecdon. The ] quantity and configuration of the panels is verifled by a visual Boral Sunmillance Prvy .. ] AN piping used in the fuel pool system is type 304 simniess steel, as is the fuel pool puiibr,vn filter and domineralizer vessels, liner plates, and the tube side of the fuel pool heat ead.i q.r. The l pumps (spent fuel cooling pumps and puitfe;un pumps) are constructed of type 316 staniess steel. l Each type of stainless steel is i::::^ tty comasion resistant. I 1 Crfticality of fuel assemblies is prMW by adequate design of fuel transfer, shipping, and storage faciuties, and by administrative control procedures. The two principal methods of pre;;..e q w A lltf are limiting the fuel assembly array size and limiting assembly interaction by fbdng the minimum . separation between assemblies and/or inserting neutron poisons between assemblies. The design basis for preventing w.L;;ty is that, considering the posable variations, there is 95% probatxBty at a 95% confidence level that the effecove multiplication factor (K,,) of the fuel assembly array will be less than 0.95 for flooded condibons as recommended in NUREG-0800. The following l conditions were assumed to meet this design bases: I
- 1. The fuel assembly contains the highest enrichment authorized without any control rods or any noncontained bumable poison and is at its most reactive point in life.
- 2. For flooded conditions, the moderator is pure water at the temperature within the design limits which ymids the largest reactivity.
- 3. The array is either infinste in lateral extent or is sumaunded by a conservatively chosen reflector, whicnever is +;-;-W:': for the design.
- 4. Mechanical uncertainties are treated by either usmg " worst case" conditions or by performing sensitivity studies and obtaining appropriate uncertainties.
Rev.17 i OSAR 3-53
MYAPC
- 5. Credit is taken for the neutron absorption in structural materials and in solid matenais added specifically for neutron absorption.
- 6. Where borated water is pnesent, credit for the dissolved boron is not taken except under >
tw=m accident cui4".u , where the double ce. e.v .cyvprinciple of NUREG 0800 is applied which requires at least two unlikely, independent, and concurrent events to produce a w.L ;;;f accusent. I For fuel storage application, water is usually present. However, the design methodology also pivvents accidental criticality when fuel assemblies are stored in the dry condition. For this case, possible source of moderation such as those that could arise dunng fire fighting operations are included in the W.d, The design base K,is 0.98, as recommended in NUREG-0800 lbr optimum moderation. The calculations which assure the criticality safety of fuel assemblies include uncertainties. The uncertainties are applied to assure that there is a 95% probability with a 95% confidence level that the specified design basis K, will not be exceeded. The total uncertamty added to the w;i ;;;f calculations include the following components:
- 1. Method Uncertainty: based on critical experiment ceirp#,.cie to establish method bias and variability.
- 2. Statistical Uncertamty: based on the statistics of the particular Monte Cario calculations.
- 3. Fabrication Uncertainly: based on manufacturing and mechanical tolerances, such as thicknesses and spacings. Either statistically- based uncertainties or " worst case" assuirpi;cre are used.
The wM;f design critana are met when the calculated effective multiplication factor plus the total uncertainty is less than the specified design basis effective multiplication factor. In the permanently defueled cer4"vun, the reactivity effects of a misplaced assembly represents the bounding case for determining the baron concentration required to maintain a 5% ak/k safety margin to w4;w.;;ty in the spent fuel pool. The analysis considered various misposition configurations inc!uding a fuel assembly laying horizontally on the top of the spent fuel racks, a fuel assembly placed sincent to or on the outsu$e comer of the fuel racks and a fuel assembly placed in the wrong spent fuel rack. A boron concentration of 670 ppm was assumed in the analysis for the postulated ] , rmsplaced fuel assembly. This was determined to be required to maintain a 5% ak/k safety margin to w.;;c ;;;f. This concentration is required whenever fuel assemblies are stored in the pool and a spent fuel pool assembly placement venfication has not been performed since the last movement of fuel assemblies in the spent fuel pool. An administrative limit is further detailed in this safety sr.Oy.;. report. Rev.17 DSAR 3-54 4
MYAPC incident Evaluation Pool water level is nommily constant except fbr relatively minor evaponWve losses. For the purposes of this evaluation, the pool water level is important from a standpoint of determinng time avaiable to the operator. The determr.dion that *T* pool water level ramens many hours or days aRar the incident is key to determirdng the severity of the event and the safety function perion'ned by SSCs. Pool water level is important to meet the design bens requrement to maintain the fuel covered with water and for pnwiding an endof4ncident water level which psovidee ad=T* shielding to mentain exposures below 10CFR 20 requirements. Consistent with the SRP criteria, the spent fuel pool and cooling, makeup and puMn hi systems design assures that, in the event of a failure in suction or retum piping, the pool level is not inadvertently drained to a level -c;-Tm;.T r f/10 fiset above the top of the active fuel. { S'w
.roin at elevations in the spent fuel pool are noted below:
Descrintion Elevation (approx.) Top of the spent fuel pool 46'0" High Waterlevel(alarm) 45' Normalwaterlevel 44' l.ow waterlevel(alerm) 43' Siphon breaker (suction side) 40'11" (Retum side) 40' ] Bottom of gate to the fuel transfer canal 12'0" Top of active fuel (approx.) Fuel peuets 21' (no higher than) Fuel pin 21'6" (no higher than) Bottom of the spent fuel pool 7' 6* The lowest elevation where piping penetrates the liner is on the cooling water retum (iniet) piping. The piping penetrates the liner at the 33.5 ft elevation and tums downward terminating at the 31.5 ft. elevation. A hypothetical break in the piping is postulated to occur on the discharge side of the pump or in the non-safety skimmer piping which causes a siphoning acbon and subsequent draining of the pool to the 31.5 ft. elevation. This condition is analyzed later in this section. (A new siphon ] anti-siphon device has subsequently been installed at El. 40'-0" on the cooling water retum piping) ] , An anti-siphon device is installed on the cooling suction outlet and is designed safety class 3. This penetrates the liner at the 40 ft elevation. The anti-siphon device e.dends up to just below the 41 ft. elevation. Therefore, the maximum level to which siphoning can hypothetically occur from this penetration is to the 40 ft. elevation. At this level, approximately 19 ft. of water remains above the active fuel. This anti-sephon devios prevents inadvertent draining of the pool and, for this reason, is designed to seismic Class I requrements. Rev.17 DSAR 3-55 l l
MYAPC O Several administrative contmis are in place to assure the continued safe storage operations. Spent fuel shippmg casks are MH.e evely precluded from being lifted over the spent fuel storage pool. Therefore, damage to the integrity of the structure from a shipping cask drop is precluded. In addition, alarm response procedures provide assurance that off-normal conditions are quicidy identified and remedied. For instance, alarms are provided to alert operators to off-normal pool conditions, including high temperature (110'F) and high/ low water level (45' and 43' elevations, i=;+1di) and radation level. Once alerted, operations personnel intervene before significant
$anges in water level or temperature can occur. In the event that the water level is increased one foot above, or falls one foot below, the normal operating level, an annuncistor would alert operations personnel. Similarly, a high temperature alarm in the control room would alert operations personnel to the high temperature condition. Plant operadng procedures require the operator to invesegate the alarm condison and inisate approprista remedial actions. This instrumentation, as was as the area radiation alarm, is useful for alerting operators of off-normal and incident conditions; however, the alarms are not credited in the spent fuel storage inadents. In the scenarios evaluated, credit was only taken for daily direct visual ir067 iiig by Operations personnel of the SFP temperature and level.
I a== of *=nt Fuel Paal chiina The following assumptions are used in this scenano:
- 1. Initial pool water temperature is 212*F,
- 2. Boil-off rate as of 12/29/97 equais 11.64 gpm (1.47 feet per day) and
- 3. Initial normal water level in the fuel pool (43' elevation),
4. Incident water level decreases to 10' above the active fuel (31' elevation). The 31' elevation (vs. the 31.5' elevation) is used because it corresponds to 10' above the fuel pellets, referred to hereinafter as the " active fuel" as pellet height is used to determine the radiological fleid and sky shine doses to receptors.
)
To analyze the time available to operators in the event of loss of spent fuel pool cooling, the pnncipal functions considered include the time it takes to boil and the boil-off rate. For ease in analyzing this event, it conservatively assumes that the temperature starts at 212*F. According to the BTP ASB 9-2 results, it would take a minimum of 8 days to reach the 31' elevation. According to passive cooling test results, it would take a minimum of 11 days to reach the 31' elevation. At ten feet above the active fuel, radiation fleid dose rates are expected to be no greater than 4 mrom/ hour (whole body) plus evaporative dose rates of no more than 2 mrom/ hour (whole body). This results in iWivivvi.; dose rate of no more than 6 mrom/ hour (whole body). An additional orgnn O Rev.14 DSAR 3-56 {
MYAPC dose rate of 2 mrom/ hour (lungs) is also expected. Historically,50 mrem /hr is the maximum dose rate permissable for fuel transfer operations. The doses received as a result of this incident are well below this limit. Sky shine dose rates at this level are approxirrcately 40 urem/hr of sky shine to a recephi 20 meters from the pool anter, and .0003 urem/hr at 610 meters (avrhinn area boundary). The above numbers are very conservative ham a radiological analysis standpoint as can be seen by the radiological analysis assumptions above. It is also conservative from the decay heat load standpoint as follows. Empirical data was gathered dunng a SFP heat up test conducted in October, 1997, in that test, the initial fuel pool temperature was 81*F and the pool level was at it's normal elevation (44 ft. elevation). The SFP cooling pumps were then secured. It took aprMT "7 70 hours for the pool water to reach 140*F (Note that there was virtually no loss of water level dudng that heat-up pedod. In addition, note that these results are conservative with respect to the passive cooling analytical results contained in section 3.3.1.3.1 which indicate 93 hours to boil). Assuming an initial SFP temperature of 140*F, BTP ASB 9-2 calcuistion results show that the tune-to-bod would take 5-;-472 34 additional
/ hours. Although not credited as part of this safety analysis, when combining the empirical and er47 4. data, a more reasonable estimate of the time to boil is approximately 100 hours.
Siohoning incident There are four penetrations which protect into to the fuel pool. These are: 1) the fuel transfer tube, ]
- 2) a piping penetration for cooling system suction (approx. 40' elevation), and 3) a pair of piping ]
penetrations for cooling system discharge (approx. 31.5' elevation)'. Failure of the transfer tube is ] not deemed credible due to the robust design (seismic Class I and safety class 3), it's passive function, and the existence of at least two passive mechanical devices preventing draindown. The piping penetration at the cooling system suction is designed with an anti-siphon device which is designed safety class 3 and seismic class 1. The lowest elevation to which siphoning can occur is 31.5 ft. elevation and is the most limiting siphoning incident. The associated piping systems connected to this penetration contain NNS piping which is not seismically designed. The evaluation of this incident assumes that the siphoning event starts immediately following the daily operations area walk-through wherein pool temperature and level is checked. The pool drains down to elevadon 31.5 feet due to siphoning caused by a non-mechanistic pipe break 2. At this level, ) l
' A pre.edsang piping penetranon (at opproxnnens s. *r) was utazed to create e retum line syphon break. ]
O 8The syphon break noted above makes this evoluenon conserveWye Rev.16
]
DSAR 3-57
MYAPC no less than 10.5 feet of water above the active fuel remains. Assuming an initial temperature of 80*F and a host load of 5.42E+6 BTU /hr, the time to bou is 39.5 hours. No credit is assumed for the low level alarm or for operations personnel to identify the existence of the dariamd fluid. Operadons personnel would detect the condidon on their deny round no later then 30 hours aner the incident and 9.5 hours prior to boRng. With 10.5 feet of pool water above the active fuel, radiological exposures to workers in the fuel bu5 ding is no greater than =-;-;-wJdr"i 2 mrom/ hour (whole body) for a person on the SJuivi directly over the pool, plus mix,.Ja whole body dose rate of no more than 2 mrom/ hour. At the time of identification, conective actions are taken to investigate the cause of the incident and start makeup to the pool. Under any credible sduation, several makeup sources could be irutisted within a few hours of identification. Not withstanding, the consequences of this incident is further analyzed assuming that corrective action is not taken immediately following identification. The assumptions are:
- a. No action is taken for another 32 hours once identified (30 hours after incident),
- b. No credit is taken for the remaining 9.5 hours to boHing (since the time of identification),
c. O d. A 25% allowance for the 24 hour surveulance is assumed (30 hours), and The water is at a boiling temperature with a boil-off rate of 11.64 gpm or 1.47 ft. per day.
- e. BTP ASB9-2 decay heatloads are used With the above assumptions, after 62 hours (30 hrs. to identify the condition + 32 hrs. to effect remedial action), the water level would be no lower than 8.5 feet above the active fuel. With 8.5 feet of pool water above the active fuel, a straight line extrapolaton of radiation field dose rates indicates that the operators on the $suiii directly over the poon% be subject to a radiation field dose rate of no more than 35 mrom/ hour whole body plus evaporative whole body dose rates of no more than 2 mrem / hour, or 37 mrom total (whole body). An additional 2 mrem / hour organ (lung) dose rate would be received. By the edge of the pool, whole body dose rate is on the order of 20 mrom/ hour.
These totals still remain within the 50 mrom/ hour maximum radiation exposure limits indicated above for fuel transfer operatons. Offsde dose rates are less than 10 urad/hr at 20 meters, and less than
.003 grad /hr at the exclusion area boundary (610 meters).
Several conservative assumptions are used in this incident. One assumption involves the 25% allowable for the routine daily round. Using the BTP ASB 9-2 heat loads, at an initial pool water temperature of 80*F, over 62 hours are aveHable from the initial event to reach a pool water level of 8.5 feet above the active fuel. This estimate is still accurate assuming an initial pool temperature of 100*F. When assuming 24 hours to discovery,62 hours is accurate up to and including an initial Rev.14 DSAR 3-58
MYAPC pool temperature of 130*F or less. Finally, using the possive cookng results, significantly more time is avadable. Using a heat load of 3.85E+6 BTU /hr, at an initial pool temperature of 80*F, the time to boil is over 93 hours. The bod <f rate is 1.05 ft. por day which a80ws for additional time to reach the 8.5 ft level above the active fuel. Other conservative assumptions in the radiological analysis assume that aN assemblies have the same operadng history as the worst case assembly. Therefore, actual doses would be lower and a reducdon in weder level below 8.5 ft could occur and stlN remam below the 50 mrom/ hour whole body does to personnel at the edge of the pool. Assuming the conservative BTP decay host loads and a bod off rate of 1.47 ftJday, it would take
^ --;;-s4T - J/ 40 hours for the level of the pool water to decrease from 8.5 feet to 6 foot above the active fuel. This cerr.eper.de to a whole body done rate of approximately 750 mrom/ hour to operators around the penmeter of the pool, and a total of 102 hours from the time of incident initiation. 6 feet of shielding above the active fuel corresponds to a dose rate of g-yc-.T:Y/ .15 urem/ hour to a receptor at the exclusion area boundary. In au cases, reasonable time is available for operators to take corrective actons to restore make up to the pool. Additionally, strict radiological mntrois are effected to minimize exposures to personnel ALARA.
Fuel Handung incident For the fuel handling incident, no credit is taken for systems or components to mitigate the incdont such as fuel building ventilation systems or control room ventNation systems. Additionally, no credit is taken for scrubbing of released gases (DF=1) and the dropped assembly is based upon the release of the fuel rod gap inventory of the worst case (highest bumup, eniici v r4 and longest operating history) ceiryosite with one year decay. The accident assumed an instantaneous puff release at ground level meteore;cgice; conditions to determine offsite exclusion area boundary and control room doses. As shown in Section 5, doses are acceptable assuming no filtration. The
-=* mal, control room and offsite radiological doses from the fuel handling incident bounds all of the fuel pool storage incidents described above.
Cask Drop Incident Movement of the cask has been carefully examined and in no case does the cask pass over systems or equipment important to safety. An analysis has revealed that a 100-ton,6-foot diameter cask dropped 42 feet straight down into the fuel stonage pool would puncture the steel liner and penetrate 1.5 feet into the 6-foot concrete floor. Leakage of 2 to 5 gpm may be expected due to the p.rir:2"T/ of the crushed concrete, backfill, and bedrock. In addition, a maximum of 2.5 gpm leakage may occur through the liner leakage detection system for each leakage zone breached as the result of the cask drop. Avadable makeup capacity is biv,J,cerey higher than the draindown rate of the spent fuel pool. Equipment location and drainage capability is such that no damage to critical equipment from this leakage would occur. DSAR 3-59 Rev.16
j MYAPC 1 The design standards and factors of safety, the documented maintenance program and operator j quemicssons, the suict suponmon of ad cook movement, and the smaN hachon of time that the cask peones over the edge of the spent fuel pool reduos the probability of a4pping fall to a tolerably low value. N-r,.t .d:esi, spent fuel shipping casks are ed...;..;...LW/ prohibited from being lifted over the fuel pool. Lams of Heat Sink l Loss of DHR cooling to the SFP heat exchanger (E-25) will result in a heat-up of the pool. DHR ] normally flows through the shou side of the fuel pool host exchanger and cools the tube side pool weher. If DHR is not available, temporary hoses may be connected to flanged connections provided ] - on the shell side of the heat exchanger to allow the alignment of altamate cooling water supplies. Loosef DHR cooling is bounded by the " Loss of Spent Fuel Pool Cooling" incidents described above. ] An abramel operating procedura describes the steps to irubate allemate cooling from a fire system supply hoes to the E-25 inist connection and a fire hose from the E-25 outlet connection to the fire pond. The fire system has ample c.pecity to provide altamate cooling to the spent fuel pool heat exchanger. De fire system also has a diesel driven fire pump. ] Evaluation and Condusion De primary function of the spent fuel pool cooling, makeup and punlicaton system is to ensure that the fuel remains covered and that suitable shielding exists to maintain radiological exposures below 10CFR 20 limits. This is pnmarily a passive design function and requires no active system intervention by the system w wiKE,.61. to prevent or trutigate the consequences of a design basis accident. In the cummt plant condition, the G;eley;c.: source terms have decayed to a level where the postulated events would not result in offsite dose limits exceeding a small fraction of the 10CFR100.11 Ilmits. The reduced heat load, and the assurance of pool integrity provide the margin required to creoit only the passive design features of the pool. Operators are provided ample time to discover the incident conditions, effect remedial actions to restore pool cooling or provide make-up prior to uncovenng the fuel, and maintain adequate shielding above the active fuel. Although boiling is pernutted as a service condition, long term boiling as a service condition is not practical due to the continuous makeup capacity required and the domineralizer temperature limits. A- ,iff, normal pool water temperatures are maintained below 110*F. The pool water is cooled using the spent fuel pool moung pumps. Additionsuy, the puirA pump may be used to cool the l Rev.17 ' DSAR 3-60
MYAPC pool water if conditions permit (i.e., depending on the decay host load, the pool temperature, DHR flow rate and the DHR inist temperature). The spent fuel pool purtAcadon system is provided to maintain water chemstry within speciRcations and is not required to prevent or trutigste the conesquences of an accident. For out of-speedication semistry condsons, conecthe acdons are taken bened upon the resuns of sample analyses or other indications. Refer to section 3.3.1.5 for the parameter values monitored and their frequency. Makeup water is provided to the spent fuel pool to compensate for loss of pool water level primar#y due to evaporation losses. Sufficient available rneio up water capacity exists through diverse sources to allow operators to take manual actions to compensate for liner leakage or natural and incident svepere^w' n rates due to boiling. Makeup capacity includes: the Pnmary Water Storage Tank (and associated pump P-SFP2), town of Wie====t water supply system, and the water storage pond ] i (and fire pump). The acthm -.venents of the spent fuel cooling, makeup and purincation system are not relied upon to assure: 1) the integrity of the reactor coolant pressure boundary, (2) the capability to shut down the reactor and maintain it in a safe shutdown condition, or (3) the capability to prevent or mitigate the consequences of an accident resulting in potendel offsite exposures mrnparable to the applicable guideline exposures set forth in 10CFR 50.34(a)(1) or 10CFR 100.11. Due to the time available to operators to provide makeup to the pool, the acthe components of the spent fuel pool cooling, make-up and purification system are not safety related nor are they credited to function during or following a design basis earthquake. The time available fogowing each of the incidents, including the worst case siphoning event, conservatively shows at least 62 hours since the initiation of the event to establish make-up without exceeding the 50 mromhour (whole body) limit normally established for fuel handling operations. The analysis also conservatively demonstrated that, with 6 feet of water over the active fuel, over 100 hours are available to an operator to provide make-up. With 6 feet of water, operators around the pool perimeter would be subject to a whole body dose rate of no more than 750 mrom/ hour. In all cases, a suitable work erwironment is available, and adequate time is avaliable to provide offsite or onsite make-up capetxBty without relying on safety related equipment. Fuel handling equipment is discussed in section 3.2.2. The results of the systems, functions, and actions credited relative to fuel handling are included below. i l Rev.17 O i DSAR 3-61
I l MYAPC The below listed SSCs are designed to Seismic Class I requirements.
- A functional ard .'$ei device on the suction side of the pool to prevent draindown,
- Fuel piadorm and hoist and yard crane
- Fuel pool concrete structure, Hner, racks and transfer tube
- Flow limiters on the liner leakage detection system The fonowing system functions were crednad in the evaluation.
- Capability of the pool liner and concrete structure to sustain long term boiling,
- CapabiNty of the transfer tube to maintain its integrity, including assocuited valves and flanges.
- Fuel platform and hoist and yard crane design safety features to assure the safe handung of fuel and casks.
. - At least one make-up water supply to the spent fuel pool is available within 24 hours after the identification of the worst case siphoning event. The fkwv rate shan be equal to or greater than the boil-off rate.
- The capability of the racks to maintain adequate specirg between the assemblies. This includes the placement of fuel in the racks in the proper region. Refer to Figures 3.3.5 and 3.34 for the proper placement of fbel in the racks as a function of bumup.
The following actions were credited during the events:
- Daily operator / handler area walk-throughs for SFP level and temperature.
- The abigty to provide make-up to the fuel pool within 82 hours fham the initiation of an loss of pool water sipets incident.
- Assuring that boron concentration was adequate prior to the movement of fuel.
- Assuring that appropriate tests and inspections were mnducted on fuel handling equipment prior to the movement of heavy loads and that such movements are administratively controlled to pin.at movements in the viciruty of safety related equipment.
