ML20236X175
| ML20236X175 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 08/03/1998 |
| From: | Maine Yankee |
| To: | |
| Shared Package | |
| ML20236X169 | List: |
| References | |
| NUDOCS 9808070179 | |
| Download: ML20236X175 (170) | |
Text
{{#Wiki_filter:28 THRU 38 - NRC HEADQUARTERS Proc. No. 0-06-1 R;v. No. 4 DOCUMENT CONTROL DESK Page 5 of 6 j [ COPY NUMBER: ATTACHMENT A MAINE YANKEE CONTROLLED DOCUMENT TRANSMITTAL FORM
- DOCUMENT
- DEFUELED SAFETY ANALYSIS REPORT - REVISION 16 l
l TRANSMITTAL ISSUE DATE: 07-29-98 TRANSMMTAL RETURN DATE*: 08-26-98 l
- 1. Please remove existing, List of Effective Pages of the DSAR MANUAL, Rev. No.15, and replace with the attached List of Effective Pages, Rev. No.16. 5 pages.
- 2. Please remove the following pages and replace with the attached pages, Rev. No.16.
Table of Contents: Iv and v. , Section 1.0: 1-3,1-4,1-5,1-7,1-8, and 1-9. Section 3.0: Entire contents is being replaced. Section 4.0: 4-11,4-3,4-9,4-12,4-13,4-14,4-15,4-16,4-17,4-18, and 4-19. Section 5.0: 5-22, 5-24, 5-25, and 5-31. Section 7.0: 7-9. Appendix A: A-34. The above listed document has been inserted into the assigned manual / file and all superseded pages have been destroyed. MANUAUFILE UPDATED BY: [ Please Print Name DATE: Signature CAUTION
- Manual Holders who do not sign and return this transmittal form to Document Control on or before the required return date may be required to return their controlled manual (s) to Document Control. Reissuance shall require Department Manager or higher management approval.
Please retum to: MAINE YANKEE ATOMIC POWER COMPANY Document Contro! Center ! P.O. Box 408 Wiscasset, Maine 04578 9808070179 980803 ~ PDR ADOCK 05000309 W PDR
MYAPC LIST OF EFFECTIVE PAGES PAGE REV REMARKS PAGE REV REMARKS PAGE REV REMARKS TABLE OF CONTENTS SECTION 1.0 SECTION 2.0 1 14 1-i 14 Table of Contents 2-i 14 Table of Contents il 14 1-ii 14 Ust of Tables 2-il 14 Table of Contents lii 14 1-iii 14 Ust of Figures 2-lii 14 Ust of Tables iv 16 11 14 2-iv 14 Ust of Figures v 16 1-2 14 2-1 14 vi 14 1-3 16 22 14 vil 14 1-4 16 2-3 14 viii 14 1-5 16 2-4 14 lx 14 1-6 14 2-5 14 Table 2.1.1 1-7 16 2-6 14 Table 2.1.2 1-8 16 2-7 14 Figure 2.1-1 1-9 16 2-8 14 Figure 2.1-2 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 Figure 2.1-5 1-13 14 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 14 Table 1.3.1 2-14 14 Figure 2.1-8 1-16 14 Table 1.3.1 2 15 14 1-17 14 Table 1.3.1 2-16 14 1-18 14 Figure 1.3-1 2-17 14 1-19 14 Figure 1.3-2 2-18 14 Figure 1.3-3 2-19 ( 1-20 1-21 14 14 Figure 1.3-4 2-20 14 14 1 22 14 2-21 14 1-23 15 2-22 14 i 2-23 14 2-24 14 i 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 l 2-30 14 Table 2.2.5 2-31 14 Table 2.2.6 2-32 14 Table 2.2.7 2-33 14 Table 2.2.8 2-34 14 Table 2.2.9 2-35 14 Table 2.2.10
, 2-36 14 Table 2.2.10 2-37 14 Figure 2.2-1 2-38 14 Figure 2.2 2 2-39 14 Figure 2.2-3 2-40 14 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 DSAR Rev.16 l _ _ _ _ - _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ - _ _ _ _ _ _ _ _ _ . _ _ _ _ _ . _ _ _ _ _ _ _ _ _
l l l l MYAPC LIST OF EFFECTIVE PAGES PAGE REV REMARKS PAGE REV REMARKS t SECTION 3.0 PAGE REV REMARKS 3-45 14 Figure 3.2-7 i 3-1 14 Table of Contents 3 46 14 Figure 3.2-8 2-47 14 3-il 16 Table of Contents 3-47 16 Figure 3.2-9 2-48 14 3-lii 16 Table of Contents 3 48 14 Figure 3.2-10 2-49 14 3-iv 14 List of Tables 3-49 14 Figure 3.2-11 l 2-50 14 3-v 16 List of Figures 3-50 14 Figure 3.2-12 2 51 14 3-vi 14 Listof Figures 3-51 14 Figure 3.2-13 2-52 14 31 14 3-52 14 2-53 14 Table 2.3.1 3-2 14 3-53 16 2-54 14 Table 2.3.2 3-3 14 3-54 16 2-55 14 Figure 2.3-1 3-4 14 3-55 14 2-56 14 Figure 2.3-2 3-5 14 3-56 14 2-57 14 Figure 2.3-3 3-6 16 3-57 16 2-58 14 Figure 2.3-4 3-7 16 3-58 14 2-59 14 Figure 2.3-5 3-8 16 3-59 16 2-60 14 Figure 2.3-6 3-9 16 3-60 16 i 2-61 14 Figure 2.3 7 3-10 14 3-61 16 l 2-62 14 Figure 2.3-8 3 11 14 3-62 14 2-63 14 3 12 14 3-63 14 , 2-64 14 3-13 15 3-64 16 ) 2-65 14 Figure 2.4-1 3 14 15 3-65 16 2-66 14 3-15 15 3-66 14 m 2-67 2-68 14 14 3 16 3-17 14 14 3-67 3-68 14 14 2-69 14 Figure 2.51 3-18 14 3-69 14 2 70 14 Figure 2.5-2 3-19 16 3-70 14 2 71 14 Figure 2.5-3 3-20 16 3 71 16 2 72 14 Figure 2.5-4 3-21 14 3-72 14 2 73 14 Figure 2.5-5 3 22 14 3-73 16 1 3-23 14 3-74 16 3-24 14 3 75 16 3-25 14 Table 3.1.1 3-76 16 3-26 14 Figure 3.1 1 3-77 16 3-27 14 Figure 3.1-2 3 78 16 3-28 16 3-79 16 3 29 15 3-80 14 3-30 14 3-81 14 Figure 3.31 3-31 14 3-82 14 Figure 3.3-2 3-32 14 3-83 14 Figure 3.3-3 3-33 14 3-84 14 Figure 3.3-4
'3-34 14 3-85 14 Figure 3.3-5 3-35 14 3-86 14 Figurs 3.3-6 3-36 16 3-87 14 Figure 3.3 7 3-37 14 3-88 14 3-38 14 3-89 14 3-39 16 Figure 3.2-1 3-90 14 1-40 16 Figure 3.2-2 3-91 14 3-41 14 Figure 3.2-3 3-92 14 3-42 14 Figure 3.2-4 3-93 14 Figure 3.3-8 3-43 14 Figure 3.2 5 3-94 16 3-44 14 Figure 3.2-6 3-95 16 I
O 3 96 16 DSAR Rev.16
l MYAPC LIST OF EFFECTIVE PAGES PAGE REV REMARKS PAGE REV REMARKS PAGE REV REMARKS 3-97 16 SECTION 4.0 SECTION 5.0 3-98 18 l 3-99 16 4-1 14 Table of Contents 5-1 14 Table of Contents 3-100 16 4-il 16 Table of Contents 5-ii 14 L4r of Tables 3-101 16 Figure 3.3-9 4-iii 14 List of Tables 5-iil 14 List of Figures 3-102 16 4-iv 14 Listof Finures 5-1 15 3-103 16 4-1 14 5-2 14 3-104 16 4-2 14 5-3 14 3 105 14 Figure 3.3-12 4-3 16 5-4 14 3 106 16 4-4 14 5-5 14 3-107 16 4-5 14 5-6 14 3-108 16 4-6 14 5-7 14 1 3-109 16 4-7 14 5-8 14 Table 5.2.1 3-110 16 4-8 14 5-9 15 3-111 16 4-9 16 5-10 14 3-112 14 Table 3.3.1 4-10 14 Table 4.4.1 5-11 15 3 113 14 Figure 3.3-13 4-11 14 5-12 14 3-114 14 Figure 3.314 4-12 16 5 13 14 Table 5.3.1 3 115 14 Figure 3.3-15 4-13 16 5-14 14 Table 5.3.2 3-116 14 Figure 3.316 4-14 16 5-15 14 Table 5.3.3 3 117 14 Figure 3.3-17 4 15 16 5-16 15 3-118 14 Figure 3.3-18 4-16 16 5 17 14 3-119 16 4-17 16 5-18 15 3 120 16 4 18 16 Table 4.6.1 5-19 14
'0' 3-121 16 3-122 16 3-123 14 4 19 16 Table 4.6.1 5-20 5-21 5-22 14 14 16 3-124 14 Figure 3.3-19 5-23 14 3-125 14 Figure 3.3-20 5-24 16 3-126 14 5-25 16 3 127 16 5-26 14 3 128 16 5-27 14 Table 5.5.1 3 129 16 5-26 14 Table 5.5.2 3 130 14 Figure 3.3-21 5-29 14 Table 5.5.2 3-131 14 Figure 3.3-22 5-30 14 Table 5.5.2 3-132 16 5-31 16 Table 5.5.2 3 133 16 5-32 14 Table 5.5.3 3 134 14 5-33 14 Figure 5.5-1 3 135 16 Figure 3.3-23 5-34 14 Figure 5.5-2 3-136 16 5-35 14 Figure 5.5-3 3-137 16 5-36 14 Figure 5.5-4 3-138 16 5-37 15 3-139 64 5-38 15 3 140 14 5-39 14 3-141 16 Figure 3.3-24 5-40 14 Table 5.6.1 3 142 14 Figure 3.3-25 5-41 14 Table 5.6.2 3-143 16 5-42 14 3-144 16 5-43 14
- 3-145 16 5A 1 14 3-146 16 SA-2 14 3 147 16 5A-3 14 A 3-148 14 5A-4 14 Table 5.A.1 Q 3-149 16 3 150 14 3-151 14
! DSAR Rev.16 l
l MYAPC LIST OF EFFECTIVE PAGES PAGE REV REMARKS PAGE REV REMARKS PAGE REV REMARKS 58 1 14 SECTION 6.0 SECTION 7.0 l 58 2 14 < 5B 14 Table 5.B.1 6-i 14 Table of Contents 7-i 14 Table of Contents 6-il 14 Ust of Tables 7-il 14 Ust of Tables ! 6-iii 14 Ust of Figures 7-iii 14 Ust of Figures 6-1 14 7-1 14 6-2 14 7-2 14 6-3 14 7-3 14 6-4 14 7-4 14 6-5 14 7-5 14 6-6 14 7-6 14 67 14 Figure 6.1-1 7-7 14 6-8 14 7-8 14 6-9 14 7-9 16 6-10 15 7-10 14 6 11 14 7-11 14 6-12 14 7-12 14 6-13 14 7-13 14 6-14 15 7-14 14 7-15 14 7-16 14 7-17 14 i i l I l k l DSAh Rev.16 l
MYAPC O LIST OF EFFECTIVE PAGES V PAGE REV REMARKS PAGE REV REMARKS APENDIX A A-l 14 Table of Contents A-47 14 Table A.2.3 A-il 14 List of Tables A-48 14 Table A.2.3 A-lil 14 List of Figures A-49 14 Table A.2.3 A1 14 A-50 14 Table A.2.3 A-2 14 A-51 14 Table A.2.3 A-3 14 A-52 14 Table A.2.3 A-4 14 A-53 14 Table A.2.4 A-5 14 Table A.1.1 A-54 14 - Table A.2.5 A-6 14 Table A.1.2 A-55 14 Table A.2.6 A-7 14 Table A.1.3 A-56 .14 Table A.2.7 A-8 14 Table A.1.4 A-57 14 Figure A.2-1 A-9 14 Table A.1.5 A-58 14 Figure A.2-2 A-10 14 Table A.1.6 A-59 14 Fgure A.2-3 A-11 14 Table A.1.7 A-60 14 Figure A.2-4 A-12 14 Table A.1.8 A-61 14 Fgure A.2-5 A-13 14 Table A.1.9 A-62 14 Figure A.2-6 A-14 14 Table A.1.10 A-63 14 Figure A.2-7 A-15 14 Table A.1.11 _ A-64 14 Figure A.2-8 A-16 14 Table A.1.12 A-65 14 Figure A.2-9 A 17 14 Table A.1.13 A-66 14 Fgure A.210 A-18 14 Table A.1.14 A-67 14 Figure A.211 p-) 5 A-19
- A-20 14 14 Table A.1.15 Table A.1.16 V A-21 14 Table A.1.17 A-22 14 Figure A.1-1 A-23 14 Figure A.12 l A-24 14 Figure A.1-3 l A-25 14 Figure A.1-4 i A-26 14 Figure A.1-5 j A-27 14 Figure A.1-6 A-28 14 Figure A.1-7 A-29 14 Figure A.1-8 A-30 14 Figure A.19 A-31 14 Figure A.1-10 A-32 14 Figure A.1-11 A-33 14 Figuia A.112 A-34 16 A-35 14 l _ A-36 14 Table A.2.1 A-37 14 Table A.2.2 l' A-38 14 Table A.2.2 A 39 14 Table A.2.2 l A-40 14 Table A.2.2 l A-41 14 Table A.2.2 A 14 Table A.2.2 A-43 14 Table A.2.2 A-44 14 Table A.2.2 ,
A-45 14 Table A.2.3 l A-46 14 Table A.2.3 d
-DSAR Rev.16 L______________________________-._________ _ _ __ _ _ _ _ _ _ _ . _ _ _ _ _
i l MYAPC Defueled Safety Analysis Report TABLE 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
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3.3.1.4 System Operation l 3.3.1.5 Monitoring and Instrumentation l 3.3.2 Fuel Handling System l 3.3.2.1 Design Basis 3.3.2.2 System Description ! 3.3.2.3 Design Evaluation l 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 i 3.3.3.4 System Operation 3.3.3.5 Monitoring and Instrumentation 3.3.4 Service Water 3.3.4.1 Design Basis 3.3.4.2 System Description 3.3.4.3 Design Evaluation 3.3.4.4 System Operation 3.3.4.5 Monitoring and Instrumentation 3.3.5 Ventilation Systems 3.3.5.1 Fuel Building Ventilation System 3.3.5.2 Control Room Ventilation System 3.3.5.3 Auxiliary Ventilation Systems l l i Rev.16 DSAR iv L___-------_-_--------------------_- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - . - - - - -
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t MYAPC O' Defueled Safety Analysis Report TABLE OF CONTENTS (continued) 3.0 FACILITY DESIGN AND OPERATION (continued) 3.3.6 Auxiliary Systems 3.3.6.1 Compressed Air 3.3.6.2 Boric Acid Makeup 3.3.6.3 Primary Water System 3.3.6.4 Primary Vent and Drain System 3.3.6.5 Radiological Waste Processing System 3.3.6.6 Fire Protection System 3.3.6.7 Meteorological Instrumentation 3.3.7 Electrical System l 3.3.7.1 Station Offsite Power l 3.3.7.2 Station Onsite Power l 3.3.7.3 Spent Fuel Pool Island and Balance of Plant Electrical l Distribution System l 3.3.7.4 Plant Conputer and Programmatic Logic Controller (PLC) l 3.4 Control of Heaw Loads Ci V DSAR v l
1 MYAPC regulations. The systems and components of the facility are designed to enable the facility to withstand the traditionally defined extemal forces that may be imposed by natural phenomena, without loss of the capability to protect the public. The protected area is enclosed within a security fence with all access controlled through a guardhouse. In addition, intrusion barriers prevent breech of the fence by vehicles. In the permanently defueled condition, two power sources are available to provide for spent fuel cooling / makeup, fire protection, security and emergency preparedness functions. 1.2.2 Fuel Handling System Fuel handling and storage facilities are provided for the safe handling, storage and shipment of both new and spent fuel and are designed to preclude accidental criticality. The fuel building houses a new-fuel unloading area, a new-fuel storage room, a spent fuel pool and the necessary cranes required for the handling of the fuel assemblies. Fuel shall not be moved within the ) Containment Building. The equipment decontamination area and the spent fuel pool support ] systems are also located in the fuel building. The spent fuel movable platform and hoist is a traveling bridge which spans the spent fuel pool and moves on rails over useable spent fuel storage locations. The hoist hook supports hand!ing tools for grappling and moving fuel assemblies at a safe depth below the operational water level. The design of the fuel handling system precludes the exposure of operating personnel to abnormal radiation fields as the result of operational transients in the spent fuel pool. Likewise, the design and operation of the fuel handling system prevents the accidental exposure of a spent fuel assembly to a position above or near the water surface. New or unirradiated fuel assemblies have been removed from the Maine Yankee site and therefore do not require a handling system. ]
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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 building adjacent to the reactor containment. The poolis constructed of reinforced concrete with a wall and floor lining of 1/4-inch thick stainless steel. The walls of the spent fuel pool are approximately 6 feet thick. The floor of the spent fuel pool rests on bedrock.- The fuel transfer tube, which is 36 inches in diameter, connects the containment refueling cavity with the fuel pool. The O Rev.16 DSAR 1-3 1 L____-______-____-_-__________
- I MYAPC transfer tube is closed by a manual isolation gate valve located in the spent fuel pool and by a blind flange located inside the containment.
Following the cessation of power opt, ration and the removal of fuel from the vessel, the spent fuel inventory consists of a total of 1432 complete spent fuel assemblies,2 partially consolidated assemblies, and 2 partially full failed fuel rod containers residing in the spent fuel pool. Additionally, the spent fuel pool contains a number of trash baskets containing filters from prior pool and refueling cavity vacuuming efforts, reactor startup neutron sources, Control Element Assemblies (CEAs), CEA plugs, incore Instrumentation (ICl) thimbies, and ICI assembly tips. ]
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The spent fuel is stored in a single tier rectilinear array of free standing modules. Each fuel assembly, failed fuel rod container, or trash basket is contained within an individual celi. Cells are grouped together to form the free standing modules or racks. These high density spent fuel storage racks are designed for the passive reactivity control of the spent fuel in storage through the use of a fixed neutron absorber material and the spacing between assemblies. Design accommodations for cooling the spent fuel while in storage in the racks is provided through the use of flow holes in the bottom of each fuel storage cell. The spent fuel storage racks are designed to maintain the fuel in a coolable, subcritical geometry during accident conditions in the pool. The spent fuel pool is filled with a borated water solution to provide a medium for cooling the spent fuel, shielding for workers and the public from normal and accidental radiation exposure, and as a means of controlling the spent fuel reactivity. The spent fuel pool cooling system removes the spent fuel decay heat stored in the spent fuel pool by circulating the borated pool water through a heat exchanger. Each spent fuel pool cooling pump , takes suction from the fuel pool, circulates the water through a heat exchanger, and retums it to ' l the fuel pool below normal water level. Decay Heat Removal (DHR) water flows through the shell ] side of the heat exchanger and cools the tube side borated pool water. The DHR water is cooled ] ; by the DHR air-cooled heat exchangers located north of the BWST dikes. In the event that the ] DHR system is not available, fire protection system water may be substituted by manually ) l connecting a hose to a flange connection on the shell side of the fuel pool heat exchanger. Rev.16 O DSAR 14
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l 1 C_ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ .
MYAPC The fuel pool cooling system also has a purification loop consisting of a pump and two filters which ] fnay be operated independently of the fuel pool cooling system. Flow from the discharge of the fuel pool purification pump is directed through the fuel pool profilter, and postfilter, in that order, while ] the retum is through the fuel pool cooling retum line. The purification pump can also be used for skimming operation or to circulate pool water through the heat exchanger. A submerged, self- ] contained domineralizer is installed in the spent fuel pool. ] f Cooling water make-up to the spent fuel pool is available from a variety of sources. These sources
- include the Primary Water Storage Tank through the Primary Water system, and, as emergency ]
sources, the fire water system and potable water from the town of Wiscasset water supply system. Boron make-up is available from the self contained, portable, SFPI boron addition tank, mixer and ] pump located in the fuel building. ] 1.2.4 Radiological Waste Treatment System I The radioactive waste treatment system is designed so the discharge of radioactivity to the environment will be minimized and at all time will be in accordance with the requirements of 10 l CFR 20.
