ML20247D345
| ML20247D345 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 05/06/1998 |
| From: | Maine Yankee |
| To: | |
| References | |
| NUDOCS 9805150025 | |
| Download: ML20247D345 (26) | |
Text
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- Proc. No. 0-06-1 28 THRU 38 - NRC HEADQUARTERS Rev No. 4 l DOCUMENT CONTROL DESK Page 5 of 6 l-t
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ATTACHMENT A L MAINE YANKEE CONTROLLED DOCUMENT TRANSMITTAL FORM i
i DOCUMENT: DEFuri pri SAFETY ANALYSIS REPORT - REVISION 15 TRANSMITTAL ISSUE DATE: 05-06-98 TRANSMITTAL RETURN DATE*: 06-04-98
- 1. Please remove the List of Effective Pages of the DSAR Manual, Rev. No.14, and replace L with the attached List of Effective Pages (5), Rev. No.15.
- 2. Please remove the following pages with Rev. No.14 and replace with the attached pages with Rev. No.15.
1-9 5-1 1-23 5-9 3-13 5-11 l 3-14 5-16 3-15 5-18 l- 3-29 5-37 L 3-65 5-38 3 6-10 3-136 6-14 (new page) l f
3-137 l i 3-138 (Fig. 3.3-23) i l
l The above listed document has been inserted into the assigned manual / file and all superseded pages l l have been destroyed.
- MANUAL / FILE UPDATED BY
l [ Please Print Name i i
l-DATE:
i Signature l
l CAUI1ON
- 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 Control Center .
P.O. Box 408 Wiscasset, Maine 04578
) 10 9905150025 990506
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l MYAPC LIST OF EFFECTIVE PAGES PAG'E REh REMARKS PAGE REV REMARKS PAGE REV REMARKS l' TABLESLGONTENT1 SECTION 1.0 SECTION 2.0 14 14 Table of Contents 14 Table of Contents 1
il 14 1-1 1-il 14 Ust cf Tables 2-1 2-il 14 Table of Contents
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i lii 14 1-iii 14 Ust of Figures 2-iii 14 Ust of Tables l
[ iv 14 11 14 2-iv 14 Ust of Figures v 14 12 14 2-1 14 l
l vi .14 13 14 22 14 vil 14 14 14 2-3 14 ;
vili 14 1-5 14 24 14 i lx 14 1-6 14 2-5 14 Table 2.1.1 l 17 14 2-6 14 Table 2.1.2 1-8 14 2-7 14 Figure 2.1 1 1-9 15 2-8 14 Figure 2.12 1-10 14 Table 1.3.1 2-9 14 Figure 2.1-3 1-11 14 Table 1.3.1 2-10 14 Figure 2.1-4 1-12 14 Table 1.3.1 2-11 14 Figure 2.1-5 1 13 14 Table 1.3.1 2 12 14 Figure 2.1-6 1-14 14 Tablo 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
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1-20 1-21 14 14 Figure 1.3-3 Figure 1.3-4 2-19 2-20 14 14 1-22 14 2-21 14 1 23 1$ 2 22 14 2-23 14 2-24 14 l 2-25 14 2-25 14 Table 2.2.1 2-27 14 Table 2.2.2 2-28 14 Table 2.2.3 2-29 14 Table 2.2.4 2 30 14 Table 2.2.5 1 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 j 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 l 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