- Appropriate cMTJeky is maintained in the pool
- A Boral Surveillance Program is implemented
- Administrative controis to prevent shipping casks from being lifted over the spent fuel pool 3.3.1.4 Fuel Storage - System Operation Normal Ocarating Soncificanons:
Fuel pool water level Between 43 ft. and 45 ft. elevations Fuel pool water level during fuel handling operations a 42.2 ft elevation ] Fuel pool temperature Between 40*F and 120*F ] O Rev.17 DSAR- 342 l
MYAPC Fuel pool cooling pumps s1800 gpm DHR Pump flow rate to E-25 s1100 gpm FH8 Normal Radiation Levels (passage ways) s2.5 mrommr(whole body) Radiation Levels during Fuel Handling operations s50 mrom/hr(whole body) watermale up Domineralized water System Ocaradon Descripnon The fuel pool cooling system is shown on Fgure 3.3-9. This syslam removes the decay heat from ] spent fuel stored in the fuel pool by circulating the pool water through a heat exchanger. The fuel pool cooling pumps (P-17 A and 178) take suction from the fuel pool at approximately 2 feet below the low level alarm (plant elevation 43 ft., or 35' 6" from the bottom of the pool). Each pump is capable of maintaining the pool watcr within normal operating parameters under all temperature conditions provided that the level is above the low level setpoint and electrical power is available. Suction flow by the spent fuel cooling pump is circulated through a heat exchanger (E-25), and retumed to the fuel pool below normal water level at the 31 ft elevation. Flow from each pump may O be independently throttled to obtain the desired flow. The fuel pool cooling pumps (P-17 A, B) are controlled locally via a local start /stop button at the pump on the 21' elevation of the Fuel Building. The fuel pool cooling system has a puiibhi loop consisting of a pump (P-85), a pre-filter (FL-2), and a post filter (FL-29) which may be c-;+_U independently of the fuel pool cooling system. The purification pump (P-85) located in the Fuel Building is controlled locally via a local start /stop switch at the local Motor Control Center on the 21' elevation of the Fuel Building. The purification pump can take suction from the cooling pump suction line during penods when cooling is not required or from the discharge of the heat exchanger wtw the cooling system is operating. Flow from the dscharge of the fuel pool purification pumps is directed through the fuel pool profilter, and/or postfilter, while the retum is through the fuel pool cooling retum line. The purtlication pump can also be used for id. .n ,g operation or to crculate and cool pool water through the heat exchanger. DHR normally flows through the shell side of the fuel pool heat exchanger and cools the tube side pool water. The desgn DHR ficw rate is adequate to assure cooling of pool water. A discussion of the DHR water rystem is provided in section 3.3.3. Rev.17 DSAR 3-63 l I
MYAPC Spent fuel makeup capacity is prowded through diverse sources. Makeup capacity includes: the Primary Water Storage Tank and the town of Wiscusset municipal water supply. Water is also avaliable from the fire pond. None of the makeup sources are safeN related. Two fire pumps are located in the fire pump house located near the water storage pond. A diesel engine drives one pump while the other is motor-driven. The busding houses the diesel fuel tank, the battenes and control board required for the diesel operation. Therefore, under severe conditions, fuel pool inventory may be restored using the 2500 gpm are pumps and the water storage pond. { During sidmmer operations, the spent fuel pool puriAcetion pump takes suction from four surface skimmers to keep the pool surface free of foreign matter. The pt iT T.u6 pump discharge is directed back to the fuel pool through the underwater retum Ene during this operation. 1 The Fuel Building hea Radiation Monitor is th M in section 4 including the response to radiation alarm conditions The Fuel Building Ventuation System is described in section 3.3.5. Operating pmcodures specify the methods and actions required to pmwde fuel pool makeup, cooling and purtscation under normal operadng conditions. Operating pmcodures also specify the methods and actions required to restore cooling temperatures and water level in the spent fuel pool under abnormal operating conditions. 4 3.3.1.5 Monitonng and instrumentation Instrumentation ; The fuel pool high or low level and the fuel pool temperature alarm conditions are annunciated in the j control room. I Pressure transmitters and gauges are prowded on the discharge of each pump in the spent fuel pool ] cooling system. This includes the differential pressure indicator located in the punscation loop which may be valved-in to measure differential pressure of each filtering component. The fuel pool heat exchanger temperature indicating transtrutter on the fuel pool cooling heat exchange outlet line is ] used to regulate the number of cooler coil fans operating in the DHR System. ] Tell-tale connections are provided to aid operations personnel in monitoring for liner weld leakage. The tell-tales discharge into the spent resin pit sump. The manually operated sump pump discharges to the Holdup Tank (TK-109).' The spent resin sump pump high level switch alarms on the PLC. ] Rev.17 , l DSAR 3-64 -
n MYAPC Upon receipt of a Spent Rosin Pit high level alarm condition, an operatur is dispatched to investigate the condition. If the alarm condition is due to liner leakage, actions are taken to restore level, as eppispri.te in accordance with operating procedures. Chemistry Spent fuel pool water ciwiney is monitored to minimize the potential effects of corrosion which i could affect the safe storage ofirW fuel. Chemistry surveillance activities are pefumsi within the specified interval below, with a maximum allowable extension not to exceed 25% of the specified { interval. The spent fuel pool water is maintained with domineralized water and addittves, as required. { The water chemistry is maintained and monitored in accordance with the following values and frequency: { Parameter l Values Nominal Frequency l pH at 25'C 4.5 to 8.5 Weekly Chloride less than 100 ppb Weekly j Total Halogens .25 ppm maximum Weekly i Boron a 670 ppm design limit ]
= 1200 ppm administrative limit MontNy* ]
Gamma Isotopic NA(for trending only) Monthly Suspended Solids NA (for trending only) Monthly 7 Days when moving Fuel
]
Although dose analyses for the various spent fuel storage and handling incidents are conservative, they do not consider the effect of an increase in dose due to particulates in the spent fuel pool water or surface contaminants. In general, maintenance of water quality is necessary to prevent degradation of the spent fuel and other stored materials in the SFP. Accordingly, the chemistry program includes sampling analysis and corrective action recommendations to protect the stored materials and minimize radioactive contamination. The underwater demineralizer and the slipstream purification pre and post filters, located in the Fuel l Pool and the Fuel Building, respectrvely, are periodically monitored to assure that activity limits, and ; differential pressures, are within = par
- ations.
I Ventilation O The ventilation systems consists of heating and ventilation and is designed to provide a suitable environment for equipment and personnel. The ventilation system utilizes fans, filters, dampers, Rev.17 DSAR 3-65
MYAPC heating elements, and ductwork to accomplish the desired effects. Refer to sechon 3.3.2 for a desenption of requirements applicable during fuel movements. The ventilation system is NNS. Monitoring Fuel pool level and temperature is visually monitored and logged nominally every 24 hours. Operators shall p.A.e.h morntor the temperature shown on the host exchanger indicator to assure the temperature is within specification. Operations shall periodically monitor fbr tell-tale liner leakage. The suction siphon breaker is monitored periodically for sgns of apparent or potential blockage. ] Mantananca The components (flanges or valves) which assure the integnty of the fuel transfer tube are safety related and are periodically inspected to ensure that they can satisfactorily perform their intended safety function.
~
The suction siphon breaker at the 40' elevation is periodicaNy inW to assure there is no ] blockage. A Boral surveillance program shall be maintained.
]
1 o Rev.17 l i DSAR 3-66
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MYAPC t 3.3.2 Fuel Handling System 3.3.2.1 Design Basis ! Danign Critada The fuel hendung design critada is as tsows:
- 1. Criticetty in now and spent fuel storage is prevented by physical design features or processes.
. A ;+: TC safe/configurallon is emphasized over proceduW controis.
- 2. Appropdate shielding is provided to rnest the requirements of 10CFR 20.
- 3. The fuel budding is condnuously monitored by area specmc detectors. Audible and visual alarms are activated at the detectorlocations and the control room for radiation levels in excess ]
of f4+ tit-J d limds.
- 4. Spent fuel storage systems are:
- a. designed to prevent or midgate accidents which could lead to the release of 440c.nt amounts of radioactivity aSecting the public health and safety,
- b. designed, fabdcated and erected to withstand, the addisonal sorces that might be imposed by natural phenomena.
'< The design basis of the fuel handling system with regard to the spent fuel handNng incident and the cask drop incident are contained in sections 3.3.1.3 and 5.0. Fuel building ventilation system requirements and radiation montoring are contained below and in section 4.0. Heavy Loads program requirements are addressed in section 3.4. DAsiGILBaait The fuel handling design critada is as fotows:
- a. Irradiated fuel shal not be consolidated until it has cooled for a period not less than 730 days after final dacharge from the reactor. Prior to conducting irradated fuel pin activities, a review is performed Jo assure that the licensing / design basis is adequately reflected in operating procedures. This includes a revow to assure that occupational doses as a result of a pin rupture are maintained ALARA consstent with the requisite design basis ventilation flow,
- b. A spent fuel shipping cask shal not be lifted over the spent fuel storage pool.
- c. . Fuel handling devces have provmons to avoid dropping orJamming of fuel assemblies during transfer operations. I
- d. Fuel hondung equipment is demgned to preclude lifting of the assembly above a height which could cause undue radiation awpaam to personnel (250 mromMr) on the platform drectly over the pool.
O Rev.17 DSAR 3-73
MYAPC
- e. The spent fuel racks are designed such that the Mg of a 2500 lb. (submerged weight) fuel ]
assembly from a height of 18" above the top of the racks wiu not incur damage which could result in w.Mii. A subsequent evaluation has shown that the racks are able to withstand the ] dropping of a 2000 lb. (submerged weight) fuel assembly from a height of 22.5 inches above ] the top of the racks without incurring damage which could result in wTMiy or spent fuel ] cooling concoms. ]
- f. Fuel handung equipment is designed to withstand the loadings that would occur during an Operational Basis Earthquake and wiu not fail so as to cause damage to any fuel elements should it occur during fuel transfer operations.
- g. Placement of assemblies in racks shall be based upon the bumup rate in Figure 3.3-5.
3.3.2.2 Fuel Handling System Description The fuel handling system provides a safe and efficient method to unload and store new fuel assemblies in the fbol building, transfer spent fuel assemblies in the fuel pool, and to ship spent fuel assemblies off-site. Equipment is also provided to transfer control element assemblies to guide tubes in each fuel assembly, as required. The refueling equipment ammgement is shown on Figure ] 3.3-8. The system is comprised of the following: Spent Fuel Movable Platform and Holst ) Yard Area Crane Communications 1 Sonnt Fuel Movable Platform and Hoist The basic structure of the movsbie platform and hoist is a traveling bridge which spans the spent fuel pool and moves on rails over any spent fuel storage i rdan, the new fuel elevator and the transfer system upending machine. A fuel hoist is mounted on the bridge structure. The hoist hook supports handling tools for grappling fuel assemblies below water. The rotation of fuel is manually controlled via grapple tool. All operations may be monitored by binocular viewing as required. CR-9 has the following three basic interlocks:
- 1. An electrical cutout that stops movement in the north direction if CR-9 gets to close to CR-6.
CR-6 has a similar interlock to prevent it from driving into CR-9.
- 2. An interlock that prevents concurrent operation of the CR-9 hoist with either the bridge or the trolley,
- 3. An interlock that limits upward travel of %, hoist (two block interlock).
Rev.17 DSAR 3-74
MYAPC Yard Area Crane The 125/20-ton yard area crane is a conventional bndge crane with a three-motor trolley and powered bridge. It was designed to the mquinunents of the Electric Overhead Crane Institute, EOCl Specification N1 lbr Class A service. The impnwed plow steel rope has a 5.5 to 1 safety factor and the remaining components generaBy have a 5 to 1 safety factor. The main hoist is powered by an allemsting cunent 30 hp wound rotor motor and operates at 3 feet per minute. armidng is provided by an eddy cunent contal brain and two spnne set, electrically reisesed holding brakes, endi with a reting of 322% of ful load anotar torque. Power to release the holding brakes is provided by the same circuit breaker that energiens the main hoist motor, thus, the holding brakes 1 are applied whenever the main hoist motor is de energized. The eddy current brake is always in l servios. A spring-operated overload swhch de energizes the main hoist motor and sets the brakes whenever a preset load is exceeded, preventing any over stress lng of the rope or other crane components. In addition, a centrifugal overspend swach de-ernrgizes the motor and sets the brakes whenever a preset lowering speed is - This, with the eddy cunent brake, provides ) i redundant protection against runaway lowering. Up travel of the main hoist is limited by a weight-operated limit switch and a goer operated limit seltch operating in series to prevent two block, while a single gear.cperated switch limits down travel so that two full tums of cable remain on the drum. Earthquake up-kick lugs are provided so that no part of the crane will leave the rails in the event of a seismic disturbance. Electrical interiocks prevent the main hoist from passing over the stored spent fuel. 3.3.2.3 Design Evaluation ~ Fuel handling equipment is designed Sommic Class 1 as it is necessary that load bearing components and brakes operate properly to prevent situations which could cause damage to the fuel pool or the fuel. The fuel platform and hoist and yard crane are designed to Seismic Class I requirements and the interiocks are fad-safe. The design of load handling equipment complies with applicable industry standards and codes. The design of the hoists and cranes coupled with the administrative controls provide assurance that a heavy load will not be dropped which could result in damage to safety related equipment and loss of required safety functions. To assure the safe handling of fuel and casks, design safety features include, as necessary: rnochanical stops, brakes, limit switches, and electrical interiocks. These design controls, toget%r with the administrative controls (e.g.,0;-isui training, W,wdic load tests, functional tests, prohibitions to lifting spent fuel shipping casks over the pool) provide assurance that the consequences of accidents will remain well below the 10CFR 100.11 and EPA PAG limits. In addition, administrative controis restrict handling of heavy loads in the vicinity of safety reisted equipment. Rev.14 DSAR 3-75
MYAPC The fuel handling SSCs designed to Seismic Class I requirements are the fuel platform and hoist and yard crane. l The following fuel handling system functions were credited in sechon 3.3.1 The fbel platibrm and hoist and yard crane design safety featmassure the safe handling of } fusi and casks. ( Administrative controis to prevent shipping casks from being ufled over the spent fuel pool ' Design controls assure that the spent fuel assembly is not lifted greater than 22.5 inches over ] the top M the rocks for loods up to 2000 lbs. And not more than 18 inches above the racks for ] lands in excess of 2000 lbs (but less than 2500 lbs). ] Design controis prevent spent fuel shipping casks from being lifted over the spent fuel pool. The following actions were credited during the events: Assuring that appropriate tests and inspections were conducted on fuel handling equipment prior to the movement of heavy loads and that such movements are adminstratively controlled to prevent movements in the vicinity of safety related equipment. Assuring the piscoment of fuel in the proper rack region. Refer to figures 3.3-5 and 3.34 for the proper placement of fuel in the racks as a function of bumup. O 3.3.2.4 System Operations Tha folkWeg operatons controls are implemented for the spent fuel movable platform and hoist:
- a. Pnor to conducting fuel handling, a complete checkout, including a load test using a dummy assembly, shal be conducted on the fuel handling crane used to handle irradiated fuel assemblies (prior to such use unless checked out in the last 18 months).
- b. Prior to conducting fuel handling operations (i.e., prior to handling irradiated fuel assemblies unless tested in the last 18 months), fuel handling equipment interiocks functional test shall ]
be conducted,
- c. The spent fuel hoist is controlled to assure that the assembly is not lifted more than 22.5 ]
inches over the top of the racks for loads up to 2000 lbs. And not more than 18 inches over ] the top of the racks for loads in excess of 2000 lbs. (but less than 2500 lbs). ]
]
i Rev.17 DSAR 3-76
MYAPC The followog mntrois are implemented regarding the yard crane
- a. All phases of fuel cask movement are under strict admrustrative control, with au movement done in accordance with written instruction and check 4f lists. The operation is under the direct supervision of a responsible member of the Mene Yankee staff, and crane operators are trained to meet the requirements of USAS B30.2-1967 Overhead and Gantry Cranes.
- b. Documented mantenance and performanos checks, induding a loaded operadonal M W h crane controis and brakes, are perfbrmed prior to each fuel cask handung evolution.
The following requirements were relocated from the todmical specificatens:
- 1. The fotowing conditions shal be satis 6ed during movement of imediated fuel within the spent fusi storage builoing:
- a. Radiation levels in the spent fuel storage area shaN be monitored condnuously.
- 2. Spent fuel shipping casks shall not be lifted over the spent fuel pool storage pool. ]
- 3. Spent fusi storage racks may be moved oiWy in acconiance with wntten procedures which ]
ensure that no rack modules ata moved over fuel assembues. 3.3.2.5 Inspection and Testing The yard area crane receives annual maintenance examinations by qualised personnel. I O Rev.17 DSAR 3-77
MYAPC O 3.3.3 Spent Fuel Pool (SFP) Decay Heat Removal (DHR) System 3.3.3.1 Design easis Danign cittaria The Spent Fuel Pool (SFP) Decay Heat Removal (DHR) system is designed to be a reimble source of heat transfer kom the SFP weenr. It removes decay host from pool water to maintain temperature within speciRestion. The system is designed to preclude the possitniity of radioactive leakage kom reaching beyond the plant boundary. Damion assis 1
- a. The secondary cooling Loop (DHR) system is an air-cooled, closed-loop cooling system. I
- b. The system piping and cuivW4. are classified as Non-Nuclear Safety /important To the DeAJeled Condition (NNS4TDC). Tha system is constructed and maintaned in accordance with appropnate industry codes and standards.
- c. The system is protected dudng periods ofi ,iukhW freezing temperature conditions.
- d. Altomate sources of cooling can be effected in the event that the DHR system is out of O service. Evaporation and pool water makeup are sources of altamate moling.
- e. The system is capable of mentaning bulk temperatures s 120*F with an outside temperature of 87'F with a duty of 2.88 rnillion BTUs per hour. ]
- f. The system is able to recover kom a fuel pool boiling incident where the water in the pdmary and secondary system may reach 212*F.
- g. Pool water bulk temperature is maintained above 40*F.
]
3.3.3.2 System Description l The SFP DHR system flow diagram is shown in Figure 3.3-9. SFP DHR provides cooling water service for spent fuel pool decay heat removal. SFP DHR is primarily filled with deionized water, with additives to protect against freezing and corrosion. Heat generated by the spent fuel is transfoned to the fuel pool water, then to the DHR system in the SFP heat exchanger (E-25) and lastly to the air by water-to-air coolers (E-SFP 1 to 6). The DHR solution is crculated by one of two DHR cooling pumps. An expansion tank and air separator are also provided. Local hardwired controis and indications are the primary means of controlling the Spent Fuel Pool leiend (SFPI) equipment. A highly reliable non-nuclear safety related computer system, referred to as the Programmable Logic Controller (PLC), is also provided. The PLC monitors, displays, and alarms certain desired DHR SFP parameters to the operators in the SFPI control room. Limited Rev.17 DSAR' 3-78 I
MYAPC l DHR loop control functions are provided by the PLC as dia==M in 3.3.3.5. The Primary Water Storage Tank provides a makeup soun:o to the DHR loop via a manual hose connection. In the event hat makeup water is required, deionized water in the PWST is pumped via P4FP2 located in me fuel building to a hoes connection in the secondary loop. SFP DHR Pumos Two parallei SFP DHR pumps are provided and each deliver a nominal flow of 1000 gallons per minute. One pump is usualy in service. The otheris an installed spars. The pur- <smal wetted surfaces are stainless sesel l Water-to-Air Coolers The DHR water-to-air cooler is cv.T--:::1 of six parallel, finned-cooled type coolers; each cooler containing three fans. The SFP DHR coolers may be bypassed to maintain system flow balance and desired heet transferrate. The heat transfer rate is controlled through the cycling of the cooler fans. Deioniand water is used in the DHR loop during the first and possibly future summers to maximize heat transfer. Ethylene glycol will be added for the winter months. The units are designed to 300 psig and 350*F. Normal operating conditions are anticipated to be -;-; usT--Y/ 80 psig and 85'F. The tubes and fins are constructed of stainless steel and aluminum, respectively. DHR Exnannion Tank The DHR bladder expensbn tank is sized to accommodate fluid expansion and contractions i resulting from temperature variations in the secondary cooling loop. The tank capacity is 211 gallons. All wetted pu"uis of the tank are constructed of stainless steel. The bladder is constructed of butyl rubber. Air Separator The air separator removes entrained air in the secondary coolant. The unit and trap are made of stainless steel. Pining System The piping and valves used throughout the system are constructed of stainless steel with the Rev.16 DSAR 3-79
MYAPC exceptions of a smal portion of carbon steel piping used at the inlet and outlet of the secondary side of E-25 and the shell side of E-25. The SFP DHR system uses corroman inhibited deionized water. A connection is prowded on the OHR loop to add corrosion inhitmors as reqused. The design of equipment in the SFP DHR comply with the following codes: Shell side of Hx (E-25) ASME, Section Vill Surge Tank ASME, Section Vill Cooier.- Cvic ..rci.i Industrial Standards Air Separator ASME, Section Vill DHR Pump: CeiTT-rci.i industrial Standards Piping, Valves and Fitting- ANSI, B31.1-1980 3.3.3.3 Design Evaluation The SFP DHR system does not perform a safety related function. Analyses demonstrate that the fuel and fuel pool structural ceTw re. are capable of sustaining boiling as a service condition. For operational considerations, pool temperatures are normally maintained below 120*F. In addition, adequate time is available to the operators to identify and remedy a low pool water level condition or loss of cooling incident. Standby power sources are not required. The secondary cooling loop flow rate is controded by the DHR cooling pumps. The typical operadng flow rate is maintained below 1100 gpm; the manufacturer's maximum recommended shed side flow rate for E-25. Flow through the heat excnanger is controlled by throttling the heat exchange discharge valve. For pump protechon, and as an indication of reduced system inventory, there is a suction pressure switch on each pump which trips the pump on low suction pressure. Six water-to-air coolers in parallel prowde the heat sink necessary to maximize heat transfer from the fuel pool. Each cooler may be bypassed to maintain system flow balance and dessed heat transfer rate or facilitate cooler maintenance activities. Over pressure protection of the secondary ] loop is also provided.
]
]
O Rev.16 DSAR 3-80
MYAPC O NW were performed to detemune the capaaty of the DHR system. The assumptions used in the analysis are as Ibliows: Fuel Pool Demgn Decay Heat Load (6/22/99) 2.88 E+6 BTU /hr ] DHR(secondaryloop) Flow rate 1000 gpm SheE Side inletTemperature 108'F ] Discharge 113*F ] Outside AirTemperature 87'F Fuel Pool Cooling Pumps Flow Rate 1700 gpm (1800 max) Tube Side inletTemperature of 116*F ] The results show that the outlet temperature of the tube side of the heat exchangerwould be 113*F.
]
Abnormal Conditions Each of the followng abnormal conditions are evaluated forits effects on the ability of the SFP DHR O- system to provide heat removal from the spent fuel pool heat ewis.gw and to preclude the possibility of radioactive leakage from reaching beyond the plant boundary. Increasing System Temperature - Changes in SFP DHR temperature may result from seasonal changes in the air temperature and humidity. Secondary coolant in the tube side is air-cooled by the coolmg fans which force amtnent air across the coil to remove heat. The DHR fans are cycled to obtain the desired heat transfer rate. Under worst case heat loads, the secondary cooling system is designed to maintain the pool temperature below 120*F. I - of SFP DHR - Loss of SFP DHR cooling would cause the pool temperature to increase. Vanous make-up sources are capable of providing water in the event of maximum evaporeevn (i.e., bod off). The normal deionized water make-up is provided from the Primary Water Storage Tank. Altemative makeup sources include the Wiscasset water supply and the fire water system. In the event that normal cooling is interrupted, alignment of altemate cooling supplies to the spent fuel pool heat exdsp may be provided by an electric or diesel driven fire water pump through flanged connections pnmded on the shell side of the heat exc v r or. The fire water pond may act as both the water source and the ultimate heat sink. Additional means include the use of the existing fire hydrant adjacent to the fire pump house which is connected to the Wiscasset water o su, . Rev.17 DSAR 3-81
MYAPC Loss of SFP DHR cooling to the spent fuel pool heat exchanger is bounded by the " Loss of Spent Fuel Pool Cooling"mcident described in Sections 3.3.1 and 5.0. An abnormal operating procedura describes the steps to supply altamate cooling from the fire system through a connection to the fire pond. The fire system has ample c.pecity to provide altamate cooling to the spent fuel pool heat exchanger.