. The radioactive waste treatment system is designed to collect, store, process, monitor and dispose of solid, liquid and gaseous radioactive wastes from the plant. The pnncipal design objective of this j system is to ensure that the general public is protected from exposure to radioactive waste l products in accordance with 10 CFR 20. The normal sources of radioactive waste in the defueled condition includes those wastes generated from decontamination and decommissioning activities and residual radioactive precipitant / contaminants resulting from the storage of spent fuel in the spent fuel pool. The radioactive waste treatment system is designed to safely process radioactive l
! wastes. 1.2.5 Radiological Waste Storage and Disposal System l Liquid waste may be mixed with a solidification agent, containerized into an approved container, solidified, and sealed in approved shipping containers and allowed to solidify before shipment to i an approved waste disposal site. Altematively, liquid waste may be released in accordance with . the limits set in the Off-Site Dose Calculation Manual. l Rev.16 DSAR 15
'O MYAPC U
The minimum exclusion radius for the site is slightly greater than 2,000 feet. The outer boundary of the Low Population Zone, as defined in the 10 CFR 100 regulations, is 6 miles from the plant. Within a 5 mile radius of the plant site, the resident population density is estimated to average 72 persons per square mile (1990). The nearest population grouping within 5 miles is situated around the Town of Wiscasset,3.9 miles NNE of the site. The town of Wiscasset has a population of about 3,340 people (1990). The surrounding towns of Edgecomb, Boothbay, Woolwich, and Westport Island, wholly or partially within a 5 mile radius, have a population of 6874 persons (1990). The city of Lewiston, 24 miles WNW from the plant site, is the nearest city with a population in excess of 25,000. The site characteristics are discussed in detail in Section 2. I 1.3.2 Structures l The major structures on the site are the reactor containment, primary auxiliary building, fuel building, turbine building, service building and circulating water pump house. Figures 1.3-1 and i fi 1.3-2 depict the major structures along with administrative office buildings, storage buildings and O other structures. The reactor containment is a steel-lined reinforced concrete cylinder with a hemispherical dome and an essentially flat reinforced concrete foundation mat. It served as a confinement barrier during plant operations and provides adequate radiation shielding for any defueled or decommissioning condition. 1 The turbine building houses the turbine generator, the component cooling water heat exchangers l and pumps, and the two diesel generators. The turbine building, and related components, are discussed in detail in Section 3. The service building contains the main control room, switchgear rooms, shops and employee facilities. The service building, and related components, are discussed in detail in Section 3. The primary auxiliary building houses equipment used for purification and processing of water from ; the reactor coolant system. The fuel building ventilation / filtration system is discussed in detail in ] Section 3. Rev.16
/3 DSAR 1-7
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t MYAPC The fuel building provides space for the storage of new and spent reactor fuel and also waste disposal equipment. The fuel building arrangement is discussed in detail in Section 3. The circulating water pump house contains the circulating water and service water pumps. Its arrangement is discussed in detail in Section 3. The seismic criteria to be used in the design of the structures and equipment in the station are described in Section 3.1. 1.3.3 Chemical Treatment in the permanently defueled condition, a self-contained portable batch tank and pump will be used ] to maintain boron concentration in the spent fuel pool. Spent fuel pool water is periodically tested to assure the desired quality. 1.3.4 Process Instruments q Temperature, pressure, flow and liquid level monitoring is provided as required to keep the operating V. personnel informed of plant and/or fuel storage conditions and to provide information from which plant processes can be evaluated or regulated. Instrument signal transmission for the plant instruments is electric. The plant gaseous and liquid effluents are monitored for radioactivity. The results of this monitoring are displayed and high values are annunciated. Area radiation monitoring stations are provided to monitor radioactivity at selected locations around the plant. 1.3.5 Shielding l Shielding is provided so that radiation exposure of personnel does not exceed the recommended limits of 10 CFR Part 20. The design of radiation shielding is dependent both on the extent of access required to a particular location and on the sources of radiation adjacent to that location. In the defueled condition, ALARA principles apply to assure that personnel have taken the necessary precautions prior to accessing or working on contaminated equipment. Rev.16 O DSAR 1-8 l l C______________________________._____ _ _ _ _ _ _ _ _ _ _ __ _ _._.__.__._____________..
l l MYAPC The plant is provided with a control room having adequate shielding to permit occupancy during j all credible accident situations. The radiation shielding in the plant, in combination with plant I radiation control procedures, ensures that operating personnel do not receive radiation exposures in excess of the applicable limits of 10 CFR Part 20. The control room is shielded to permit continuous occupancy following any accidental release of radioactivity resulting from a design basis accident. It should be noted however that, in the defueled condition, control room shielding is not required due to the lack of a significant source term from any of the design basis accidents in addition, control room ventilation is not credited in the safety analyses. i 1.3.6 Electrical Equipment - The plant electrical supply is provided by the auxiliary station service transformer (X-14) connected through an oil circuit breaker to the 115 kV transmission lines from the Central Maine Power Company system. In addition, auxiliary station service transformer (X-16) is connected through a fused disconnect switch directly to the 115 KV transmission lines. In the event that off-site power is interrupted, an on-site diesel generator is available for standby power. Independent offsite and onsite electrical power is provided for the nfe storage of spent ] fuel. Following a loss of offsite power to the spent fuel pool cooling system, and considering the significantly diminished decay heat load of spent fuel in the pool, ample time is available for i i operators to initiate altemate means of cooling or makeup for the spent fuel pool prior to substantial l heatup or inventory loss. Batteries are installed to supply any required de power. 1.3.7 System Flow Diagrams l Flow diagrams for each plant system are incorporated in the appropriate section of this report. The symbols used in these diagrams are shown in Figures 1.3-3 and 1.3-4. i r Rev.16 O V DSAR 1-9 L________.______. _ _ _ . . . . _
MYAPC I O SECTION 3.0 i FACILITY DESIGN AND OPERATION TABLE OF CONTENTS Section IJ.UR 2a91 3.1 D esign C ntana.... .. ... ...... .. . . .... .. . . . . . . . .. .. . . .. . . . . . . . . . . . . . . . . . .. . . .... .. . . . . . . . . 3-1 3.1.1 Conformance with 10CFR 50 Appendix A General Demgn Criteria 3.1.2 Classdication of Structures, Systems, and Components 3.1.2.1 SSCs important to the Defueled Condition 3.1.2.2 Wind, Missile, and Tornado Loadings 3.1.2.3 Water Level (Flood) Design ! 3.1.2.4 Seismic Design j 3.2 S tru ctu res . . . . . . ... . .. .. . . . ... . . .. .. . . . . . . ... . . . . . . . . . . . . . . . . . . . . .. . . .. . . . . . . . . .. . . .. 3-28 3.2.1 Fuel Building 3.2.1.1 General 3.2.1.2 Fuel Unloading Area
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3.2.1.3 New Fuel Storage 3.2.1.4 Spent Fuel Pool 3.2.1.5 Fuel Storage Racks 3.2.2 Storage Buildings 3.2.2.1 Underground RCA Storage Bunker 3.2.2.2 Radiation Controlled 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 Sernce Building 3.2.3.1 Control Room Area i 3.2.4 Turbine Building 3.2.5 Primary Auxiliary Building l 3.2.6 Serwce Water intake Structure 3.2.7 Fire Pump House 3.2.8 Masonry Walls Rev.14 O DSAR 3-i
MYAPC SECTION 3.0 FACILITY DESIGN AND OPERATION l 1 TABLE OF CONTENTS - l Section Iltla ERER 3.3 Systems 3.3.1 Fuel Storage.. . . . .......... ...................... ........... ....... .. .... . ..... ......... .... 3-52 j 3.3.1.1 Design Basis 3.3.1.2 System Description ! 3.3.1.3 Design Evaluation i i 3.3.1.4 System Operation l 3.3.1.5 Monitoring and instrumentation 3.3.2 Fuel Handling System.......................... ...................................... 3-88 ] 3.3.2.1 Design Basis 3.3.2.2 System Description p V 3.3.2.3 Design Evaluation Inspection and Testing 3.3.2.4 3.3.2.5 Monitoring and Instrumentation ; 3.3.3 Spent Fuel Pool (SF) Decay Heat Removal (DHR) Removal ) ) System........................................................................................3-94 ) l l 3.3.3.1 Design Basis 3.3.3.2 System Description 1 3.3.3.3 Design Evaluation 3.3.3.4 System Operation 3.3.3.5 Monitoring and instrumentation 3.3.4 Service Wa ter S ystem........................... ... . . . .. . .. . . .. . .. . .. . . ........ . ...... 3-102 ] 3.3.4.1 Design Basis 3.3.4.2 System Description 3.3.4.3 Design Evaluation 3.3.4.4 System Operation f 3.3.4.5 Monitoring and Instrumentation
- Rev.16 DSAR 3-il l
MYAPC SECTION 3.0 FACILITY DESIGN AND OPERATION TABLE OF CONTENTS Section Iida Eaga 3.3.5 Ventilation S ystems........... ....... ............ . . ....... ...... .. ..... ... .. ... .. .. ... . 3-106 ] 3.3.5.1 Fuel Building Ventilation System 3.3.5.2 Control Room Ventilation System 3.3.5.3 Auxiliary Ventilation Systems ! 3.3.5.3.1 Turbine Building Ventilation 3.3.5.3.2 Primary Auxiliary Building Ventilation 3.3.5.3.3 Service Building Ventilation ] 3.3.6 Auxilia ry Systems.......... ............................ .. .... ....... .. .. . ..... .. ... . ...... 3-1 1 1 ] 'O 3.3.6.1 Compressed Air 3.3.6.2 Boric Acid Makeup 3.3.6.3 Primary V!ater System 3.3.6.4 Primary Vent And Drain System 3.3.6.5 Radioactive Waste Processing System 3.3.6.6 Fire Protection System 3.3.6.7 Meteorological Instrumentation 3.3.7 Electrical S ystems... ......... ...... ......... . ... . .. ... . ..... .. .. . .. . .. . ..... . .. . . . . . . 3- 133 ] 3.3.7.1 Offsite Power 3.3.7.2 Onsite Power 3.3.7.3 Spent Fuel Pool Island and Balance of Plant Electrical ] Distribution System ] 3.3.7.4 Plant Computer and Programmatic Logic Controller (PLC) ] 3.4 Con trol of Heavv Loads ............... ................................. ... ..................... 3-148 ] O Rev.16 DSAR 3-iii
MYAPC SECTION 3.0 FACluTY DESIGN AND OPERATION UST OF TABLES Table No. Ilga 3.1.1 Earthquake Dampeg Factors 3.3.1 Compressed Air System - Equipment Description and Ratings l O l 1 l O DSAR 3-iv Rev.14
i MYAPC SECTION 3.0 FACILITY DESIGN AND OPERATION LIST OF FIGURES Fiaure No. Hg 3.1-1 Response Spectra for 0.05g Maximum Ground Acceleration 3.1-2 Response Spectra for 0.10g Maximum Ground Acceleration 3.2-1 Fuel Building Arrangement 3.2-2 Fuel Building Arrangement 3.2-3 Service Building 3.2-4 Service Building 3.2-5 Main Control Room 3.2-6 Turbine Building - Elevation 21' 3.2-7 Turbine Building - Elevation 39' 3.2-8 Turbine Building - Operating Floor 3.2-9 Primary Auxiliary Building Arrangement 3.2-10 Primary Auxiliary Building Arrangement O v 3.2-11 Service Water intake Structure 3.2.12 Fire Pump House 3.2-13 Fuel Transfer Tube 3.3-1 Typical Fuel Assembly 3.3-2 Typical Fuel Spacer Grid ; 3.3-3 Control Element Assembi'/ 3.3-4 Control Element Assembly 3.3-5 Spent Fuel Pool Assembly Placen:ent Umitations 3.3-6 High Density Spent Fuel Pool Layout For The Two Region Pool 3.3-7 Fuel Pool Cooling System 3.3-8 Refueling Equipment Arrangement 3.3-9 Fuel Pool Cooling Piping ] 3.3-12 Service Water System 3.3-13 Compressed Air System 3.3 14 Compressed Air System l DSAR 3-v
MYAPC SECTION 3.0 FACIUTY DESIGN AND OPERATION LIST OF FIGURES Fiaure No. M 3.3-15 Cornpressed Air System - Turbine Building 3.3-16 Compressed Air System -Turbine Building 3.3-17 Compressed Air System - Auxiliary Building i 3.3-18 Compressed Air System - Auxiliary Building l 3.3-19 Primary Vent And Drain System j 3.3-20 Primary Vent and Drain System 3.3-21 Fire Protecten System 3.3-22 Fire Protechon System ) . 3.3-23 One Une Diagram - Transmission and Utility Interconnections with MY J 3.3-24 One une Diagram - Auxiliary Power System 3.3-25 One Une Diagram - 125 Volt DC Emergency Buses and 120 Volt AC Vital Buses l l I l h Rev.14 DSAR 3-vi l
MYAPC SECTION 3.0 l FACILITY DESIGN AND OPERATION Section 3.0 discusses the design and operation of the structures, systems, and components required to safely store fuel. It also discusses supporting systems such as ventilation and auxiliary i systems used to safely store fuel or support decontamination and decommissioning activities. 1 ( 3.1 Design Criteria h l 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 regulatory requirement. Nevertheless, as the following discussion 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 permits issued prior to 'O May 21,1971 were evaluated on a plant-specific basis, and backfitting the current General Design 'V Criteria of Appendix A to 10CFR 50 would provide little or no safety benefit. (Reference 1, l SECY-92-223) On August 7,1997, Maine Yankee certified in accordance with 10CFR 50.82 that the company had permanently ceased power operation and that all fuel was removed from the reactor vessel (Reference 2, MN-97-89). With the docketing of these certifications, the Maine Yankee license no longer allowed operation of the reactor or placing of fuel in the reactor vessel. The AEC 1967 Design Criteria listed below are relevant to the permanently defueled plant condition. ANALOGOUS 1971 CRITERIA NUMBER CRITERIA l Group I - Overall Plant Requirements Quality Standards 1 1 l Performance Standards 2 2 Fire Protection 3 3 Sharing of Systems 4 5 Record Requirement 5 1 O Rev.14 i DSAR 3-1
1 MYAPC l O Group I!! - Nuclear and Radiation Controls i Control Room 11 19 Instrumentation and Control Systems 12 13 Monitoring Radioactive Releases 17 64 Monitoring Fuel and Waste Storage 18 63 Group Vill - Fuel and Waste Storage System Prevention of Fuel Storage Criticality 66 62 Fuel and Waste Storage Decay Heat 67 61 Fuel and Waste Storage Radiation Shielding 68 61 Protection Against Radioactivity Release from j Spent Fuel and Waste Storage 69 - l Group IX - Plant Effluents l Control of Release of Radioactmty to the Environment 70 60 1 Group I - Overall Plant Requirements Criterion 1 - Quality Standards Those systems and components of reactor facilities which are essential to the prevention of accidents which could affect the public health and safety, or to mitigation of their consequences, , I shall be identified and then designed, fabricated and erected to quality standards that reflect the importance of the safety funcbon to be performed. Where generally recognized codes or standards on design, materials, fabrication, and inspection are used, they shall be identified. Where adherence to such codes or ste .Jards does not suffice to assure a quality product in keeping with the safety function, they "Aall be supplemented or modified as necessary. Quality assurance programs, test pecedues, and inspection acceptance levels to be used shall be identified. A : showing of sufficiency and applicability of codes, ste7dards, quality assurance programs, test procedures, and inspection acceptance levels used is required. (This Criterion is directly analogous to Criterion 1 of 10CFR 50, Appendix A,1971, except that the 1971 version addresses records retention. Maine Yankee record retention requirements are addressed in Section 5.10 of the Technical Specifications or the Quality Assurance Program. Design, fabrication, and construction records are addressed by Criterion 5 of this section.) Rev.14 DSAR 3-2 l
i
,_ MYAPC Resconse:
Those systems and components which are essential to the prevention of accidents which could affect the public health and safety, or to mitigation of their consequences, have been designed, fabricated, and erected to quality standards that reflect the importance of the safety function to be performed. Generally recognized codes and standards on design, materials, fabrication, and inspection have been used. These were supplemented to reflect current practices. The descriptions of the systems and components to which this criterion applies include the codes and l other standards 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. l
References:
Sections 0 and 4 : Criterion 2 - Performance Standards Those systems and components of reactor facilities which are essential to the prevention of n accidents wh!ch could affect the public health and safety, or to mitigation of their consequences, () shall be designed, fabricated, and erected to phance standards that will enable the facility to w^ stand, without ioss of the capability to p w t tie public, the additional forces that might be
. ~ josed by natural phenomena such as earthquakes, tomadoes, flooding conditions, winds, ice, and other local site effects. The design bases so established shall reflect: (a) appropriate consideration of the most severe of these natural phenomena that have been recorded for the site and the surrounding area, and (b) an appropriate margin for withstanding forces greater than those recorded to reflect uncertainties about the historical data and their suitability as a basis for design.
(This Cri%rion is directly analogous to Criterion 2 of 10CFR 50, Appendix A,1971, except that the 1971 version addresses consideration of appropriate combinations of the effects of normal and accident conditions with the effects of the natural phenomena and the importance of the safety functions to be performed.) Resconse: l Systems and components which are essential to the prevention of accidents which could affect the public health and safety, or to mitigation of their consequences, are designed, fabricated and erected to performance standards that enable the facility to withstand, without loss of the capability G Rev.14 DSAR 3-3
MYAPC O to protect the public, the additional forces that might be imposed by natural phenomena such as earthquakes, tomadoes, flooding conditions, winds, ice, and other local site effects. The design
. bases so established reflect appropnate consideration for the most severe natural phenomena that
, have been recorded for the site and sunounding area, and appropriate margin for withstanding l forces greater than those recorded to reflect uncertainties about the historical data and their I suitability as a basis for design. ! l l l Or: April 17,1979, an earthquake of approximate magnitude 4 occurred about 10 kilometers west of the plant. On January 9,1982, an earthquake of approximate magnitude 5.75 occurred in Central New Brunswick. Subsequently, the Ucensee joined with the Regulator in a program to assess the j seismic ruggedness of the Maine Yankee plant. This program is referred to as the Seismic Design l Margins Program (SDMP). This program is fully described in NUREG/CR-4826, dated March 1987. l On March 26,1987, the Regulator issued a Safety Evaluation Report which concluded that all issues . associated with the seismic design were considered resolved (Ref. 3).
References:
Sections 2 and 3. j Criterion 3 - Fire Protection The reactor facility shall be designed (1) to minimize the probability of events such as fires and explosions, and (2) to minimize the potential effects of such events to safety. Noncombustible and fire resistant materials shall be used whenever practical througnout the facility, particularly in areas l containing critical portions of the facility such as containment, control room, and components of engineered safety features. (This Criterion is directly analogous to Criterion 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 minimize the adverse effects of fire on structures, systems, and components important to safety taking into account the decommissioning plant condition and j activities, i O Rev.14 ! DSAR 3-4
I i l l MYAPC m ! (J 1 Critenon 5 - Record Requirement Records of the design, fabrication, and construction of essential components of the plant shall be i maintained by the reactor operator or under its control throughout the life of the reactor. l (This Criterion, in combination wi'h Cnterion 1 of this section, is analogous with Criterion 1 of 10CFR 50, Appendix A,1971.) Besconse; I I Records of design, fabrication, and construction of components are being maintained for the duration of the license. Design calculations are in the possession of Stone & Webster and Yankee Atomic. All other required design, fabrication, and construction information is in the possession of ! Maine Yankee. 1
References:
Section 6 and the Quality Assurance Program GROUP 111 - NUCLEAR AND RADIATION CONTROLS I
/Q Criterion 11 - Control Room The facility shall be provided with a control room from which actions to maintain safe operational status of the plant can be controlled. Adequate radiation protection shall 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 facility without radiation exposures of personnel in excess of 10CFR 20 limits. It shall 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.
Resoonse: l This criterion is met only to the extent that radiation exposures of personnel in excess of 10CFR 20 limits cannot occur based on the available source term from any credible accident. Sufficient time is available to allow operators to restore cooling or mckeup to the spent fuel pool and maintain exposures well below the limits of 10CFR 20. Controls for spent fuel pool cooling, makeup and purification are located near the equipment and are not in the control room.
References:
Sections 3,5, and 7 ( Rev.14 DSAR 3-5 l l
l MYAPC . Cntanon 12 - Instrumentation and Control Systems l Instrumentation and controls shall be pnmded as required to monitor and maintain variables within prescribed operating ranges. l Resoonse: 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 ; operation. In the permanently defueled condition, the principal variables to be monitored include l fuel pool level, and temperature. Boron concentration is determined via sampling. ! i
References:
Sections 2,3, and 4 l Critarion 17 - Monitonng Radioactive Releases l 1 ! Means shall be provided for ire 4A,g the containment atmosphere, the facility emuent discharge paths, and the facility environs for radioactivity that could be released from nomial operations, from lh anticipated transients, and from accident conditions. (This Criterion is analogous to Critation 64 of 10CFR50, App. A,1971, except that the 1971 version
. addresses spaces containing components for recirculation of loss-of-coolant accident fluids.)
l l 1 l Resoonse: l l ! i The means for monitoring radiation levels in the spent fuel pool nnd concentrations on-site is ] 1 l provided. The sensitivity and range of this equipment is adequate for operating and anticipated l l ' transients in the permanently defueled condition. Emuent discharge paths where a potential release of radioactive material exists are monitored. Monitoring equipment has sufficient sensitivity and range for operating and anticipated transients in the permanently defueled condition. Instrumentation is provided for monitoring radiation levels in the spent fuel pool area. ]
]
Rev.16 O DSAR 3-6 , l
l MYAPC l An environmental surveillance program has been established. The facility's environmental surveillance program provides effective monitoring of radioactive material released from the plant.