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MYAPC l LIST OF EFFECTIVE PAGES I PAGh! REV REMARKS PAGE REV REMARKS PAGE REV REMARKS 2-47 14 SECTION 3.0 3-45 14 Figure 3.2-7 2-48 14 3-46 14 Figure 3.2-8 2-49 14 3-i 14 Table of Contents 3-47 14 Figure 3.2-9 2-50 14 3-il 14 Table of Contents 3-48 14 Figure 3.2-10 2-51 14 3-iii 14 Table of Contents 3-49 14 Figure 3.2-11 2-52 14 3-iv 14 Ust of Tables 3-50 14 Figure 3.2-12 2-53 14 Table 2.3.1 3-v 14 Ust of Figures 3-51 14 Figure 3.213 2-54 14 Table 2.3.2 3-vi 14 Ust of Figures 3 52 14 2-55 14 - Figure 2.3-1 3-1 14 3-53 14 2-56 14 Figure 2.3-2 3-2 14 3-54 14 2-57 14 Figure 2.3-3 3-3 14 3-55 14 2-58 14 Figure 2.3-4 3-4 14 3-56 14 2-59 14 Figure 2.3-5 3-5 14 3-57 14 2-60 14 Figure 2.3-6 3-6 14 3-58 14 2-61 14 Figure 2.3-7 3-7 14 3-59 14 2-62 14 Figure 2.3-8 3-8 14 3-60 14 2-63 14 3-9 14 3-61 14 2-64 14 3-10 14 3-62 14 2-65 14 Figure 2.4-1 3-11 14 3-63 14 2-66 14 3-12 14 3-64 14 2-67 14 3-13 15 3-65 15 2-68 14 3-14 15 3-66 14 2-69 14 Figure 2.5-1 3 15 15 3-67 14
,.- 2-70 14 Figure 2.5-2 3-16 14 3 <58 14 4 2-71 14 Figure 2.5-3 3-17 14 3-69 14 2 72 14 Figure 2.5-4 3-18 14 3-70 14 2-73 14 Figure 2.5-5 3-19 14 3-71 14 3-20 14 3-72 14 3-21 14 3-73 15 3-22 14 3-74 14 3-23 14 3-75 14 3-24 14 3 76 14 3-25 14 Table 3.1.1 3-77 14 3-26 14 Figure 3.1-1 3-78 14 q 3-27 14 Figure 3.1-2 3-79 14 !
3-28 14 3-80 14 3-29 15 3-81 14 Figure 3.3-1 3-30 14 3-82 14 Figure 3.3-2 3-31 14 3-83 14 Figure 3.3-3 !
i 3-32 14 3-84 14 Figure 3.3-4 3-33 14 3-85 14 Figure 3.3-5 3-34 14 3-86 14 Figure 3.3-6 3-35 14 3-87 14 Figure 3.3-7 3-36 14 3-88 14 3-37 14 3-89 14 3-38 14 3-90 14 ;
3-39 14 Figure 3.2-1 3-91 14 3-40 14 Figure 3.2-2 3-92 14 l 3-41 14 Figure 3.2-3 3-93 14 Figure 3.3-8 ;
3-42 14 Figure 3.2-4 3-94 14 ;
3-43 14 Figure 3.2-5 3-95 14 3 44 14 Figure 3.2-6 3-96 14 '
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a I MYAPC LIST OF EFFECTIVE PAGES
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PAGE REV REMARKS PAGE REV REMARKS PAGE REV REMARKS i 3-97 14 1
3-98 14 SECTION 4.0 SECTION 5.0 3-99 14 Figure 3.3-9 3-100 14 Figure 3.3-10 4-1 14 Table of Contents 5-1 14 Table of Contents 3-101 14 Figure 3.3-11 4-il 14 Table of Contents 5-ii 14 Ust of Tables 3-102 14 4-iii 14 Ust of Tables 5-iii 14 Ust of Figures 3-103 14 4-iv 14 Ust of Figures 5-1 15 3-104 14 4-1 14 5-2 14 3-105 14 4-2 14 5-3 14 3-106 14 Figure 3.3-12 4-3 14 5-4 14 3-107 14 4-4 14 5-5 14 3-108 14 4-5 14 5-6 14 3 109 14 4-6 14 5-7 14 3-110 14 4-7 14 5-8 14 Table 5.2.1 3-111 14 4-8 14 5-9 15 3-112 14 4-9 14 5-10 14 3 113 14 4-10 14 Table 4.4.1 5-11 15 3-114 14 4-11 14 5-12 14 3-115 14 Table 3.3.1 4-12 14 5-13 14 Table 5.3.1 3-116 14 Figure 3.3-13 4-13 14 5-14 14 Table 5.3.2 3-117 14 Figurs 3.3-14 4-14 14 5-15 14 Table 5.3.3 3-118 14 Figure 3.3-15 4-15 14 5-16 15 3-119 14 Figure 3.3-16 4-16 14 Table 4.6.1 5-1? 14 3-120 14 Figure 3.3-17 4-17 14 Table 4.6.1 5-18 15