]
Lankage IntdOut of the System - Spent fuel pool weher is penodically morutored for radiation levels. In addition, the purification loop assures that radiation levels in the pool are minimized to preclude the existence of sigrvficantly contammated water. Lankage from the pnolinto the SFP DHR system is unilkely since the SFP DHR system is mantained at a higher proc 4ure than the fuel pool cooling system. Ifleakage were to occur through the host exdienger, the makage would be diluted by the volume of SFP DHR and result in ani ' ;yJcent amount of SFF DHR System contamination. In onder to identify potential leakage, the SFP DHR water is periodically sampled for radioactive contammants. Leakage into the SFP DHR system may be indicated by low pressure drop across ] the heat exchanger, or by loss of pool water level. Freeze pici.ction and corrosson control is effected in the secondary loop by the use of ethylene glycol and dipotassium phosphate, respectively. O Loss of Off-Site Power - if offsite electrical power is lost, the SFP DHR pumps will shutdown. To restore power to the SFP DHR pumps, the onsite diesel generator may be manually started. As described in Sechon 3.3.1, there is sufficient time to restore power. 3.3.3.4 System Operation The fuel pool heat exchanger E-25 transfers heat from the primary (tube) side pool water to the secondary (shell) side SFP DHR cooling loop. The DHR pumps circulate water from the fuel pool heat exchanger to the water-to-air coolers and back to the shell side of the fuel pool heat exchanger. The DHR bladder expansion tank accommodates fluid expansion and contraction resulting froin temperature variations and an air separator removes entrained air in the secondary coolant. The pumps are located in the fuel building at elevation 21 feet. One pump is normally operating and the otheris an installed spare. The water-to-air coolers reject heat from the DHR system into the surrounding atmosphere by the use of finned tubes and forced air. The coolers are located in a diked area. A gabled roof over the diked cooler reduces the collection of precipitation inside of the dike. The diked cooler area is released to storm drains after it is determined that the fluid is compliant with waste discharge O Rev.17 DSAR 3-82
l l MYAPC specifications. The operation of the cooler fans are staggered to adjust for heat rejection rate vanations due to variation in ambient air temperature. Flow is provided to all the coolers whether or not the fans are operating. Cooler bypasses are installed to allow for maintenance. During winter operation some of the fans may be in standby as radiant and natural convection heat transfer rmy remove sumcient heat from the secondary system. The DHR pumps are provided with a local "stop run" control switch. One pump is likely to run condnuously year round; however, the pump may be secured to prevent cu. cow;;c.g, as required. Pump run-time is a function of the decay heat in the spent fuel pool. As the decay heat decreases over time, the pump may be isolated more frequently to p event overcooling. 3.3.3.5 Monitoring and instrumentation There are two automatic actions performed by the DHR System instrumentation (automatic trip of ] ) the DHR pumps on low suction pressure and cyding on/off the fan coolers) and they are both non- ] safety related. {
] 1 l
Both DHR pumps are provided with suction and disdiarge pressure transmitters and gauges. The ] suction pressure transmetters provide a low pressure alarm in the contro! room and initiate a DHR ] pump tHp via the PLC in the event that a low suction pressure condition exists. In addition, pump O discharge pressuis transmitters provi$e continuous indication in the SFPI control room. The water-to-air cooler fans are controlled from a temperature instrument on the fuel pool heat ] exchanger outlet. The temperature indicating transmitter information is fed to the PLC. The PLC ] automatically cycles the cooling fans to attain the desired cooling rate. In the event of a programmable logic controller malfunction or out-of-service condition, the capability exists to disable the system and locally secure or initiate control of DHR pumps and fan coolers. A primary to secondary differential pressure switch is provided for local indication and low differential pressuru alarm in the SFPI control room. The system is monitored for proper concentrations of a corrosion inhibitor (dipotassium phosphate) and the degree of freeze protection capability. In order to identify potential leakage, the SFP DHR water is periodically sampled for radioactive containments. O Rev. ,z DSAR 3-83 2
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MYAPC 3.3.4 venesson Systems
]
venseman systems == designed to provide a suitable environment for equipment and pasonnel. These systems pnwide homeng and air condWoning. The vendadon systems utman fans, Stars, dampers, heedng elements, coadng elemords, and dud work to accompush the desired eNocts. In the rm conkoded areas, the fuel building and the primary auxSlary building, outdoor air is suppded to these simasures. Air is enheueled in greater quantities then it is supplied to.maintG, the budding at a nogetvo preneure and ensure that the general flow of air is into the structure. The exhausted air is discharged post radiation monitors in the primary vent stack or Fuel Building exhaust duct. t
- c- 1 Tet :.d =r- =' -. t r--. of V raWinn Exhaust Treatment Svstam The Ventilation Exhaust Treatment System includes au -W designed and installed to reduce radioacthe motorialin particulate form in emuents by passing ventgotion through HEPA filters for the purpose of removing partkadetes from the exhaust stream prior to release to the environment.
Sue systems are not considered to have any effect on noble gas amuents. Er.s.r.d Safety Feature (ESF) atmospheric cleanup systems are not considered to be VentRetion Exhaust Treatment Systems components. Inanaction and Testing Fuel Budding Ventilation System inside the fuel budding is the fuel storage pool. The stored fuel wiR remain in the building until it is transferred to dry-casks or off-site. The fuel building will remain independent of other systems, ] structures, or ceTipur ,6 undergoing the decommissioning process. In this regard, the fuel building ventilation system wiH remain operational. The fuel building ventilation system has Administrative Controis established to ensure the following in=W and Testing requirements are satisfied: Requirements for an operable ventilation system Requirements for demonstrating operability Requirements for testing and test frequency Rev.17 ) DSAR 3-85
)
MYAPC i Control Room and Auxiliary Ventuadon Systems The remaining vendadon systems consist of the Control Room Venendon and AuxWary Vendadon Systems. These systems are in buk5ngs where f+x , .! i-.;.4 activmes wW occur. Maine Yankee has committed to the intent of NUREGdCR 0130. These vendadon systems wHl be subjec to the testing and opombility requirements of DSAR, Chapter 7, DECOMMISSIONING. 3.3.4.1 Fuel Building VentNetion System
]
'3.3.4.1.1 Design Basis
]
The fuel building ventilation system is designed to: maintain the operability of the fuel building ] equipment during normal operating w4r.'wr., ensure that air flow is from outside into the building] to prevent unmonitomd release of radiadon, and ensure that exhaust air is continuously monitored. 3.3.4.1.2 System C::=;-tion
)
p 1he pronously existing fuel building ventilation system has been replaced with one designed to support the needs of the spent fuel poolisland. Variable speed exhaust fan HV-SFP1 draws from 2000 cfm to 12,000 cfm of cooling air through a louver and filter assembly in the northwast wau, and discharges the air through a duct mounted on the exterior of the fuel building east wou. Before it is discharged to the outside, the exhaust air passes through HEPA filters to remove any particulme materiais. Because the exhaust system is once through, the fuel building is always maintained at a siight negadve pressure. The headng and ventNetion equipment in the fuel building is designed to minimias moisture condensation on the walls and the roof and to limit the space temperature to a maximum 95'F and a rmnimum of 60'F. A radiation .r-,rui 9 system is installed in the fuel building exhaust duct which continuously monitors the discharge air to identify any potential releases. Fan FN-SFP1 supplies unconditioned building air to the heat exchanger cubicle to aid in cooling the components located there. 3.3.4.1.3 Design Evaluation
]
Since the consequences of the Fuel Handling Accident are significantly below 10CFR100.11 limits without toldng credit for the building ventilation, the fuel buRding ventilation system is not designated to mitigate the consequences of a Fuel MarxNing Accident. Therefore, the fuel building ventliation system is not safety related. Rev.17 DSAR 3-86
I O ~~ 3.3.4.2 Auxiliary Vendiation Systems
]
This secdon h- the vendiation sysimms at the pient. The vendiation systems include: primary auxNimry bunding, front anicekontrol room buliding, wart bunding, containment building, containment apray buiding, and RCA builting ventNodon sysimms. These ventNation systems do not perform any safety funedons in the permanently defueled condition. 3.3.4.2.1 Primary AuxNimry Building Ventilation System
]
Ventilation air to PAB areas at elevationsjt1 feet and 11 feet is suppued by HV-1. This unit supplies ] approximately 30,000 cfm of unconditioned mored outside air to these spaces. Supply unit HV-2 has ] been removed from service and its inlet duct has been modified to allow introduction of fresh air ] through the louvered opening to the 36 foot elevation. In addition, this exposed opening serves as ] a pressure relief pathway in the event that HV-1 trips dudng normal operation. ] The PAB exhaust is separated into two sutW: PAB shielded areas and PAB miscellaneous areas. PAB shielded exhaust is h from elevations 36 feet and 21 feet. The shielded exhaust is Altered by a HEPA filter assembly (previously referred to as the PAB tray filter) and discharged by fan FN-1B to the pdmary vent stack. PAB miscellaneous exhaust is filtered through a HEPA filter assembly (previously referred to as the PAB safety class filter) and discharged by fan FN-1A to the prinwery vent stack. Either fan may also be a5gned to exhaust the Containment or both areas of the PAB Each exhaust fan has a nominal capacity of 27,000 cim. Exhaust fan FN-2, which previously served the 36 lbot elevation, has been removed from service. The supply and exhaust airflows are balanced to ensure that the PAB is mantained at a shght negative pressure relative to the outside, as well as to the Service Building and the new HP ct apcint waikway. 3.3.4.2.2 Containment Building Ventilation System ] The Containment Building is ventilated by unconditioned outside air drawn through the containment personnel hatch, and through the HV-9 housmg, by fan FN-1 A or FN-18. The effluent air is filtered through a HEPA filter assembly and discharged to the primary vent stack. HV-9, previously the containment supply air unit, has been removed from service, and its filters and steam heating coils have been removed to decrease the resistance of the airflow path. Because the exhaust system is once-through, the contamment is always maintained at a slight negative pressure relative to the outside. FN-1 A and is may also be abgned to exhaust the PAB. O Rev.17 DSAR 3-87
MYAPC 3.3.4.2.3 Containment Spray Budding Ventiladon System
]
The Containment Spray Building is ventilated by exhaust fan FN-44A. Unconditioned outside air l's drawn ine the building through the HV-7 bypass duct. AAer cooling the spaces in this building, the airis passed thmugh HEPA Ritors and discharged b the primary vent stack. Supply unit HV-7 has been removed from service, and exhaust fan FN44B wm not be m-powered. The Containment . Spray Building is maintained at a slight negadvs pressure taladve to the adjacent spaces as a result of the once-through ventilation system design. 3.3.4.2.4 RCA Busiding Ventilation System ] The RCA Building will continue to be used for procesang radiological waste during f+x ,,,, ' "x,;rg. Supply unit HV4 provides unconditioned filtered outside air to this building, and exhaust fan FN-30 { exhausts this space. This fan also exhausts the trash compactor in the LSA Building. The air i exhausted from the RCA Building is passed through HEPA fMors and discharged to the primary vent stack. The supply and exhaust airflows are balanced to ensure that the RCA Building is maintained at a negative pressure relative to the outside. Building heat is provided by four thermostatically-controlled electric unit heaters. i 3.3.4.2.5 Front Office / Control Room Building Ventilation System ] The Front Office Building (Gatehouse) is served by a central air handling unit, AC-3, two roof-mounted heat pumps, and five exhaust fans. AC-3 is a six zone unit which conditions (filters and heats or cools) and recirculates the building air, while also introducing a continuous stream of fresh air to all the building spaces it serves, including the Spent Fuel Pool Island Control Room. ] There are no automatic signals which isolate ventuation to the Spent Fuel Pool Island Control Room. ] in the permanently defueled condition, the reduced source term (Gection 5) precludes the need for ] automatic isolation of the Control Room ventilation system. The filter bank in AC-3 consists of ] Wla roughing filters, and heat is provided by a multi-stage electric heating coil installed in the ] unit's discharge pienuim. Supplementary building heat is provided by electric baseboard heaters. ] Ceiling exhaust fans in the PBX room, the Spent Fuel Pool Island Control Room assure positive ] airflow to these spaces. ]
]
3.3.4.2.6 Administration Building (WART Building) ] 3 The Administration Building houses the body count room, the chemistry count room, and the ] relocated RCA d,esprArit on the first floor, and offlos spaces on the second and third floors. Each ]
. Rev.17 ]
DSAR 3-88
MYAPC floor is ventilated, cooled, and heated by its own dedicated HVAC system. The PAB ventilation ] system maintans the PAB at a negative pressure relative to the RCA cr@it walkway. This ) assures that any potentially i=*A-? ^/ contaminated airin PAB cannot enter the walkway or the ) RCA checkpoint.
]
O l l 1 Rev.17 DSAR 3-89
MYAPC 3.3.5 Auxiliary Systems ] This section discusses auxiliary sysimms supporting plant openning miuipment. These systems do not perform any safety reisted funcuans in the permaneney defueled condulon. 3.3.5.1 Boric Add Makeup ) Design cntaria crtucemy in new and spent fuel storage shan be prevented by physical systems or processes. Such means as geometricauy safs configurations shan be emphasized over procedural controis. svatem ommaintion A self-contained batch tank mixer and pump unit is provided to increase boron concentration in the ] pool, if required. Danign Evalunhan irradiated spent fuel is stored under water in a reinforced concrete pool, lined with stainless steel. Fuel assembues are spaced and the racks are fabricated so that criticauty is precluded. Although O 'a - ' r i '" a ii o r av *e ' e. 'a r i i 6 maintain the irradiated fuel subcritical. r ' ir e < a i e '- . J soric acid concentranon requimments are only applicabie in the ame patod immediately paceding fuel handling movements, dunng fuel handling movements, and up to the time that an assembly piacoment verification is performed. Boric acid makeup capability is provided to account for pool water dilution due to the addition of domineralized water makeup used to compensate for losses of borated water due to pump seal, maintenance activities or other minor pool water leakage paths. The boron concentration required as a result of the analyzed " misplaced assembly" incident is discussed in section 3.2. Bulk pool water boric acid concentration generally increases under routine or non-routine (boil-off) evaporative losses in the pool. Accordingly, when routine dommeralized water is added to the pool to account for boil-off or natural evaporation, the addition of boron should not be necessary. Major losses of pool water due to draindown, er dilution of boron due to flooding, are extremely unlikely, but are the only methods through which sigruficant makeup water is added to dilute the pool water. For the reasons stated above, loss of boron concentration under normal storage conditions is not a significant event from the standpoint of criticality. Rev.17 DSAR 3-90 l 1 i
NC Significant dilutions can occur as a result of adding dommeralized water to compensate for a draindown event or from flooding from a neartry pipe break wrd.;iic g water. Flooding due to natural phenomena is not pan =Na due to the elevadon of the spent fuel pool. To preclude acudirg design oradminstrative controis are implemented. Design considerations may { include one or a combination of the following elevation and location of significant water sources, I berms, freeze pie;.ction (e.g., heat tracing, heating, draining), drains, instrumentation or valves in ppng systems. For instance, due to its elevation and location, a failure of the pnmary water storage f tank would have no flooding affect on the pool. Similerty, the fire protection system is comprised only { of detection equipment and is limited only to extemal fire suppression capability. Therefore, a fire piet.ctien pipe break or inadvertent actuation of the system causing a flooding condition is not possible. I l if a primary water pipe break (e.g., hose nozzle) occurred in the fuel building at the 46' elevaton, the water would drain to the fuel building sump or into the fuel pool. During routine rounds, the operator ] would identify the leakage visually or by excessive sump pump run time via an integrator, in the event that the break was significant, sump pumps would pump to the Waste Holdup Tank (TK-109). ] in addition, a berm exsts around the perimeter of the pool to preclude draining of water into the pool. Flood levels exceeding the height of the berm is prevented by the existence of other leakage paths (e.g., doors, drains, etc). In the event that, for any reason, water level in the pool was to rise, a pool water high level alarm would alert operators to the conditen. In both cases, operators would be disp.kJ ed to the area to investigate the cause of the condition and effect ceri.ctive actions, as appropriate. Administrative controis include maintaining a temperature in the fuel building to preclude pipe freezing. Routine Opei.ivr walk-throughs can also be credited. The defense-in-depth of the design and administrative controls precludes floodmg from being a significant incident relative to boron dilution. The design and administrative controis, including reactivity control via the Boral plates, water moderator and spacing of the assemblies, provide reasonable assurance that a provision to add boron via manual means is sufficient to meet technical specification requirements. The following SSCs were credited in the above discussion:
. Fuel building sump pump and high level alarm
. Fuel building sump pump integrator
- Spent resin pit sump pump and high level alarm
. Waste hoidup tank (TK-109) ] ;
. Fuel pool high level alarm O Rev.17 DSAR 3-91
l I MYAPC Tests and insametions Tests and inspections of instrumentation sensors and alarms are penodicauy conducted in accordance with plant procedures. Dew operator rounds are conducted to identify any flooding condnions, excessive sump pump run time (Integrator), and potential freezing conditions. 3.3.5.2 PrimaryWaterSystem ] The primary water makeup system provides a source of domineradzad water for use by various ; locations in the plant. DomineraEand water is stored in the primary water storage tank. The primary ] water storage tank is prevented fmm theezing by a combination of wan insulation and an electric hosting element. Makeup to the primary water tank is provided som the Wiscasset potable water system. The potable water is processed through a truck-mounted domineraHzer system prior to bemg pumped to the primary water storage tank. chemistry periodicah samples the output of the demeurailmer system to ensure proper water queuty. The following systems, cwww.ents, or areas are supplied by the primary water system:
. Spent fuel cask decontammation area
. Fuel Building hose connections Spent Fuel Pool Island (SFPI) make-up ]
3.3.5.3 Primary Vent and Drain System ] The majority of the Primary Vent and Drain System has been abandoned IAW the Maine Yankee ] decommesioning process. Portons of the onginal system have been retamed and rqH:onfigured IAW ] MY DPR 98-003 to provide for the safe and limited collection, transfer and retention of potentially ] radioactive waste. The system re-con 6guration provides for collecting sources of waste fmm the ] Spent Rosin Pit and Fuel Building Sumps and transfer of these wastes to the Waste Hoidup Tank ] (TK-109). ] 3.3.5.4 Radioactive Waste Processmg System ] The radioactive waste piMg system was originaNy designed to coNect, store, process, monitor, ] O and dispose of solid, liquid, and gaseous wastes from the plant. Sections 4.4 and 4.5 provide details Rev.17
]
]
DSAR 3-92 1
1 l i 1 1 MYAPC O of the Uquid Waste Treatment and Solid Waste Treatment Systems as reconfigured for the ] t l
- h. . ;c,r;rg. '
]
] l The Baron Recovery, Liquid and Gaseous Wastes Systems no longer serve a function in the
] I C+x . ' ' --is and have been abandoned by Maine Yankee. The Decommesioning Operating ]
Contractor has reconAgured the Duratek Uquid Waste iwM4 skid to support wasts water ] processing fham decommusioning activities. q
] l O
1 { O Rev.17 DSAR 3-93 !
MYAPC t% U 3.3.5.5 Fire Protection System
]
Fire detection and suppression systems are provided to minimize the adverse radiologica consequences of fires at the plant.
]
]
The fire suppression water system consists of the fire pond; fire pumps, and distribution ]piping associated sectionalizng control or isolation valves. These valves include yard hydrant curb valves
]
and the first valve aheed of the water flow alarm device, which is provided for those sprinider and standpipes which are normally pressurized with water. Water is supplied to inside fire j suppression systems either by underground piping, or by connection to nearby fire hydrants ]using fire hose.
]
1 Supply water for the fire pie 6crw system is contained in the fire pond. Two 2500 gpm,115 psi ] rated fire pumps, one electric and one diesel, take suction from the fire pond and discharge to the ] fire system loop. The motor driven fire pump will start automatically when system pressure] drops to =pp-047 ^ 'y 90 peig. The diesel fire pump will start automatically when system pressure] drops to ep;-wkc^ly 80 peig. A pressure maintenance system corgJ.T.4 of a pump, hydro-pneumatic ] tank and air compressor is designed to maintain system pressure between approximately 100] and 110 psig.
]
]
The discharge of the fire pumps is routed to the yard loop, as shown on figures 3.3-21 and 3.3-22.
]
The loop consists of 12 inch undeiyvund piping which encircles the plant. Eight fire hydrants tap ] off of the loop, seven of which are provided with adjacent hose houses. Six of the hydrants have
]
block valves which pommt isolation of those hydrants from the fire mains.
]
]
Vertical and horizontal fire barriers have been established in designated areas of the plant. Fire ] doors and fire dampers have been provided as needed in these ,'0couciis. Mechanical and electncal
]
pentrations are generally sealed with fire retardant materials within or near the fire barrier.
]
3.3.5.5.1 Fire Detection Systems
]
Fire detection systems are installed in the following area:
)
1.
]
Fuel Building (Drurnming Room and Heat Exchanger Cubicle)
- 2. ]
RCA Storage Building 3.
]
LSA Buking
]
- 4. Fire Pump House 5.
]
Low Level Waste Storage Building O ] Rev.17 ] DSAR 3-94
MYAPC O 6. Wart Budding
]
- 7. X-16 Station Service Transformer
]
- 8. Staff Building
]
]
The detection systems will send a signal to the Control Room, whidi is a constan0y manned locadon. ] Manuel pun stations are provided lomily withui al of these buildings except the Fire Pump House and ] the Fuel Buuding.
]
1 ~ 3.3.5.5.2 Hose Stations and Hose Houses
]
]
Eight hoes houses are located exterior to the plant and are supplied from the 12-inch undergroung ] 1 yard loop. The hoses are of various sizes and lengths prh to the hydrants, depending ] upon the location.
]
1 Home stations are located in interior areas of the plant. Each station consists of a water shut-off valve
]
and walkmounted home rack. The hoses are of various lengths, depending upon the locations and ] eedt is fitted with a spray nozzle. A wet pipe system provides hose stations for the Staff BuNding off ] fkom the yard loop. Dry manual hoes stations protect the followmg areas: ) 0- 1. Turbine Building 1
]
- 2. Service Building
]
- 3. Wart Building
]
- 4. Fuel Building l
]
- 5. RCA Storage Building
)
- 6. LSA Building
]
- 7. PAB
]
1 Water is supplied to the PAB hose stations by operating a valve at the X-16 station service ] transformer deluge hooder in the PAB. Water to al other locations is supplied by a combination of ] valve line-ups and connection via hose to a nearby fire hydrant. ] 1 3.3.5.5.3 Wet Pipe Sprinkler Systems
]
A wet pipe sprinkler system is located in Warehouse 2/3. The piping is filled with water at all times. ] The sprinkler heads have links designed to melt at various temperatures from 165'F to 286*F, ] depending upon the specific location. An alarm check valve in the supply line to this system will ] activate a Control Room alarm when flow is sensed. ] O Rev.17 DSAR 3-95
MYAPC O 3.3.5.5.4 Deluge Sprinidw System
]
]
A deluge sprinider system protects the X-16 station service transformer. The spray header supplies ] open nozzles which spray water onto the transkumor. The deluge valve fior the transformer spray ] header is operated automaticely by a signal ftom heat sensors at the transformer. The valve is also
]
operated by mmate manual put stations orlocaly at the valve by monuouy operating a handie. ] 3.3.5.5.5 Dry Manuel Spdnider System
]
)
Dry manuel sprinider systems protect the following areas:
]
]
- 1. Turbine Building
]
- 2. Welding Shop
]
- 3. Service Building
)
- 4. Fuel Building
]
- 5. RCA Storage Building
]
- 6. LSA Building
]
]
O The piping to spriniders is rmrmany depresourland. The sprinider heads have links designed to melt at various tempxatures from 165'F to 286*F, depending upon the specdic location. Water is
]
]
supplied to these areas by a combination of valve Ene-ups and connection via hose to a nesty fire ] hydrant.