References:
Section 4. I Criterion 18 - Monitorina Fuel and Waste Storace Monitoring and alarm instrumentation shall be provided for fuel and' waste storage and handling areas for conditions that might contribute to loss of continuity in decay heat removal and to radiation exposures. Resconse: 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 audiole alarms may be used to monitor radiation exposure. l The spent fuel pool is equipped with high and low liquid level alarms, and a high temperature alarm l t/ which will 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 Criterion 66 - Prevention of Fuel Storace Criticality i Criticality in new and spent fuel storage shall be prevented by physical systems or processes. Such means as geometrically safe configurations shall be emphasized over procedural controls. s Resoonse: Criticality is prevented by geometrically safe configurations. New fuel is stored dry in a steel and concrete building at an elevation at which flooding with water is impossible. Rev.16 O DSAR 3-7
I MYAPC Irradiated spent fuel is stored under water in a reinforced concrete pool, lined with stainless steel. l Fuel assemblies are spaced and the racks are so fabricated that criticality is precluded. Although the water in the pool is generally borated, neither soluble boron nor control rods are required to keep ever: unirradiated fuel assemblies subcritical. The boron concentration required as a result of the analyzed " misplaced assembly" incident is discussed in section 3.3 and 5.2. I
Reference:
Section 3.3. Criterion 67 - Fuel and Waste Storace Decav Heat j Reliable decay heat removal systems shall be designed to prevent damage to the fuel in storage I facilities that could result in radioactivity release to plant operating areas or the public environs. l \ Resoonse: The spent fuel pool is designed to maintain the water level safely above the spent fuel assemblies j at all times. The fuel pool outlet pipe, which serves as the suction pipe tc the fuel pool cooling l l pumps, is siphon protected 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 installed to increase this limit to no less than 19' above the active ] l fuel. Pool water level may be restored through diverse (offsite and onsite) supplies of fresh or demineralized water. Emergency makeup water and cooling is also available from the fire pond.
Reference:
Section 3. Criterion 68 - Fuel and Waste Storace Radiation Shieldino l Shielding for radiation protection shall be provided in the design of spent fuel and waste storage facilities as required to meet the requirements of 10CFR 20. I (Also refer to Criterion 67 of this DSAR section.) l 2 Rev.16 O DSAR 3-8 l
r I t MYAPC O-l l l The radiation shielding (water) of the spent fuel pool is designed to provide adequate personnel ! protection. The waste storage and processing facilities are shielded as required to protect i personnel from exposures in excess of regulatory requirements. in addition to shielding design, l implementation of the Radiation Protection Plan requirements assures that doses to personnel performing work in these areas are maintained ALARA.
Reference:
Section 3 and 4. l L Criterion 69 - Protection Anninst Rmdinardvity Ral==== frorn Soent Fuel and Waste Stornos Containment of fuel and waste storage shall be provided if accidents could lead to release of undue amounts of radioactivity to the public environs. l t
Response
I Fuel and waste storage areas are designed to preclude the inadvertent release of undue amounts l . of radioactivity. Considering the 5;ymntly diminished source terms in the permanently defueled l condition, there are no credible accidents resulting in doses to the public approaching the 10CFR 100 limits. All spent fuel and waste storage systems are conservatively designed with ample margin to prevent the possibility of gross mechanical failure which could release significant amounts of l radioactivity. Backup systems such as floor and trench drains are provided to collect potential l leakages to preclude the release of radioactive materials to the environment. Personnel are rigorously trained and administrative procedures are strictly followed to reduce the potential for human error in addition, radiological limits on waste storage systems are established i to assure that credible accident conditions will not result in doses to the public which approach 1 rem l (whole body) exposure over a 2 hour-limited duration accident. \ l The consequences of a fuel handling incident are presented in Sections 3 and 5. In this analysis, j lt is demonstrated that undue amounts of radioactivity are not released to the public. i
References:
Sections 3,4,5 and 7. ! Rev.16 l DSAR 3-9 l l
i l l MYAPC GROUP IX - PLANT EFFLUENTS Critarion 70 - Control of R-1-- of RmW-*vity to the Environment l The facility design shall include those means necessary to maintain control over the plant radioactive effluents, whether gaseous, liquid, or solid. Appropriate holdup espacity shall be provided for retention of gaseous, liquid, or solid effluents, particularly where unfavorable environmental conditions can be expected to require operational limitations upon the release of radioactive affluents to the envuonment. In all cases, the design for radioactivity control shall be justified (a) on the basis of 10CFR 20 requirements for normal operation and for any transient situation that might reasonably be anticipated to occur and (b) on the basis of 10CFR 100 dosage level guidelines for potential reactor accidents of exceedingly low probability of occurrence except that reduction of the recommended dosage levels may be required where high population densities or very large cities can be affected by the radioactive effluents. Resoonse: The plant radioactive waste control systems (which include the liquid, gaseous and solid radweste systems) are designed to limit the off-site radiation exposure during normal operation to levels below limits set forth in 10CFR 20. I
References:
Sections 4. I p Rev.14
. \.)
DSAR 3-10
MYAPC 3.1.2 Classification of Structures, Systems, and Components Structure Classification l The plant structures and process systems are dassified according to their function and the degree l of integrity required to protect the public from uncontrolled releases of radioactive byproducts. Structural design criteria include two classes of buildings:
- 1. Class i Structures Class I structures were designed in accordance with the " Building Code Requirements for Reinforced Concrete," ACl-318-63, including increases allowed for stresses produced by earthquake loads in combination with other appropriate loads. Where steel is utilized, it is designed in accordance with AISC " Specification for the Design, Fabrication and Bection of Structural Steel l for Buildings." A Class I structure will mcintain its integrity during the hypothetical earthquake where the combination of the normal operating loads and the seismic stresses do not exceed 90%
of the yield strength. These structures are primarily constructed with massive reinforced concrete and the earthquake loads are not a major factor in the design. l V) Class I structures and equipment are designed to remain functional during an operational basis earthquake (ground acceleration 0.05g) and maintain fuel pool integrity during the more severe design basis of the hypothetical earthquake (ground acceleration 0.10g). In addition, some of the i Class I structures are designed so that damage will not result from tomado winds or missiles. l l The structures, systems and components which have been designed to Class I seismic l requirements are listed below: ! Reactor Containment j Reinforced concrete substructure l l Reinforced concrete superstructure j Reactor containment fuel transfer tube l Soent Fuel Pool Fuel assemblies Reinforced concrete structure, partly below ground Steel superstructure o Rev.14 DSAR 3-11
)
,- 3 MYAPC I )
V Spent fuel storage rack Fuel building - yard crane steel support structure, portion within spent fuel pool building only Fuel building - yard crane Fuel handling platform and hoist I Anti-siphon device Flow limiters on the liner leakage detecticr. system
- 2. Class 11 Structures Class il structures are usually designed to the requirements of the Uniform Building Code. Class 11 structure.s were not designed for OBE and DBE. They were designed for dead load plus live load plus wind in accordance with AISC " Specification for the Design, Fabrication and Erection of Structural Steel for Buildings" and " Building Code Requirements for Reinforced Concrete", ACI-318-63, Part IV-A, utilizing 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 cf both, in addition, shake space between all q adjacent Class I strLetures is provided for in 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 buildings in the event of a disturbance. Dismantlement of Seismically Desioned Structures. Svstems and Comoonents For SSCs formerly designed to Seismic Class I requirements but not credited for performing a Seismic Class I function in the defueled condition, the following criteria apply prior to performing dismantlement operations:
- a. Declassification of components shall be performed in accordance with appropriate engineering and design procedures and processes.
- b. When declassifying an SSC, a 10CFR 50.59 evaluation shall be performed if:
- 1) the safety classification is described in the FSAR, or
- 2) it's failure in a seismic event could affect a Seismic Class I component described in the FSAR in such a manner as to cause an unanalyzed incident or an accident with offsite doses exceeding 230 mrem (whole body) or 260 mrem (organ dose).
- c. When declassifying an SSC, a 10CFR 50.54 evaluation shall be performed if the classification is described in the OOAP.
KJ Rev.14 ( DSAR 3-12
l r MYAPC l
- d. SSCs shall be designed to seismic Class I requirements if, during a seismic event, its failure has the potential to drain the fuel pool water level lower than 10 feet above the active fuel.
3.1.2.1 Structures, Systems and Components important to The Defueled Condition (ITDC) General On August 7,1997, Maine Yankee certified per 10 CFR 50.82 that the company had permanently ceased power operation and that all irradiated fuel had been permanently ) removed from the reactor vessel (Reference 2). This is a permanent, non-revocable certification that changed Maine Yankee's licensing basis by no longer allowing fuel in the reactor vessel and no longer allowing power operation. The license basis for the majority of Structures, Systems and Components (SSCs) associated with nuclear safety has been changed. Those SSCs which only performed a reactor safety function (i.e., SSCs which do not support a spent fuel or radiation protection safety function) need no longer be maintained under nuclear grade controls. l V SSC classification involves a determination that an SSC is, or is not, safety-related'. SSCs classified as safety-related are treated differerstly by regulation tien other SSCs.2 l For a plant undergoing decommissioning, tne only SSCs which meet the definition of safety-related' are the spent fuel pool stnJctura, storage racks, fuel transfer tube, and spent fuel cooling system syphon breaks. This jesults in two areas of interest: l l l l 1. safety related ssCs are &"- J upon to remain funcnonal during and following design basis events to ensure: a) the integnty of the reactor cc141 p"4 es boundary; b) the capability to shut down the reactor and maintain it in a safe shutdown condition; and c) the cab'% = / event or mitigate the consequences of accidents that could result in potennal offsite exposures comparable to tr ' N of 10 CFR 100.
- 2. 10 CFR 50 Appendix B notes that'The perenent requirements of this appendix apply 03 all activities affecting the safety.
related funedons of ." ssCs. 3 The Arst two parts of the safetyve6ated de6 ninon (reactor coolant pressure boundary, and capability to achieve and maintain safe attutdown) do not appty to a decommissioning plant, gtven the recense restnctions of 10 CFR 50.82. The third part of the safety-related aefiniton (acodent consequences comparable to 10 CFR 100 guidehnes) also does not apply. At Maine
. Yankee. the consegunces associsted wim the design / license basis events applicable to decommissioning are nearty three orders of magnitude kwer than Part 100 guidelines and lower than me EPA protecDve accon guide limit.
DSAR 3-13 Rev.15
MYAPC >O L./ 1) Maine Yankee's " nuclear grade" processes are based largely upon quality assurance (10 CFR 50 Appendix B) requirements. Reclassifying all SSCs as non-safety related could lead to the elimination of most management controls in situations where maintaining ! rigorous management controls is intended, l l 2) Maine Yankee recognizes that certain functions remain important to safety in the defueled cc,odition. It is necessary to reclassify SSCs in order to proceed with decommissioning. Strictly followtg regulatory requirements in reclassification results in elimination of most of the current management controis, which is contrary to management's intent. Thus, in order to provide an
. enhanced engineering controls above that mandated by regulatory requirements, an artificial j classificadon system termed "Important to the Defueled Condition" (ITDC) is introduced.
l The following concems are addressed within this classification:
- SSCs which support a fuel safety or radiation protection safety function, and L = identification of enhanced management and engineering controls are maintained on SSCs classified as ITDC.
1 it is noted that SSCs which are not defined as within the ITDC classification, or otherwise ! designated as safety class, are eliminated from the license basis. It is not the intention of implementing the ITDC classification to reclassify components previously defined as NNS as ITDC. Additionally, there may be other SSCs to which a level of enhanced quality or engineering oversight has been applied, but do not meet the intent of the ITDC classification. The following criteria are used to determine which SSCs are designated as ITDC: ) L Criterion 1. The SSC is essential to the normal operation of the safe storage, control, or maintenance of the spent nuclear fuel or safe handling of radioactive waste. Criterion 2. The SSC is essential in preventing postulated accidents or incidents involving the safe storage, control, or maintenance of the spent nuclear fuel or safe handling of radioactive waste. O DSAR 3-14 Rev.15 I
l MYAPC Ik /~} Cnterion 3. The SSC historical dassification is the direct result of an outstanding l commitment to the USNRC which remains applicable to safe storage, control, l l or maintenance of the spent nudear fuel; or safe handling of radioactive waste. t Criterion 4. The SSC satisfies a requirement based in regulations which remain applicable
- l. to safe storage, control, or maintenance of the spent nuclear fuel; or safe l handling of radioactive waste as defined in the Maine Yankee licensing basis.
This indudes any SSC which is independently required by the Umiting Conditions of Operation -(Sechon 3) of the Technical Specifications.' l A positive response to any criterion indicates that an SSC is ITDC. Authorizations. Restrictions and Limitations on use of the SSC rar l==ification criteria. Tre SSC redassification criteria will be used as a basis to change various Maine Yankee pro Passes, provided that the change involves an SSC that is non-ITDC and, provided that plant procedures contain an acceptable method for approving the change. The following kinds of
" software" changes associated with non-ITDC SSCs are allowed:
. SSC classifications
! - drawingt,
= calculations, l
l
. procedures
. nonconforming items and corTective actions (Leaming Bank)
. extemal industry operating experience reports
= commitments l
l . open work orders (in process at the time the decision was made to decommission the plant)
. the application of 10 CFR 50 Appendix B criteria provided it does not represent a reduction .
in commitment. { Use of these criteria does not authorize:
- a. Activities creating new hazards or initiators not already recognized as part of the current ,
license basis (e.g., decontamination or decommissioning of major components defined in ( ( , 10 CFR 50.82) { O 4 The IToC evaluation assures that me apprognate regulatory change meenanism is used for effectng the change. ) I DSAR 3-15 Rev.15
)
MYAPC
- 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.
- c. Changes to Technical Specification requirements applicable to the current mode of operation.
- d. Changes to regulations, license conditions, nales, and permits until such time that relief is granted from the regulating authority. However, it may provide the basis for requesting l
relief from the regulations, license conditions, rules, and permits.
- a. Changes to commitments.- Application of the commitment change process is required to l change commitments.
l l f. Changes to the OQAP. However,it may provide the basis for initiating a change to the OQAP.
- g. Changes to the ODCM. However, it may provida the basis for initiating a change to the l l
ODCM.
- h. Changes to the Emergency Plan. However, it may provide the basis for initiating a change i to the Emergency Plan.
l l. Changes to the Security Plan. However, it may provide the basis for initiating a change j to the Security Plan. J. Changes to the Fire Protection Plan. However, it may provide the basis for initiating a change to the Fire Protection Plan.
- k. Changes to the Radiation Protection Program. However, it may provide the basis for i initiating a change to the Radiation Protection Program.
Boundaries and Interfaces for ITDC SSCs I l SSCs identified as ITDC that require " availability", must meet the following criterion:
"A system, subsystem, train, component, or device is "available" or will have " availability" when it is capable of performing its specified function (s)." j Implicit in this definition 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.
Er,cir.;; red Requirements for ITDC SSCs A higher level of quality is maintained for ITDC components to assure that the capablity exists to I reliably meet the performance expectations and requirements. The controlled list of ITDC Rev.14 . DSAR 3-16
MYAPC O components is govemed by plant procedures. ITDC components are not safety-related components and are not required to satisfy 1COFR50 Appendix B requirements. Although not l- required by regulation, the following criteria is developed and applied, as appropriate, to ITDC ! SSCs to assure continued reliability:
- a. Design Control
! Measures will be invoked to assure applicable regulatory requirements, license basis, and l design basis information is correctly translated into specifications, drawings, procedures and instructions. These measures shall include provisions to assure that appropriate quality standards are specified and included in design documents and that deviations from such standards are controlled. Design changes, including field changes will be subjected to engineered design control measures commensurate with the importance of the SSC.
- b. Procurement Document Control I Measures will be invoked to assure that applicable regulatory requirements, design basis, and other requirements which are necessary to assure adequate quality are suitably included or referenced in the documents for procurement of material, equipment, services,
- c. Instructions, Procedures, and Drawings l Activities affecting SSCs will be prescribed by documented instructions, procedures, or
- drawings, of a type appropriate to the circumstances and will be accomplished in accordance I
with these instructions, procedures, and drawings. Instructions procedures, and drawings will include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished.
- d. Control of Purchased Material, Equipment, and Services Measures will be invoked to assure that material, equipment, and services conform to the procurement documents. These measures shall include provisions, as appropriate, for source l
evaluation and selection, objective evidence of quality fumished, inspection at the source, and examination upon delivery,
- e. Inspection Inspection of activities affecting quality will be invoked and executed to verify conformance with the documented instructions, procedures, and drawings for accomplishing the activity.
Rev.14 O DSAR 3-17 1
i MYAPC O
- f. Handling, Storage and Shipping Measures will be invoked to control the handling, storage. shipping, cleaning and preservation of material and equipment in accordance with work and inspechon instructions to prevent damage or deterioration. i
- g. Test Control Surveillance testing wiH be established for SSCs to ensure that the SSCs perform satisfactorily commensurate with the importance of its intended safety function. ,
L I
- o p h. Measuring and Test Equipment j- Appropriate controls wiu be invoked to assure that measuring and test devices used on SSCs
[ are properly controlled, calibrated and adjusted at specified periods to maintain accuracy within necessary limits.
- l. Corrective Action l i
l Measures will be invoked to assure that conditions adverse to quality are promptly identified j and corrected. In the case of significant conditions adverse to quality, the measures will
\
assure that the cause of the condition is determined and correchve action is taken to preclude i 1 repetition. ; i l l 3.1.2.2 Wind, Missile, and Tomado Loadings ! The Maine Yankee facility is capable of withstanding the effects of severe winds or tomadoes without loss of capability 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 extemal 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 utility pole 35 ft long,14 inches in diameter, weighing 50 lb/cu. ft. (1850 lbs , total), and traveling 150 mph, or, (2) a 1-ton automobile traveling at 150 mph. The structures housing spent fuel and adjacent structures are designed to resist the combined effects of tomado wind load, pressure drop and missile loads to produce the most critical loading I condition. The original design code for the spent fuel pool and other Class I structures was ACI 318-63. The allowable stresses for shear and flexure are defined by the criteria in sections 1600 and 1700 of ACI 318-63. This specifically included the appropriate capacity reduction factor and Rev.14 DSAR 3-18 l.
l i l l MYAPC i allowable stresses. The Fuei Building is designed for protection against wind and tomado as i ! follows: Reinforced concrete structure i Steel superstructure Fuel building - yard crane steel support structure, portion within spent fuel pool building only 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 effects of tomado and missiles. A very substantial degree of added tomado and missile protection is afforded by the below-grade construction of the fuel pool. The nominal grade level (20' elev) is approximately at the same elevation as the top of the active fuel. The steel framing above the poolis designed for tomado loadings such that it will not fall into the pool and damage fuel assemblies. ! l The loss of water from the storage pool unoer the effect of a tomado would result in a maximum ) calculated loss of five feet of water, primarily due to vortexing. Given that the normal level of the j spent fuel pool is at the 44 ft. elevation and approximately 23 ft. above the active fuel, this water i loss is not significant as it is bounded by the siphoning incident.
- O
- 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 protection reservoir are tomado protected and ] l l screened from the full effects of the tomado by the dike. This is a very reliable makeup source of ) l water to the pool. i 3.1.2.3 Water Level (Flood) Design i 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 Maine Yankee Atome Power Station on the Sheepscot River estuary when the probable maximum hurricane is taken as the design basis meteorological event. The investigation is based upon the parameters of the probable maximum hurricane as defined by U.S. Weather Bureau Report HUR 7 97, Interim Report- Meteerciegicei Characteristics of Probable Maximum Hurricane, Atlantic and Gulf Coasts of the United States and discussed in section 2.2.6. Rev.16 O DSAR 3-19
MYAPC l This investigation shows that the maximum water levels at the Maine Yankee Power Station due ! to the probable maximum hurricane are predicted to be at Elevation 19.,9 feet and Elevation 21.4 feet on the plant site and screen well structure, respectively. These levels are based upon the ! simultaneous occurrence of the maximum storm surge, maximum predicted astronomical tide, an l initial rise in mean sea level, estuarine amplification, the probable maximum flood in the Sheepscot watershed, maximum waves in Montsweag Bay and existence of a channel restriction at the former Cowseagan Nanows Causeway. 3.1.2.3.2 Snow Section 2.2.5.1. Table 2.2.5 shows average snowfall statistics for Portland which are considered l
)
to be representative of the site. Structural loading and capacity reduction factors for buildings required in the defueled condition provide ample margin to accommodate snow loading. l 3.1.2.3.3 lee, Glaze, and Temperature Extremes Ice loading is discussed in Sections 2.2.5 and 2.3.4 of this report. Glaze and Ice storms usually q occur in the months October through April with an average frequency of 1 to 3 storms per year. ) b ice thicknesses of .25 inches or .5 inches will likely occur every year whereas .75 inches is likely to occur at least every three years. Structural loading and capacity reduction factors for buildings required in the defueled condition provide ample margin to accommodate ice loading. l l The circulating water intake structure is used as the intake structure for both service water and circulating water systems. The intake structure is protected from ice as it will not form at the intake channels to a depth to obstruct intake flow. Circulating water provides no spent fuel storage cooling or makeup functions. It may be used, if elected, to dilute planned and conirolled waste l discharges. The service water system performed a safety class function during power operation but does not support a safety class function in the defueled condition. ] i 3 l 1 3 As discussed in section 2.2.4, the average January temperature is about 22*F with between 10 and 20 days of sub-zero temperatures occurring yearly. Temperature data representative for the site is provided in Table 2.2.2. During extended periods of freezing temperatures, it is possible that Rev.16 DSAR 3-20
i MYAPC O freeze damage could occur to water filled piping in buildings no longer maintained due to the
~ defueled condition. Tne primary concem regarding freezing is the affect on the integrity of the r ol.