, 3-121 14 Figure 3.3-18 5-19 14 (O'~') 3-122 3 123 14 14 5-20 5-21 14 14 3-124 14 5-22 14 3-125 14 5-23 14 3 926 14 5-24 14 3-127 14 Figure 3.3-19 5-25 14 3-128 14 Figure 3.3-20 5-26 14 3 129 14 5-27 14 Table 5.5.1 3-130 14 5-28 14 Table 5.5.2 3-131 14 5-29 14 Table 5.5.2 3-132 14 5-30 14 Table 5.5.2 3-133 14 Figure 3.3-21 5-31 14 Table 5.5.2 3-134 14 Figure 3.3-22 5-32 14 Table 5.5.3 3-135 14 5-33 14 Figure 5.5-1 3-136 15 5-34 14 Figure 5.5-2 3-137 15 5-35 14 Figure 5.5-3 3-138 15 Figure 3.3-23 5-36 14 Figure 5.5-4 3 139 14 5-37 15 3-140 14 5-38 16 3-141 14 5-33 14 3-142 14 5-40 14 Table 5.6.1 3-143 14 5-41 14 Table 5.6.2 3-144 14 Figure 3.3-24 5-42 14 j 3-145 14 Figure 3.3-25 5-43 14 l 3-146 14 5A-1 14 3-147 14 5A-2 14 3-148 14 5A-3 14 3 149 14 5A-4 14 Table 5.A.1 l( O)
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REMARKS PAGE REV REMARKS PAGE REV REMARKS-SB-1 14 SECTION 6.0 SECTION 7 0 58-2 14 5B-3 14 Table 5.8.1 6-1 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 Ustof Figures 7 -111 14 Ust of Figures G-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 75 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 14 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 lO i
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MYAPC LIST OF EFFECTIVE PAGES O PAGE REV REMARKS PAGE REV REMARKS l APENDIX A A 47 14 Table A.2.3
! A-48 14 Table A.2.3 l A-1 14 Table of Contents A-49 14 Table A.2.3 l i A-il 14 List of Tables A-50 14 Table A.2.3 A-iii 14 List of Figures A-51 14 Table A.2.3 A-1 14 A-52 14 Table A.2.3 l A-2 14 A-53 14 Table A.2.4 l A-3 14 A-54 14 Table A.2.5 A-4 14 A-55 14 Table A.2.6 A-5 14 Table A.1.1 A-56 14 Table A.2.7 A-6 14 Table A.1.2 A-57 14 Figure A.2-1 A7 14 Table A.1.3 A 58 14 Figure A.2-2 A-8 14 Table A.1.4 A-59 14 Figure A.2-3 A-9 14 Table A.1.5 A-60 14 Figure A.2-4 A-10 14 Table A.1.6 A41 14 Figure A.2-5 A-11 14 Table A.1.7 A42 14 Figure A.2-6 A 12 14 Table A.1.8 A43 14 Figure A.2-7 A-13 14 Table A.1.9 A44 14 Figure A.2-8 A-14 14 Table A.1.10 A45 14 Figure A.2-9 A-15 14 Table A.1.11 A46 14 Figure A.2-10 A-16 14 Table A.1.12 A47 14 Figure A.2-11 A-17 14 Table A.1.13 A-18 14 Table A.1.14 A-19 14 Table A.1.15 i A 20 14 Table A.1.16 i- A-21 14 Table A.1.17
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A 22 14 Figure A.1-1 A-23 14 Figure A.1-2 A-24 14 Figure A.13 A-25 14 Figure A.1-4 A-26 14 Figure A.15 A-27 14 Figure A.1-6 A 28 14 Figure A.17 A-29 14 Figure A.1-8 A 30 14 Figure A.1-9 A-31 14 Figure A.110 1 A-32 14 . Figure A.1-11 A-33 14 Figure A.112 A-34 14 A-35 14 l A-36 14 Table A.2.1 i I
A-37 14 . Table A.2.2 A-38 14 Table A.2.2 A-39 14 Table A.2.2 A-40 14 Table A.2.2 A-41 14 Table A.2.2 A-42 14 Table A.2.2 A-43 14 Table A.2.2 J A 44 14 Table A.2.2 !