]
Rev.17 O DSAR 3-96 1
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Figure 3.3-21 Rev. 17 - FIRE PROTECTION SYSTEM 3 Maine
- m - ,- - .,
Yankee w,- 1
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W% s=, = = t.ya FIG. 3.3-22 REV.17 -
- =.~.:::.=r.::'- -
FIRE PROTECTION SYSTEM Maine Yankee
./ #
s
MYAPC 3.3.5.6 Metec ciegicalInstrumentation ] The on-site meteorological data collection program was designed to meet the intent of Regulatory Guide 1.23, Revision O. The .'.t ^::-14-fM Monitonng Program is diamanad in detailin Appendix A. ~ 1he it ^ E -34 l monitoring system utilzos a guyed 200-foot tower located on-site. Instrumentation on the tower is located on booms at the 33-foot and 195-foot levels. Wind measurements are observed at heights of 35 feet and 197 feet above the tower base. Wind speed is measured with transmitters and anemometer cup assemblies. Wind direction is measured with wind direction transmitters and wind vanes. Both systems have starting speeds of less than 1.0 miles per hour. Temperature sensors are provided to measure ambient temperature difference between the 32 foot ] and 194 foot. These data are obtained from resistance temperature sensors and Model 414L linear ] bridges. Ambient temperature is also measured by this system for the 32-foot level. A rain gauge is installed on the ground near the base of the tower. A digital recording system is the primary data enilar+ inn mechanism for the Maine Yankee
't:^ EA:-gical System. Through the use of a programmable logic controller, each meteorological ]
parameteris scanned approximately once every second and compiled into 15-minute averages. The ] digital data are available for display on terminals located in the Control Room. EOF /TSC. ]
]
]
The programmable logic controller system is used as the pnmary data collection and storage ] mechanism for the Maine Yankee meteorological system. ] Rev.16 O DSAR 3-99 1 1
MYAPC 3.3.6 Electrical Systems
]
The station electrical system consists of two independent systems, one for the original plant equipment and systems and the second to support long term plant loads. These long term loads include the Spent Fuel Pool Island and stodon loads required dunng the duxxnmissioning process. Sections 3.3.6.1 through 3.3.6.3 descrees the plant electrical distribution system. Section 3.3.6.3 ) desertes Spent Fuel Pool Island and long term plant load electrical distrtudon systems. 3.3.6.1 Offsite Power
]
Offsite power is supplied to Maine Yankee via a single 115 kV transmission line. This line is connected to the New England 345 kV power grid at the Mason and Surounoc Stations and other 115 kV lines in Topsham. This line provides two power feeds to Maine Yankee, through X-16 and X-5. ] The feed to X-16 is through a fused deconnect switch. X-16 steps the 115 kV down to 4160 volts
]
and it is connected to two 2.5 MVA stepdown transformers, X-16A and X-168. X-16A steps the 4160 ] volts down to 480 volts to provide power to the SFPI loads. X-168 steps the 4160 volts down to 480 ] volts to power the remairung loads required to remain functional during the desviii ;cning . The ] feed to X-5 is through a fused disconnect switch. X-5 is used as a source of construction power. Figure 3.3.23 provides line arrangement. The omute power system is a reliable source of electrical power for plant equipment. In the defueled condition, no active systems meet the critons for safety related systems or miWents as the consequences of accidents are significantly lower than the limits of 10CFR100.11. In the event of an intenuption in power, the robust design of the passive systems assure the continued safe storage of fuel. I Rev.17 DSAR 3-100
- 5 4 2 = I 3 e i
\"<- .- x I
i i -
. , . - c w u C
.~ i.
- s :
- : l
- : i
~
s ! !
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- e : :
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- 2 : -
- . :. I i /-< i. .
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_x
- FIGURE 3.3-23 115kV transmission Rev. 17 1
. i
MYAPC O 3.3.6.2 Station Onsite Power
]
3.3.6.2.1 Auxillery PowerSystems
]
If the offsile power sourts is lost, a standby diesel generator may be used to power desined plant loads (Figure 3.3-24).
]
]
]
]
]
1 3
]
1 4ao voit System There are two 480 voit bus sections, one dedicated to the SFPI and the other for balance of plant ] load. Normal supply to each of these sections is through individual 4160/480 volt, of immersed self- ] O cooled, transibrmers connected on the high voltage side to the 4.16 kV bus. Refer to Section 3.3.6.3 foradditional detail.
]
]
j 1 3.3.6.2.2 Lighting and Heat Tracing System
]
Ughting Svstam Normal lighting for the Control Room, Spent Fuel Building, and Administration (WART) thm1 ding is ] supplied from the 480V electric distribution system through single-phase 480V/120-240V dry-type ] transformers. Battery back-up lighting is provided in vital areas for access, egress, or at pieces of ] equipment which may need to be operated during emergency conditions.
] j Hast Tracing System I q
l Each of the lines containing water (i.e.PWST) subjected to freezing conditions is heat traced.
]
O Rev.17 DSAR 3-102
MYAPC 3.3.6.2.3 Commumcadons Systems ] Normal and emergency communication systems are'desenbod in the Maine Yankee Emergency Plan. These systems include those systems required to contact extemal emergency management agencies, oscials and other govemment entities, and the general public. 3.3.6.2.4 Power and Control Cables ] cable Sizing and Radng AI cables are sized to operate within their normal rating and temperature rise by using conservative margins with respect to their cummt carrying capacities and insuistion properties. Cable insulation was selected with due consideration to the radiation, temperature rise and humidity conditions. The 480 voit power cables are insulated for at least 600 voit service and are suitabis for use indoors and,auldoors in wet and dry locations. Single conductor cables are Jacketed, while three<:onductor cables are triplexed orJN All control cables are rated for at least 600 voit service. ] Cable Routing Cables are separated when routed as medium voltage, low voltage, control, and instrument cables, when pi 1 l. Cable separation within the same group, but for redundant equipment or separate protective instrument channels is not required in the permanently defueled condition. There is no requirement for redundant electrical equipment. Cables are identified with their designation at regular intervals along their route and at raceway transitions Cable Susmotibility to Fire By routing cables in such a way as to avoid combustible materials, as wen as carefully sizing and placing them in their trays, the chance of an electncal fire developing is minimized. Rev.17 DSAR 3-103
MYAPC O Control cables will not ignits from overioading or grounds, since the maximum fault is insufficient to heat the insulation to the liash point. Fire in control cables can, therefore, only occur as the result of another fire. Good W :':::$g and sufficient care in cable routing reduces to a mirumum the chanons of extemal fires. Power cables can carry sumcient fault current to reach the flash point of the cable insulation: however, protective relaying on the =rJc circuits will respond to fault currents and open the cin:uit before enough heat has occurred to damage the cable insulation and start a fire.
}
O I 1 I O Rev.17 DSAR l 3-104 i
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MYAPC O 3.3.6.3 Spent Fuel Pool Island and Balance of Plant Electrical Distribution System ] The Spent Fuel Pool Island and Balance-of-Plant electrical distribution system (SFP/ BOP) pnwides power fbr the equipment to support the Spent Fuel Pooling Island operation and plant buildings and systems which wiu roman operationalduring the t:, ... ' ' -e.g process. The SFP/ BOP electrical system is suppiled by the 6.3 MVA,4 kV tertiary winding of the original Maine Yankee station service transibrmer X-16, shown in Figure 3.3-24, from the 115 kV system described in section 3.3.6.1. ] The SFP/ BOP electrical system is comprised of the following equipment: Station Service transformer X-16 and transformer prve is pronded by line fuses installed on the 115 kV dead-end tower ] adjacent to X-16. The 4 kV tertiary winding is used to power two outdoor 480 V unit, substations Bus-SFP and Bus-BOP. Each unit substation is equipped with a set of 400 ampere fused disconnect switches, and a 2500 kVA,4160 - 480/277 V oil-filled tir.r.fviir.er. The Staff Building is powered ] directly from the BUS-BOP fused disconnect at 4160 volts.
]
The unit substations are located on a concrets slab just to the west of transformer X-16. 480 V Bus-SFP The unit substation Bus-SFP is equipped with two electrically operated and three manually operated breakers. The substation is used to supply the following loads: MCC-SFP1 Fuel Building MCC (previously MCC-11B) MCC-SFP2 Fuel Building MCC (new) MCC-SFP3 Office Building MCC (pnwicusly MCC-138)
- 345 kV Switchyard Relay House 480 V Bus-BOP The unit substation for Bus-BOP is equipped with two electrically operated and five manually operated breakers. The substation issued to supply the following loads: ]
]
. WART Building ]
e info Center ]
. MCC 128 (Fire Pump House) ]
. P4 (Fire Pumps) ]
]
O Rev.17 DSAR 3-106
r MYAPC Standhv Power System A nominal 250 kWm standby diesel generator is provided which can provide power to limited SFP ] leiend loads. The diesel engine is serviced by a 1500 gallon fuel tank mounted on the trailor and is
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maintained wnh a 24 hour nominal fuel supply at fbH loed. The desel generator is equipped with a ] plug stored on the diesel traBer that is hard wired at the diesel and rminocts to a receptado located l at Bus-SFP. The diesel is designed to be manuaNy started and loaded to re-establish SFP island power. 1 The diesel breaker on bus-SFP is equipped with a mechanical intertock which wiu prevent accidental closure with the bus normal supply breaker closed. The diesel generator unit is provided with 480 V ac supplies which is used to maintain the diesel ] bettery charger, engine compartment hester, engine block bester, Ighong, etc. Provisions have been ] made to ensure that any diesel fuel leakage is contained on/In the trailer. These include a double woued fuel tank and a bermed area around the diesel engine and fuel line. Uninterruptible Power Sunolv Unintenuptible Power Suppues (UPSs) are provided to support continuous operation for the SFP island Programmable Logic Controis (PLCs) and security system. The UPSc are rated for four hour operation. This wiu provide adequate time to reestablish ac power from the stand-by diesel generator. 120 VAC Distroudian Svstam and Ilahtina 120 VAC power is provided using original and new lighting panels. Emergency lighting for the Fuel ] Building is provided by self<:ontained, battery-powered emergency lighting units. Raceway and Cable Svatem Since the equipment, raceways and cables assocated with the SFP island are intended to be operational through most of the decommissioning activities, a means of identification is essential. Rev.17
]
O -----'-<'.--===-<.ro<<-..<> > DSAR 3-107
O ""' Orange, day-glow orange, Iridescent brand, etc. paint, tape, tags, labels, etc. is used on electncal components. Raceways and/or supports are painted orange, and existing tray side rails and conduits are pointed orange at 3 to 5 foot intervals. Cables and equipment idendlication tags have an orange background. The load name plates fbr loads ==m with the SFP island are orange. The fbliowing cable service separation is maintaned in the raceways systems associated with the SFP island. ' Large 480 VAC power (cables requinng spacing) 480 VAC power,120 V and motor status crcuits by wul be installed in the same raceway without physical separation. Instrumentation circuits ; Control circuits Instrumentation. Contml and Alarms Motor status indication is provided locally and, in some cases, at the SFP island Control Room. Pump motor control is provided locally with remote stop capability also provided in the SFP island Control Room. Electrical and Spent Fuel Pool cooling system parameters and alarms are provided in the SFP island l Control Room. Idag Treina
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Heat tracing panel HT4FP-A and its tron.fvin ser, located in the Spent Fuel Building, is powered from ] MCC-SFP2 to support SFPI heat tracing loads. A common heat tracing trouble alarm is provided in the SFPI Control Room. The heat tracing on the Spent Fuel Pool Island diesel fuel oil !ines and the
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Decay Heat Removal (DHR) structure drain have local trouble indication.
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3.3.6.4 Programmatic Logic Controller (PLC) ] The PLC contains a microprocessor based controner which accepts equipment status information, operating parameter data, and instrumentation outputs. Information is provided to the PLC and is compared with set operating limits used in controNing equipment automatic functions. The plant operatons monitor the PLC through an Operator Work station and have the capability to operate the plant equipment through the PLC. O Rev.17 DSAR 3-108
1 MYAPC The Programmable Logic Controller (PLC) monitors effluents from the Spent Fuel Pool Island (SFPI), operating equipment of the SFPI, MET tower data, and performs some limited control functions of the equipment. The data is prcmded to a Operator Workstation in the New Control Room (NCR) via a serial connection from the PLC and to the Local Area Network (LAN) for display and control on workstations. Additional displays are provided in the Techrucal Support Center (TSC) and any other workstation connected to the LAN. The same data display screen appear on al workstations; however, only the Operator workstation in the new control room has the capability to provide equipment control. . The physical components of the system are the: 1 The PLC is located in the new control room. The PLC is designed to be able to operate vnthout operator interface. The PLC also communicates with the field devices in this system and has an interface to receive the MET tower data. 2 2-1/0 cabinets located in the spent fuel pool area. These cabinets contain the system interfaces for input and output signals in the Fuel Building.
- 3. 2-PowerTrac modules in the SFPI switchgear and 1-PowerTrac module in the BOP Switch ]
Room for monstoring the electrical parameters. ]
- 4. Operator Station located in the new control room. This PC is has control and server software wtuch interface with the PLC to provide control and indication of the SFPI equipment and to serve the LAN for display to the other PCS.
- 5. A modem at the new Control Room connected to the PLC and a modem at the Montsweag ]
Reservoir Pump House connected to a Micro PLC. This provides control and supervisory ] alarm for the fire pond fill pumps. ]
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- 6. 2-Genius 1/O modules at the Fire Pump House for inputing status of the fire water system. ]
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- 7. 2-Genius 1/O modules at the New Plant Ventilation Stack Sampling Skid for inputing the PVS ]
system status. ] Rev.17 DSAR 3-109 1
e MYAPC O Sadion 13 - References
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1 usNRC t.atter to MYAPC dated March 16,1999, Amendment No.162
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4 I O Rev.17 D&AR 3-110
MYAPC 3.4 Controinf Heavy Loads The controis for the handling of heavy loads at Mane Yankee satisfy the guidelines of NUREG 0612
" Control of Heavy Loads at Nuclear Power Plants" as documented in References 2 through 7 (Phase 1). In Reference 9, the NRC concluded that Phase 11 actions submitted in Reistence 8 were not required to reduce risks associmbed with the handung of heavy loads. While not a requirement, the i NRC encouraged the impiamentadon of any actions identified in Phase il regarding the handng of heavy loads considered appropnote. Some of the controls and actions described in References 1 through 9 remain applicable to the control of heavy loads at Maine Yankee with the reactor permanently shutdown and defueled. '
The design of load handung eqtspment at Mene Yankee complies with appucable industry standards and codes. Administrative controis restnct handung of heavy loads in the vicinity of equipment important to the defueled condition. The design of the hoists and cranes coupled with the admmistrative controis provide assurance that a heavy load will not be dropped which cousd result in damage to equipment i.Tv t.4 to the defusied condition. Load handling equipment that was considered under the NUREG-0612 review that remains applicable to the defusied condition includes:
- fuel building yard crane (CR-3)
- fuel building crane (CR-6) a mobile platform-hoist (CR-9) 1
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These load handling systems, with the exception of the mobile platform-hoist as described below, were evaluated against the seven gtadolines established by the NRC in NUREG-0612. These seven guidelines consisted of:
- 1. Safe Load Paths
- 2. Load Handling Procedures
- 3. Crane OperatorTraining
- 4. Special Lifting Devices
- 5. Lifting Devices (Not Specifically Designed)
- 6. Cranes (Inspection, Testing, and Maintenance)
- 7. Crane Design.
The two specal lifting devices identified by Maine Yankee for the reactor vessel head and the reactor l intomais are no longer important to the defueled condition. The conclusions and requirements resulting from the evaluation of the remaining guidelines (1 through 3 and 5 through 7) remain \ Rev.17 DSAR 3-111
MYAPC applicable for the load handhng systems specified above (CR-9, CR-3 and CR4), as described in References 2 through 7.
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]
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Fuel Buildinp Yard Crans fCR-3)- The travel path of the fuel building yard crane is depicted in Reference 4. Also shown is the spent fuel cask lay down area in the spent fuel pool. In 1975 the Commission reviewed Maine Yankse's analysis of a postulated spent fuel cask drop accident in the spent fusi pool and concurred wah our evaluston that no safety related equipment was beneath the path for cask travel. Additionsuy, CR-3, which would be used in handling spent fuel casks, was modified to improve reliability, including the addition of limit switches and a main hoist equalizer sheave assembly to provide overtoed sensing of main hoist hook loads. The Rmit switches were installed to prevent movement of any load over spent fuel in the pool. The Commission concurred with Maine Yankee's actions and found that prod.;siw to prevent a postulated spent fuel shipping cask accident are acceptable. Nonetheless, spent fuel shipping casks shall not be lifted over the spent fuel storage pool. In addalon, spent fuel storage racks may be moved only in accordance with procedures which ensure that no rack modules are moved over fuel assemblies. O Fuel Building Crane (CR4)-The travel path of the fuel building crane, the path of the fuel pool coohng pump power cables and the entical volume associated with the fuel pool cooling system are depicted in Referena 4. The fuel pool cooling pump power cables are routed beneath a structural guder and are thus protected from a load drop. The fuel pool cooling system is entirely beneath the new fuel storage area, separated from the fuel building crane by one or rnore floors. The critical area l Is cieerfy marked using safety striping and waming signs and the handling of loads within this area is administratively controlled. i unhiin Platform-hai=+ (CR-9)-The rnobile platform and hoist is used to move fuel assemblies and is operated over the spent fuel pool and on rail extensions to the north of the SFP. This two ton rated hoist generally does not lift loads greater than the weight of a new fuel assembly with handling tools. CR-9 has design features which assure that of etiva of the mobile platform-hoist can be conducted with very little probability of a load drop. Two upper limit switches of different types prevent two-blocking from occurring. Two load holding brakes of different types, each rated at 150% of the hoisting motor's full load torque, assures that the load can be dependably retained. The platform superstructure was designed to withstand an earthquake of magnitude .10g honzontal and .06 vertical while under rated load without loss of structural integrity or function. The hoist is also seismically qualified. Interiocks prevent trolley movement while the hoist is being operated. Rev.17 , DSAR 3-112
MYAPC Unless completed within the last eighteen months, a complete dux:kout of CR-9, induding a load test { and a functional test of the refueling system interlocks, shall be conducted prior to using CR-9 to handle irradiated fuel assembhos. l O Rev.14 O DSAR 3-113
MYAPC Section 3.4 -
References:
- 1. USNRC Letter to MYAPC dated December 22,1980 Control of Heavy Loads
- 2. MYAPC Letter to USNRC cated September 18,1981 (FMY-81-141) Control of Heavy Loads, Section 2.1 submittal.
- 3. MYAPC Letter to USNRC dated July 7,1992 (MN-82-131) Control of Heavy Loads
- 4. MYAPC Letter to USNRC dated August 27,1982 (MN-82-169) Control of Heavy Loads
- 5. MYAPC Letter to USNRC dated December 7,1982 (MN-82-242) Control of Heavy Loads
- 6. MYAPC Letter to USNRC dated October 6,1983 (MN-63-221) Control of Heavy Loads, Phase i Report
- 7. USNRC Letter to MYAPC dated December 30,1983 Control of Heavy Loads, Safety Evaluation Report
- 8. MYAPCO Letter to USNRC dated March 19,1984 (MN-84-22) Control of Heavy Loads, Phase 11 Report
- 9. USNRC Letter to MYAPC dated June 28,1985 Completion of Phase 11 of " Control of Heavy Loads at Nuclear Power Plants" NUREG-0612 (Generic Letter 85-11)
Rev.14 O DSAR 3-114
F MYAPC Control and Area Shielding Shielding is provided for the main control room to ensure an adequate control room environment i is established. Refer to Sechon 3.0 for a discussion of the control room ventilation.
- - Yard Shielding The boron waste storage tanks are shielded for personnel pici.ct;cn.
L 4.2.2 Health Physics i Personnel requiring unescorted access to radiologically controlled areas are given training in Radiological Health and Safety. Personnel requiring escorted access to radiologically controlled areas are given training commensurate with potenbal radiologmal health protection problems in the . radiologically controlled areas to be frequented. Administrative controls are established to assure that procedures and requirements relating to radiation protection are followed by station personnel. These procedures include a radiation work permt system. Work on systems or in locations where
- exposure to radiation or radioactive materials is W to occur requires an appropnate radiation l
! work permit before work can begin. The radiological hazards associated with the job are determined ! and evaluated prior to issuing the permit. Personnel Monitoring Eauioment
- . PersonnelinviJiciing equipment, consisting of TLD badges and electronic alarming dosimeters, is I issued to and wom by personnel within the Radiologically Controlled Area. Personnel monitoring ,
equipment is also available on a day-to-day basis for visitors not assigned to the station who have ) occasion to enter the Radiologically Controlled Areas. Records of complete radiation exposure history are obtained for the current year for individuals prior to allowing those individuals to exceed 10 percent of the applicable limit in one year. Personnel Contamination Control Anti-cent.ni; nation clothing is wom as necessary to protect personnel from radioactive contaminabon. Respiratory protechon may be used to keep the total effective dose equivalent Al. ARA. Rev.17 DSAR 4-3 l
MYAPC Change Area Facilities A change area is provided where personnel may obtain clean anti-contamination clothing required for station work. Monitoring equipment is also provided to aid in the decontamination of personnel. l Temporary change areas are piW,ded when required. - Decontammatiop Facilities 1 An equipment decontammation facility will be provided if required to support decommissioning. l 4 Access Control l To prevent inadvertent access to high radiation areas and very high radiation areas, waming signs, visualindicators, barricades and locked doors are used as necessary. Administrative procedures , are written to control access to these areas. Laboratory Facilities The station includes a ietu.iuiy with adequate facilities and equipment for detecting, analyzing and measuring radioactivity, and for evaluating any radiological problem that may be anticipated. Counting equipment, such as a multi-channel analyzer and liquid scintillabon counters are provided in an appropriately designed counting room. Environmental sample analyses are conducted by outside laboratories. Health Physics Instrumentation Portable radiabon survey instruments are provided for use by trained personnel. A sufficient number of instruments for detecting and measuring radiation are available. Monitoring instruments and count rate malers are located at exits from the radiologically controlled areas. These instruments are intended to assist operating personnel in preventing contaminston from being spread into clean areas. Monitonng instruments are available at various locations within the Radiologically Controlled
' Area for contamination control purposes. Portable radiation monitoring instruments are available to measure radiation dose rates and airbome concentration within the control room. The station has permanently installed remote radiation and radioactivity monitoring equipment in various locations, as described in Section 4.6.