Pool water freezing is not possible provided that the aggregate decay heat load of the stored assemblies is reasonably high. Assurance is provided through routine operator rounds which monitor and log the temperature of the pool water. Adjacent buildings and rooms containing significant water sources which could potentially affect the safe storage of fuel are either temperature controlled to preclude freezmg, or the water source is appropriately heat traced or drained, as necessary. This includes adequate administrative or design controls for protecting the integrity of the fuel transfer tube. 3.1.2.4 Seismic Design 3.1.2.4.1 Design Basis All structures and elements of the plant are designed in accordance with sound engineering practice and are considered capable of withstanding seismic forces corresponding to a ground acceleration of at least 0.03g, in addition to normal loads, without damage or loss of function. In addition, all structures and components of the plant which are important from the standpoint of nuclear safety and damage which could affect the health and safety of the public, i.e., " Class I" portion, are designed to meet the following criteria: j
- 1. The design earthquake is based on a ground acceleration of 0.05g, and this portion of the plant shall be capable of operating through such earthquake.
- 2. The hypothetical earthquake is based on a ground acceleration of 0.1g, and this portion c7 the plant shall be capable of performing its intended safety function under an earthquake. ,
- 3. A spectrum analysis is used, with appropriate conservative damping factors.
3.1.2.4.2 Design Data i l While 0.04g was found to be the maximum probable ground acceleration at the base of the structures on the site,it was decided to round this upward to 0.05g horizontal for design purposes. The values for seismic design of Class I structures and components are as stated under " Design Bases." In Class I structures 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 loads plus the hypothetical earthquake do not exceed the yield stress of the affected Rev.14 DSAR 3-21
MYAPC materials. Earthquake stresses are based on a horizontal ground acceleration and a vertical ground acceleration of two-thirds of the horizontal, with the two acting simultaneously. 1 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 Figure 3.1-1 for the design earthquake and Figure 3.1-2 for the hypothetical earthquake. The curves are derived from the "Housner Spectrum" normalized to 0.05g for the design earthquake and 0.10g for the hypothetical earthquake. The design response spectra (Figures 3.1-1 and 3.1-2) are a specification of the level of seismic design acceleration, or displacement, as a function of natural period of vibration and damping level. The response spectrum analysis is applied to all category I structures and components and groups thereof whose responses may be interdependent, considering their natural period and using appropriate damping factors as listed on Table 3.1.1. Class I structures and components are designed in the following general manner: O 1. An analysis is made to determine the natural periods of vibration of the structure using b equivalent lump mass systems or distributed mass systems as is considered appropriate. In these analyses, periods and mode shapes are determined for each lumped mass mode. These data then define participa; ion factors for each structure. Where structures are supported on their own foundations, foundation displacements are considered in determining natural periods and participation factors. It should be noted, however, that Class I structures at this site are founded on granite gneiss. Accordingly, 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 specific natural period of the structure or component being considered is determined from Figure 3.1-1 using appropriate damping factors. The horizontal component of the ground acceleration is taken directly and the vertical component is taken as two-thirds of the horizontal value. These components are considered as acting simultaneously.
i
- 3. For certain structures, and especially for vibratory systems of a highly complex nature, such as a piping system, use of the maximum response value (peak of the curve) corresponding to the appropriate damping factor may be elected in performing the stress analysis of the system.
- 4. A tabulation of typical damping factors which are used for various vibratory systems important Rev.14 DSAR 3-22
l l MYAPC 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 Figure 3.1-2. Refer to section 3.1.2 for a listing of SSCs required to be designed to Seismic Class I requirements in the defueled condition. SSCs which were proveusly designed to Class I criteria but are not listed in section 3.1.2, are not credited for performing a safety function (Class I ) in the permanently defueled condition, and therefore, are no longer required to be designed to Class I requirements.
. 3.1.2.4.3 Seismic Design, Qualification, and instrumentation As of March 24,1986, new Class I systems, structures, and components will be designed and qualified to the seismic demands as defined by a 0.18g NUREG/CR-0098 50th percentile Ground Response Spectra (GRS). Analytical qualification will be to the SEP allowable stress levels and damping values listed in Reference 4 (i.e., for piping, damping = 3% or PVRC, allowable stress =
l 2.4Sh, no OBE) up to any interface with existing structures, systems, and components. Seismic adequacy may also be demonstrated through similarity by comparison to the documented performance of equipment in natural earthquakes (Reference 5), or simulated earthquakes on testing machines. Maine Yankee's original seismograph installation consists of two triaxial strong motion accelerographs. A vertical seismic trigger, set at approximately 0.01g will have the device operating fully within 0.1 seconds of the initial"P" wave and will continue to operate 10 seconds after the last motion of a seismic event is detected. Each device is operated by a self-contained ! battery supply and'contains enough 70mm photographic film to operate for 25 minutes. l
- One accelerograph is located on bed rock adjacent to the containment base mat at elevation -1S'-
L 9". The other instrument is bolted to a heavy shelf which in tum is bolted to the outside of the reactor containment wall at approximately elevation 34'. The two instruments will be in the same vertical plane. The vertical separation of 50' constitutes a significant fraction of the structure's height. O Rev.14 I DSAR 3-23
- Section 3.1 References-
- 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-97-89 " Certification of Cassation of Power Operation and l Permanent Removal of Fuel from the Reactor, dated August 7,1997.
1
- 3. Letter: ' Seismic Demgn Maryns 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 March 26,1987.
l
- 5. USNRC Letter to MYAPCO, dated February 19,1987, (Generic Letter 87-02).
1 O i l-l l Rev.14 Dt A 3-24
i, MYAPC TABLE 3.1.1 EARTHQUAKE DAMPING FACTORS Percent of Crttical Damping DESIGN HYPOTHETICAL EARTHQUAKE EARTHQUAKE Reactor Containment 2.0 5.0 Reinforced concrete structure, other than 2.0 5.0 containment, founded on soil or rock Reinforced concrete supporting structure, not 2.0 5.0 founded on soil or rock Steel-framed structures, including supporting structures and foundations Bolted or Riveted 3.0 5.0 Welded 1.0 2.0 I Reactor vessel, intemals and control rod drives Welded Assemblies 1.0 1.0 Bolted Assemblies 3.0 3.0 Mechanical equipment, including pumps, fans and 2.0 2.0 similar items Piping Systems 1.0 2.0 l i l l t Rev.14 DSAR 3-25
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MYAPC 1 3.2 Structures 3.2.1 Fuel Building 3.2.1.1 General The principal function of the fuel building is to provide a location for the safe storage of new and q spent fuel assemblies. The building houses a new-fuel unloading area, a new-fuel storage room, I a spent fuel pool and the necessary cranes required for the handling of the fuel assemblies.~ The I spent fuel pool cooling system heat exchanger, the fuel pool cooling pumps ar'd the fuel pool purification pump are located in the area adjacent to the solid waste disposal equipment.- The equipment decontamination area and the spent fuel pool support systems are also located in the fuel building. The fuel building arrangement is shown on Figures 3.2-1 and 3.2-2. 1 I 3.2.1.2 Fuel Unloading Area New fuel was shipped to the site in two-element shipping casks. A five-ton overhead crane in the fuel building was used to unload the shipping casks. The spent fuel pool purification system filters
,- (used during plant operational phases) are located in shielded cubicles below the fuel unloading ]
Q area. Shield slabs are removed from the fuel unloading floor to replace expended filter cartridge elements. I l 3.2.1.3 New Fuel Storage Area The new-fuel storage room is designed for storage of 160 fuel assemblies. The fuel room is located over the solid waste disposal area and the spent fuel pool cooling pumps and heat exchanger. The fuel rack consists of guide sleeves symmetrically located on the floor at Elevation 31 ft.1-1/2 in. and through the ceiling of the new-fuel room at Elevation 44 ft. 6 in. The fuel room I floor has a drain opening located over the spent fuel pool cooling equipment cubicle. The floor opening prevents flooding of the new-fuel storage area. The spent fuel pool new-fuel elevator l winch is also located in the new-fuel storage room. l 3.2.1.4 Spent Fuel Pool l Cooling of the spent fuel assemblies during the radioactive decay period is accomplished in a stainless steel lined reinforced concrete pool filled with borated water. Space is provided in the pool to place the spent fuel shipping cask. The poolis serviced by means of the yard crane, as well p as a moveable plattorm with hoist. A new-fuel area adjoins the spent fuel pool. The pool is V Rev.16 DSAR 3-28
i l I MYAPC l - p designed to safely resist the hypothetical earthquake or tomado, as well as the applied loads of the ' water and fuel. The pool has a reinforced concrete floor founded on rock and sidewalls 6 feet thick which extend from 12 feet 6 inches below ground grade to 26 feet above ground grade. The concrete is reinforced with #11 bars at 12 inch center to center spacing with a yield strength of 40,000 psi. The concrete has a 28 day minimum compressive strength of 3,000 psi. The reinforced spent fuel pool was originally designed in accordance with ACl-318-63 to resist the appropriate dead, live, hydrostatic and maximum hypothetical seismic loadings. The structure was reanalyzed, in support of EDCR 92-111, to demonstrate the acceptability of installing the new high density spent fuel storage racks. As part of the preliminary decommissioning activities, 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 212*F. Complete details of this evaluation are contained in References 3.2-1 and 3.2-2. The pool is completely lined 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 tell tale pipes, {} v each with a flow limiter at the end of the pipe in the event of a malfunction of a liner weld, the leakage through each telltale is limited to less than 2.5 gpm. The liner is designed as a ASME Section Ill, Division 2, Paragraph CC-3720, Liner, Table CC-3720-1, Service Category, Membrane. The plate materialis ASTM A240, Type 304 stainless steel. Liner Anchors are designed to ASME Section Ill, 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 E309 (carbon to stainlet.' steel) with a minimum tensile strength of 81,000 psi. The fuel transfer tube was originally designed as safety class 2; however, since the containment integrity design basis is not applicable in the defueled condition, it has been reclassified as safety class 3. It consists of a 36-inch OD,3/8 inch thick, ASTM A312 TP304, stainless steel pipe installed 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 joints, backed up by a packed slip joint to compensate for any differential movement. I Structural steel supports a superstructure of protected metal siding which encloses the pool. The (m steel framing above the pool is designed for earthquake and tornado to prevent it from falling into O the pool and damaging fuel assemblies. The masonry wall at the south end of the fuel building is DSAR 3-29 Rev.15
MYAPC O not designed for certain wind or earthquake loadings, and, therefore, an evaluation of the consequences of a wall collapse was performed. The analysis demonstrated adequate spent fuel 4 pool cooling capability and structural rack integrity. l 3.2.1.5 Fuel Storage Racks The new and spent fuel pool structures including fuel racks are designed to withstand the anticipated earthquake loadings as Class I structures in accordance with the guidance of Regulatory Guide 1.29. Analyses show that the racks will perform their intended function under both seismic and load drop loadings in accordance with Regulatory Guide 1.124 and NUREG-0800. The design ensures that during the event, rack-to-rack and rack-to-wall interaction is appropriately considered. Structural material used in the rack design is ASME Section 11, SA 240, Type 304 stainless steel. The design considered thermalloads induced by an operating temperature of 154*F. Subsequently 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 all computer aided mechanical analysis. The design considered impact loads from a fuel element dropped 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. The dropping of objects over the O k storage array was conservatively analyzed by assuming that the dropped object is twice the weight of a sta'idard assembly. Suberiticality and a coolable geometry are maintained and damage to the stored fuelis minimized. The racks consist of individual storage cells joined into a rack module. The racks are a single tier, rectilinear array of free standing modules, not anchored to the pool walls, floor or adjoining racks. Each rack module is provided 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 plate 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 ll contains 21 racks spaced on a minimum of 9 ir'ch centers. Spent fuel storage racks may be moved only in accordance with written procedures which ensures that no rack modules are moved over fuel assemblies. 1 I c Rev.14 DSAR 3-30 l l* l l
1 MYAPC 7 3.2.2 Storage Buildings (G 3.2.2.1 Underground RCA Storage Bunker The underground RCA storage bunker, also referred to as the high rad bunker, is located within the protected area about 120 feet northwest of containment. The bunker is a reinforced concrete structure and is partially buried below yard grade. The bunker is 27.5 feet by 16 feet and approximately 12 feet high. The top of the bunker is about 5 feet higher than the surrounding yard i grade. The bunker is divided into five internal compartments each separated by 18 inch 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 liquids collected in the bunker to a sump. Liquids collected in the sump would be pumped to the spent resin pit sump in the nearby RCA storage building through an underground pipe. The spent resin pit sump discharges to the Aerated Drain System where any liquids would be collected for processing. The bunker provides temporary storage for radioactive wastes before they are moved to the low level waste storage building, described in Section 3.2.2.5, for longer term storage. The outside yard crane is available to move waste containers in and out of the bunker. Waste is typically stored O in the bunker for up to about one year to allow for some decay before being placed in the low level b waste storage building until arrangements are made for permanent off-site disposal. 3.2.2.2 Radiation Controlled Area (RCA) Storage Bui! ding This RCA storage building, adjacent to the fuel building, houses the Duratek skid waste processor, a decontamination area, and the waste solidification system. The waste solidification system includes: the waste resin storage tank (TK-85), waste holdup tank (TK-95), resin holdup tank (TK-109), the resin transfer pump (P-144), resin holdup tank dewaterir.g pump (P-145), and dewatering pump (P-123). Waste storage and holdup tanks are appropriately shielded. The area provides space for solidification agent storage and a shipping container filling station. The solidification agent is thoroughly mixed to a homogeneous mixture which is then sealed in approved shipping containers and allowed to solidify befere shipment to an approved waste disposal site. This area is also useo as an equipment decontamination facility. The ventilation system can be operated to maintain the waste processing and equipment decontamination area under a slightly negative pressure. i fs Rev.14 DSAR 3-31 l l
MYAPC 3.2.2.3 Low Specific Activity (LSA) Storage Building This building houses the LSA compactor and serves as the storage building for LSA containers to l keep them free of damage or deterioration. It extends from the south wall of the RCA building to the containment, but is not structurally attached to either. The building is designed non-nuclear safety. The LSA compactor compresses LSA material. It is vented into the RCA filtered ventilation system. The LSA sump pump discharges to the spent resin pit sump in the RCA Building. 3.2.2.4 Warehouse This building is located outside the RCA. The building is used to store replacement components l and equipment. l 3.2.2.5 Low Level Waste Storage Building The low level waste storage building is located on the plant site, outside of the protected area. The building is 154 feet by 64 feet by approximately 25 feet high from ground to roof line with floor elevation at 26 feet. The outer walls of the building consist of steel siding with one foot thick concrete shield walls inside the siding to a height of 16 feet. A truck bay is included to allow access for shipping waste containers. The building provides for interim on-site storage of low level waste and for storage of contaminated equipment. I 3.2.3 Service Building i The service building floor plan is shown on Figures 3.2-3 and 3.2-4. This building houses the l warehouse and tool crib, the instrument and plant offices, the secondary chemistry laboratory, and the RCA machine shops. The control room, the personnel service area, lockers, washrooms, i lunchroom, are also located on the first floor of the service building. The building floor plan shows l the layout of the offices and shops servicing within the service building. The RCA and secondary services are separated in such a way so as to effectively control traffic through a health physics check point. A section of the building is built up to provide two floor areas that house the cable tray and the switchgear areas at Elevation 35 feet 0 inches and 45 feet 6 inches. l Rev.14 DSAR 3-32 i i L
4 MYAPC 3.2.3.1 Control Room Area 1 The main control room, located in the service building, contained many of the controls and instrumentation necessary to monitor and control various areas and equipment in the plant. The main control room is designed to be available at all times. Safe occupancy of the main control room during an abnormal condition is provided in the design of the service building. Adequate shielding and ventilation are used to maintain tolerable radiation and temperature levels in the main control room. The decayed source term in the defueled condition, in conjunction with the location and design of the control area, provides 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 possibility of a condition which could lead to possible inaccessibility or evacuation. In the event that this area becomes inaccessible, the controls for spent fuel pool l cooling and makeup, water treatment and waste disposal, are located at control stations remote from the main control room. The main control room is shown in Figure 3.2-5. The main control room arrangement gives consideration to the fact that certain systems normally require more operator attention than do others. The main control board is the central item in the main control room. The control board has a bench section and a vertical section. Annunciators are provided so that the control room operator is made aware of any deviation from normal conditions at remote control stations. Radiation monitoring system audible alarms are provided in . O the control room. Output information from the computer is displayed in the main control room for use of the plant operating personnel. The control room operators hierface with the plant process computer through a standard computer terminal keyboard and color monitor. The keyboard also contains "Special Function" keys which allow the operator to access frequently used applications by depressing a single key. Other applications are accessed through a variety of menu options so that the operatcrs only have to select an option and depress the transmit key to execute programs. Terminals of equal functionality are also located in the TSC and EOF for use by support personnel in the event of any situation which required activation of these centers. Printers are used to provide permanent operating records. One printer is used to record alarmed parameters and sequence-of-events logs (other devices provide trending capability), and one printer will print various plant logs as they are requested by the operators. Power to operate the plant computer system is supplied by a static inverter which is connected to Rev.14 O DSAR 3-33
MYAPC I " h one of the station batteries. The inverter is independent of station service AC power sources and thus the availability of computer functions is assured during plant transients or a loss-of-station service power. I 1 \ A continuous indication of noble gas fission product concentrations in the primary vent stack is i provided in the control room and meets the following criteria:
- 1. Measurement and indication is in cpm or mR/hr convertible to pCi/ce;
- 2. The systems are not designed to meet the redundancy requirements as defined in IEEE 279;
- 3. Radiation levels are continuously recorded in the control room during an accident. Tne major equipment associated with the plant computer system is located in the computer room. This computer equipment consists of termination cabinets, signal conditioning components, analog to digital converters, central processing units and bulk data storage devices. Additional support devices (console terminals, line printers, communication controllers and display generators) are also located in the computer room. Additionally, two cabinets for special terminations and the mounting of special hardware, including power supplies, converters, latching contacts and chassis for resistance temperature detector bridges are also located in i V the computer room.
l The main control room has three independent communication systems. One system consists of telephones which ara leased from the local telephone company. These telephones and several outside trunk lines service the station for outside calls. This system may or may not be available under emergency conditions. The second system is a communication and voice paging system and interconnects the entire station. The third system is sound-powered, with telephone Jacks and interconnecting wires at each major control point for test purposes. This system does not rely on any power source; therefore, it is available at all times. 3.2.4 Turbine Building The turbine building houses the secondary plant components and systems. Though the turbine building is not a Class I seismic structure, it b designed for wind forces which are greater than the seismic forces as determined from a combined seismic analysis of the service building, control room and the turbine hall. The turbine hall will remain intact during the DBE. The building is provided with I three operating levels: Rev.14 O DSAR 3-34 l l l l l l L________________.___
MYAPC First Floor Elevation 21 feet 0 Inches The main equipment located on this level as shown Figure 3.2-6 consists of the main condenser, condensate, heater drain and steam generator feed pumps, vertical feedwater heaters and the water treatment plant. None of the aforementioned systems or components are required in the defueled condition of the plant. Enclosures for the auxiliary boilers, diesel generators and for the control room are adjacent to the turbine building first floor operator level. The first floor level is extended to include the service building area. Mezzanine Floor Elevation 39/35 Feet 0 inches i l J The heater drain receivers, gland steam condenser and air ejectors are located on the mezzanine ! level. The principal function of this floor is to provide convenient access to the operating equipment and valve operating stations. None of the aforementioned systems or components are required in the defueled condition of the plant. The mezzanine floor plan is shown in Figure 3.2-7. Turbine Generator Floor Elevation 61 Feet 0 Inches The third floor area is occupied with the turbine generator, the moisture separators and the second , point feedwater heaters. The feedwater vertical heat exchangers project through and above the floor level. None of the aforementioned systems or components are required in the defueled condition of I the plant. The third level of the turbine building is shown in Figure 3.2-8. i 1 3.2.5 Primary Auxiliary Building (PAB) i Function ' i The PAB is located within the radiation control area (RCA) and is shown on Figures 3.2-9 and 3.2-10. Qescriotion The PAB has a concrete foundation on bedrock 9 feet below ground grade, with 2-foot thick reinforced ! concrete exterior walls extending 16 feet above ground level. Radioactive equipment and tanks within the structure are shielded as required with concrete walls. The concrete cubicle provided to shield the volume control tank extends above the roof of the building. All exterior and interior concrete walls and slabs are designed to safely resist the hypothetical earthquake. Rev.14 O DSAR 3-35
MYAPC Primary Auxiliary Buildina Eauioment The building contains the aerated drain and the hydrogenated drain tanks, the waste gas, and liquid waste disposal system described in Sections 3.3.6 and 4.4 of this report. The larger tanks of the waste disposal system are the boron waste storage and the test tanks which are located on the north and south side of the PAB. The containment pipe tunnel connects to the building at the 11-foot 0-inch floor elevation, and the associated piping and valves occupy most of the floor area. The lower level of the PAB contains the liquid transfer pumps, the auxiliary charging pump end the boric acid mix tank. The three charging pumps of the charging and volume control system are located on the second or ground floor of the building at Elevation 21 feet 0 inches. The upper floor plan view is shown in Figure 3.2-9. The containment purge fan, the PAB ventilation fan and the particulate and charcoal filter of the ventilation system are located on this floor. The waste gas compressors, the waste gas decay drums and the volume control tank are also located on the upper floor level. This { floor is provided with removable slabs over the filter and demineralized cubicles which are built up from the lower PAB floor at Elevation 11 feet 0 inches. The upper floor operating area is also provided with a monorail for the handling of expended filter elements. 3.2.6 Service Water Intake Structure The circulating water pumps and service water pumps are located in this structure. The building floor plans at Elevation 7 feet 0 inches and 21 feet 0 inches are shown on Figures 3.2-11. A concrete barrier separates the service water pumps from the circulating water pumps and piping. This barrier, in conjunction with high water alarms and circulating water pump trips, protects the service water systems from the effects of flooding should a rupture of the circulating water system occur.