A-45 14 Table A.2.3 A-46 14 Table A.2.3 O DSAR Rev.15
) The, plant is provided with a control room having adequate shielding to permit occupancy during all credible accident situations. The radiation shielding in the plant, in combination with plant radiation control procedures, ensures that operating personnel do not receive radiation exposures l 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.
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 be 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. Offsite and onsite electrical power is provided for the safe 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 fuelin the pool, ample time is available for operators to initiate altemate means of cooling or makeup for the spent fuel pool prior to substantial heatup or inventory loss.
Batteries are installed to supply any required de power.
1.3.7 System Flow Diagrams 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.
l DSAR 1-9 Rev.15 ;
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,i (j 1.5 MaterialIncorocrated Bv Reference Certain program documents and associated topical reports or analyses have been incorporated into the DSAR by reference and are listed in each section as appropriate. This documentation may include information developed by Maine Yankee, as well as Yankee Atomic, ABB-CE, Westinghouse, Stone and Webster, and other organizations.
Some documentation that is incorporated by reference continues to be updated to assure that the information used is the latest available. These documents include the following:
- 1. Quality Assurance Program
- 3. Security Plan
- 5. Off Site Dose Calculation Manual ]
- 7. Post Shutdown Decommissioning Activities Report ]
- 8. Technical Specifications ]
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b Each of these programs and plans may be modified as necessary in accordance with the regulatory and Maine Yankee requirements identified in section 6.
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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 60.82 that the company had permanently ceased power operation and thct 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)
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V need no longer be maintained under nuclear grade controls.
SSC classification involves a determination that an SSC is, or is not, safety-related'. SSCs classified as safety-related are treated differently by regulation than other SSCs.2 j l
For a plant undergoing decommissioning, the only'SSCs which meet the definition of safety- j related8 are the spent fuel pool structure, storage racks, fuel transfer tube, and spent fuel ] i cooling system syphon breaks. This results in two areas of interest: ]
- 1. safety related ssCs are those relied upon to remain functional during and following design basis events to ensure: a) the !
integnty of the reactor coolant pressure boundary; b) the capability to shut down the reactor and maintain it in a safe shutdown f condition; and c) the capability to prevent or mitigate the consequences of accidents that could result in potential offsite 1 exposures comparable to the guidelines of 10 CFR 100. !
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- 2. 10 CFR 50 Appendix B notes that'The pertnent requirements of this appendix apply to all activities affecting the safety- ,
related functions of..? ssCs. I
- 3. The first two parts of the safety.related definition (reactor coolant pressure boundary. and capability to achieve and maintain safe shutdown) do not apply to a decommissioning plant. given the license restrictions of 10 CFR 50.82. The third part of the safety.related definition (accident consequences comparable to 10 CFR 100 guidelines) also does not apply. At Maine Yankee the consequences associated with the design / license basis events applicable to decommissioning are nearty thee l orders of magnitude lower than Part 100 guidelines and lower than the EPA protective action guide limit.