Rev.17 mm ==v and Medical Proureins DSAR 4-4
p. 1 MYAPC 4.4.2 System Description The liquid waste system is divided into two subsystems: hydrogenated and aerated liquid waste. 1
-The hydrogenated liquid waste system is part of the boron recovery system and is not used in the permanently defueled condition. With the decontammated evolullons completed, the Primary Drain , Tank (PDT) no longer serves a system functional purpose. However, mention of the Primary Drain Tank is retained for recognition of its basis in the DSAR Sechon 5.6 for projected gaseous radioactivity release (Table 5.6.1) and projected doses (Table 5.6.2) from a PDT rupture.
Uquid waste is collected in a waste hoidup tank prior to pic-1::' ,g. Liquid waste is held in the l waste holdup tank for processing through a domineralizer using an ion exchange process. l In the waste water domineralization system, the water is processed through a series of vessels which use a combination of filtration and ion exchange processes. When full, the test tank is l sampled and then its contents are pumped to the forebay. Depleted waste water domineralization l system filtration and ion exchange media is transferred either directly to a high integrity container or to a resin storage tank for storage and eventual shipment off-site. 4.4.3 Design Evaluation The Waste Holdup Tank has a capacity of 5,835 gallons. The waste water dommeralization system l capacities are sufficient to process the liquid waste at the average rates expected during l decommissioning activities. t The effluents from the liquid waste system are discharged to the test tank. The test tank will be j sampled and analyzed for radioactivity prior to release. If the activity levels of the test tank exceed the allowable level stated in the ODCM, the liquid may be retained for holdup or further processing. Effluents are discharged to the forebay through a trip valve which will close on high activity from l the liquid waste disposal monitor. I i Rev.17 O DSAR 4-9
MYAPC TABLE 4.4.1 LIQUID WASTE TREATMENT COMPONENT DESCRIPTION l Duratek Booster Pump P-DT-1 Design Pressure, psig 150 Design Temperature 200 Capacity, gpm 225 i Material Bronze Code None Duratek Vessels No.1 through 5 Design Pressure, psig 150 Design Temperature 200 Volume, ft8 No.1 50 No. 2-5 30 Material O Code 316 SS ASME Vill Waste Holdup Tank: Tank 109 l Design Pressure 4 psig l Design Temperature 150 psig l Capacity 5835 gal l Material 304 SS l l l Rev.17 O DSAR 4-10
l j l MYAPC t] 4.5 - SOLID WASTE TREATMENT 4.5.1 Design Basis The solid waste treatment system is designed to store, process, monitor and dispose of solid radioactive wastes from the plant. The principal design objective is to insure that the general public is protected from exposure to radioactive waste products in accordance with 10CFR20. The normal { sources of radioactive wastes are activated corrosion products and fission products generated I during previous plant operabon, wastes generated from maintaining spent fuel pool water chemistry, and wastes generated during decommissioning activities. 1 4.5.2 System Description { 4.5.2.1 Filter Handling l The fuel pool system filters are removed from service wMn the pressure drop across the filters becomes excessive or when the radiation level exceeds a predetermmed level. The expended filter ( cartridges are moved to a shipping cask filter container or to storage in the underground RCA l
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storage bunker or the low level waste storage building. Small, low activity filter cartridges are l l packaged for ultimate disposalin approved containers. In each case, the procedure conforms to l DOT regulations for shippog to an NRC approved disposal site, i l 4.5.2.2 Solid Wastes l Noncompressible solid wastes such as contaminated metallic materials and highly contaminated solid objects are placed inside approved shipping containers and stored until they are disposed of at an NRC approved disposal site. 4.5.3 Design Evaluation The solid waste treatment system provides adequate handling and storage capabilities for continued spent fuel pool operation and to support decommissioning activities, as required. Rev.17 U.A DSAR 4-11
MYAPC O 4.6 RADIATION MONITORING SYSTEMS 1he radiation monitoring system prtwides a means of continuously and remotely montoring radation levels at key locations in the permanently defueled condition. The systems consist of permanently installed morntoring devices with a program for portable monitors, sampling, and data analyses. The Fuel Building exhaust fans and the Fuel Building monitoring system are fed from the SFP power l system with a stand by power supply. The Primary Vent Stack exhaust fans and sampling system l are fed from the BOP Power System. The ODCM contains the administrative controls and Emits l associated with the radiation montors. When a radiation monitor is inoperable, rnonitonng is performed by collecting and analyzing grab samples. 4.6.1 Design Bases The objechves of the radiation monitoring system include: l Provide an earty waming of a plant malfunction resulting in increased radiation levels. O -
=aa' tar r oia ciiv ei ca re exceed spected linuts.
ia ta aviraa- at ao ravio a -iao ir coac air aoa- ,
. Wam pomonnel ofincreasing radiation which could result in a radiation health hazard.
4.6.2 - System Description 4.6.2.1 Primary Vent Stack Sampling l The primary vent stack sampling system withdraws a continuous, isokinetic flow sample of effluent l from the pnmary vent stack, and routes the flow through a sampling skid. This skid allows the flow l l to be continuously sampled for particulates, and is periodically sampled for tritium gas, and the l l sample flow is then discharged to the primary vent stack. The skid also allows for attemate l sampling from by-pass connections in the event of equipment malfunction. ] O Rev.17 DSAR 4-12
i l MYAPC ! 4.6.2.2 Process Monitoring System l The following requirements for radiation process and effluent monitors have been relocated from l Technical Specifications by Proposed Change PC-207: Instrument Operation and Source Checks: l
- a. Dailv* Check: Intemal test signals used to sock instrument operation. The Liquid Waste l Effluent Monitor performs a self-diagnostic check without operator action. l l j
- b. Quarterly
- Functional Test Expose the detector with either an intomal or an extemal j radiation source or an electronic signal to verify instrument operabon.
I
- c. 18 Month Calibration: Spannes to known radiation source. l
- When required to be operable I i
4.6.2.3 Fuel Building Ventilation Exhaust Radiation Monitoring Skid l , I This equipment takes a continuous isokinetic air sample from the Fuel Building ventilation exhaust l duct and, using a vacuum pump draws it through a sealed system to particulate cartridge, a i continuous gas monitor, a tritium collodor system, and a automatic flow control as identified below. The sample is then discharged inside the Fuel Building. The fuel building process radiabon monitoring system provides ear 1y waming of a plant malfunction; l woms personnel of increasing radiation which could result in a radiation health hazard; channels l include locally operated check sources, readout and alarm in the spent fuel pool control room and l are recorded by the PLC-SFP-01. The PLC indication consists of a log count rate meter and an j alarm. All channels also include readout and alarm at the detector station. l Particulata Cartridge A cartridge assembly, which collects a particulate sample from the isokinetic sample flow, is periodically removed and checked for activity. Continuous Gas Monitor A beta scintillation detector mounted within a fixed volume shielded chamber continuously measures the ventilation exhaust gas sample activity. This measurement is used in off-site dose projections performed per the ODCM. Rev.17 DSAR 4-13 l
l MYAPC l Tntium Collector Subsystem The tritium maar*w system draws a gas sample from the main sampling system using a separate vacuum pump, and then sends it through a tritium collector cartndge, and a manual flow indicating controller. The sample is retumed back to the main sampling system, downstream of the tritium maar*r system sample point. The tritium c.di@ is removed and checked for tritium on a periodic { basis. I Automatic Flow Controner The automatic flow controller adjusts the radiation monitoring skid sample system flowrote to adjust for changes in the ventilation exhaust duct flowrate. A flow sensor installed in the duct sends a flow signal to the radiation monitoring skid electronics which, in tum, sends a control signal to the automatic flow controller to set the required flowrate. A sample flow measurement device integral to the flow controller, feeds back a flow signal to the radiation monitoring skid electronics, 4.6.2.4 Liquid Waste Disposal System l l This channel continuously monitors test tank effluent during releases by means of a detector l mounted in an off-line liquid sampler. The system has a control function that causes the waste l emuent flow valve to close in the event of high activity. This detector is located downstream of the l last possible point of radioactivity addition. Although this monitor performs and provides information l similar to the other process monitorn, there are some minor differences. The check source funcbon l la performed at the monitor skid. There is no visual indication of the source check counts. 1 Annunciation is provided should the unit fail to detect property. The unit also performs self . l diagnostics. As with the source check function, there is no indication unless the diagnostics fail. l Should either of these malfunchons occur, the unit will provide annunciation. Readout and alarm l indications are localindications. l 4.6.2.5 Area Radiation Monitoring System l l The area radiation monitoring system measures, indicates, and annunciates radiation and radioactivity in the Fuel Building. This system wams personnel of increasing radiation and radioactivity and provides early waming of fuel pool inventory loss. The following surveillance requirements for area radiation monitors have been relocated from ; Technical Specifications by Proposed Change PC-207: O '- Ta roa i a =="dia" a a ti d a#ri"a -av - "t =< irr di
- e r" ' *it"i" ta a "i Rev.17 DSAR 4-14
MYAPC fuel storage building: Radiation levels in the spent fuel storage building shall be monitored continuously.
- 2. Surveillances
) a. Daily
- Check: Intemal test signals used to check instrument operation.
l
- b. Quarterly
- Functional Test Expose the detector with either an infomal or an extemal radiation source or an electronic signal to venfy instrument operation.
- c. 18 Month Calibration: Exposure to known radiation source.
- When required to be operable i
The monitor channels display and alarms in the Spent Fuel Pool control room, is recorded by the l the PLC-SFP-01 and has local readout and alarm. Adjustments of alarm points, high voltage and l other variables are made locally for the Spent Fuel Pool monitor. The channel is equipped with a l calibration check source operated locally for the Spent Fuel Pool monitor. 4.6.3 Design Evaiustion O The process and area radiation monitoring systems wR ensure that releases, planned or accidental will be identified and monitored. The radiation monitoring systems will provide alarms to prevent overexposure of personnel due to unexpected high radiation and/or contamination levels. These monitors are listed on Table 4.6.1. 2 O Rev.17 CSAR 4-15
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C\ MYAPC V SECTION 5.0 ACCIDENT ANALYSIS TABLE OF CONTENTS Section D[g Eagt 5.1 In trod uction . . . .. .. . . . . . . . . . . . .. . . . . . . . . . . . . . .. . . . . .. .... ... . . .. .. . ... ... ...... ........ 5-1 .. .. ... 5.2 Soant Fuel Critimaity Analvses .......................... ..... ............. ....... 5-2 5.2.1 Misplaced Assembly 5.2.2 Dropped Assembly 5.2.3 Assembly Adjacent to the Racks. 5.2.4 Assembly in the Comer of the Racks 5.2.5 Boron Dilution 5.3 Fuel Hand fina Accid ent .. .... ......... ...................... .... ............ .......... 5-9 5.4 Soant Fuel mek D roo .. .......... ........................ . ... ......... ........ 5-16 5.5 Soent Fuel Pool Accidents ................................ ...... ............ ......... 5-17 5.5.1 Loss of Spent Fuel Pool Cooling 5.5.1.1 Blocked / improper Cell Flow 5.5.1.2 Loss of Forced Flow 5.5.1.3 Loss of Heat Sink 5.5.2 Loss of Spent Fuel Pool Inventory 5.6 Low Level Waste Retease Incident ...................... ... ...................... 5-37 5.6.1 Radioactive Waste Gas System Leaks and Failures 5.6.2 Radioactive Uquid Waste System Leaks and Failures 5.6.3 Low Level Waste Storage Building Accident s Accendices Appendix SA Summary of Parameters Used for Evaluating the Radiological Effects of Accidents. ....... . .................. 5.A-1 Appendix SB Atmospheric Transport and Diffusion ! Characteristics for Accident Analysis ...... ....... ........... 5.B-1 Rev.14 l I DSAR 5-i l
MYAPC SECTION 5.0 LIST OF TABLES Inhia_No. Ela 5.2.1 Spent Fuel Pool Criticality Accident Analysis Results 5.3.1 Bounding inventones for Radiological Analyses 5.3.2 Maine Yankee Post Shutdown Fuel Handling Accident EAB Doses 5.3.3 Maine Yankee Post Shutdown Fuel Handling Accident CR Doses as of December 6,1997 5.5.1 Blocked Assembly Cooling Analyses Results 5.5.2 Calculated Spent Fuel Pool Time to Boli 5.5.3 Projected Radiological Dose Consequences of Boiling in the Spent Fuel Pool 5.6.1 Deleted
]
5.6.2 Projected Radiological Doses from Primary Drain Tank Rupture 5.A.1 Equilibrium Core Fission Product inventories 5.B.1 Dilution Factors for Accident Analysis at Maine Yankee O Rev.17 O DSAR 5-il
O MYAPC SECTION 5.0 UST OF FIGURES Figum Nct Iltlg 5.5-1 Spent Fuel Pool Boilof Rate ' 5.5-2 Loss in Spent Fuel Pool Level as a Function of Tirne After Shutdown (Bodo# Losses Only) 5.5-3 Projected Radiation Fields at 25 Feet Above Active Fuel as a Function of Water Depth and Decay Time 5.5-4 Projected Skyshine Radiation Fields at Various Hortzcatal Distances fnxn the Center of the Spent Fuel Pool as a Function of Water Depth at 1 Year After Pfant Shutdown O i Rev.14 O DSAR 5-iil
MYAPC SECTION 5.0 ACCIDENT ANALYSIS 5.1 .i.....l.,..i Earlier sections of this report describe the major systems and cGnponents of the plant from the perspective of safe spent fuel handling, spent fuel storage, and other decommissioning activities as would be appropriate to a permanently defueled plant. This section uses the previous infunretion and examines the potential consequences of accdonts and incidents, notwithstanding the precautions taken to prevent their happening, to assess the adequacy of the plant design in minimizing or mitigating potential consequences of such l occurrences. Additionally, the accident analyses presented in this section provide assurance that the health and safety of the pubhc is protected from the consequences of even the most severe of the hypothetical incidents analyzed. '- With the permanent defueling of the Maine Yankee facility and the certification of the ra-Han of operations, the postulated accidents assoanted with reactor operation are no longer applicable and O need not be considered. Likewise, the unirradiated nuclearfuel has been removed from the Maine ] Yankee site and therefore accidents involving new fuel assemblies are also no longer applicable. However, those accdonts associated with the storage or handling of irradiated fuel or radioactive waste storage or procesang remain applicable and are discussed within this section. The general classificaton of accidents for the permanently defueled condition are limited. These groupings are !Med as follows:
- 1. Inadvertent criticality of the stored spent fuel,
- 2. Fuel assembly handling accident,
- 3. Spent fuel shipping cask drop in the spent fuel pool.
- 4. Loss of spent fuel decay heat removal capability,
- 5. Loss of spent fuel poolinventory,
- 6. Radioactive release from a subsystem or component, or
- 7. Low level waste storage accident.
The evaluation of these accidents is based on a conservative set ofinitial conditions and analytical methodologies. The key selected initial conditions are from worst case operating conditions Rev.15 DSAR 5-1
MYAPC established by consderation of regulatory limits, Technical Specification requirements (limits), or administrative controls. The key initial condsbons and analysis assumptions are identified in the appropriate section for each acodont analysis. Analysis methodologies and related information associated with the radiological analyses is provded in the Appendices to this section. 5.2 Spent Fuel Criticality Analysee The family of spent fuel criticality analyses applicable to the defueled condition of the plant was p.Jv ic i.d for the third Maine Yankee spent fuel pool raracking in 1992. These calculations, with 4 the exception of the spent fuel pool boren dilution analysis, were used to justify the installation and use of the high density spent fuel racks cummtly residing, and in use, in the spent fuel pool. The analyses were initially described and presented to the NRC in Reference 1. The analyses were accepted by the NRC in Reference 2. The minimum spent fuel pool temperature applicable to ] these analyses is a=====ad to be 40'F in Reference 3.
]
The design of the spent fuel racks is provided in Reference 1. The spent fuel is stored in a single tier rectilinear array of free standing modules consisting of groupings of cells with similar design characteristics and parameters. Each fuel assembly is stored within an individual cell. The I differing designs of the modules are identified as either Region I or Region 11 racks. Region I racks are designed to as,v,TTivdate the more reactive fuel; Region 11 accommodates the less reactive fuel. Technical Specification condebons define the placement of fuel into the differing regions. The applicable regulations, guides, and standards pertaining to the design and assessment of criticality safety for spent fuel storage include the following:
- 1. General Design Catenon 66 - Prevention of Criticality in Fuel Storage and Handling. Maine Yankee is licensed to the Interim Design Criteria issued in 1967.
- 2. NUREG-0800, USNRC Standard Review Plan, Section 9.1.2, Spent Fuel Storage and Section 9.1.1, New Fuel Storage.
- 3. ANSI /ANS 57.2-1983, Design Requirements for Spent Fuel Storage Facilities at Nuclear Power Plants, Section 6.4.2.
- 4. ANSI /ANS 57.3-1983, Design Requirements for New Fuel Storage Facilities at LWR Plants, Sechon 6.2.4.
These guides and standards require that for spent fuel racks, the maximum calculated K,, , including margin for uncertainty in calculational methods and mechanical tolerances, be less than O DSAR 5-2 Rev.17
MYAPC or equal to 0.95 with a 95% probability at a 95% confidence level. The criticality analysis of the high density spent fuel storage racks demonstrates that this criteria is satisfied. i in order to assure that the true reactivity will always be less than the calculated reactivity, the l followng conservative assumptions were made in calculating the enticality safety limits for spent
. fuel storage in the spent fuel pool:
- 1. The spent fuel pool water contains a minimal boron concentration well below the concentration administratively and operationally maintained at the Technical Specifications. See Table 5.2.1 for the equivalent boron concentration to maintain the spent fuel pool K,, s 0.95. Soluble boron is not credited in the design analysis of the racks for fuel storage under non-accident conditions.
- 2. The spent fuel pool is conservatively analyzed at a bulk water temperature of 40*F.
]
- 3. The Boral" fixed neutron absorber panel loadings in the spent fuel racks are at the minimum specified amount of 0.020 grams of BSper square centimeter.
- 4. The rack and pool structural matenais neutron absorption effect is neglected.
- 5. An infinite fuel array with no radial or axial neutron leakage is assumed.
O 6. All accdont calculations are performed with new, unirradiated fuel at an initial enrichment of 4.5 w/o U". The inadvertent placement of a Region I assembly into a Region 11 rack assumed the most reactive spent fuel assemblies stored in Region 11 racks.
- 7. Mechanical and calculational uncertainties were maximized in the conservative direction.
Analysis of the reactivity effects of fuel storage in Regions I and ll of the spent fuel pool was performed with the two dimensional, multigroup transport theory computer code, CASMO-3. Independent verification calculations were performed with a Monte Carlo technique using the l KENOV.a computer code package and the 123-group nuclide cross section library prepared with the NITAWL-S code. To minimize the statistical uncertainty of the KENOV.a calculations,1.2 million neutron histories in 2000 generations of 600 neutrons each were accumulated in each calculation. Exponence has shown that this number of histories is sufficient to assure convergence of the KENOV.a reactivity calculations. The PDQ-7 fine mesh diffusion theory code was used to analyze abnormal configurations associated with accident conditions. The postulated criticality accident situations are of the fuel handling variety. They consist of an assembly dropped horizontally on top of the racks, an assembly inadvertently placed adjacent to DSAR 5-3 Rev.17 l
MYAPC the sides or comer outsde of the racks and a Region I assembly inadvertently placed in a Region 11 rack. These abnormal situations were analyzed with PDQ-7 with cross section input from CASMO-3. In all cases, it was assumed that Region ll contained spent fuel at 4.5 w/o and a bumup of 30 GWD/MTU and that Region I contained all fresh unirradiated fuel at 4.5 w/o. 5.2.1 Mispiaced Assembly For the misplaced assembly, it is assumed that a Region I assembly is placed in a Region ll rack. As with the other ecc,w, it is assumed that the assembly is fresh, unirradiated fuel at 4.5 w/o and is surrounded by spent fuel of 4.5 w/o at 30 GWD/MTU. The results of this analysis, as presented in Table 5.2.1, subcate an accident K, value of less than 0.95 with a required minimum boron concentration of 241 ppm. 5.2.2 ' Dropped Assembly An assembly divw de on top of the spent fuel racks is assumed to be laying horizontally on top of O the racks. In this condition, the dropped assembly is more than 20 inches above the top of the active fuel in the spent fuel racks. However, for this analysis it is assumed that the assembly is j touching the top of the active fuel. It is further assumed that the dropped assembly is fresh, unirradiated fuel at 4.5 w/o. The dropped assembly analysis is performed for both Region I and Region 11 racks. The results of this analysis, as presented in Table 5.2.1 for both Region I and Region 11 racks, indicate an accident K, value of less than 0.95 with a required minimum boron concentration of 362 ppm. 5.2.3 Assembly Adjacent to the Racks it is possible to inadvertently place an assembly adjacent to the racks. The side of the Region ll storage cells along the outside of a spent fuel rack are not designed to contain a BORAL* sheet. l I It is assumed that a fresh fuel assembly of 4.5 w/o is placed at the side of the racks. Using these conditions and assumptions, the analysis is poderirred for both Region I and Region 11 racks. The results of this analysis, as presented in Table 5.2.1 for both Region I and Region 11 racks, indicate an accdont K, value of less than 0.95 with a required minimum boron concentration of 460 ppm. DSAR 5-4 Rev.14
r p MYAPC
%-.J 5.2.4 Assembly in the Comer of the Racks it is possible to inadvertently place an assembly adjacent to the comer on the outside of the racks.
The side of the Region il storage cells along the outside of a spent fuel rack does not contain a BORAL* sheet. It isassumed that the assembly is fresh, unirradiated fuel at 4.5 w/o. The analysis is perfomied for a Region I - Region ll comer and for a Region II - Region ll comer. These are the most limiting comer conditions that exist in the Maine Yankee spent fuel pool. The results of this analysis, as presented in Table 5.2.1 for both Region I and Region ll racks, indicate an accident K,,value of less than 0.95 with a required minimum boron concentration of 670 ppm.
]
5.2.5 Boron Dilution The Boron Dilution accident analyses associated with the permanently defueled condition of the Maine Yankee plant is based on the inadvertent addition of unborated water to the spent fuel pool. It is only intended that this sid,C demonstrate the magnitude of dilution required to approach the O crit!cality safety analysis limits from a given set ofinitial conditions. The results and conditions of this analysis should not used as indications of safety limits, Technical Specification interpretations, or operating restrictions. The acceptance enteria for this transient, in accordance with the initial conditions of the wd,2, provides assurance that the spent fuel pool boron concentration remains above the minimum required accident concentration of 670 ppm as derived for the preceding ] criticality analyses. Even though this analysis is performed using the criticality accident derived minimum required concentration of 670 ppm, the normal spent or unitradiated fuel storage in the pool is designed to ] maintain the K , s 0.95 for a boron concentration of 0 ppm (Reference 1 and 2). To assess the potential dilution of the spent fuel pool, a calculation was performed with the following initial conditions:
- 1. A spent fuel pool volume of 50,000 cubic feet of water is assumed. This volume corresponds to a initiating water depth of 32.5 feet of water. The nominal operational water depth in the spent fuel pool is 36.5 feet with a low level alarm at 35.5 feet.
O DSAR 5-5 Rev.17
MYAPC
- 2. One Galf of the assumed spent fuel pool operating volume is available for diiution.1his assumption conserva0vely accounts for the incomplete mixing of unborated and borated water.