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3.2.7 Fire Pump Building i The fire pump house is shown on Figure 3.2-12. It is located near the water storage pond and houses the equipment servicing the liquid portion of the fire protection system. The building houses two fire l pumps, a pressure maintenance pump and a hydro-pneumatic tank. A diesel engine drives one pump while the other is motor-driven. The building houses the diesel fuel tank, the batteries and control board required for the diesel operation. Rev.16 O V i DSAR 3-36 l I l
i MYAPO i 3.2.8 Masonry Walls Masonry walls in the vicinity of safety-related equipment have been evaluated to determine whether they will withstand all postulated design loads (seismic, tornado, hurricane, attached equipment, etc.). Table 3.2.1 lists the walls evaluated and notes those which are fully qualified and those which are assumed to fail, but such failure will not result in unacceptable damage to equipment. Bulletin 80-11 Category Definitions: Category 3: Walls whose collapse will not affect safety-related equipment. This list has been revised to address only those walls that are required in the permanently defueled condition. ROOM REFERENCE WALL EQUIPMENT BUILDING WALL ID LOCATION DRAWING LOCATION PROTECTED CATEGORY Fuel FB 44 1 Spent Fuel 11550-FA-12A South Wall Spent Fuel 3 I Pool Spent Fuel Purification Elev 44'-6" Building and Cooling Q Return Lines and spent fuel storage , racks. I 4 L 1 Rev.14 O DSAR 3-37 l__-____________________-_--_______________________.
MYAPC Section 3.2 -
References:
1
- 1. YNSD Letter 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 Calculation No. MYC-2001," Analysis of SFP Structure for Elevated Pool Water Temperature," dated November 25,1997.
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MYAPC
.r. l 3.3 Systems 3.3.1 Fuel Storage 3.3.1.1 Fuel Storage Design Basis Design Cntena The fuel storage design criteria are as follows:
- 1. Criticality in new and spent fuel storage is prevented by physical design features or processes.
A geometrically safe configuration' is emphasized over procedural controls.
- 2. Appropriate shielding is provided to meet the requirements of.10CFR 20.
- 3. The fuel building is continuously monitored by area specific detectors. Audible and visual
' alarms are activated at the detector locations and the main control panel for radiation levels in excess of predetermined limits.-
4.~ Spent fuel storage systems are:
- a. designed to prevent or mitigate accidents which could lead to the release of significant amounts of radioactmty affecting the public health and safety,
- b. designed, fabricated and emeted to withstand, the additional forces that might be imposed by natural phenomena.
The spent fuel pool cooling, make-up, and purification system design criteria is as follows:
- 1. Prevent damage to spent fuel which could cause radioactive release to plant operating areas or the public environs.
- 2. Maintain adequate level of poolwaterto provide shielding for radiation protection.
- 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 liquid level alarm, and a high temperature alarm which will indicate loss of continuity in decay heat removal.
Desian Basis I-The fuel storage and handling design basis is as follows:
- a. Store 1432 complete spent fuel assemblies,2 partially consolidated assemblies, and 2 partially full failed fuel rod containers in the spent fuel pool in addition, other irradiated components such as CEAs are stored in the fuel assemblies. The total maximum decay heat load to be stored in the pool is s5.96 x 10' BTU /hr. This assures that the calculated time-to-boil and boil-Rev.14 DSAR 3-52
1 MYAPC off rates are adequate to provide ample time for operators to take remedial actions in the event of a loss of forced cooling incident.
- b. The design basis for preventing criticality is that, considering the possible variations, there is 95% probability at a 95% confidence level that the effective multiplication factor (K,,) of the fuel assembly array will be less than 0.95 for flooded conditions without credit for borated water,
- c. A minimum boron concentration of 670 ppm was assumed in the analysis for the postulated ]
misplaced fuel assembly. A higher administrative limit is required by technical specifications. The minimum concentration was determined to be required to maintain a K.,less than .95 for the most reactive accident condition. This concentration is required whenever fuel assemblies are stored in the pool and a spent fuel pool assembly placement verification has not been j' performed since the last movement of fuel assemblies in the spent fuel pool. The analysis considered various disposition configurations including a fuel assembly laying horizontally on the top of the spent fuel racks, a fuel assembly placed adjacent to or on the outside comer of the fuel racks and a fuel assembly placed in the wrong rack.
- 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 exceeds evaporative losses. Makeup supplies are available in the event of an extended loss of offsite power incident.
- e. The new and spent fuel pool structures, including fuel racks, are designed to withstand the anticipated earthquake loadings as Class I structures. The spent fuel pool is lined with stainless steel to ensure against loss of water.
l
- f. The spent fuel racks are designed to Seismic Class I requi ements, capable of sustaining a temperature of 212*F, and able to withstand the dropping 'l a fuel assembly from a height of 18" above the top of the racks without incurring damage which could result in criticality.
- 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 full tomado wind pressures on the bare structural steel frame)
Rev.16 DSAR 3-53
i ! MYAPC
- h. The dose rates at the surface of the SFP from spent fuel assemblies do not exceed 2.5 mrem /hr in fuel building passageways during normal storage (i.e., 50 mrem /hr during fuel ;
handling 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 area. l 1 ! i. Siphon protection is provided to preclude inadvertent draining of the SFP to a level lower than l approximately 10 feet above the top of the active fuel. Additional margin (to approximately 19') ] was provided by the installation of a branch syphon break. ] ! j. The minimum spent fuel pool water bulk temperature shall be 68'F.
- k. Spent fuel storage racks may be moved only in accordance with written procedures which ;
ensures that no rack modules are moved over fuel assemblies. l 3.3.1.2 Fuel Pool Storage System Description The fuel pool storage and handling system is designed to prevent or mitigate the consequences i of accidents associated with the storage of fuel. A description of the fuel and irradiated n components is presented below followed by the structures, systems, and components supporting V the safe storage and handling of the spent fuel and irradiated components. A description of the spent fuel handling system equipment is contained in section 3.3.2. l 3.3.1.2.1 Fuel Ascemblies (Typicai) Figure 3.3-1 shows the details of the fuel assembly. Except for the two consolidated assemblies, all assemblies consist of five guide tubes, nine fuel spacer grids, upper and lower end fittings ^ tel rode and bumable absorber rods, if required. The number and type of bumable absorber rods in each type of assembly were cycle 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 end fittings and the spacer grids are welded or mechanically attached to all five guide tubes. A sleeve is provided in the upper region of the guide tube. The sleeve is made of stainless steel and is chrome plated on the ID. The sleeve is positively located within the guide tubes and components of the upper end 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. Rev.16 DSAR 3-54
i l MYAPC l (] A hold-down device is incorporated into the upper end fitting and features a hold-down plate which ! V acts on the underside of the fuel alignment plate (Figure 3.3-1). The hold-down plate is loaded by l coil springs which are located at each of the guide tube posts. The inconel springs are positioned l at the upper end of the assembly so that the spring load combines with the assembly weight in counteracting the upward hydraulic forces. The springs are sized and spring preload selected, j such that a sufficient net downward force will be maintained for all normal and anticipated transient flow and temperatures conditions. The fuel assembly upper end fitting is a cast stainless steel structure. It serves as an attachment for the guide tubes and as the lifting fixture. The fuel rod spacer grids (Figure 3.3-2) maintain the fuel rod pitch over the full length of the fuel j I rods. Typical grids are fabricated from preformed strips interlocked in an egg crate fashion and welded together. Each fuel rod is supported by springs and arches opposite the springs. The j springs press the rod against the arches to restrict relative motion between the grids and the fuel rods. The springs and arch positions may be reversed from grid to grid to provide additional i restriction to relative motion. The perimeter strips also contain springs and arches in addition 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 discrete bumable absorber rods. A brief description of each type of rod is contained below. FuelRod The fuel rods consist of UO 2pellets, a compression spring and in the ABB-CE and EXXON design spacer discs, all encapsulated within a Zircaloy-4 or ZlRLO tube. The UO2 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 criteria regarding clad stresses and strains, and transfer of heat from the pellets. The compression spring located at the top of the fuel pellet column maintains the column in its proper position during handling and shipping. It also provides support for the clad in the plenum region to prevent local buckthg. In the ABB-CE and EXXON designs, an alumina spaceris located at the upper end of the fuel pellet stack to insulate the plenum region and to prevent UO2 chips from entering the plenum region. In fpd Rev.14 l DSAR 3-55
MYAPC ABB-CE fuel, an alumina spacer is located at the bottom of the fuel pellet stack to reduce lower end cap temperature. The plenum above the pellet column provides space for axial thermal expansion of the fuel column and for accommodation of fission gas. Discrete Bumable Absorber Rods Description Discrete bumable absorber rods are mechanically similar to fuel rods except that they each contain a column of bumable poison pellets instead of fuel pellets. The poison material is alumina with uniformly dispersed boron carbide particles. The poison material is a 122.7 inch long column approximately centered within the 136 to 137 inch long active fuel region. Components stored within the fuel assembly guide tubes in the spent fuel pool are: control element assembly and neutron source assembly. A description of each is provided below. Control Element Assembiv The CEA (shown in Figure 3.3-3) compnsed of five inconel tubes 0.948 inches in diameter. Each tube (finger) is sealed by welded end caps. A gas expansion space is provided to limit maximum tubes stress due to intemal pressure developed by the release of helium gas and moisture from the O boron carbide. The overall length of the control elements assembly is approximately 161-5/16 inches. Four fingers are assembled in a square array around the centrally located fifth finger. The fingers are joined by an upper end fitting. i Most CEAs are full-strength and comprise five active fingers containing B,C pellets. The four outer fingers on all CEAs and all five fingers on most CEAs have silver-Indium-cadmium filled tips. Some CEAs are designed as part-strength CEAs, utilizing stainless steel pellets in place of B,C in selected fingers. The full strength and part-strength CEAs are described in Figure 3.3-4. : I t'mtmn Source Assemblies l
. Between two and four neutron source assemblies were installed in the reactor for each cycle. Each source may include both a startup and sustainer source, orjust a sustainer source. The sources are held in vacant CEA guide tubes by means of protrusions which fit into slots cut into the top of the upper end fitting posts.
l O Rev. ', , DSAR 3-56 u_________________-_____ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
MYAPC 3.3.1.2.2 Fuel Storage System Description i This section identifies the SSCs which comprise the fuel storage system. A description of each is contained below.
. Fuel pool
- Transfer tube and isolation valve
. Spent fuel storage rackn
. 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 building adjacent to the reactor containment. The poolis constructed of reinforced concrete with a wall and floor lining of 1/4-inch thick stainless steel. The liner ensures another level of defense 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 Q
U approximately 59,116 ft8. The fuel pool is protected from being drained down to a level approximately 10m feet above the ] active fuel assuming a rupture of any pipe normally connected to the pool. The consequences of the i siphoning incident is discussed in section 3.3.1.3. The anti-siphon device on the suction side of the fuel pool cooling pumps is safety class 3, seismically designed and periodically inspected for debris. The desigri of the pool, in conjunction with analyses (MYC-2001) concluding that the MY spent fuel pool concrete structure and liner are capable of resisting thermal stresses due to prolonged elevated pool water temperatures of 212' F without failure, provides reasonable assurance that the pool water level will not be affected by credible incidents other than evaporative losses due to normal and loss of forced cooling incidents. Transfer Tube and Isolation 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
'As noted in Section 3, a branch syphon break has been installed which increases the margin to approximately 19*. ]
Rev.16 DSAR 3-57
L MYAPC independent mechanical devices; by an isolation valve (FP-21) located in the spent fuel pool and a !\ blind flange seal at the refueling canal end. Spent Fuel Storace Racks The'26 stainless steel, poison curtain storage racks provide for the storage of a maximum of 1758 l fuel assemblies. The fuel pool is separated into two regions.' Region I contains five racks to store up to 228 fuel assemblies, spaced on a minimum of 10.5 inch centers and is designed for initial
~
enrichments up to 4.5 weight percent U-235. Region il contains 21 racks to store up to 1530 fuel assemblies, spaced on a minimum of 9 inch centers. Figure 3.3-6 shows the high density spent fuel pool layout for the two region pool. The spent fuel is stored in borated water. Water is necessary to provide shielding and cooling for removal of the decay heat being released by the spent fuel assemblies. The new and spent fuel pool structures including fuel racks are designe ; i withstand the anticipated earthqu-ke loadings as Class I structures. Refer to Figure 3.3-5 for a description of
- restrictions for fuel which may be placed in each region.
! The racks are a single tier, rectilinear array of free standing modules and are seismically designed. The design ensures that during the event, adequate distances between the assemblies is maintained to prevent a criticality event and to prevent adverse interaction between the pool structure and the racks. Structural material used in the rack design is ASME Section 11, SA 240, ~,'ype 304 or ASTM l A240 Type 304L stainless steel. , 4 The poison material selected as a neutron absorber is Boral; a cermet composite material made of i boron carbide particles in a Type 1100 aluminum matrix. Boral Panels consist of 2 outer sheets of 1 type 1100 aluminum that clad a sintered plate of boron carbide in a type 1100 aluminum matrix. i New Fuel Storaae l The new fuel storage area, located at elevation 31 feet in the fuel building, contains facilities to l receive, handle, and store new fuel assemblies 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 defueled condition, there are provisions for storing 160 fuel assemblies in the new fuel storage area. New fuel may not be placed in the pool. I i Rev.14 DSAR 3-58
MYAPC
-3.3,1.2.4 Fuel Pool Cooling, Makeup and Purification System Description Fuel Pool Heat Exchanaer The fuel pool heat exchanger is a cross-flow heat exchanger of the shell-and-tube design. The heat exchanger is designed to cool the fuel pool water based on the following conditions:
Shell Side Tube Side 104*F in and 111*F out 114*F in and 110*F out ] 3.33 x 108 BTU /hr heat transfer 3.33 x 108 Btu /hr heat transfer ] The fuel pool heat exchanger is designed for 150 psig and 225'F on both the shell and tube sides. The tubes are 304 stainless steel, and the shell is of carton steel. The above values are based on ] a tubeside flow of 1700 gpm, a shellside flow of 1000 gpm and an outside temperature of 87*F. ]
]
]
1 Fuel Pool Coolino Pumos v There are two fuel pool cooling pumps designed for maximum efficiency at 772 gpm per pump at a total pump head of approximately 80 ft. The pumps may be operated at other flows and heads. The casing is designed for 125 psi at 250*F. The intemal wetted surfaces of the pumps are type 316 stainless steel or an equivalent cast form. The pumps are installed for parallel operation but only one is necessary for normal and incident heat load conditions. i Fuel Pool Purification Pumo The fuel pool purificsdon pump is designed for 250 gpm. The casing is designed for 200 psig at 250*F. The intemal wetted surfaces of the pump are type 316 stainless steel or an equivalent cast form. The pump is used for fuel pool cooling when the heat load is low, and it can also be used for l skimming operations. skimmers are provided to collect precipitant on the water surface and prevent it's build-up on the pool wall surface.
)
Fuel Pool Filters The fuel pool prefilter and post filter are of the cartridge type and are located in the ] Rev.16 DSAR 3-59 l l
MYAPC purification loop to remove the particulate or undissolved solids. The prefilter retains 95% of the particles nominally 2 micron, or larger, at a flow of 200 gpm. The vessel is constructed to Type 304 stainless steel and is rated at 200 psig and 260'F. The post filter retains 95% of the particles ] nominally 5 micron, or larger, at a flow of 200 gpm. The vessel is constructed to Type 304 stainless steel and is rated at 200 psig and 250*F. Fuel Pool Demineralized
]
A mixed bed, underwater demineralized is located in the spent fuel pool to remove dissolved solids ] from the water by ionic exchange with the resin. The vessel is constructed primarily of Type 304 ] , stainless steel and has a design pressure of 50 psig. The vessel rests on the fuel pool floor at the ] I 7.5 ft. elevation and circulates fuel pool water, at a nominal rate of 100 gpm, using an intemally ] mounted pump and motor. Fuel pool water is drawn into the vessel by the pump, through two inlet ] l connections, and through the process bed. Water is discharged from the vessel through a hose ] ; assembly which distributes the water to a remote area of the fuel pool, ensuring proper pool ] l circulation. The demineralized is sized to contain approximately 30 cubic feet of process media. ] i Fuel Pool Svstem Valves
)
x All valves in the fuel pool system are austenitic stainless steel. The gate and check valves are butt-welded. The diaphragm and ball valves are socket-welded. Fuel Pool System Pioino All the piping used in the fuel pool system is type 304 stainless steel with welded connections ! l throughout, except for flanged connections at the pump suction and discharge. Code Requirements The design of all components in the fuel pool cooling system complies with the following codes and regulations: l Fuel Pool Heat Exchanger Tube side ASME Section ill Class C Paragraph UW-2(a) of Section Vill applies Shell side ASME Section Vill Rev.16 DSAR 3-60
- j. MYAPC Fuel Pool Cooling Pumps No code !
Fuel Pool Purification Pump No code ! Fuel Pool Filters ASME Section ill Class C Paragraph UW-2(a) of Section Vill applies Underwater Demineralized ASME Section Vill, Not Stamped ] Fuel Pool System Piping ANSI B31.1 l- 3.3.1.3 Design Evaluation 1 ! Several considerations and analyses determine the safety of fuel pool handling and storage i operations. These include: Thermal Analysis Radiological Analysis l Fuel Assembly Structural and Material Considerations Fuel Pool SSC Structural and Material Considerations Criticality Analysis 'O U The passive design features of the pool prevent significant loss of pool water and assure that sufficient time is provided for operators to identify any incident condition and take the necessary ; action to restore cooling, and, if required, provide makeup to the spent fuel pool. The design and administrative controls assure that the fuel is safely stored and the public and workers are adequately protected. Several incident scenarios are described below. These include: No pumps running (loss of forced cooling) Loss of heat sink Siphoning Fuel Handling incident Cask Drop Incident To determine the significance of these events, each of these incidents are reviewed in terms of the j design criteria and design basis requirements. ' An evaluation is performed for each incident to determine the time available to the operator to discover the incident and take remedial actions. The consequences of each event are then evaluated in terms of pool water level and the radiological consequences. Structures, systems, and components, credited in preventing or Rev.16 j DSAR 3-61 l
MYAPC
' mitigating the event are then classified according to its safety function. Similarly, activities credited to mitigate the condition are then defined. SSCs which are safety class are contained in section 3.1.