- 3. A single spent fuel pool cooling pump is operational to asast in mocing the pure and borated water.
- 4. In order to avoid overflowing the spent fuel pool as the dilution occurs, it is conservatively assumed that an equal amount of borated water is draned from the pool concurrent with the addition of pure water. I
- 5. The K,, of the spent fuel pool remains s 0.95, as consistent with the conservatisms associated with the other w ;;;y analyses discussed in this f
section.
- 6. The spent fuel pool initial boron concentration is assumed to be 1800 ppm.
Note: This value is intended to represent initial pool conditions for the purposes of illustrating the effect of a spent fuel pool boron dilution event. This value does not represent a safety limit. The Boron Dilution analymisp%lates that a dilubon of unborated water in a volumeof 12,500 cubic feet (greater than 93,500 gallons) would be required to lower the pool boron concentration O- to the 800 ppm level. This amount of water addition co- gw-4s to the replacement of almost 8 feet of pool level of the existing borated water with unborated water. Based on the conservative nature of the Boron Dilution analysis, and the operational margins associated with the massive dilutions required to approach the minimum boron concentrations required under accident conditions, this analysis indicates that there is sufficient margin for operational intervention to recognize and terminate any credible boron dilution event. 1 O DSAR 5-6 Rev.14
MYAPC sacnon 5.2 Rainmnces
- 1. Laner, Maine Yankee to the NRC, Proposed Technical Spedlicadon Change No.177: Maine Yankee Spent Fuel Pool Roracidng", MN-934, dated January 25,1993.
. 2. Lamar, NRC to Maine Yankee,"leeuence of Amendment No.144 to Fac5ty Opendhg License No. DPR-36, Maine Yankee Atomic Power Station (TAC No. M85794)", NM-94-27, dated March 15,1994. I
- 3. Memorandum, G.M. Solen (DE&S) to R.P. Jordan (Maine Yankee), " Maine Yankee SFP ]
Evaluation for a Minimum Temperature of 40T', RP-MY-98-0009, dated March 5,1998. ] O J 4 O DSAR 5-7 Rev.17 l 1
\
MYAPC TABUE 5.2.1 SPENT FUEL POOL CRITICALITY ACCIDENT ANALYSIS RESULTS 1 .I
$$AmeMmmt$ MIAppEsmhis4* MCalamisend N'#Cadenhibid%#
- V$$ ** & fM@$E) $[MQ ?M
- EsaE : +@Fa@ 2%MGWai "W&s?@ WWM5Mi Dropped Region i Rads 0.9229 0.9862 318 M Regen 11 Racks 0.9104 0.9823 362 Misplaced Region i Racks 0.9309' 1.0230 460 Adjacent '
Region il Racks 0.9103 0.9934 422 As W Misplaced Region I and 0.9249 1.0427 607 Comer Region 11 Rad
^
. :::14 Comer ComerofTwo 0.9069 1.0261 670 }
Region ll Racks Misplaced Regen 1 0.9124 0.9460 241 Assiembly Assembly into Region il Racks (1) Non-accident base calculation of K, assuming 0 ppm Boron in the spent fuel pool. I (2) Accident base calculation of K, assuming 0 ppm Boron in the spent fuel pool. I (3) The relationship between the K, for accident conditons and the Equivalent Baron Concentration is dependant upon the irCg of the accident within the computer code suite and will not be the same between accident classifications. (4) The equivalent boron concentration is that concentration required to offset the nsactivity incnnsee for the accxient condulon analyzed, thus resulting in an accxient condition K, of less than 0.95 (0.93 or 0.91 (accident dependant) when uncertainties are taken into account). ]
\
DSAR 5-8 Rev.17
MYAPC 5.3 Fuel Handling Accidard The purpose of this section is to assess andcipstod spent fbel pool fbel handling operations in order to anive at the accident which would resut in w-Ci conservative off-site and control room radioactive release e5ects. Fuel handgng incidents e,'G ;;.4 in the containment are not appscoble to the permanendy defumied ocndition. Fuel handling operation W with the use of a spent fuel cask are addressed in section 5.4. The Ikeuhood of a fuel handling inddent in the spent fuel pool is minitruzad by impiamentation of appropriate and long standing administradve controis and physical limitations imposed on fuel handling operations. All fuel hancsing operations are conducted in accordance with prescribed . procedures under the direct surveigance and supervimon of quellflod personnel. The fuel handling equipment and facdity are demgned for the transfer and handling of a single fuel assembly at any time, and movement of equipment when heric;;.s the fuel is administradwely restricted. The fuel handling manipulators and hoists are designed so that fuel cannot be raised above a position which pnmdes adequate shield water depth for the safety of operadng personnel. In the spent fusi pool, the design of fuel storage racks and manipulator equipment, in conjuncdon
\ with appropriate administrative controls, is such that
- 1. Fuel is always mentaned by rnochenical restraint. Fuel at rest is positioned by positive restreints in a subcritical geometrical array, with no credit for boric acid in the water.
- 2. Fuel can be manipulated only one assembly at a time.
- 3. Violation of procedures by placing one fuel assembly in juxtaposition with any i' group of assemblies in racks wiu not result in enticality.
- 4. The spent fuel shipping contaner does not pass over spent fuel during transfer ]
operations. The fuel assembly is immersed continuously while in the spent fuel pool. Adequate cooling of fuel during underwater handling is provided by convective heat transfer to the surrounding water.1he fuel handling equipment and spent fuel pool are described in detail in Section 3. Inadvertent C::.---;--i rit of the fuel assembly from the fuel handling equipment is prevented by design, mechanical, and proceduralinteriocks. Consequently, the possibility of dropping and damaging a fuel assembly is unikely. The combmetion of these safeguards make the probability of a fuel handling inodont verylow. DSAR 5-9 Rev.15
MYAPC For the purpose of establishing an upper limit on the amount of fuel damage resulting from a fuel handling incident, it is assumed that the fuel assembly is dropped during handling. The number of ruptured fuel rods which would result depends on several variables induding the Idnotic energy at impact and fbel assembly onentation at impact. However, to assure that the most conservative ] mee is considered, it wil be assumed that al rods in the dropped assembly fail upon imped. In this case, any assion products esemping som the spent fuel pool would be avadable for reisese to the atmosphere via the fuel building venesman system. The evaluation of this accident, as appicable to the permanently defueled condition of Maine Yankee, is based on the release of the fuel rod gap inventory of the " worst" (highest bumup, highest initial ennchment, longest operating history) fuel assembly presently stored at Maine Yankee. The radionuclide gap inventories of the " worst case fuel assembly" are based on a hypothetical composite fuel assembly consisting on the worst gap inventories from those assemblies with maximum bumup. Calculation of the fuel gap inventory is based on a one year decay period, effective December 6,1997. Table 5.3.1 provides the bounding fuel rod gap radiological inventones assumed in this analysis. These radiological nudides represent the long sved isotopes as appropriate for spent fuel at the one year decay period. The principal assumptions utilized in the development of this analysis for the original operating ] l plant Control Room are:
]
- 1. A minimum of 19 feet of water is above the hypothetical failed assembly.
- 2. The fuel assembly that is dim releases 100% of the assembly maximum gap radiological inventory of Kr", l*, l*, Xe*"', and Xe".
- 3. Assembly gap radionuclide inventories consist of 10% core iodine (except l*
at 30%), and 10% core noble gases (except Kr" at 30%). These gap inventones are released to the pool water with a subsequent release of 100% of the noble gases. The iodine releases from the pool water include a credit for a Decontamination Fador (DF) of 75 based on the 19 feet of water for iodine scrubbing. In addition, both the Exdusion Area Boundary (EAB)an& Control Room (CR) doses were evaluated for a condition of no credit for pool water iodine scrubbing, i.e., a DF of 1.
- 4. No fuel overheating occurs, and thus, no signdicant solid fission products are released. I
- 5. No credit is taken for the isolation of the spent fuel building. The release is .
modeled as an instantaneous puff release at accident (95 percentile) ground level meteorological condibons. Rev.17 DSAR 5-10
i O 6.
"'^'
The free air volume of the CR is 38,000 cubic feet.
- 7. Normal CR vendistion is adher 120 or 900 cubic feet per minute for the duration of the accident.
- a. No credit is assumed for CR isoladon or air Stration.
The radiological analyses was performed using the code ELISA, Reference 1. The results of this analysis for the EAB are presented in Table 5.3.2. These results show that the projected doses Rom the fuel handing acculent are insigni6 cent in comparison to the 10 CFR 100 limits and for less than the Ermranmental Protection Agency Protective Action Guidelines (PAGs). i ' The calculated doses for the CR portion of the analysis at the CR ventilation intakes and inside the CR are presented in Table 5.3.3. These results show that the anticipated doses are within the 10 CFR 50, General Design Criteria 19 dose limds. In May 1998, Control Room functions were relocated to the Spent Fuel Pool Island Control ] t, Room in the Gate House / Front Omco Building. A rowwaluation of the control Room dose ] as a result of the new location was performed. The principle di5erences in assumptions use ] in this re enalysis from that of the old Control Room are:
]
- 1. The volume of the SFPI Contml Room is +;-;cueT ' 'i 6400. ]
- 2. Normal CR ventilation is 1100 cubic feet per minute for the duration of the ] I Accident. ] )
- 3. The distance tom the Fuel Buildng is +;-;-04T:Ey 113 meters and not within ]
! the wake of the Containment. '
]
} ]
Since the dose rates at the airintakes are proportional to the X/Q values, the dose rates for ] the new SFPI Control Room are +;-;-04T ^ y 60% of the values in Table 5.3.3. The ] dl5srenons in volumes and air flows have a negligible impact. (Reference 2) ] Section 5.3 References
- 1. Yankee Atomic Electric Power Company computer code ELISA, calculation YC-313, dated February 24,1994.
- 2. Memo: M. Johnson, MYAPC to J. McCann, DE&S; Review of DSAR Change 98-43; ]
MWJ-99-01; January 25,1999. ] Rev.17 DSAR 5-11
i MYAPC i Table 5.3.1 IBOUNDING SPENT FUEL INVENTuiiss FOR RADIOLOGICAL ANALYggnao 4 54#i#i@4sMC42I100CLIDEsPlisg@yM )$$ n ~s! & atu m a n a r v , 2 u .. MM { I-129 1.85 E42 ! l-131 6.60 E-09 Kr-81 3.85 E-07 Kr-65 4.04 E+03 Xe-129m 5.97 E-14 Xe-131m 5.86 E-06 Xe-133 8.06 E-16 (1) Inventories include a 5% uncertainty factor. 4 (2) On a per assembly basis. e Rev.14 1 9 DSAR 5-12 '
I i MYAPC Table 5.3.2 MAINE YANKFF POST SHUTDOWN FUEL HANDLING ACCIDENT EXCLUSION AREA BOUNDARY DOSES
?
Spent Fuel Poollodine DF = 75: A;if~DoomyTimedy ThyroktDbse $ t x;}4 a n -;EfilmilhemialeBody~,~b %; SkinDose n ,,. a ," mg.. ,.:y~ -n yM. n msgwyp;e;. no$pkw . may; < '; j% hh -. jsg D
@L'^lNh,;ijg6$ NN$ WOR bh5fjik NY s $
p;g ?@O'.kf % gpp;ff; *:( Q ( g %)p g$ g$y jh$m$g$gidg$NN gg 1 1.95 E-04 2.27 E-04 3.17 E-02 5 1.63 E-04 1.65 E-04 2.30 E-02 { 10 0 1.20 E-04 1.66 E-02 { 10 CFR 100 Umsts 300 25 n/a EPA PAGs 5 1 50 Spent Fuel Poollodine DF = 1:
.J. Decay Tkne ... F %ThyroidDosee > ;EffectheWhole Body < ':iSida DoseJ@a
+'% '- -n v :*n' ,
. n. evyy~~ +.>cy
' -p wm y. . :. 'p.';' . 3, ,
. .. :.= ,, - n .. ..n . ,
My(( {;-b . n x s > iSN Y:fi '#582 d~'f W II@:! " f T;Y@ g.h$,n,
+'#
s 1 6.12 E-03 4.05 E-04 3.17 E-02 5 6.09 E-03 3.43 E-04 2.30 E-02 10 6.06 E-03 2.98 E-04 1.66 E-02 10 CFR 100 Umits 300 25 n/a EPA PAGs 5 1 50 i Rev.14 DSAR 5-13
MYAPC Table 5.3.3 MAINE YANKEE POST SHUTDOWN FUEL HANDLING ACCIDENT CONTROL ROOM DOSES AS OF DECEMBER 6.1997 Spent Fuel Poollodine DF = 75: M Location s /E3s it" @Th
. ."3^4%
r- *yroid
~*L:wsDoseG@5 c: WE5ectiveVIAsole$ 9 V s;2SIds: D o m e W :9
+-
us. itt&+>'2%;g% ".??.,. s, . ; f ~:t# [: g
'MQL D , .by 0;; ::we-wWavem%dd. t$y,;.. g :
s.
- u.ep w a w uus.L S .
sa? . w ;- yC h w m:rzn d:xfI.T ^I f m^' .'-:y ' W~s;.W.),w'i fyp.:c:g[$g. v^? A F. k. t
.y . ..f a m:w.-/Si.: 'N ,y<- msw mp
.+
E,se e 34 i m' ' w' d. [
^ b D$ bM$$$+&$1:,$ ?ll% @$y@:$$${"$:y ykk,$* ;'*'yl w W Y & fiL.
khi;f
. . > xkk. hht:lky,_;h$e, .jN kN At CR Ventilation 4.3 E-03 7.20 E-03 3.50 E-01 System Intakes inside CR - 120 cfm, 1.6 E-03 1.40 E-03 3.50 E-01 ,
No isolation inside CR - 900 cim, 1.6 E-03 1.40 E 03 3.50 E-01 No isoletion 10 CFR 100, 30 5.00 N/a GDC 19 Umits Spent Fuel Poollodine DF = 1:
't. Location 2.~.4.#" 3,. Thyroid
. ~
- e. . , . .
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At CR Ventilation 6.9 E-02 9.20 E-03 3.50 E-01 System Intakes inside CR - 120 cfm, 6.6 E-02 3.30 E-03 3.40 E-01 No Isolation inside CR - 900 cfm, 6.6 E-02 3.30 E-03 3.40 E-01 No isolation i 1 1 Rev.14 5.4 Sonnt Fuel Cask Dron - I DSAR 5-14 J
o ,q MYAPC V Spent fuel shipping casks are designed, as por the requirements of 10 CFR 71, to withstand a free fall of 30 foot onto an unyielding surface. For this reason, radiological consequences of a MW spent fuel cask drop medient outside the spent fuel pool are not requred if potential drop distances are less than 30 feet. The design of the Maine Yankee spent fuel cask transfer system is audi that the cask drop distance is less then 30 iset whenever the cask is not diracey over the spent Ibel pool. It is concluded, therefore, that an evaluation of the spent fuel cask drop accident outside of the spent fuel pool is not riaca==ary. Operations with the spent fuel shipping container are demgned not to para over spent fuel storage racks or spent fuel assemblies during cask loading or fuel transfer operations. At the cunent time, Maine Yankee is prohibited from ilRing a spent fbel shipping cask over the spent fuel pool. Therefore, an accu $ent analysis of a spent fuel cask drop in the spent fuel pool is not required and ] does not supply safety analysis limits.
]
DSAR 5-15 Rev.15
i MYAPC ; 1 5.5 Sonnt Fumi Pool Accidenta The category of potendal acddents, other than the previously discussed w My'acddents, mamar4=hwi with the spent fuel pool while in the permanently defueled condition are limited to those events related to the lose of pool cooling or the loss of water inventory. The spent fuel pool and the associated cooling system are designed to: (1) maintain the wolorin the spent fuel pool at a workable operating temperature with a full contingent of spent fuel assemblies, (2) mentain fuel dien.g integrity in the event all forced cooling is lost and cooling occurs by baigng at the surface of the spent fuel pool, with boil off losses being made up by a supply of makeup water, and (3) mentain suflident coadng of fuel assemblies in the event a fuel assembly or other object is dropped and rests across the top of one or more assembly locations. The design of the spent fusi pool and coosng system is discussed in section 3. In the permanently defueled condibon, the sole requirement with respect to coolmg, is to assure adequate decay heat removal capability in maintaining the water level in the spent fuel pool so that the spent fuel assemblies remain covered. The design of the spent fuel pool is such that a loss of coolant below the top of the fuel is not considered to be a credible accident. Events do exist which can result in loss of fon=ed spent lbel cooling or reduce the water inventory in the spent fuel pool available for cooling. However, the worst case event, using the decay heat levels at the end of December,1997, allows adequate time (a minimum of a week from nominal spent fuel pool conditions) to establish pool makeup capability to ensure that fuel elements remain covered and the resultant radiation doses are low. 5.5.1 Loss of Spent Fuel Pool Cooling The loss of spent fuel pool cooling family of analyses consist of: (1) the postulated blockage of the top of individual storage cells, (2) the loss of forced flow to the spent fuel pool heat exchangers, and (3) the loss of the spent fuel pool heat exchanger heat sink. The accidents associated with the loss of spent fuel pool cooling are dependant on the determination of the decay heat loadings in the spent fuel pool. For these analyses, the decay heat was determined through application of the NRC Branch Technical Position ASB 9-2 with an uncertainty factor of
+10%for decay times in excess of 10' seconds as is consistent with the NRC Standard Review Plan, Section 9.1.3, " Spent Fuel Pool Cooling and Cleanup System". The decay heat calculations were C . ;di determmed assuming prior plant operation at 2700 MWt for the most recently discharged spent fuel. Actual power operation was limited to 2440 MWt for the last batch of fuel used ]
for powerproduction at Mame Yankee. ] DSAR 5-16 Rev.14
WMC in late October and earty November 1997, Mane Yankee performed passive cooling tests in the spent fuel pool to obtain data in assesang the pool hootup rate, host losses due to evaporation at elevated temperatures, and accuracy of the +% of spent fuel decay host. Conservative assumptions were used for inclusion of the fuel buildng heet losses in this assessment. The results of these tests indicate that possive (i.e.: no forood flow through the cooling system) cooling of the spent fuel pool will result in a steady bulk water temperature of between 190 and 200*F with water surface wind speeds between 3 and 5 miles per hour. When compared to the analytical predictions of the spent fuel decay heat using the NRC Branch Technical Poellion ASB 9-2, these results demonstrate substantial margin to the decay heet analysis assumpoons presented in this secuan. 5.5.1.1 Blocked / Improper Ce8 Flow The design of the spent fuel racks (Reference 1) is such that the top of the rack cells may be blocked and sufficient cooling of the fuel assembly is still assured. Analyses of various types of flow blockage of the cell exit have been performed to demonstrate e.f f.dury preservation of the stored fuel in a coolable geometry under these conditions. Two types of cell blockage were considered. The first assumed that a fuel assembly was laying honzontally across the limiting spent fuel storage cell. Conservatively neglecting flow upwards through the horizontal pins, this event results in a not blockage of the storage cell exit area of 79%. The second flow blockage scenario involves the piacoment of a 8 by 10 foot section of masonry wall from the south end of the spent fuel pool onto the top of the storage cells. This event blocks 100% of multiple cell exit openings and requires the cooling of the spent fuel through the 1 inch diameter flow holes located near the top of the celle. Specific analyses were performed using the RETRAN code (Rsfarence 2) in analyzing the blockage of cells to conservatively predict the limiting fuel coil coolant conditions in verifying that localized boiling , or that the onset of Critical Heat Flux (CHF) will not occur in the individual fuel storage cell. The limsung cod was then evaluated to detemune the peak fuel pin surface cladding and fool pellet temperatures. These er 'i; .; were performed as part of Reference 1. The acceptance criteria established for these analyses are as follows: i
- 1. The calculated peak cod coolant exit temperature does not exceed the local l saturation temperature of the cooient at the top of the active fuel. I
- 2. The calculated local peak ciedding and fuel pellet temperatures are sufficiently '
cool to ensure long term integnty. O DSAR 5-17 Rev.15
MYAPC
- 3. In the case of multiple blocked cells, the acceptance cdteria is the avoidance of reaching CHF onset and the assurance that the fuel remains coolable.
The ionowing key assumptions were lrww into the RETRAN modeling to pedu.in the calculadon of a conservative peak col exit cooient temperakse:
- 1. One dimensional fluid and vapor flows.
- 2. All decay heat calculations used the NRC Branch Technical Position ASB 9-2.
- 3. The spent fuel was assumed to be no more than 6 days removed from the time of reactor shutdown. Fuel for the multiple blodcod cet analysis utuized a 180 day cooling period since piacoment of the hattest spent fbal is administratively controlled to the north and of the spent fuel pool and the masonry waN section initiating the blockage is at the south and of the spent fuel pool. All spent fuel currently being stored in the pool has undergone at least a 365 day cooling period as of December 6,1997.
- 4. The fuel was assumed to be of the " debris resistant" type with higher inlet spacer grid flow resistance.
5. O 6. No heat transfer between cats was atoned. The temperature of the coolant was assumed to be at the bulk water temperature of the pool at 154'F. No credit was taken for the coolant retuming from the heat ewherg.c at a lower tanperature.
- 7. Coolant exiting the cells is assumed to mix completely with the bulk fluid above the spent fuel storage racks.
- 8. AH cells are assumed to contain a relatively high powered assembly with bumups of 70,000 MWD /MTU, and assembly powers of 64.97 MBtu/hr.
1
~
The methodology utilized for this analysis is based on an adjacently stored 20 assembly model of the fuel in the spent fuel pool. To evaluate detailed coolant flow to a cell, the spent fuel pool is wr ::LLei reduced to a two-dimensional model. Credit is taken only for coolant flowing downward between the fuel rack and pool waH and comespondingly upward through the storage cell. No credit is taken for flow from or to a third dimension. Flow is requwed to transverse the floor from the wall until reaching the farthest cell from the wall. Thus, the limiting region is determined by identifying the cells furthest from the wall, drawing flow from the nanowest gap between the pool wall and rack, and containing the hottest fuel. Rev.14 - OSAR 5-18
^
MYAPC The analytical results for the heiiu.Ed.i fuel assembly blockage show the peak cell exit temperature occurs in the farthest cell from the pool wall. These temperatures remain subcooled and thus, ensure that the fuel remains in a <miaNa state with acceptable cladding temperatures. The multiple cell blockage sr.dy.i. results indicate that CHF is avoided and the fuel remains in a cootable state with cladding temperatures at or below 400*F. The detailed predicted temperature results from these analyses are provided in Table 5.5.1. De results from these analyses indicate that the acceptance critana has been met and that coolability of the stored spent fusi under partially blocked, or simultaneous multiple cell blocked, condisons is assured. 5.5.1.2 Loss of Forced Flow 1 The Loss of Forced Flow analysis is intended to assess the impact and postulated consequences associated with the loss of forced cooling to the spent fuel pool. De function of forced cooling is to remove the decay heat generated by the spent fuel. The spent fuel pool cooling system design and operation is described in section 3. There is no initiating mechanism identified as the cause of this incident. Rather, the analysis simply assumes that cooling flow to and from the spent fuel pool has been terminated. Initiation of this incident from an inadvertent inventory loss or syphorung from the spent fuel pool is discussed in section 5.5.2. With the loss of forced flow to/from the spent fuel pool, the water in the pool will absorb the decay heat generated by the spent fuel and, without restoration of cooling, increase in temperature of the , pool to the boiling point. Heat transfer from the pool at elevated temperatures occurs by conduction l through the pool walls and surface, convection across the surface of the pool and boil-off or evaporation of water from the pool surface. In this analysis, the time to increase the spent fuel pool temperature to the point of bulk boiling defines the available times for operator compensatory actions. Therefore, the results of this analysis are defined as the time to boil and, because boiling will result in a loss of pool inventory, the boiloff rate. The principal assu.Tptions used in the dr.d.cyT d of this analysis are as follows:
- 1. Initial spent fuel pool conditions:
Free space water volume 13,742.3 cubic feet Rev.14 DSAR 5-19
MYAPC Free space watermess 1,620,000 pounds Water temperatures 80,100,120,140, and 154*F Waterlevels 31,40, and 43 foot elevations (Note that the 43 foot elevation represents the spent fuel pool low level alarm setpoint)
- 2. Immediate loss of flow frum both spent fuel pool cooling pumps and the fuel pool purification pump. -
- 3. The mass of the stored spent fuel, fuel storage racks, and associated water volumes are assumed to remain at the incident initianon temperature and do not absorb decay heat.
- 4. It is assumed that when the free space water reaches an enthalpy cuci pe,.dk.g to 212*F at atmospheric pressure, bulk boiling has been reached. Hyd.v.i.isc pressure effects on water enthalpy are neglected.
The results from this .ci,a are presented in Table 5.5.2. This table contains the time to boil
+dananal results for p i c.:1;--J-:- on initial water levels and temperatures for decay heat levels ranging from late December 1997 through December 2000. The nunimum time *W to
, reach boiling fham a normal water level (43 feet)and temperature (100*F) is in excess of 57 hours as oflate December 1997. Thecse tabulated results, along with the resultant boiloff rate and calculatrd loss in spent fuel pool level per day of boiling, provide assurance that the total loss of all spent fuel pool cooling, even at reduced water inventory levels and conservative decay heat values, wm allow sufficient time for appropriate operator actions in the restoration of cooling and makeup. The spent fuel pool bolloff rate and boiloff inventory losses as functions of the fuel decay times are provided in Figures 5.51 and 5.5-2. The resultant radiological doses associated with the boiling of water in the spent fuel pool have been calculated and are provided in Table 5.5.3. The results of this calculation indicate that the projected dose rates are less than 2 mrom/ hour within the fuel building and may be considered negligible outside of the fuel building. Exclusion Area Boundary Doses (2 hour) are ceiculated to be 2.23 E-03 mrum. Based on the above, it is concluded that the Loss of Forced Flow incident for spent fuel pool cooling does not constitute an excessive risk or radiological consequence for eNher the workers of Maine Yankee or the general public. Rev.14 DSAR 5-20
MYAPC 5.5.1.3 Loss of Heat Sink The consequences of a i=m loss of the spent fuel pool host sink are similar to and bounded by the ions ofIbroad now analysis in the spent fusi pool. In the Laos of Heat Sink incident, cooling to the shou side of the spent fuel pool heet endungeris . _. , presumed ioet and the water temperature begins to imremse. This event is trieny discumend in l Reference 1. Multiple sources of bedcup cooling or spent fuel pool makeup are av=EmNa Forinstance, the piant are protection system is avalieble fbr connecdon to the heat exchanger via edsting two emergency f cooung connecdons. Additionsuy, with the long spent fusi pool hestup times, ;,,-;>- . -, of temporary host sink cooling through the use of temporary pumps, are hoses, and the Sheepscot River water is possible. In either case, mitigation of the loss of heat sink cooling would be e8ected. Spent fuel pool maksi& si available through a number of sources. Norrnal or routine makeup to the spent fuel pool is accompelshed by P-SFP2 and domineralized water from the PWST. ] AdditionaNy, there are at least three primary grade water hose connections in the vicinity of the spent fuel pool. Each connection is capable of providing +;-i-sai,T ti 20 gpm makeup flow to the spent fuel pool via hoses. In an emergency situation with no other means of makeup available, makeup from the fire mein system is available. Using one or more homes, makeup rates in exoses of 150 gpm are avaNable. Additionady, a hoes connection to the town domestic water supply could be utelzed in an emergency ] situation. 1 Reccesy of the cooling to the shell side of the spent fuel pool heat exchanger and/or establishing appropriate levels of makeup to the spent fuel pool would be =W to occur well within the spent fuel pool heetup times discussed in section 5.5.1.2. Potential radiological consequences of this transient are the same or bounded by that deAned in section 5.5.2. Based on the above, it is concluded that the Loss of Heat Sink incident for spent fuel pool cooling does not coneslute an ====ve risk or radiological consequence for other the worters of Marm Yankee or the general public. . Rev.16 DSAR. 5-21
MYAPC 5.5.2 Loss of Spent Fuel Poolinventory There are two types of pool inventory loss incidents for consideration consistent with the permaneney defueled condition of Malae Yankee. Lama of Caoung inventary The first, a loss of water inventory asso'ciated with the boling of water in the spent fuel pool due to a loss of cooling or host sink, resuNs in a minimal rate of water loss. The calculated bodoff rate at a spent fuel decay host level W with the fbol decay date at the end of December,1997, is calculated to be 11.64 gun. This boiloff rate corresponds to a loss of 1.47 feet of the spent fuel pool level per day. Later dates and lower decay heat levels result in lower boiloff rates and lower inventory losses. Figure 5.5-1 IEustrates the bolloff rate as a function of the time following reactor shutdown. l The cu,T pe,, ding loss of inventory resulting from a boiloff of the spent fuel pool wateris shown in Figure 5.5-2. The radiological consequences of boding in the spent fuel pool are shown in Table 5.5.3. These consequenons are minimal for plant workers exposed to the fuel building environment (i.e., less than 2 mrom/hr radi=Nart fields projected). In this inventory loss inddent, the extremely slow loss of water due to boiloff is easily ceiiveE':1for using the previously desaibed mekaup sources and appropriate operator actions. No additional actions or compensating trutigaNon activities are required fbr this type of a loss of spent fuel pool inventory incident. Inadvertant Svohoning inventory Loss 4 The second Loss of Spent Fuel Pool Inventory inckient, an inadvertent syphoning of spent fuel pool water through the pool cooling water inlet or outlet, results in a more serious and rapid drainage of water from the spent fuel pool. Both the spent fuel pool cooling water inlet (6" diameter une at an 31.5 foot elevation) and the pool cooling water outlet (8" diameter line at an 8 foot elevation, with a syphon break at the 40 foot,11 inch elevation ) are subject to inadvertent syphorung through a postulated breakage in smaN section of Non-Nucteer Safety grade common piping. Instistion of these transients also results in establistung the irutial pool level conditions for the Loss of Spent Fuel Coonng incidents (see Table 5.5.2). Although the spent fuel pool cootng water outlet is located at the 8 foot elevation, the piping desgn
^
'nce pw,-- z a passive syphon break at the 40 foot,11 inch elevation, thus resulting in the potential i
Rev.14
-CMW 5-22
MYAPC drainage of approxsnotely 2 feet of spent fuel poolwater below the low level alarm setpoint. At that level, the syphon break would introduce air into the cooling water outset pgang and the syphoning would terminate. The loss of 2 feet of spent fuel pool water (approximately 23,500 gallons) in the fuel building would alert the operators to a mishap and conective actions would be indisted in accordance with those identifled previously. Concunent with the loss of pool inventory due to the syptn,a 4, the ceu".4 How to the spent fuel heat exchanger would also be lost. Based on a nommel pool water temperature of 100*F, an inlual pool level at 40 foot, and the decay host levels mamariated with the and of December 1997, in excess of 51 hours are available fbr compensatory actions prior to the iruttation of pool boEng. Sullicient time is avadable for the operators to initiate conective actions in the restoration of cooling and makeup. I The original spent fuel pool cooling water inlet design, however, did not lin.npuie a syphon break. ] In this assessment, it was assumed that the spent fuel pool drains to the 31.5 foot elevation, a loss of ] 12.5 feet of spent fuel pool level (147,500 gallons spilled to the fuel building). The result of this spdiage not only temunstes the spent fuel pool cooling (as the syphon break in the outset piping is uncovered), but also results in +;-;-wet ry 10 feet of water covering the spent fuel in the storage colis. I Makeup to the spent fuel pool and recovery of the syphoning of water through the cooling water inlet piping would be the same as that paviously discussed. Based on a nominal pool water temperature of 100*F, an initial pool leve! at 31 feet, and the decay heat levels associated with the end of ] December 19g7, in avemma of 33 hours are available for compensatory actions prior to the initiation of pool boiling. Sufficient time is avadable for the operators to initiate conective actions in the restoration of coohng and ma,keup. In order to enhance the time available for opemtor action, a branch syphon break has been added to ] the cooling retum line. The syphon break has an elevation of approximately 40 feet which increases ] the avadable response time to essentially the same as those listed above for the assumed spent fuel ] pool cooling water outset line break.
]
The radiological consequences assometed with the loss of spent fuel pool k;;,L,y have been doened. The inadvertent liquid releases due to the loss of inventory incident are bounded by the analyses discussed in section 5.6. The antiopated radiological dose rates in the fuel building are presented in Figure 5.5-3 and the skyshine dose assessments in Figure 5.54. Based on the above discussion, it is concluded that the Loss of Spent Fuel Pool Inventory incidents do not constitute excessive risks or radiological consequences for either the workers of Maine O Yankee or the gennral public. Rev.16 i DSAR 5-23 i l
MYAPC P Sadion 5.5 Rafarances
- 1. W, Maine Yankee to the NRC, W Technical S;-=*2% Change No.177: Mene Yankee Spent Fuel Pool Rarackingf, MN-93@, dated January 25,1993.
- 2. Report N - A Prognen for Transient L,ird i.ulici Analysis of Cornplex F12 syneerrW. EPRI NP-1850CCM A.
O Rev.14 DSAR 5-24 !
MYAPC n Table 5.5.1 RLOCKEn Ammpusal Y COOLING ANALYMER RERULTS
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MYAPC Table 5.5.3 O Faoaecreo aioiotoeicit ooss coussoussces 9 OF BOILING IN THE SPENT FUEL POOL LKeW:LOCGON: AAABara ~WLDOSETOLUNGS?KW% 44DOSETOWHOLEBODYA . Exclusion Area Boundary 3.21 E-06 Rom 2.23 E-06 Rem (2 hour dose) Control Room 1.70 E-05 Rem /hr 1.21 E-05 Remhr Fuel Building 1.80 E-03 Rem /hr 1.28 E-03 RemMr The primary contributor to these doses is H8, contributing about 58% of the projected lung dose and about 82% of the Effective Dose Equivalent Whole Body. i O I 1
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MYAPC - 5.6 Low Level Waste Rel==== l=ide.a. 5.4.1 Radioactive Waste Gas System Leaks and Failures
]
The projected radioactive gas release informaiksi and projected doses that are applicable to the ] permanently defueled condition are defined in sechons 5.3, 5.5, and 5.6. 5.6.2 Radioactive Liquid Waste System Leaks and Failures Because of the continued storage of radioac6ve fluids at Maine Yankee during the permanently defueled ] condition, the analysis of radioactive liquid waste system leaks and failures remains essentially the same as that used as the licensing basis for the plant during power operations. The use of the full power transient is used as representative of the maximum dose consequences associated with the offsite receptor due to a liquid and gaseous release. This transient is not a credible accident scenario in the defueled condition. it is not intended to be the sole basis upon which the retention of systems or components is required. Acbons leading to transfer of a radioactive fluid from a system to the environment or to another system require positrve operator control and morutonng. O ^ii onerator aciions re2uired are performed in accoreance wiin wnt'en operatiaa n,ocedur instruction, checidists and allowable release information. With these considerations in mind, the probability of the accidental release of radioachvity from the radwaste systems as a result of operator erroris minimal. The purpose of this section, however, is to consider postulated liquid waste system single component leaks or failures in order to arrive at the postulated incident which could have maximum off-site effects. Method of Analysis: A. RA .= to the Atmnenhare P-mus Releases) For the purpose of establishing an upper limit on the activity released from a single coinpererit failure in the liquid waste system, it is assumed that the primary drain tank fails, releasing its total inventory. The primary drain tank failure has been selected since this tank has the highest inventory of dissolved noble gases and halogens during operabonal periods. The release takes place as a liquid spill on the floor of the compartment in the weste processing building where the tank is located. Radsoac8vity is released to the atmosphere fium noble gases and halogens evolved from the spilled liquid. O Rev.17 DSAR 5-35
)
l MYAPC p" The following conservative assumptions are used in conjunction with the meteorological g and dose assumptions given in Appendices SA and 58. I
- 1. Eighty percent of the primary drain tank's 8,150 gallon capacity is filled with I undecayed, un-degasified primary reactor coolant (with activity concentrations at Technical Specrfication limits of 1.0 pCl/g Dose Equivalent 1-131 and 100/E pC1/g).
- 2. One hundred percent of the tank's inventory is spilled and all of the noble gases and 1 percent of the radioiodines are available for direct release through the i building ventilation system to the environment.
- 3. Duration of the release is 1 hour.
B. Potential Releases to the Groundwater Table Postulated liquid spills escaping concrete structures may be released to the site groundwater table. The groundwater table at the Maine Yankee site, however, flows towards Back River and Montsweag Bay, both of which are tidal saltwater estuaries. It is concluded, therefore, the potable grourd*ss and surface water supphes would I not be affected by radioactrve liquid spills and no dose evaluation for this postulated event is required. Analysis Results: The potential whole body extemal exposure doses and thyroid inhalation doses to an individual are summanzod in Table 5.6.2. The doses which have been calculated for radioactive liquid waste system failures are below small fractions of the values in the applicable regulation,10 CFR 100, " Reactor Site Criteria" and the EPA Protective Action Guidelines. I Rev.17 DSAR 5-36
I MYAPC Table 5.6.2 PROJECTED DOSES FROM PRIMARY DRAIN TANK RUPTURE Dose Point Thyroid Dose (rem) Whole Body Dose (rem) 2-Hour Site Boundary 5.8 - 2* 2.3 - 1 1 30-Day Low Population Zone 4.2 - 3 1.7 - 2
- 5.8 - 2 = 5.8 x 104 O
l I l O Rev.14 DSAR 5-37
MYAPC 5.6.3 Low Level Waste Storage Building Accident The temporary storage of low level waste (LLW) on the Maine Yankee site during both periods of power operations and the current permanently defueled condition is administratively controlled within the existing Low Level Waste Storage Building (LLWSB). This building was designed and constructed to safety store up to a 5 year inventory (volume and activity) of the low level waste generated during penods of normal and refueling operations. All waste stored in the LLWSB is packaged and ready for near term offsite shipment. The design of the LLWSB is discussed in section 3. There are no credible initiating events to cause an accident in the LLWSB. The bounding accident for the radiological impact analysis is defined as the dropping of a highly loaded spent resin liner within the building, resulting in the liner failure, spillage of the spent resin, and the islease of a fraction of the radioisotopic contents in a cloud. The contents of this cloud form the basis for determining the radiological source term at the site boundary. The key analysis assumptions for this incident are as follows:
- 1. The liner is loaded with spent resin at the allowable Department of Transportation Low Specific Activity (LSA) limits por 40 CFR 173.403, except for the isotopes of l'8 , l'8', and Co".
{O 2. The amount of l'8 is assumed to be equivalent to 0.08 Curies per cubic meter, which is in excess of the limits allowed by 10 CFR 61 LLW burial criteria.
- 3. The amount of l is assumed to be 800 Curies in the liner.
- 4. The amount of Co"is assumed to be 1,627 Curies in the liner.
- 5. The liner contains 148 cubic feet of spent dewatered resin at 40 lbs/ cubic feet.
The spent resin liners are stored in High Integrity Containers (HICs) which are designed to survive a 20 foot drop while fully loaded. Although the LLWSB crane height does not exceed the 20 foot height, it is conservatively assumed that the loaded HIC falls and breaks open, thus spilling the spent resin. It is assumed that 1% of the activity of the liner non-mechanistically forms an aerosol and that 10% of the aerosol is non-mechanistically released outside the LLWSB. This aerosol acts as a " puff" release in assessing the potential doses at the EAB. The activity of the release is assumed to be dispersed over an arc of 225 degrees at 700 meters from the LLWSB. SANDIA National Laboratories (report SAND 87-2808, August 1988) notes that this type of accidental release and dispersion does not result without the assistance of a fire source to disperse the contents of the spelled resin. However, there is an extremely low fire loading in the LLWSB and the postulation of a i fire while moving a loaded HIC is beyond credible. O Rev.14 DSAR 5-38
MYAPC O ota riaiti ti"9 at r x -im o for aet "ti i< aioio9'c 'r i a ta- v=. ^ - ot or seismic events, tomados, floods, and hydrogen gas generation and explosions on the potential to yield a more severe radiological release have not identified a more severe case than the dropping of a loaded HIC. The calculation of the doses at the EAB was performed in accordance with the NRC Regulatory Guide 1.109, Appendix C, using the code ATMODOS. The results of this calculaton for a two hour penod at the EAB are as follows:
- 1. Maximum Organ Dose (Teen Lung) 262 mrom
- 2. Thyroid Dose 70.52 mrom
- 3. Whole Body Dose 7.08 mrem The principal radionuclide contributors to the maximum organ doses are:
Nuclide Dose (mrem) Co" 76.6 Ce* 58.0 Ag* 29.3 Sb* 16.7 Sb5 11.9 Based on these dose calculations, the doses which have been calculated for the limiting incident in the storage of LLW in the LLWSB, the dropping of a fully loaded HIC with spent resins at or above regulatory limits, are below small fractions of the values in the applicable regulation,10 CFR 100, " Reactor Site l 3 Criteria" and the EPA Protective Action Guidelines. I O l Rev.14 l l l DSAR 5-39 ,
MYAPC APPENDIX SA
SUMMARY
OF PARAMFTERS USED FOR EVALUATING THE I;j - - RADIOLOGICAL EFFECTS OF INCIDENTS i I. Data and Assumptions Used to Estimate Radioactive Sources from Postulated incidents. On August 7,1997, Maine Yankee certified to the NRC that it would no longer operate and that fuel had been removed from the reactor vessel and placed in the spent fuel pool. Therefore, certain information below is being retained for historical purposes. A. Power Level The power level is considered when determining the fission product inventory levels of the spent fuel assemblies. B. Sum-Un Operation at 2700 MWt is the maximum power level obtained during the operation of the Maine Yankee station. Equilibrium core Inventories for this power level for selected radionuclides are given in Table 5.A.1. Noin: Actual fuel assembly power history levels, prior to permanent plant shutdown on December 6,1996, were used to evaluate post plant shutdown spent fuel pool incidents. C. Eggent of Fuel Perforated , Discussed in individual incident analysis. D. Eglease of Activity by Nuclide Discussed in individual incident analysis. E. lodine Fractions Discussed in individual incident analysis. O Rev.14 DSAR SA-1
MYAPC
- 11. Atmospheric Dispersion Data A. Site Boundary and Low Pooulmunn 7nne (LPZ) Distances The distance to the site boundary is 610 meters (0.38 miles) and the distance to the LPZ is 9,600 meters (6 mdes).
B. Atmonoheric Disoarsion Factors (X/Q) values See Appendix 58. Ill. Dose Data P A. Method of Dose Calculations Thyroid, whole body, and skin doses are evaluated using an approved hand calculation metho'dology or the Duke Engineering & Services, Inc. computar codes referenced below (Ref. 1,2,3, and 4). Whole body doses and the gamma portion of the skin dose are calculated by assuming the receptor is emersed in a either a semi-infinite plume or finite plume, whichever is appropriate. l O in seaerai. aii the doses from extemai radiatiea cio the ihr,oid. whoie bodv. and sain) are evaluated using a product of the activity concentration at the receptorlocation of interest Ed the appropriate Dose Conversion Factors (DCPs) as taken from ICRP 30. These DCPs are consistent with those provided in References 5 & 6 below. Thyroid doses due to inhalation are the product of the activity concentration at the appropriate receptor location, the iodine radionuclide DCPs, and the post accident breathing rates as defined in Reference 7 below. l B. Doses Doses are discussed in individual incident analyses. I O Rev.14 , DSAR 5A-2
MYAPC
- 1. ELISA, "A Computer Code for the Radiological Evaluation of Licensing and Severe Accidents at Light-Water Nuclear Power Stations," April 1991.
- 2. DIDOS-Ill, "A Three-Dimensional Point-Kamel Shielding Code For Cylindrical Sources,"
1980.
- 3. ORIGEN 2.1, " Isotope Generation and Depletion Code Matrix Exponential Method," Oak Ridge National Laboratory, CCC-371,1991.
- 4. RADFLEX, "A Two-Dimensional Shielding Code for the Determination of Skyshine Radiation From Point-Isotropic Gamma Sources," 1982, S. Federal Guidance Report No.11. " Limiting Values of Radionuclide. Intake and Air !
Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," ) Oak Ridge National Laboratory,1988. ] l
- 6. Federal Guidance Report No.12. "Extemal Exposure to Radionuclides in Air, Water, and O soii. o k aiose " tiea i' bor torv.4993.
- 7. USNRC Regulatory Guide 1.4, " Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water Reactors," 1974.
1 i O ' Rev.14 DSAR 5A-3
MYAPC Table 5.A.1 EQUILIBRIUM CORE FISSION PRODUCT INVENTORIES" (Halogans and Noble Gases) thg;Uda Core Activity (Cl) Gap Fraction (%) Gao Activity (Cl) 1-131 6.56 + 7* 10 6.56 + 6 l-132 9.64 + 7 10 9.64 + 6 l-133 1.51 + 8 10 1.51 + 7 l-134 1.77 + 8 10 1.77 + 7 l-135 1.40 + 8 10 1.40 + 7 l-136 7.02 + 7 10 7.02 + 6 l-129 1.81 + 0 10 1.81 - 1 Kr43m 1.18 + 7 10 1.18 + 6 Kr-85m 2.94 + 7 10 2.94 + 6 Kr45 7.03 + 5 30 2.11 + 5 Kr 5.73 + 7 10 5.73 + 6 O kree Kr49 8.oe + 7 1.04 + 8
'o 10 e oe + e 1.04 + 7 Xe-131m 3.83 + 5 10 3.83 + 4 Xe-133m 3.62 + 6 10 3.62 + 5 Xe-133 1.51 + 8 10 1.51 + 7 Xe-135m 4.07 + 7 10 4.07 + 6 Xe-135 3.02 + 7 10 3.02 + 6 Xe-137 1.36 + 8 10 1.36 + 7 Xe-138 1.34 + 8 10 1.34 + 7
- 6.56 + 7 = 6.56 x 10 7
" Based on operaten at 2683 MWt. Limiting spent fuel, as produced in the last cycle of operation, was at a nominal power level of 2440 Mwt.