A discussion of the analyses and considerations pertaining to the defueled condition is contained
- below followed by a description of the incidents and the consequences.
l 3.3.1.3.1 Analyses and Considerations l l Thermal Analvsis l The heat load as a function of time is provided in Table 5.5.2. The reduced heat load on 12/29/97
- l. is calculated to be 5,420,000 BTU /hr. which approximately represents a 75% reduction in heat load from the previously analyzed full core off-load scenario when the plant was operating. The revised residual decay energy release rate was calculated using the assumptions and formulations of Branch Technical Position (BTP) ASB 9-2 " Residual Decay Energy for Light Water Reactors for Long-Term Cooling" and is consistent with the guidance contained in the dtandard Review Plan (NUREG-0800),
7 section 9.1.3. It also considers the long term uncertainty f' action for cooling times greater thar,10 seconds and are very conservative. iO Time to Boll In this analysis, the " time to boil" evaluation was performed for three levels (elevations) of water, each at five initial pool temperature conditions. The approximate elevations shown in the following table: Elevation Height from Height above Description l ( l (ft.) pool floor (ft.) active fuel (ft.) ) 31.5' 23.5' 10' This elevation corresponds to lowest point on the cooling water retum piping into the q I SFP. 40' 32.5' 19' This elevation corresponds to the level at which the cooling water suction piping penetrates the liner. 43' 35.5' 22' This elevation corresponds to the water level one foot be:ow the normal pool water (i.e. the low level instrument setting) O Rev.14 DSAR 342 i
MYAPC i o i ) The time to boil is also a function of the initial temperature of the pool water. Accordingly, for each elevation, time to boil is calculated at initial pool water temperatures of 80*F,100*F,120*F,140*F, and 154*F. Time to boil was calculated for each of the three water levels for differing heat loads as l a function of time after shutdown and is shown in Table 5.5.2. A spent fuel pool heat load test, hereinafter refen ed 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, convection, and evaporation results in a significantly longer " time-to-boil" and lower " boil-off" rate. It is therefore more representative of pool cooling than the BTP ASB 9-2 calculation results. As a nale-of-thumb, the representative " passive cooling" results are approximately 71% of the BTP ASB 9-2 results. An example of the differences are illustrated in the table below. Initial wats-level = elev. 43' BTP ASB 9-2 Results Passive Cooling Results Final water level = elev. 31' 12/29I97 12/29/97 initial Pool Temp. = 80'F 5.42E+6 BTU /hr 3.85E+6 BTUIhr Time to Boil 67.36 hrs. ; 2.8 days 93.5 hrs. ; 3.9 days Boil off Rate 11.64 gpm ;1.47 ft/ day 8.3 gpm ; 1.05 ft/ day A)
! Days to reach 210 ft. above active 8.16 days 11.42 days fuel l
Radiological Analvsis An evaluation was performed to determine the upper limit radiation fields in the fuel building and at surrounding onsite and offsite locations resulting from gamma radiation from the spent fuel in the fuel pool. This evaluation was performed as a function of both decay and water level. The results are contained in Table 5.5.3. The analyses are conservative for the following major reasons:
- 1. The SFP cross-sectional area was assumed to be loaded with discharged assemblies, 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 shine radiation was based on the assumption that the gamma rays escaping from the fuel pool had a spectrum identical to the uncollided spectrum originating within the fuel, thus o allowing the radiation to propagate to farther distances.
U Rev.14 l DSAR 3-63 l i _ _ _ _ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ _ .
MYAPC 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 platform.
- 2. A water depth of about 11 feet over the fuelis needed for a dose rate less than 1 mrem /hr on the platform immediately above the pool.
- 3. For 23 feet of water above active fuel, the dose rate on the over-pool platform is less than .0003 prad/hr, l
- 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 fmm the SFP center) is approximately 15 mrem /hr:
this sky shine radiation level drops to approximately S rem /hr at 610 meters (exclusion area boundary). The siphoning incident assumes that siphoning occurs from the cooling water retum piping inlet and represents the worst case scenario. This incident results in a worst case pool water level approximately 10 feet above the active fuel'. Radiological doses are illustrated in the table below. ] Water Level at 10 Feet Above the Active Fuel Decay time Sky shine at Radiation field on platform Radiation field on 46' since 12/6/96 610 meters directly above pool elevation beside pool (EAB) 1 year .0003 urem/hr < 4 mrom/hr < 2 mrem /hr 2 years data not < 2 mrom/hr < 1 mrem /hr available 3 years .00007urem/hr < .8 mrem /hr < .4 mrem /hr Radiological consequences are discussed for each incident resulting in a loss of water level. Consistent with the operating plant requirements, dose rates around the fuel pool can normally be expected to be less than 2.5 mrem / hour during normal spent fuel pool storage conditions. During fuel handling operations, the maximum dose rate to the operators is administratively maintained below 50 mrem / hour. This limit is preserved when analyzing the incidents below. Radiological effects are discussed in terms of direct radiation field, sky shine, pool surface evaporation, and in the l~ case of the fuel handling incident, the decontamination 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
~
]
O 'As noted in Section 3 a branch syphon break has been installed which increases the margin to approximately 19'. Rev.16 DSAR 3-64
MYAPC 1.8 mrem / hour (organ-lungs). The decontamination factor is a function of water level above the fuel ] pin. lodine is, not a factor unless there is a breach of the fuel pin (e.g., fuel handling incident). l Even then, the short half life of iodine coupled with the fuel decay time minimizes the source term !' dose affects contributed by radio-iodines. l Fuel Assembly Structural and Material Considerations l The fuel assemblies are designed to maintain their structural integrity under steady-state and transient operating conditions, as well as under normal handling, shipping, and refueling loads. The design takes into account differential thermal expansion of fuel rods, thermal bowing of fuel rods and CEA guide tubes, irradiation effects, and wear of all components. Mechanical tolerances and clearances were established on the basis of the functional requirements of the components. All components, including welds, are highly resistant to corrosion in the reactor and fuel storage environment. Each core was made up of 38,192 Zircaloy 4 or ZlRLO clad rods in 217 assemblies. The rods contain slightly enriched uranium in 'na form of sintered UO2 pellets, bumable absorber, or water-filled rods. The principle design structural criteria for the fuel rods is that the predicted permanent strain of the cladding is less than 1.0% during the fuel lifetime. I ! Fuel Pool SSC Structural and Material Considerations The fuel pool wall and floor slab is constructed of 6' 0" thick reinforced concrete with a wall and floor lining of 1/4-inch thick stainless steel to ensure against loss of water. The new and spent fuel pool j i structures including fuel racks are designed to withstand the anticipated earthquake loadings as seismic Class I structures. The spent fuel pool liner is designed QAR, A structural evaluation was performed to determine the dead load and hydrostatic force affect on the SFP concrete wall, floor stab, and stainless steel liner when the pool is subjected to an elevated pool water temperature of 212*F. Even when revising the thermal load factor, ACI allowables were met l for a differential temperature of 186*F. Further qualitative discussions showed that even if the ACI l allowables were exceeded (i.e.,long term loss of all SFP cooling during the coldest possible winter f temperatures), the water retaining capacity of the pool would be maintained since the pool side concrete remains in compression and the liner has sufficient ductility 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 capacity to resist the forces and moments generated by the self [ weight, water pressure, and thermal effects due to the boil-off of the pool. The load factors assumed l for the evaluation are for a service load condition and therefore does not restrict the elevated temperature condition to a, one-time only, limited duration incident. \ DSAR 3-65 Rev.16
MYAPC f The spent fuel racks are designed to Seismic Class I requirements, constnacted of ASME Section f3 V 11, SA240 Type 304, or ASTM A240 Type 304L stainless steel. The steel alloy is reasonably corrosion resistant to the oxidizing effects of most electrolytes at low concentrations. The steel is susceptible to corrosion in acidic solutions (pH less than 7.0) containing chloride or fluoride anions and can lead to pitting of the material. Control of the water impurities are provided by the SFP purification demineralized and filters and a chemical control program to assure that impurities are ! minimal and will not affect the structural integrity of the racks, liner, piping, vessels, exchangers and l pumps. The neutron absorber poison material is "Boral"; a cermet composite material made of boron carbide particles in a Type 1100 aluminum matrix. This composite is highly durable and heat resistant. Boral panels consist of two outer sheets of type 1100 aluminum that clad a sintered plate of boron carbide in a type 1100 aluminum matrix. The type 1100 aluminum material imparts sufficient corros'u. . resistance by forming an aluminum oxide layer on its surface when exposed to oxidizing agents. This oxide layer is stable in environments with a pH range from 4.5 to 8.5. The Boral panels are installed snug between the outside wall of the storage cells and the 304 stainless steel sheath that is welded to the wall. Vent holes at the corners of the stainless steel sheath create a sufficient vent path for l any potential hydrogen produced by a water-aluminum reaction. In addition, a Boral surveillance program has been established to characterize the performance of the Boral panels. The chemical control program and the Boral surveillance program provide reasonable assurance of the integrity of the system components and the capability of the components to perform their intended function. All piping used in the fuel pool system is type 304 stainless steel, as is the fuel pool purification filter and demineralized vessels, liner plates, and the tube side of the fuel pool heat exchanger. The l pumps (spent fuel cooling pumps and purification pumps) are constructed of type 316 stainless steel. I Each type of stainless steel is reasonably corrosion resistant. Criticality Criticality of fuel assemblies is precluded by adequate design of fuel transfer, shipping, and storage facilities, and by administrative control procedures. The two principal methods of preventing criticality l are limiting the fuel assembly array size and limiting assembly interaction by fixing the minimum l I separation between assemblies and/or inserting neutron poisons between assemblies. The design basis for preventing criticality is that, considering the possible variations, there is 95% j i O Y l l l l DSAR 3-66 Rev.14 1 l [ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _
MYAPC i probability at a 95% confidence level that the effective multiplication factor (K,) of the fuel assembly
\ array will be less than 0.95 for flooded conditions as recommended in NUREG-0800. The following conditions were assumed to meet this design basis:
- 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 yields the largest reactivity. I l 3. The array is either infinite in lateral extent or is surrounded by a conservatively chosen reflector, whichever is appropriate for the design.
- 4. Mec%nical uncertainties are treated by either using " worst case" conditions or by performing sensitivity studies and obtaining appropriate uncertainties.
- 5. Credit is taken for the neutron absorption in structural materials and in solid materials added l specifically for neutron absorption.
- 6. Where borated water is present, credit for the dissolved boron is not taken except under postulated accident condition, where the double contingency principle of NUREG-0800 is applied which requires at least two unlikely, independent, and concurrent events to produce a criticality I
accident. For fuel storage application, water is usually present. However, the design methodology also prevents accidental criticality when fuel assemblies are stored in the dry condition. For this case, possible source of moderation such as those that could arise during fire fighting operations are included in the analysis. The design basis K,is 0.98, as recommended in NUREG-0800 for 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 uncertainty added to the criticality calculations include the following components:
- 1. Method Uncertainty: based on critical experiment comparisons to establish method bias and variability.
- 2. Statistical Uncertainty: based on the statistics of the particular Monte Carlo calculations.
- 3. Fabrication Uncertainly based on manufacturing and mechanical tolerances, such as thicknesses and spacings. Either statistically- based uncertainties or " worst case" assumptions are used.
O C l l DSAR 3-67 Rev.14
MYAPC The criticality design criteria 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 condition, the reactivity effects of a misplaced assembly represents the bounding case for determining the boron concentration required to maintain a 5% ak/k safety margin to criticality in the spent fuel pool. The analysis considered various disposition coriJgurations including a fuel assembly laying horizontally on the top of the spent fuel racks, a fuei assembly placed adjacent to or on the outside comer of the fuel racks and a fuel assembly placed in the wrong spent fuel rack. A boron concentration of 663 ppm was assumed in the analysis for the postulated misplaced fuel assembly. This was determined to be required to maintain a 5% ak/k safety margin to criticality. This concentration 's required whenever fuel assemblies are stored in the pool and a spent fuel pool assembly placement verification 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 analysis report. Incident Evaluation Pool water level is normally constant except for relaHvely minor evaporative losses. For the purposes () of this evaluation, the pool water level is important from a standpoint of determining time available to the operator. The determination that adequate pool water level remains many hours or days after the incident is key to determining the severity of the event and the safety function performed by SSCs. Pool water level is important to meet the design basis requirement to maintain the fuel covered with water and for providing an end-of-incident water level which provides adequate shielding to maintain exposures below 10CFR 20 requirements. Consistent with the SRP criteria, the spent fuel pool and cooling, makeup and purification systems design assures that, in the event of a failure in suction or retum piping, the pool level is not inadvertently drained to a levet approximately 10 feet above the top of the active fuel. Significant elevations in the spent fuel pool are noted below: Description Elevation (approx.)
- Top of the spent fuel pool 46'0" High Water level (alarm) 45' Normal water level 44' j 43' I Low water level (alarm)
I Siphon breaker (suction side) 40'11" (g) DSAR 3-68 Rev.14 l l l
1 MYAPC Retum side 31' 6" Bottom of gate to the fuel transfer canal 12'0" ( Top of active fuel (approx.) Fuel pellets 21' (no higher than) Fuel pin 21'6" (no higher than) Bottom of the spent fuel pool T 6" i The lowest elevation where piping penetrates the liner is on the cooling water retum (inlet) 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 cause:: a siphoning action and subsequent draining of the pool to the 31.5 ft. elevation. This condition is analyzed later in this section. 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 extends 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-siphon device prevents inadvertent draining of the pool and, for this reason, is , ]n designed to seismic Class i requirements. Several administrative controls are in place to assure the continued safe storage operations. Spent fuel shipping casks are administratively 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 quickly identified and remedied. For instance, aiarms are provided to alert operators to off-normal pool conditions, including high temperature (110*F) and high/ low water level (45' and 43' elevations, respectively) and radiation level. Once alerted, operations personnel intervene before significant changes 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 annunciator would alert operations personnel. Similarly, a high temperature alarm in the control room would alert operations personnel to the high temperature condition. Plant operating procedures require the operator to investigate the alarm condition and initiate appropriate remedial actions. This instrumentation, as well 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 incidents. In the scenarios evaluated, credit was only taken for daily direct visual monitoring by Operations personnel of the SFP temperature and level. O Rev.14
'Q DSAR 3-69 l
I
MYAPC fO lV . Loss of Soeilt Fuel Pool Coolina l L The following assumptions are used in this scenario: ( 1, ' Initial pool water temperature is 212*F, l 2. Boil-off rate as of 12/29/97 equals 11.64 gpm (1.47 feet per day) and
- 3. Initial normal water level in the fJel pool (43' elevation), j l
- 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 field and sky
- i. shine doses to receptors.
l: 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
- l. cooling test results, it would take a minimum of 11 days to reach the 31' elevation.
At ten feet above the active fuel, radiation field dose rates are expected to be no greater than 4 mrom/ hour (whole body) plus evaporative dose rates of no more than 2 mrem / hour (whole body). This results in radiologeal dose rate of no more than 6 mrem / hour (whole body). An additional organ dose rate of 2 mrem / hour (lungs) is also expected. Historically,50 mrem /hr is the maximum dose rate permissible 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 approximately 40 rem /hr of sky shine to a receptor 20 meters from the pool center, and .0003 rem /hr at 610 meters (exclusion area boundary). I I l ( The above numbers are very conservative from 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 during 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 l elevation (44 ft. elevation). The SFP cooling pumps were then secured. It took approximately 70 I hours for the pool water to reach 140*F (Note that there was virtually no loss of water level during that heat-up period. 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 calculation results I Rev.14 l DSAR 3-70 L
l l I MYAPC l ~ (q/ show that the time-to-boil would take approximately 34 additional hours. Although not credited as l part of this safety analysis, when combining the empirical and analysis data, a mere reasonable estimate of the time to boilis approximately 100 hours. Siohonino incident There are four penetrations which project 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 ] l 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 I , daily operations area walk-through wherein pool temperature and level is checked. The pool drains ,V down to elevation 31.5 feet due to siphoning caused by a non-mechanistic pipe break 2. At this level, ) i no less than 10.5 feet of water above the active fuel remains. Assuming an initial temperature of 80*F and a heat load of 5.42E+6 BTU /hr, the time to boil is 39.5 hours. No credit is assumed for the low level alarm or for operations personnel to identify the existence of the displaced fluid. Operations personnel would detect the condition on their daily round no later than 30 hours after the incident and 9.5 hours prior to boiling. With 10.5 feat of pool water above the active fuel, radiological exposures to workers in the fuel building is no greater than approximately 2 mrem / hour (whole body) for a person on the platform directly over the pool, plus evaporative whole body dose rate of no more than 2 mrem / hour. At the time of identification, corrective actions are taken to investigate the cause of the incider t and start makeup to the pool. Under any credible situation, several makeup sources could be inidated within a few hoeirs 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: i
' A pre-existing piping penetration (at approximate El. 40') was utilized to create a return line syphon break. ] j
'The syphon break noted above makes this evaluation conservative ]
1 b Rev.16 l l DSAR 3-71 l
MYAPC
- 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 boiling (since the time of identification),
- c. A 25% allowance for the 24 hour surveillance is assumed (30 hours), and i
- d. The water is at a boiling temperature with a boil-off rate of 11.64 gpm or 1.47 ft per day.
- e. BTP AS89-2 decay heat loads are used l
With the above assumptions, after 62 hours (30 hrs. to identify the condition + 32 hrs. to effect !* remedial action), the waterlevel would be no lower than 8.5 feet above the active fuel. With 8.5 feet , l of pool water above the active fuel, a straight line extrapolation of radiation field dose rates indicates l that the operators on the platform directly over the pool would be subject to a radiation field dose rate of no more than 35 mrem / hour whole body plus evaporative whole body dose rates of no more than 2 mrem / hour, or 37 mrem total (whole body). An additional 2 mrem / hour organ (lung) dose rate , 4 would be received. By the edge of the pool, whole body dose rate is on the order of 20 mrem / hour. These totals still remain wdhin the 50 mrem / hour maximum radiation exposure limits indicated above for fuel transfer operations. Offsite dose rates are less than 10 rad /hr at 20 meters, and less than
.003 urad/hr at the exclusion area boundary (610 meters).
- . Several conservative assumptions are used in this incident. One assumption involves the 25%
i 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 available 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 pool temperature of 130*F or less. Finally, using the passive cooling results, significantly more time is available. 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 boil-off rate is 1.05 ft. per day which allows for additional time to reach the 8.5 ft level above the active fuel. Other conservative assumptions in the radiological analysis assume that all assembiies have the same operating history as the worst case assembly. Therefore, actual doses would be lower and a reduction in water level below 8.5 ft could cc. cur and still remain below the 50 mrem /hourwhole body dose to personnel at the edge of the pool. Assuming the conservative BTP decay heat loads and a boil-off rate of 1.47 ft./ day, it would take j approximately 40 hours for the level of the pool water to decrease from 8.5 feet to 6 feet above the ) active fuel. This corresponds to a whole body dose rate of approximately 750 mremihour to operators around the perimeter 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 approximately .15 ; prem/ hour to a receptor at the exclusion area boundary. In all cases, reasonable time is available i for operators to take corrective actions to restore make-up to the pool. Additionally, strict radiological I controls are effected to minimize exposures to personnel ALARA.
- Rev.14 1
DSAR 3-72 l
MYAPC Fuel Handlina incident For the fuel handling incident, no credit is taken for systems or components to mitigate the incident such as fuel building vent'lation systems or control room ventilation 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, enrichment and longest operating history) composite with one year decay. The accident assumed an instantaneous puff release at ground level meteorological conditions to determine offsite exclusion area boundary and control room doses. As shown in Section 5, doses are acceptable assuming no filtration. The occupational, control room and offsite radiological doses from the fuel handling incident bounds all of the fuel pool *.torage incidents described above. l Cask Droo incident 1 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 storage 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 O permeability 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. Available makeup capacity is significantly higher than the draindown rate of the spent fuel pool. Equipment location and drainage capability is such that no damage to critical l equipment from this leakage would occur. l The design standards and factors of safety, the documented maintenance program and operator qualifications, the strict supervision of all cask movement, and the small fraction of time that the cask passes over the edge of the spent fuel pool reduce the probability of a tipping fall to a tolerably low value. Notwithstanding, spent fuel shipping casks are administratively prohibited from being lifted l over the fuel pool. Loss of Heat Sink 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 shell side of the fuel pool heat exchanger and cools the tube side pool water. 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 attemate cooling water supplies. Loss of DHR cooling is bounded by the " Loss of Spent Fuel Pool Cooling" incidents described above. ] DSAR 3-73 Rev.16 c _ _______ _ - - _ - _ _ _ _ _ - _ - _ _ _ _ _ _ _ _ _
MYAPC An abnormal operating procedure desenbes the steps to initiate attemate cooling from a fire system supply hose to the E-25 inlet connection and a fire hose from the E-25 outlet connection to the fire pond. The fire system has ample capacity to provide attemate cooling to the spent fuel pool heat exchanger. The fire system also has a diesel driven fire pump. The effects of this incident bound a loss of a service water pump and the required response actions are identical. Evaluation and Conclusion The primary function of the spent fuel pool cooling, makeup and purification system is to ensure that the fuel remains covered and that suitable shielding exists to maintain radiological exposures below 10CFR 20 limits. This is primarily a passive design function and requires no active system intervention by the system components to prevent or mitigate the consequences of a design basis accident. In the current plant condition, the radiological source terms have decayed to a level where the postulated events would not result in offsite dose limits exceeding a smail fraction of the 10CFR100.11 limits. The reduced heat load, and the assurance of pool integrity provide the margin required to credit 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 uncovering the fuel, and maintain adequate shielding above the active fuel. O Although boiling is permitted as a service condition, long term boiling as a service condition is not practical due to the continuous makeup capacity required and the demineralized temperature limits. Accordingly, normal pool water temperatures are maintained below 110* F. The pool water is cooled using the spent fuel pool cooling pumps. Additionally, the purification pump may be used to cool the ] pool water if conditions permit (i.e., depending on the decay heat load, '.he pool temperature, DHR ] flow rate and the DHR inlet temperature). ] The spent fuel pool purification system is provided to maintain water chemistry within specifications and is not required to prevent or mitigate the consequences of an accident. For out-of-specification chemistry conditions, corrective actions are taken based upon We results of sample analyses or other i 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 primarily I due to evaporation losses. Sufficient available make-up water capacity exists through diverse l l Rev.16 s DSAR 3-74 l l t_ _ . _ _ _ . - - - . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ . . _ .
MYAPC sources to allow operators to take manual actions to compensate for liner leakage or natural and incident evaporation rates due to boiling. Makeup capacity includes: the Primary Water Storage Tank (and associated pump P-SFP2 or primary water pump P-24A), town of Wiscasset water supply ] system, and the water storage pond (and fire pump). The active components of the spent fuel cooling, makeup and purification 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 safc shutdown condition, or (3) the capability to prevent or mitigate the consequences of an accident resulting in potential offsite exposures comparable 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 active 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 following 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 mrem / hour (whole body) limit normally established for j 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 mrem / hour. In all cases, a suitable work environment is available, and adequate time is available to provide offsite or onsite make-up capability without relying on safety related equipment. Fuel handling equipment is discussed in secten 3.2.2. The re.ults of the systems, functions, and
. actions credited relative to fuel handling are included below. l The below listed SSCs are designed to Seismic Class I requirements.
. A functional anti-siphon device on the suction side of the pool to prevent draindown, l
. Fuel platform and hoist and yard crane
. Fuel pool concrete structure, liner, racks and transfer tube l
. Flow limiters on the liner leakage detection system i l
The following system functions were credited in the evaluation.
. Capability of the pool liner and concrete structure to sustain long term boiling,
. Capability of the transfer tube to maintain its integrity, including associated valves and flanges.
j Fuel platform and hoist and yard crane design safety features to assure the safe handling of fuel ! and casks. Rev.16 O DSAR 3-75
MYAPC
. 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 flow rate shall be equal to or greater than the boil-off rate.
. The capability of the racks to maintain adequate spacing between the assemblies. This includes the placement of fuel in the racks in the proper region. Refer to Figures 3.3.5 and 3.3-6 for the proper placement of fuel 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 ability to provide make-up to the fuel pool within 62 hours from the initiation of an loss of pool water siphoning incident.
. Assuring tha; boron concentration was adequate prior to the movement of fuel.
. Assuring that appropriate tests and inspections were conducted on fuel handling equipment prior to the movement of heavy loads and that such movements are administratively controlled to prevent movements in the vicinity of safety related equipment.
. Appropriate chemistry is maintained in the pool
. A Boral surveillance program is implemented
. Administrative controls to prevent shipping casks from being lifted over the spent fuel pool 3.3.1.4 - Fuel Storage - System Operation Normal Operating Specifications:
Fuel pool water level Between 43 ft. and 45 ft. elevations Fuel pool water level during fuel handling operations 44.2 ft elevation ] Fuel pool temperature Between 68'F and 120*F ] Fuel pool cooling pumps s1800 gpm ] DHR Pump flow rate to E-25 s1100 gpm ] FHB Normal Radiation Levels (passage ways) s2.5 mrem /hr (whole body) l Radiation Levels during Fuel Handling operations s50 mrem /hr (whole body) Water make-up Demineralized Water i t System Operation Description The fuel pool cooling system is shown on Figure 3.3-7. This system removes the decay heat from spent fuel stored in the fuel pool by circulating the pool water through a heat exchanger. The fuel O DSAR 3-76 Rev.16
0 MYAPC 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 water 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 be independently throttled to obtain the desired flow. Tne 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 purification loop consisting of a pump (P-85), a pre-filter (FL-2), and a post filter (FL-29) which may be operated 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 cooiing pump suction line during periods when cooling is not required or from the discharge of the heat exchanger when the cooling system is operating. Flow from the discharge of the fuel pool purification pumps is directed through the fuel pool refilter, and/or postfilter, while the retum iis through the fuel pool cooling retum line. The ] purification pump can also be used for skimming operation or to circulate 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 design DHR flow rate is adequate to assure cooling of pool water. A discussion of ] the DHR water system is provided in section 3.3.3. ] Spent fuel makeup capacity is provided through diverse sources. Makeup capacity includes: the Primary Water Storage Tank and the town of Wiscasset municipal water supply. Water is also ] available from the fire pond. None of the makeup sources are safety related. Two fire pumps are l l located in the fire pump house located near the water storage pond. A diesel engine drives one l pump while the other is motor-driven. The building houses the diesel fuel tank, t'he batteries and control board required for the diesel operation. Therefore, under severe conditions, fuel pool i Rev.16 O ! DSAR 3-77
,,,,m,,,a
MYAPC inventory may be restored using the 2500 gpm fire pumps and the water storage pond. l During skimmer operations, the spent fuel pool purification pump takes suction from four surface skimmers to keep the pool surface free of foreign matter. The pu#wini pump discharge is directed ack to the fuel pool through the underwater retum line during this operation. , The Fuel Building Area Radiation Monitor is discussed in sechon 4 including the response to radiation l alarm conditions. The Fuel Building Ventilation System is described in section 3.3.5. Operating procedures specify the methods and actions required to provide fuel pool makeup, cooling and puMwiicn under normal operating conditions. Operating procedures also specify the methods and actions required to restore cooling temperatures and water level in the spent fuel pool under abnormal operating conditions. l 3.3.1.5 Monitoring and Instrumentation Instrumentation IO There are two automatic actions performed by the spent fuel pool cooling and cleanup system ] instrumentation (automatic trip of the DHR pumps on low suchon pressure and cycling on/off the fan ] coolers) and they are both non-safety related. The fuel pool high or low level and the fuel pool ] temperature alarm conditions are annunciated in the control room. l Pressure gauges provided on the discharge of each pump in the spent fuel pool cooling system. This includes the differential pressure indicator located in the purification loop which may be valved-in to measure differential pressure o' each filtering component. The fuel pool heat exchanger temperature j indicator on the fuel pool cooling heat exchange outlet line is used to regulate the number of cooler ] coil fans operating. ] 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 aerated drain tank. The spent resin sump pump high level switch alarms on the PAB Panel. Upon receipt of a Spent Resin Pit high level alarm condition, an operator is dispatched to investigate , the condition. If the alarm condition is due to liner leakage, actions are taken to restore level, as appropnate in accordance with operating procedures. An alarm condition due to a high level in the Aerated Drain Tank would further prompt operations personnel to investigate the cause of the alarm O condition. Rev.16 DSAR 3-78
MYAPC Chemistrv Spent fuel pool water chemistry is monitored to minimize the potential effects of corrosion which could affect the safe storage of irradiated fuel. Chemistry surveillance actuties are performed 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 demineralized water and additives, as required.
The water chemistry is maintained and monitored in accordance with the following values and frequency: Parameter Values Nominal Frequency pH at 25'C 4.5 to 8.5 Weekly Chloride less than 100 ppb Weekly Total Halogens .25 ppm maximum Weekly Boron a 663 ppm design limit a 1720 ppm administrative limit Monthly Gamma Isotopic NA (for trending only) Monthly Suspended Solids NA (for trending only) Monthly r Although dose analyses for the various spent fuel storage and handling incidents are conservative, they do not consider the effect of an increase in oose due to particulate 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 i program includes sampling analysis and corrective action recommendations to protect the stored I
materials and minimize radioactive contamination. The underwater domineralizer and the slipstream purification pre and post filters, located in the Fuel ] Pool and the Fuel Building, respectively., are periodically monitored to assure that activity limits, and ] differential pressures, are within specifications. Ventilation l The ventilation systems consists of heating and ventilation and is designed to provide a suitable l environment for equipment and personnel. The ventilation system utilizes fans, filters, dampers, heating elements, and ductwork to accomplish the desired effects. Refer to section 3.3.2 for a description of requirements applicable during fuel movements. The ventilation system is NNS. l 1 O Rev.16 DSAR 3-79
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l: MYAPC I i 3.3.2 ' Fuel Handling System 3.3.2.1 Design Basis Desian Criteria L The fuel' hand ling design criteria is as follows:
- 1. Criticality in new and spent fuel storage is prevented by physical design features or processes.
lL A geometrically safe configuration is emphasized over procedural controls.
- 2. Appropriate shielding is provided to meet the requirements of 9CFR 20,
- 3. The fuel building is continuously monitored by area specific detectors. Audible and visual alarms are activated at the detector locations and the main control panel for radiation levels in excess of predetermined limits.
' 4. Spent fuel storage systems are:
- a. designed to prevent or mitigate accidents which could lead to the release of significant
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amounts of radioactivity affecting the public health and safety, l
- b. designed, fabricated and erected to withstand, the additional forces that might be imposed E by natural phenomena.
LO The design basis of the fuel handling system with regard to the spent fuel handling incident and the cask drcp incident are contained in sections 3.3.1.3 and 5.0. Fuel building ventilation system requirements and radiation monitoring are contained below and in section 4.0. Heavy Loads program requirements are addressed in section 3.4. l
; Design Basis I
i! The fuel handling design criteria is as follows: l a. Irradiated fuel shall not be consolidated until it has cooled for a period not less than 730 days after final discharge from the reactor. Prior to conducting irradiated fuel pin activities, a review l: is performed to assure that the licensing / design basis is adequately reflected in operating l procedures. This includes a review to assure that occupational doses as a result of a pin rupture are maintained ALARA consistent with the requisite design basis ventilation flow.
- b. A spent fuel shipping cask shall not be lifted over the spent fuel storage pool.
. c. . Fuel handling devices have provisions to avoid dropping or jamming of fuel assemblies during ]
transfer operations. ) Rev.14 )
)
C ; DSAR 3-88 l
MYAPC
- d. Fuel handling equipment is designed to preclude lifting of the assembly above a height which could cause undue radiation exposure to personnel (250 mrem /hr) on the platform directly over the pool.
- e. The spent fuel racks are designe d such that the dropping of a fuel assembly from a height of 3 18" above the top of the racks will not incur damage which could result in criticality.
- f. Fuel handling equipment is designed to withstand the loadings that would occur during an Operational Basis Earthquake and will 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 fuel building, transfer spent fuel assemblies in the fuel pool, and to *p spent fuel I assemblies off-site. Equipment is also provided to transfer control element assemblies to guide tubes in each fuel assembly, as required. A periscope may be used for underwater close inspection of fuel assemblies and control element assemblies. The refueling equipment arrangement is shown on Figure 3.3-8. The system is comprised of the following:
/" Spent Fuel Movable Platform and Hoist Yard Area Crane Communications Soent Fuel Movable Platform and Hoist The basic structure of the movable platform and hoist is a traveling bridge which spans the spent fuel pool and moves on rails over any spent fuel storage position, 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: i 1. An electrical cutout that stops movement in the north direction if CR-9 gets to slose to CR-6. CR-6 has a similar interlock to prevent i from driving into CR-9.
- 2. An interlock that prevents concurrent operation of the CR-9 hoist with either the bridge or the l trolley.
- 3. An interlock that limits upward travel of the hoist (two block interlock).
Rev.14 DSAR 3-89 l ! -- - _ - - - - - ---------------_-----------_a
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C/ Yard Area Grane The 125/20-ton yard area crane is a conventional bridge crane with a three-motor trolley and powered bridge. It was designed to the requirements of the Electric Overhead Crane Institute, EOCl Specification #61 for Class A service. The improved plow steel rope has a 5.5 to 1 safety factor and the remaining components generally have a 5 to 1 safety factor. The main hoist is powered by an alternating current 30 hp wound rotor motor and operates at 3 feet per minute. Braking is provided by an eddy current control brake and two spring set, electrically released holding brakes, each with a rating of 322% of fullload motor torque. Power to release the holding brakes is provided by the same circuit breaker that energizes the main hoist motor; thus, the holding brakes i are applied whenever the main hoist motor is de-energized. The eddy current brake is always in l service. A spring-operated overload switch de-energizes the main hoist motor and sets the b akes whenever a preset load is exceeded, preventing any over stressing cf the rope or other crane components. In addition, a centrifugal overspend switch de-energizes the motor and sets the brakes whenever a preset lowering speed is exceeded. This, with the eddy current brake, provides redundant protection against runaway lowering. Up travel of the main hoist is limited by a weight- , m operated limit switch and a gear-operated limit switch operating in series to prevent two-block, while a single gear-operated switch limits down travel so that two full turns 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 interlocks prevent the main hoist from passing over the stored spent fuel. 3.3.2.3 Design Evaluation Fuel handling equipment is designed Seismic 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 interlocks are fail-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: mechanical stops, brakes, limit switches, and electrical interlocks. These design controls, together with the administrative _ controls (e.g., operator training, periodic load tests, functional tests, prohibitions to lifting spent fuel shipping rasks over the pool) provide assurance that the consequences of accidents will remain r3 Rev.14 O ( DSAR 3-90 l
l l MYAPC ; I well below the 10CFR 100.11 and EPA PAG limits. In addition, administrative controls restrict handling of heavy loads in the vicinity of safety related equipment. The fuel handling SSCs designed to Seismic Class I requirements are the fuel platform and hoist and yard crane. The following fuel handling system functions were credited in section 3.3.1 ; The fuel platform and hoist and yard crane design safety features assure the safe handling of fuel and casks Administrrave controls to prevent shipping casks from being lifted over the spent fuel pool Design controls assure that the spent fuel assembly is not lifted greater than 18 inches above ! the racks. Design controls 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 j prior to the movement of heavy loads and that such movements are administratively controlled l( to prevent movements in the vicinity of safety related equipment.
. Assuring the placement of fuel in the proper rack region. Refer to figures 3.3-5 and 3.3-6 for the proper placement of fuelin the racks as a function of burnup.
3.3.2.4 System Operations
- The following operations controls are implemented for the spent fuel movable platform and hoist
- a. Prior to conducting fuel handling, a complete checkout, including a load test using a dummy l
assembly, shall 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), a refueling system interlocks functional test shall be conducted.
- c. The spent fuel hoist is controlled to assure that the assembly is not lifted more than 18 inches over the top of the racks.
- d. Administrative controls are established to prevent movements in the vicinity of safety related equipment.
Rev.14 DSAR 3-91 ; l i 4
I l MYAPC i
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The following controls are implemented regarding the yard crane
- a. All phases of fuel cask movement are under strict administrative control, with all movement done in accordance with written instruction and check-off lists. The operation is under the direct supervision of a responsible member of the Maine Yankee staff, and crane operators are trained to meet the requirements of USAS B30.2-1967, Overhead and Gantry Cranes. ,
- b. Documented maintenance and performance checks, including a loaded operational test of the crane controls and brakes, are performed prior to each fuel cask handling evolution. ;
The following requirements were relocated from the technical specifications:
- 1. The following conditions shall be satisfied during movement of irradiated fuel within the spent fuel storage building:
- a. Radiation levels in the spent fuel storage area shall be monitored continuously.
- 2. When irradiated fuel which has decayed less than 60 days is in the cpent fi ni pool, the spent fuel pool ventilation system shall be operating and discharging through o . idPA and charcoal absorber filter train during either:
- a. Fuel movement within the spent fuel pool, or
- b. Crane operation with loads over the spent fuel pool.
Remedial Actions: With the spent fuel pool ventilation system inoperhie, suspend all operations involving movement of fuel within the spent fuel pool or crane operation with loads over the spent fuel pool.
- 3. Spent fuel shipping casks shall not be lifted over the spent fuel pool storage pool.
- 4. Spent fuel storage racks may be moved only in accordance with written procedures which ensure that no rack modules are moved over fuel assemblies.