Rev.14 DSAR SA-4
MYAPC O APee~ Dix 5B ATMOSPHERIC TRANSPORT AND DIFFUSION CHARACTERISTICS FOR INCIDENT ANALYSIS Conservative and realistic short-term CHl/Q values for a ground-level release have been computed for various time intervals at the swa mirvi area boundary (EAB), a crcle with a radius of 610 m, and the outer boundary of the low popuistion zone (LPZ), a circle with a radius of 9,600 m. Meteorological data collected on-site from January 1,1979 through December 31,1979 were used in the analysis. l Dilution factors were computed for each sequential hour of measured meteorological data and for l receptors positioned in each of the 16 downwind sectors. These hourly CHl/Q values were calculated using a modification of the Gaussian dispersion model outlined in Regulatory Guide 1.145. Plume conteriine values were used to determine the short-term dilution factors (up through 8 hours), and sector average values were used for the longer-term dilution factors. Using the hourly CHl/Q values calculated above, average CHl/Q values for each downwind sector were then determined for ser===ive overlapping time intervals of 1,2,8,24,96 and 720 hours corresponding to time periods of 0 to 1 hours,1 to 2 hours, O to 8 hours,8 to 24 hours,1 to 4 days and 4 to 30 days, respechvely. For each selected downwind sector and interval size, the averaging process began with the first hourly dilution value on record and was then repeated for the same interval size starting with each subsequent hour of dispersion data. In the averaging process, the only non-zero values within a given time interval which were considered in evaluating the average dilution factor for the interval were those hour's during which the wind was blowing into the downwind sector of interest. The average CHl/Q values I were then classified into groups as a function of interval size and downwind sector, and corresponding cumulative frequency distributions of non-zero values for each group were prepared. The CHl/Q value which was exceeded 0.5 percent of the total time was then determined for each sector j and the maximum value chosen as the maximum sector conservative CHl/Q. The median CHl/Q value for each sector was also determined and the maximum value among all the sectors was chosen as the ; maximum realistic CHl/Q. Overall site incident dilution factors were also computed for each of the time intervals of interest. These parameters were determined by Arzt developing arrays of sector-dependent CHl/Qs (averaged over selected time intervals) and then forming an equivalent sector-4ndependent array consisting of the maximum CHl/Os at equivalent locations in the sector dependent arrays. These maximum CHl/Qs were then used to form an overall-site cumulative distribution, from which the values at desired percentile points were determined. Rev.14 DSAR 5.B-1
MYAPC The CHl/Q value which was exceeded no more than 5 percent of the total tine was then determined and classified as the overall site conservative CHl/Q. The median CHl/Q value was used to represent the overall site realistic CHl/Q. l For each time interval, the CHl/Q value used for inodont evaluations is the higher of either the maximum sector of the overall site values. These are hsted in Table 5.B.1. O O Rev.14 DSAR 5.B-2 l l
MYAPC Table 5.B.1 DILUTION FACTORS FOR INCIDENT ANALYSIS AT MAINE YANKEE
- CHl/O (sec/m8)
Time Interval Conservative Realistic
- 1. Exclusion Area Boundary 0 - 1 hour 5.40 -4" 9.92 - 5 (610 meters) 1 - 2 hours 3.11 - 4 6.51 - 5 2 - 8 hours 1.56 - 4 4.56 - 5 8 -24 hours 5.68 - 5 2.19-5 1 - 4 days 2.87 - 5 1.55 - 5 4 - 30 days 1.36 - 5 1.07-5
- 11. Low Population Zone 0 - 1 hour 3.88 - 5 2.58 - 6 (9,600 meters) 1 - 2 hours 2.15- 5 1.91 - 6 2 - 8 hours 1.01 - 5 1.30 -6 8 - 24 hours 1.09 - 6 3.67- 7 1 - 4 days 5.46 - 7 2.66 - 7 4 - 30 days 2.42-7 1.87 - 7 Ill. Control Room 0 - 1 hours 5.88 - 3 1 - 2 hours 4.69- 3 2 - 8 hours 2.08 - 3 8 - 24 hours 1.15-3 1 - 4 days 5.97 - 3 4 - 30 days 4.43-3 l
Based on on-site meteorology from January 1979 - December 1979. l 5.40 - 4 = 5.40 x 10d O l DSAR 5.B-3 Rev.14 i
MYAPC SECTION 6.0 CONDUCT OF OPERATIONS TABLE OF CONTENTS Section Iltle East 6.1 Ra= nonsibili ty and Orc-rh a n .. .............. .. .. . ... . .. . . .. .. . .. . . .. ... .. ........ . ... . .... . 6-1 6.1.1 Duties and Responsibilities of the Operating Staff Personnel 6.1.2 Duties and Responsibilities of the Support Staff 6.2 Tech nical S oecifications ............. .... .............. . ... .... ..... .. .. . . . . . . . . . . . . . . . .. . . . . ... 6-8 6.3 Trainina................................................................................................... 6-8 6.4 Procedures.............................................................................................. 6-8 6.5 Proarams................................................................................................. 66 6.5.1 Emergency Plan , t 6.5.2 Security Plan i 6.5.3 Fire Protection Program 6.5.4 Fitness For Duty 6.5.5 Offsite Dose Calculation Manual 6.5.6 Quality Assurance Program 6.5.7 Process Control Program 6.6 R eview a n d Au d i t . . . . . . . . . . . . . . . . . . . . .. . . .. . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-12
.............. l i
6.4.1 Deleted l { 6.4.2 Deleted l l 6.4.3 Deleted l 3 Rev.17 DSAR 6-i
MYAPC SECTION 6.0 LIST OF TABLES Tabis No. Illie None O Rev.14 O DSAR 6-il
MYAPC SECTION 6.0 UST OF FIGURES Figure No. Iltla 6.1-1 Facility Organization O Rev.14
.i O DSAR 6-iii
)
1 l
i l MYAPC O ; U SECTION 6.0 l l CONDUCT OF OPERATIONS l Maine Yankee Atomic Power Company is responsible for all aspects of plant operation, including maintalaing the plant in the permanently defueled condition through termination of the existing j Operating Ucense. As a portion of this responsibility, Maine Yankee is also responsible for employing and training qualified personnel to operate the plant. j Technical or managerial assistance in the saWinn of this responsibility may be provkied through a number of extemal resources. Such assistance is provided by Entergy Nuclear, Inc, Stone & Webster Engineering Corporation, ABB-Combustion Engineering, Inc., Westinghouse Electric Company, and Duke Engineering & Services (formerly Yankee Atomic Electric Company, Nuclear Services Division). Other support, as required, may be arranged from time to time. l 6.1 Responsibility and Organization The functional organization and key lines of responsibility for the Maine Yankee plant staff are shown in Figure 6.1-1. j G U The on-site organization includes the technically trained personnel necessary to support all aspects of plant operation. Each member of the facility staff meets or exceeds the minimum qualifications of Regulatory Guide 1.8, dated September 1975. 6.1.1 Duties and Responsibilities of the Operating Staff Personnel Ooerations Director The Operations Director directs the daily plant activities and is responsible for the overall safe and efficient operation of the plant. Additionally, the Operations Director is responsible for the compliance of operations with the requirements of the facility license and Technical Specifications. This responsibility includes oversight of the Operations, Radiation Protection, Maintenance and Chemistry Departments and the Security and Training functions. The Operations Director may delegate these responsibilities to the appropriate Manager (s) in his absence. The Operations Director reports directly to the President and Chief Nuclear Officer. DSAR 6-1
MYAPC Manager. Ocarations Denartment/ Maintenance Department l The Manager, Operatens/ Maintenance directs the activities of the Opershons Department including l
' the operating shifts, personnel in training, and staff.
The Yc+;=, ~ Operations / Maintenance is responsible for E--C ,g the activities of the l Operations Department and plant operation, with other plant functions. Additionally, the Manager, Operations is also responsible for maintaining plant operation records as required by the facility license and Technical Specihuns. The Manager, Operations / Maintenance Department is l responsible for the training and qualifications of the plant operatens personnel. The Manager, Operations / Maintenance directs and coordinates the scheduling and supervising l mechanical, electrical, and instrument and control maintenance work with other plant functions in l the handling of routine and non-routine maintenance assignments. l l , The Manager, Operations / Maintenance is responsible for routine inspections, preventative l maintenance practicos and record keeping as required by the facility license and Maine Yankee j q policies and practices. Additionally, the Manager, Operations / Maintenance is responsible for the j training and qu alifications of maintenance personnel. l The Manager, Operations / Maintenance reports to the Operations Director. l Manager. Radiation Protection Department The Manager, Radiation Protection directs and coordinates the actmbes of the radiological controls and radiological program activities, including overall responsibility of implementation of the ALARA Program, at Maine Yankee. As such the Manager, Radiation Protection is responsible for storage and handling of radioactive material, radiaton safety, and the radioactive waste program. Additenelly, the Manager, Radiation
- PfcItection is responsible for the training and qualifications of radiation protection personnel.
The Manager, Radiation Protection reports to the Operations Director Rev.17
.DSAR 6-2 l
r MYAPC Chemistry Supervisor l Chemistry Supervisor is responsible for developing, maintaining and implementing the Radiological l Emuents Control Program, Hazardous Material cnd Waste programs, Uquid Radiological Waste processing, and providing general chemistry support to other departments. The Chemistry Supervisor reports to the Manager, Radiation Protection Department. l O O Rev.17 DSAR 6-3 l
MYAPC 6.1.2 Duties and Responsibilities of the Support Staff Vice President Adminsstration & Fmance The Vice President Administration & Finance has oversight responsibility for the development, maintenance, and implementation of policies, practices, and procedures addressing Budgets, Cash Management, Trusts, Information Technology, Accounting, and Administrative Support. The Vice President Admirustration & Finance reports to the President, Chief Nuclear Officer. Dractor Nuclear Safety & Regulatorv Affairs l The Director Nuclear Safety & Regulatory Affairs has overall responsibility for activities required to l maintain the permits and licenses required for the plant including production, mamtenance, and interpretation oflicensing documents. This includes oversight of the plant's design / licensing basis, compliance with applicable federal, state and local laws and regulations, and the company's interaction with state and federal technical regulatory agencies, including Emergency Preparedness. Additionally, the Director Nuclear Safety and Reguletory Affairs is responsible for ensuring l h implementation of the Quality Assurance Program, Corrective Action Program, and Worker Concems Program. The Director Nuclear Safety & Regulatory Affairs reports to the President, Chief Nuclear Officer. l l l l 1 Mananar. On=lity Procri.rris The Manager, Quality Programs is responsible for providing quality control coverage of maintenance and modification activities and developing and maintaining the Quality Assurance Program. The Manager, Quality Programs has the authority and independence to identify quality problems; initiate, recommend, or provide solutions to quality problems through designated channels; and verify implementation of solutions to quality problems. The Manager, Quality Programs also has the authority and responsibility to initiate stop work orders to responsible management, as necessary, for any condition adverse to quality. Rev.17 DSAR 6-4
MYAPC e
. Additionally, the Manager, Quality Programs has overall responsibility for development, maintenance, and implementation of, the Corrective Action Program.
- The Manager, Quality Programs drectly reports to the Director Nuclear Safety & Reguistory Affairs l Additionally, the Manager, Quality Programs is a functional report to the President, Chief Nuclear Officer.
Vice Pr==Want Da -.m&=hnina l The Vice President Decommissioning is responsible for developing and implementing projects and l programs necessary to ensure the safe and economical decommissioning of the plant. , The Vice President Decommissioning reports to the President, Chief Nuclear Officer. l Engineerina Manaaer l The Engineering Manager is responsible for providing engineering assistance to the plant operations l and support staffs, inillation and implementation of key plant modifications, maintenance of the plant design basis, design configuration data and documentation, and oversight and direction of the Civil / Seismic Program. The Engineering Manager reports to the Vice President Decommissioning. l Vim President - Leaal and Govemment Affairs l The Vice President, Legal and Govemment Affairs coordinates and/or provides legal services to l Maine Yankee. The Vice President, Legal and Govemment Affairs acts as Maine Yankee's lead l legislative representative in the supervision and direction of outside legal work. The Vice President, Legal and Govemment Affairs represents Maine Yankee before judicial and l regulatory bodies and is responsible for development, preparation and participation in these proceedings; assists in developing Maine Yankee's position on matters before the Legislature in consultation with other members of management; advises and assists Maina Yankee with wiiviience with statutes, rules, regulations, licenses, permits and decisions by regulatory and other govemmental bodies; researches legal issues, assembles factual data and documentation, prepares and drafts legal Rev.17 DSAR 6-5
~
O documents, analyzes and organizes documentary evidence; works with other members of management in the development of strategies and policies to meet corporate objectives; drafts, reviews, and approves contracts and other legal documents; retains and directs services from outside counsel, as deemed necessary. The Vice President, Legal and Govemment A# airs is responsible for the develcpment, maintenance, l implementation, and oversight of policies, pracbces, and procedures addressing human resource management which meet the needs of Maine Yankee and comply with applicable local, state, and federal regulations. The Vice President, Legal and Govemment A# airs reports to the President, Chief Nuclear Officer. l Rev.17 O DSAR 6-6
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Mame Yankee Atomic Power Station is govemed by the Technical Speedications which are provided as Appendix A to Operating License No. DPR-36, Docket No. 50-309. 6.3 Training Programs are conducted to train plant personnel. Key technical and operating personnel receive on-site classroom or guided self-study and on-the-job trainng Appropriate plant personnel receive instruction in emergency plan and radiation protection procedures. Specialized training in specific areas conducted by the equipment manufacturers or other vendors is utilized as necessary. Training on a continuing basis is used to maintain a high level of proficiency in the' staff. 6.4 Procedures Written procedures are required for maintenance, repair, or operational activities related to the O. structures, systems and components which are safety related (Safety Class 1, 2, or 3) or identified as important To the Defueled Condition (ITDC) as defined in Section 3.1. In addition, written procedures are required for the movement of any spent fuel storage racks, to ensure that no rack modules are moved over fuel assemblies. Written procedures shall be established, implemented and maintained per Technical Specifications. l l 6.5 Programs 6.5.1 Emergency Plan j The Maine Yankee Emergency Plan is docketed as a separate document. Changes to the Emergency Plan are evaluated under 10 CFR 50.54 (q) which allow changes to be made without regustory approvalif these changes do not decrease the effectiveness of the plan and the plan, as changed, continues to meet the standards of 10 CFR 50.47(b) and the requirements of 10 CFR 50 Appendix E. A report of such changes is required to be submitted to the NRC within 30 days after the change is made. l Rev.17 DSAR 6-8
MYAPC 6.5.2 Security Plan The Maine Yankee Physical Secunty Plan is docketed as a separate document and is required by 10 CFR 50.34(c),50.34(d), and 10 CFR 73. Changes to the Physical Security Plan are evaluated under 10 CFR 50.54 (p) which allows changes to be made without regulatory approval if these changes do not decrease the safeguards effectiveness of the plan. A report of such changes is required to be submitted to the NRC within two months after the change is made. l 6.5.3 Fire Protection Program Upon docketing of the permanently defueled certifications specified under 10 CFR 50.82 (a) (1), only the requirements of 10 CFR 50.48(f) apply to Maine Yankee. These requirements address the potential for fires which could cause the release or spread of radioactive materials (i.e., which could result in a radiological hazard) including the prevention, detachon, control and extinguishing of such fires such that the risk of fire-induced radiological hazards to the public, environment and plant personnelis minimized, n The Maine Yankee Fire Protection Program describes how Maine Yankee complies with and meets d the objectives of 10 CFR 50.48(f) and describes the fire detection and suppression systems. The Fire Protection Program includes provisions for periodic assessments to ensure that the Program is being maintained and is appropriate throughout the various stages of facility decommissioning. A fire suppression water system consists of: a water source (s); gravity tank (s) or pumps; and dis +,ribution piping with associated sectionalizing control or isolation valves. l Changes to the Fire Protection Program are evaluated under 10 CFR 50.48(f)(3) which allows changes to be made without regulatory approvalif these changes do not reduce the effectiveness of fire protection for facilities, systems and equipment which could result in a radiological hazard, taking into account the decommissioning plant conditions and activities. 6.5.4 Fitness For Duty Program l As a result of the permanently shutdown condition of the plant and the 10 CFR 50.82(a)(1) certifications, the NRC has concluded that the Fitness for Duty Program rule,10 CFR 26, no longer applies to Maine Yankee, Reference 1. O' . Rev.17 DSAR 6-9
MYAPC 6.5.5 Offsite Dose Calculation Manual The Maine Yankee Offsite Dose Calculation Manual (ODCM) is defined by Technical Specifications to contain the methodology and parameters used in the calculation of off-site doses resulting from radioactive gaseous and liquid emuents, in the calculation of gaseous and liquid emuent monitoring Alarm / Trip Setpoints and in the conduct of the Environmental Radiological Monitoring Program. The ODCM shall also contain the Radioactive Emuent Control and Radiological Environmental Monitoring Programs and descriptions of the information that should be included in the Annual Radiological Environmental Operating and Annual Radioactive Emuent Release Reports. Changes to the ODCM shall be made in accordance with Maine Yankee Technical Specifications and may be made if the change will maintain the level of radioactive emuent control required by applicable regulations and not adversely impact the accuracy or reliability of emuent, dose, or setpoint calculations. 6.5.6 Quality Assurance Program The Maine Yankee Quality Assurance Program is docketed as a separate document and is required v' by 10 CFR 50.54(a). Changes to the Quality Assurance Plan are evaluated under 10 CFR 50.54 (a) which allows changes to be made without NRC approval if these changes do not reduce the commitments in the program description previously accepted by the NRC. These changes must be submitted to the NRC in accordance with the requirements of 50.71(e), FSAR update requirements. 6.5.7 Process Control Program The Process Control Program (PCP) contains the current formulas, sampling analyses, tests and determinations to be made to ensure that processing and packaging of solid radioactive wastes I DSAR 6-10 Rev.15
. .. . , .L, MYAPC based on demonstrated processing of actual or simulated wet solid wastes will be accuiTpished in such a way as to assure compliance with 10 CFR Parts 20,61, and 71, State regulations, burial ground requirements, and other requirements goveming the disposal of solid radioactive waste. Dry l active waste (DAW) such as compacted trash and contaminated components are not included in l the scope of the PCP. Written procedures are established, implemented, and maintained covering the key activities of the Process Control Program.
Changes to the PCP shall be documented and records of reviews performed shall be retained as required by Technical Spoolfications. This documentation shall contain: I a) Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change (s) and b) A determination that the change will maintain the overall conformance of the solidified waste product to existing requirements of Federal State or other applicable regulations. O i O Rev.14 DSAR 6-11
MYAPC 6.6 Review and Audit The Review and Audit functions, the responsibility and the committees are desenbod in l Section I of the Quality Assurance Program. I O l
@ .Rev.17 l 1
DSAR 6-12 1 i
MYAPC l l
References:
- 1. Letter, USNRC to Maine Yankee, " Fitness for Duty Programs (10 CFR 26) for Maine
' Yankee Atomic Power Station", dated January 12,1998.
O l i O DSAR 6-13 Rev.15 l
MYAPC Q U 7.1.1 Decommissioning Approach The following is an overview of the Maine Yankee decommissioning plans. The detailed planning for each activity will be completed prior to the initiation of that activity. The list of activities discussed below is preliminary and items may be modified or eliminated based on detailed planning and cost benefit studies. 7.1.1.1 Planning Planning and preparation for decommissioning will include the following general types of activities: Development and selection of the decommissioning organization. Review and reclassification of systems, structures, and components. Review and revision of plant programs and procedures.
. Review and revision of the plant licensing basis.
Design of a long term approach to spent fuel poc! cooling with isolation of the spent fuel pool from the remainder of the plant. Preparation of detailed decommissioning procedures and cost estimates. 7.1.1.2 Site Characterization A detailed site characterization has been conducted. Surveys will estabhsh the contamination and ] radiation levels throughout the facility. This information will be used in developing the detailed procedures to ensure that the contaminated materials are removed and to ensure that worker exposure is maintained ALARA. For the same purpose, surveys of the outdoor areas have been ] performed to confirm the locations of known contaminated soil and to identify any previously unknown contaminated soils. 7.1.1.3 Decontamination Exterior and interior surfaces of components will be decontaminated. The objectives of the decontamination effort are to minimize the radiation exposure to personnel, and to minimize the amount of radioactive waste. Rev.17 r ( u DSAR 7-2 1
n MYAPC 7.1.1.3.1 Chemical Decontamination Chemical decontamination methods have been used to clean the interior surfaces of systems, such ] as RCS and associated systems. Decontamination methods will involve standard processes with well understood chemical interactions, and the resulting waste has been administratively controlled ] and disposed in accordance with plant procedures and applicable federal and state regulations. 7.1.1.3.2
- Ace Cleaning Several methods will be used on extenor surface contamination. These may include: sweeping compounds, household detergents, and high velocity water jets. If the surface contamination '
cannot be removed, the surface material may be removed for disposal as radioactive waste. 7.1.1.3.3 Structural Material Removal Material surfaces, such as concrete, may be contaminated to a depth of several centimeters. Surfaces which cannot be cleaned will be removed and packaged for disposal as rad'osctive waste. Several methods are available to remove concrete and steel. These methods include: concrete surface scrabbling, flame cutting, blasting, thermic lance cutting, core boring, and rock splitting. Each area will be evaluated to determine the appropriate method for removal of materials. Some ] material will be rubblized and buried on site if it meets applicable regulatory requirements for final ] site release. ] 7.1.1.4 Major Decommissioning Activities A major decommissioning activity is any activity that results in permanent removal of major radioactive components, permanently modifies the structures of the containment, or results in dismantling components for shipment containing Greater-Than-Class-C waste in accordance with 10 CFR 61.55. The major activities are summarized as follows:
- Disassembly, segmentation, and packaging of the reactor intemals.
. Removal of the steam generators and the pressurizer. Extemal surfaces will be decontaminated and all openings will be seal-welded closed. These components will act as their own disposal containers. . Segmentation of Greater-Than-Class-C components for storage with the spent fuel. . Segmentation and packaging of the reactor vessel, or preparation of the vessel for shipment intact.
Decontamination, segmentation, and disposal of RCS and other larger-bore piping. Rev.17 DSAR 7-3
MYAPC A.2 Current on-site "" :+:- =+=1 Prwie.T. A.2.1 Introduction
'The on-site meteorological data conectional program was upgraded in late 1976 to meet the intent of Revision 0 of Regulatory Guide 1.23. This report describes the current on-site monitoring program and presents wind and statulity data summaries for one full year of operation; January 1, 1979 through December 31,1979. A discussion of the data summaries is included, and a comparison is made between the initial (July 1967 - June 1968) and the current (January 1979 -
December 1979) data bases. It is concluded that results from both programs are compatible, and that both programs produced data bases which are representative of site meteorology. A.2.2 Description of the Monitoring Program The current meteorological monitonng system utilizes a guyed 200-foot tower located on-site as shown in Figure A.2-1. Instrumentation on the tower is located on booms at the 33-foot and 195-foot levels. Wind measurements are observed at heights of 35 feet and 197 feet above the tower base (two feet above both booms). Both wind speed and wind direction are measured at each height. Ambient temperature difference is measured on the tower between 32 feet and 194 feet (one foot below both booms). Ambient temperature is also measured by this system for the 32-foot level. A digital recording system is the primary data collection mechanism for the Maine Yankee Meteorological System. The data collection mechanism is provided by the Control Room PLC. ] l A.2.3 Results i The data base used to compile the 1979 data summaries which follow represents hourly averaged I data digitized from strip charts (the 1979 data base was collected before the plant computer began collecting data digitally). The 1979 data recovery rates, which are well above the Regulatory Guide 1.23 goal of 90 percent, are presented in Table A.2.1. Rev.17 O DSAR 4 A-34 l l}}