3.3.2.5 Inspection and Testing , The yard area crane receives annual maintenance examinations by qualified personnel. l Rev.14 rx DSAR 3-92
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MYAPC d 3.3.3 Spent Fuel Pool (SFP) Decay Heat Removal (DHR) System ] 3.3.3.1 Design Basis ]
]
Desion Criteria ] The Spent Fuel Pool (SFP) Decay Heat Removal (DHR) system is designed to be a reliable source ] of heat transfer from the SFP water. It removes decay heat from pool water to maintain temperature ] within specification. The system is designed to preclude the possibility of radioactive leakage from ] reaching beyond the plant boundary. ]
]
Desion Basis ]
]
- a. The secondary cooling Loop (DHR) system is an air-cooled, closed-loop cooling system. ] j
- b. The system piping and components are classified as Non-Nuclear Safety /important To the ]
Defueled Condition (NNS/ITDC). The system is constructed and maintained in accordance ] with appropriate industry codes and standards. ] cs, c. The system is protected during periods of prolonged freezing '.emperature conditions. ] (,) d. Alternate sources of cooling can be effected in the event that the DHR system is out of ] service. Evaporation and pool water makeup are sources of a temate cooling. ]
- e. The system is capable of maintaining bulk temperatures s120'F with an outside temperature ] ;
of 87'F with a duty of 3.33 million BTUs per hour. ]
- f. The system is able to recover from a fuel pool boiling incident where the water in the primary ]
and secondary system may reach 212*F. ]
- g. Pool water bulk temperature is maintained above 68'F. ]
3.3.3.2 System Description The SFP DHR system flow diagram is shown in Figure 3.3-9. SFP DHR provides cooling water ] 4 service for spent fuel pool decay heat removal. SFP DHR is primarily filled with delonized water, ] with additives to protect against freezing and corrosion. Heat generated by the spent fuel is ] transferred 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 circulated by one of two ] DHR cooling pumps. An expansion tank and air separator are also provided. ] Local hardwired controls and indications are the primary means of controlling the Spent Fuel Pool ] l [ ' island (SFPI) equipment. A highly reliable non-nuclear safety related computer system, referred to Rev.16
]
) DSAR 3-94 u___ _ _
MYAPC 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 ] DHR loop control functions are provided by the PLC as discussed in 3.3.3.5. ]
]
The Primary Water Storage Tank provides a makeup source to the DHR loop via a manual hose ] connection. In the event that makeup water is required, deionized water in the PWST is pumped ] via P-SFP2 located in the fuel building to a hose connection in the secondary loop. ]
]
SFP DHR Pumos ] Two parallel SFP DHR pumps are provided and each deliver a nominal flow of 1000 gallons per ] minute. One pump is usually in service. The other is an installed spare. The pump internal wetted ] surfaces are stainless steel. ] Water-to-Air Coolers ] The DHR water-to-air cooler is composed of six parallel, finned-coiled type coolers; each cooler ] p containing three fans. The SFP DHR coolers may be bypassed to maintain system flow balance ] V and desired heat transfer rate. The heat transfer rate is controlled through the cycling of the cooler ] fans. ]
]
Deionized water is used in the DHR loop during the first and possibly future summets to maxim;ze ] 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 approximately 80 psig and 85'F. ] The tubes and fins are constructed of stainless steel and aluminum, respectively. ] DHR Exoansion Tank ] The DHR bladder expansion tank is sized to accommodate fluid expansion and contractions ] resulting from temperature variations in the secondary cooling loop. The tank capacity is 211 ] gallons. All wetted portions of the tank are constructed of stainless steel. The bladder is ] constructed of butyl rubber. ] Air Seoarator ] The air separator removes entrained air in the secondary coolant. The unit and trap are made of ] l stainless steel. ] Rev.16 ] ) l DSAR 3-95 l w___________
MYAPC Picino System ] I The piping and valves used throughout the system are constructed of stainless steel with the ] exceptions of a small 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 corrosion inhibited deionized water. A connection is provided on the ] DHR loop to add corrosion inhibitors as required. ] Codes 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 ] Coolers Commercial Industrial Standards ] Air Separator ASME, Section Vlli ) DHR Pumps Commercial Industrial Standards ] !,( J' Piping, Valves and Fittings ANSI, B31.1-1980 ] v 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 components 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. ] 3 The secondary cooling loop flow rate is controlled by the DHR cooling pumps. The typical operating ] flow rate is maintained below 1100 gpm; the manufacturer's maximum recommended shell side flow ] rate for E 25. Flow through the heat exchanger is controlled by throttling the heat exchange ] discharge valve. For pump protection, and as an indication of reduced system inventory, theb 6 ] a suction pressure switch on each pump which trips the pump on low suction pressure. ] ! 1 i Six water-to-air coolers in parallel provide the heat sink necessary to maximize heat transfer from ] the fuel pool. Each cooler may be bypassed to maintain system flow balance and desired heat ] O Rev.16 b t ! DSAR 3-96 l l
l i MYAPC transfer rate or facilitate cooler maintenance activities.' Over pressure protection of the secondary ] l loop is also provided. ] f 1 !. Calculations were performed to determine the capacity of the DHR system. The assumptions used ] in the analysis are as follows: ] Fuel Pool Design Decay Heat Load (6/27/98) 3.33 E+6 BTU /hr ] DHR (secondary loop) ] Flow rate 1000 gpm ] l Shell Side inlet Temperature 104*F ] l Discharge 111*F ] Outside AirTemperature 87*F ] Fuel Pool Cooling Pumps Flow Rate 1700 gpm (1800 max) ] ! Tube Side Inlet Temperature of 114*F ] The results show that the outlet temperature of the tube side of the heat exchanger would be 110*F. ] i Abnormal Con'ditions Each of the following abnormal conditions are evaluated for its effects on the ability of the SFP DHR ] system to provide heat removal from the spent fuel pool heat exchanger and to preclude the possibility of radioactive leakage from reaching beyond the plant boundary. Increasina Svstem Temperature - Changes in SFP DHR temperature may result from seasonal ]- i changes in the air temperature and humidity. Secondary coolant in the tube side is air-cooled by ] the cooling fans which force ambient 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 Loss of SFP DHR - Loss of SFP DHR cooling would cause the pool temperature to increase. ] Various make-up sources are capable of providing water in the event of maximum evaporation (i.e., ] boil-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. ]
l Rev.16 DSAR 3-97
MYAPC n C In the event that normal cooling is intenupted, alignment of altemate cooling supplies to the spent ] fuel pool heat exchanger may be provided by an electric or diesel driven fire water pump through ] flanged connections provided on the shell side of the heat exchanger. 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 ] supply. ]
]
Loss of SFP DHR coo!ing to the spent fuel pool heat exchanger is bounded by the " Loss of Spent ] Fuel Pool Cooling" incident described in Sections 3.3.1 and 5.0. An abnormal operating procedure ] describes the steps to supply altemate cooling from the fire system through a connection to the fire ] pond. The fire system has ample capacity to provide altemate cooling to the spent fuel pool heat ] ; exchanger. Since the fire system contains a diesel driven fire pump, and ample time is available ] to effect remedial measures, no onsite or offsite emergency electrical power is required to provide ] alternate spent fuel pool cooling. ] Leakage into/Out of the System - Spent fuel pool water is periodically monitored for radiation levels. In addition, the purification loop assures that radiation levels in the pool are minimized to preclude l the existence of significantly contaminated water. Leakage from the poolinto the SFP DHR system ] (q is unlikely since the SFP DHR system is maintained at a hiaher pressure then the fuel pool cooling ] , system. If leakage were to occur through the heat exchcoger, the leakage would be diluted by the ] volume of SFP DHR and resultin an insignificant amount of SFP DHR System contamination. In ] j order to identify potential leakage, the SFP DHR water is periodically sampled for radioactive ] l , contaminants. Leakage into the SFP DHR system may be indicated by an increasing surge tank ] level, low pressure drop across the heat exchanger, or by loss of pool water level. ] Freeze protection and corrosion controlis effected in the secondary loop by the use of ethylene ] glycol and dipotassium phosphate, respectively. ] 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 Section 3.3.1, there is sufficient time to restore power. ] t 3.3.3.4 System Operation j 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 ] A b Rev.16 DSAR 3-98
MYAPC
, ~y Q 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 from
]
]
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 ] other is an installed spare. ] j l ! 1 ] 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 cooiers are located in a diked / roofed area. A gabled roof ] over the diked cooler reduces the collection of precipitation inside of the dike. The diked cooler area ] ) I is released to stcrm drains after it is determined that the fluid is compliant with waste discharge ] I specifications. The operation of the cooler fans are staggered to adjust for heat rejection rate ] ; variations due to variation in ambient air temperature. Flow is provided to all the coolers whether ] l 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 adiant and natural convection heat transfer may ]
)
remove sufficient heat from the secondary system. ] i l
]
The DHR pumps are provided with a local "stop-run" control switch. One pump is likely to run ] l continuously year round; however, the pump may be secured to prevent overcooling, as required. ] u A*y 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 prevent overcooling. ] 3.3.3.5 Monitoring and Instrumentation A DHR pump suction pressure switch is provide for each pump. The switches provide a low ] pressure alarm in the control room and initiate a DHR pump trip via the PLC in the event that a low ] suction pressure condition exists. In addition, pump discharge pressure transmitters provide ] continuous indication in the SFPI control room. ]
]
The water-to-air cooler fans are controlled from a temperature indicator sensor on the fuel pool heat ] exchanger outlet temperature. The temperature sensor and transmitter information is fed to the ] PLC. The PLC automatically cycles the cooling fans to attain the desired cooling rate. ]
]
In the esent 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. ] l ] A primary to secondary differential pressure switch is provided for localindication and low differential ] pressure alarm in the SFPI control room. ] g Rev.16 V DSAR 3-99 l
MYAPC The system is monitored for proper concentrations of a corrosion inhibitor (dipotassium phosphate) ] and the degree of freeze protection capability. ] 3 in order to identify potential leakage, the SFP DHR water is penodically sampled for radioactive ] l containments. ]
]
A drain system is provided and sampled for contaminants prior to release. ] I L . U l l l t' [ .. I Rev.16 I' DSAR 3-100 L I
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FUEL POOL COOLING PIPING rr c - 9808070179-I Maine Yankee
. _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ __ f
MYAPC I\ J 3.3.4 Service Water 3.3.4.1 Design Basis The service water system is designed to be a reliable source of cooling for the component cooling system. System piping and components are designed in accordance with applicable codes. The ] system provides adequate heat removal capacity for the pe manently shutdown and defueled condition. The system also provides for the proper dilution of radioactive effluent releases. The pumps, heat exchangers and interconnecting piping are designed to meet seismic criteria. 3.3.4.2 System Description { General Figure 3.3-12 shows the flow diagram of the service water system. The service water system is an open loop cooling system which provides the heat sink for heat from the plant auxiliary systems. It performs this function by pumping sea water through the component cooling heat exchangers. v
]
3 Two service water pumps are available to dilute effluent discharge, as necassary. The system is ] l also provided with an Amortap system to the heat exchanger, sampling equipment, and mussel control features. Service Water Pumos The service water pumps are vertical, single stage, turbine pumps each having a rated flow of ; 10,000 gpm at a total dynamic head of 66 feet. The motors are rated at 250 horsepower. The two service water pumps may be lined up to discharge to the north and/or the south header in various ! combinations. Secondarv Comoonent Coolino Heat Excharp.en ] Service water cooling flow from the supply cross-connect header through the heat exchanger tubes I t O, () Rev.16
- DSAR 3-102 l
MYAPC . l (O1 absorbs heat from the component cooling water flowing through the heat exchanger shells. Curved elbow mesh and bar screens are installed in the inlet channe8 heads of both cort ponent cooling heat exchangers to prevent mussel shells or other debris from becoming lodged in the heat exchanger tubes. The inlet screens in E-5A are constructed of titanium, and those in E-5B are constructed of ] j monel. The titanium and monel screens are sit.ilar, although not identical in design. The curved screen design, provides the ability to flush the debris from the screen without disassembly of the l heat exchanger inlet. Service Water Pioino The discharge of the service water pumps is routed to the component cooling heat exchangers ; through either the north or the south header. Check valves are installed in the supply headers in the turbine building to prevent backflowin the event of pump or piping failures. Cross-connect lines are provided between the supply headers at the discharge of the pumps and just upstream of the heat exchangers. These cross-connect headers and their associated isolation valves provide flexibility to maximize system reliability.. Service water flow leaving the heat exchangers enters a common discharge header and is retumed to the Back River via the circulating water system seal l pit and forebay. A major portion of the interconnecting cast iron pipe has been lined with a resin ; O U material under a process referred to as cured-in-place-pipe (CIPP). ! Codes and Standards Piping Flanged ANSI- A21.2 Mechanical Joint ANSI- A21.6 Valves Butterfly AWWA (Spec. C504-66) or ANSI B31.1, B16.5, and B16.34 Gate, Globe & Check ANSI- B31.1 3.0.4.3 Design Evaluation Each service water pump has a rated flow of 10,000 gpm at a total dynamic head of 66 feet. A ] single pump pmvides cooling for the control room air conditioning and the diesei generator coolers. ] The system provides no safety-related function. Rev 16 DSAR 3-103 4 1 i 1
4 MYAPC r The service water system consists or two service water pumps supplying the component cooling Q) water heat exchangers. The design of the system ensures the availability of service water following a single piping or expansion joint failure. The North and the South header may be isolated from each other and fed by various combinations of service water pumps to supply various combinations of component cooling heat exchangers. Check valves in the supply headers prevent backflow from one header to the other. Manually operated isolation valves in the service water pump header and the service water supply cross-connect header to the component cooling water heat exchangers provide flexibility in alignment combinations. Water level alarms and a dawatering pump are provided in the Service Water Pump Motor Bay to minimize the possibility of flooding the service water pump motors. Abnormal Conditions Las of Site Power - If power is lost to Bus 7 and 8, the ser/ ice water pumps will nct operate. To ] reatore power to the service water pumps, a diesel generator may be started. ] 3.3.4.4 System Operation Normal plant cooling requirements can be met by operating one service water pump supplying [] V cooling to one component cooling water heat exchanger. The service water pumps may be operated from the main control board or the protected switchgear room. Manually operated isolation valves in the service water pump header and the seMee water supply cross-connect header to the component cooling water heat exchangers may be operated to provide various alignment combinations of pumps and heat exchangers. 3.3.4.5 Monitoring and instrumentation Service Water pump status is indicated by lights and by motor smmeters on the main control board. Pressure transmitters located in the service water pump discharge headers are indicated and alarmed on the main control board. Level switches have been provided which annunciate a hyh water level on the floor of the pump house at three inches. l The system is monitored for pH, sediment content, oil c'ntent and for radiation levels by obtaining l samples on a periodic basis. Rev.16 l p L) DSAR 3-104 l L____________.____________-__
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,q 3.3.5 Ventilation Systems V
Ventilation systems are designed to provide a suitable environment for squipment and personnel. These systems provide heating and air conditioning. The ventilation systems utilize fans, filters, dampers, heating elements, cooling elements, and duct work to accomplish the desired effects. In the radiologically controlled areas, the fuel building and the primary auxiliary building, outdoor air is supplied to these structures. Air is exhausted in greater quantities than it is supplied to maintain the building at a negative pressure and ensure that the general flow of air is into the structure. The exhausted air is discharged past radiation monitors in the primary vent stack or Fuel Building ] exhaust duct. ] Relocated Technical Specification Definition of Ventilation Exhaust Treatment System The Ventilation Exhaust Treatment System includes all systems designed and installed to reduce radioactive material in particulate form in effluents by passing ventilation through HEPA filters for ] the purpose of removing particulate from the exhaust stream prior to release to the environment. ] Such systems are not considered to have any effect on noble gas effluents. Engineered Safety ] Feature (ESF) atmospheric cleanup systems are not considered to be Ventilation Exhaust Treatment Systems components. O l y) Insoection and Testing ]
]
Fuel Building Ventilation System ] 3 Inside the fuel building is the fuel storage pool. The stored fuel will remain in the builrf.ng until it is ] transferred to dry-casks or ff-site. The fuel building will remain independent of other systems, ] structures, or components undergoing the decommissioning process. in this regard, the fuel ] building ventilation system will remain operational. The fuel building ventilation system has ] Administrative Controls established to ensure the following inspection and Testing requirements are ] satisfied: 3
)
. Requirements for an operable ventilation system ]
. Requirements for demonstrating operability ]
Requirements for testing and test frequency !
]
]
! 3 l 3 Rev.16 ) o V DSAR 3-106 I J
MYAPC (O) Control Room and Auxiliary Ventilation Systems ]
] ,
The remaining ventilation systems consist of the Centrol Room Ventilation and Auxiliary Ventilation ] Systems. These systems are in buildings where decommissioning activities will occur. Maine ] l Yankee has committed to the intent of NUREG/CR 0130. These ventilation systems will be subject ] to the testing and operability requirements of DSAR, Chapter 7, DECOMMISSIONING. ] 3.3.5.1 Fuel Building Ventilation System l 3.3.5.1.1 Design Basis The fuel building ventilation system is designed to: maintain the operability of the fuel building ] ! equipment during normal operating conditions, ensure that air flow is from outside into the building ] l to prevent unmonitored release of radiation, and ensure that exhaust air is continuously monitored. 3.3.5.1.2 System Description , The previously existing fuel building ventilation system has been replaced with one designed to ] i (D 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 northwest wall, and ] discharges the air through a duct mounted on the exterior of the fuel building east wall. Before it is ] discharged to the outside, the exhaust air passes through HEPA filters to remove any particulate ] materials. Because the er.naust system is once through, the fuel building is always maintained at ] a slight negative pressure. The heating and ventilation equipment in the fuel building is designed to ] minimize moisture condensation on the walls and the roof and to limit the space temperature to a ] maximum 95'F and a minimum of 60*F. A radiation monitoring 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. ] 3.3.5.1.3 Design Evaluation ] Since the consequences of the Fuel Handling Accident are significantly below 10CFR"0.11 limits ] without taking credit for the building ventilation, the fuel building ventilation system is not designated ] to mitigate the consequences of a Fuel Handling Accident. Therefore, the fuel building ventilation ] system is not safety related. Ih I V Rev.16 I DSAR 3-107 I
l MYAPC O . V 3.3.5.2 Control Room Ventilation J 3.3.5.2.1 Design Basis The control room ventilation system is designed to ensore habitability for personnel and proper ; cooling for equipment under normal and accident conditions. I 3.3.5.2.2 System Description The control room air conditioning and ventilation system consists of air conditioning unit AC-1B, ] exhaust fan FN-15, path way to provide fresh air make-up and associated ductwork. The AC-1 A air ] conditioning unit is no longer available to provide conditioned air, however, the fan which is integral ] to AC-1 A may be aligned to supply fresh air to the control room if the AC-1B unit fails. No credit for ] control room isolation is taken in the safety analyses and the doses to operators are still well below ] General Design Criterion (GDC) 19 limits. Therefore, the control room ventilation system no longer ] has an isolated operating mode. Not having an isolated operating mode eliminates the need for ] recirculation filters, filter booster fans FN-11A and FN-118, the breathing air inlet filters and the ] breathing air fans FN-7A and 78. Therefore, these components have been abandoned and are no ] ! longer part of the control room ventilation system. ] The temperature in the control room may vary from 65'F to 85'F. Ind:cating and control instruments will continue to function within design accuracy in ambient temperatures up to 110*F. The control room ventilation system is designed to preclude a total loss of cooling capability. The system consists of a 100% capacity air conditioning unit. If the air conditioner fails, then with control room ] doors opened, the system will be cperating on straight ventilation to maintain Control Room ] temperature less than 110*F. ] Normal Mode of Ooeration ' An air conditioning unit supplies filtered, temperature controlled air to the control room. If AC-1B is ] i not available, the fan which is integral to AC-1 A may be aligned to supply fresh air make up to the ] Control Room. ] Isolated Mode of Ooeration in the permanently defueled condition, the rduced source term (Section 5.0) precludes the need for ; f lp automatic isolation of the control room ventilation system. V Rev.16 l DSAR 3-108 l
'MYAPC (3
V 3.3.5.2.3 Design Evaluation The control room ventilation system was designed to handle a loss of coolant accident. Therefore, it is sufficiently designed to accommodate the significantly reduced source term in the permanently defueled conditbn. The consequences of the accidents in the defueled condition are significantly below 10CFR100.11 limits. In addition, no credit for control room ventilation isolation is taken in the safety analyses since doses to operators are maintained below GDC 19 limits without isolation. In ] the event of evacLation, controls related to spent fuel cooling, makeup and purification are located in areas near the equipment. No ventilation systems are credited to prevent doses to control room personnel or the public for the ] credible spent fuel handling and storage incidents. Analyses indicate that 3 mrem (whole body) dose ] is expected if the control room ventilation system is left in the existing mode of operation. "No action" ] in the event of this incident would result in a maximum 2-hour dose of approximately 3.3 mrem ] assuming an air intake rate between 120 and 900 cfm. Additionally, strict radiological controls are ] effected to minimize exposures to personnel ALARA. ] 3.3.5.3 Auxiliary Ventilation Systems () This section discusses the ventilation systems at the plant. The ventilation systems include: turbine building, primary auxiliary building, and the service building ventilation systems. These ventilation systems do not perform any safety functions in the permanently defueled condition. 3.3.5.3.1 Turbine Building Ventilation System Three separate ventilation systems supply a total of 200,000 cfm to the turbine room. Each fan delivers approximately 67,700 cfm of cutside air or a mixture of outside cir and recirculated air, regulated by dampers. An additimal 205,000 cfm can be induced into the turbine room by operating gravity intake louvers located at ground floor elevations. The exhaust of the turbine room is through gravity ventilators located on the roof. 3.3.5.3 Primary Auxiliary Building Ventilation System The PAB ventilation is supplied by two separate units. HV-1 supplies 31,350 cfm of outside air to the areas at elevations 21 feet and 11 feet. This unit has provisions for heating the air when required. HV-2 supplies 32.550 dfrn of outside air in summer and has provisions for recirculating air f O N- Rev.16 l DSAR 3-109 i
I, MYAPC l l U ir winter. This unit can heat the full 32,350 ofm of air in winter. HV-2 supplies air to elevation 36 fret. l The PAB exhaust is separated into two categories: PAB shielded and PAB miscellaneous. PAB shielded exhaust is collected from elevations 36 feet and 21 feet. The shielded exhaust is filtered by charcoal and HEPA filters in the PAB tray filter assembly, and discharged by fan FN-1 A or FN-1B to the primary vent stack. Each fan has a capacity of 27,000 dfm. PAB miscellaneous exhaust is collected from elevations 36 feet,21 feet and 11 feet. The miscellaneous exhaust is filtered through a deep-bed charcoal and HEPA filter assembly to FN 1 A and discharged to the primary vent stack. In this configuration, the PAB shielded exhaust must be aligned to fan FN-1B The heating and ventilation equipment in the PAB operates automatically to maintain space temperatures between 60*F and 95'F. Space temperatures are limited to 104*F in normally l unattended areas which house electrical machinery and heat producing equipment. 3.3.5.3.3 Service Building Ventilation System 7 Ventilation to the Service Building is supplied by two units HV-5 and HV-8. HV-5 has a capacity of ] J 13,000 cfm while HV-8 hcs a capacity of 5000 cfm. The HV units are equipped with filter, preheaters, ] reheaters, and dampers to supply filtered and heated air to the various spaces of the Service ] Building. Where additional heat is required, electric and/or steam space heaters are provided. ] Numerous exhaust fans are installed to ensure all spaces of the building are ventilated. Exhaust fans ] are inctalled at such capacity to ensure that the Service buildinp is maintained at a slightly negative ] pressure. All Service building exhaust is continuously monitec d for radiation by an Air Particulate ] Detector (APD). ]
]
Part of the Service Building houses the Chemistry laboratory and is supplied by 5800 cfm of ] conditioned outside air. This supply system has provisions for heating and cooling. Auxiliary steam ] supplies the heating element while cooling is supplied by AC-2. 6400 cfm of air is exhausted through ] laboratory hoods by two fans each designed to accomodate full exhaust flow. ]
]
Administrative controls are established that will allow for the elimination of the ventilation ] requirements for the Service building. When the various Service Building spaces are no longer ] required, the supply / exhaust uquipment will be removed from service. After all the ventilation ] equipment has been secured then the Service Building APD will be removed from service. ]
]
[ 7m Rev.16 l DSAR 3-11J i
MYAPC /G V 3.3.6 Auxiliary Systems ' I This section discusses auxiliary systems supporting plant operating equipment. These systems include: compressed air, boric acid makeup, primary water, primary vent and drain, rad waste processing, primary component cooling, fire protection, and meteorological monitoring. These j systems do not perform any safety related functions in the permanently defueled condition. 3.3.6.1 Compressed Air The compressed air system supplies oil-free air for the instrumentation, control and service air requirements throughout the plant. Three 322 scfm full capacity instrument air compressors are provided. Each unit is capable of supplying air at 100 psig at a flow rate sufficient to meet plant l demands. Two compressors may run to meet heavy load conditions. Air from the three compressors
!s directed to two after coolers / moisture separators and two air receivers. The service air system carries air to the turbine building, primary auxiliary building, service building, and the fuel building.
The instrument air header is piped to prefilters, dryers, and postfilters, then distributed throughout the plant for instrumentation and control purposes. / Each of the three control and service air compressors is a skid-mounted self-contained unit, each d with its own starter controls. Under normal operating conditions, the three compressors are operated from a common system controller. The system controller monitors system pressure from a single l transducer located on the air sensing line connected between the air receivers and will load and unload compressors to meet plant air demand needs. One compressor should provide sufficient air to meet air demands. Each compress or is equipped with instrumentation and controls to monitor key operating parameters. Compressorand system controller malfunctions are indicated locally and via alarm in the Main Control Roorn. The compressed air system is shown on Figures 3.3-13 through 3.3-18. Equipment description and code requirements are provided in Tabte 3.3.1. t r U Rev.16 DSAR 3-111
l MYAPC TABLE 3.3.1 l l COMPRESSED AIR SYSTEM l l EQUIPMENT DESCRIPTION AND RATINGS Air Compressors (3) l Type Two-stage, oil-free, air-cooled, rotary lobe Design capacity (each), scfm 322 , Design discharge pressure, psig 100 l Motor 100 hp,3600 rpm,3-phase,60 cycle,460 V l After coolers (3 - 1 per compressor) l Type Shell and Tube Separator Cyclone type i 1 Air Receivers (2) Type Vertical Design pressure, psig 125 Capacity, cu ft (each) 50 Code ASME Vill lD instrument Air Dryers (2) Alumina desiccant, nonheat reactivated '\ Type Capacity (each), scfm 300 Outlet moisture content with saturated -40*F dew point at 100 peig l air inlet at 35 - 100*F Code ASME Vill Instrument Air Dryer Filters Air Dryer Prefilters (2) Capacity at 100 psig, scfm 300 Particle size filtered 10 microns and larger Air Dryer Postfilters (2) l Capacity at 100 psig, scfm 300 Particle size filtered 3 microns and largt - Piping and Valves Upstrearn of dryer Carbon steel piping & steel & bronze valves Downstream of dryer Copper tube and bronze valves Code q ANSI B31.1 (" Rev.14 DSAR 3-112
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