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Maine Yankee's " nuclear grade" processes are based largely upon quality assurance (10 l 1)
( CFR 50 Appendix B) requirements. Reclassifying all SSCs as non-safety related could l lead to the elimination of most management controls in situations where maintaining rigorous manageme.nt controls is intended,
- 2) Maine Yankee recogr,izes that certain functions remain important to safety in the defueled condition. I It is necessary to reclassify SSCs in order to proceed with decommissioning. Strictly following regulatory requirements in reclassification results in elimination of most of the current ]
management controls, which is contrary to management's intent. Thus, in order to provide an ]
enhanced engineering controls above that mandated by regulatory requirements, an artificial ]
classification system termed "important to the Defueled Condition (ITDC) is introduced. ]
The following concems are addressed within this classification: )
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. SSCs which support a fuel safety or radiation protection safety function, and ]
. Identification of enhanced management and engineering controls are maintained on SSCs ]
classified as ITDC.
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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. ]
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The following criteria are used to determine which SSCs are designated as ITDC:
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 ] 1
- the safe storage, control, or maintenance of the spent nuclear fuel or safe ]
I handling of radioactive waste. ]
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DSAR 3-14 Rev.15 l
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The SSC historical classification is the direct result of an outstanding V Criterien 3. ]
commitment to the USNRC which remains applicable to safe storage, control, ]
or maintenance of the spent nuclear fuel; or safe handling of radioactive waste. ]
Criterion 4. The SSC satisfies a requirement based in regulations which remain applicable ] ;
to safe storage, control, or maintenance of the spent nuclear fuel; or safe ]
i handling of radioactive waste as defined in the Maine Yankee licensing basis. ]
This includes any SSC which is independently required by the Limiting ]
Conditions of Operation (Section 3) of the Technical Specifications.' ]
A positive response to any criterion indicates that an SSC is ITDC. ] l Authorizations. Restrictions and Limitations on use of the SSC reclassification criteria.
The SSC reclassification criteria will be used as a basis to change various Maine Yankee processes, provided that the change involves an SSC that is non-lTDC 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
. drawings,
. calculations,
. procedures
. nonconforming items and corrective actions (Learning Bank)
. extemal industry operating experience reports
. commitments
. 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)
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/* 4 The ITDC evaluation assures that the apprepnate regulatory change mechanism is used for effecting the change. ]
DSAR 3-15 Rev.15
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as a moveable platform with hoist. A new-fuel area adjoins the spent fuel pool. The pool is
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I 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 ACI-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.
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l V 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, 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 stainless 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 j integrity design basis is not applicable in the defueled condition, it has been reclassified as safety j 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 l and is we'ded to the fuel pool stainless steel liner. The outer pipe is fitted with bellows expansion i (3 l
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C' Even then,' the short half life of iodine coupled with the fuel decay time minimizes the source term dose affects contributed by radio-locines.
Fuel Assembiv Structural and Material Considerations The fuel assemblies are designed to maintain their structural integrity under steady-state and transient operating conditions, as wt 11 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 tne basis of the functional requirements of the components.
All components, including welds, are highly resistant to corrosion in the reactor snd 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 the form of sintered UO 2 pellets, bumable absorber, or water-filled rods. The principle des gn structural criteria for the fuel rods is that the predicted permanent strain of the ciadding is less than 1.0% during the fuel lifetime.
Fuel Pool SSC Structural and Material Considerations The fuel pool wall and floor slab is con: tnacted of 6' 0" thick reinforced concrete with a wall and floor lining of 1/4-inch thick stainless steeT to ensure against loss of water. The new and spent fuel pool structures including fuel racks are de signed 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 cetermine the dead load and hydrostatic force affect on the SFP concrete wall, floor slab, and stain, ass 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 for a differential temperature of 186'F. Further qualitative discussions showed that even if the ACI allowables were exceeded (i.e.,long tem 11oss of all SFP cooling during the coldest possible winter temperatures), the water retaining capacity of the pool would be maintained since the pool side concrete remains in compression and the 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 ,
l concrete walls and base slab have adequate capacity to resist the forces and moments generated j by the self weight, water pressure, and ther nal effects due to the boil-off of the pool. The load !
factors assumed for the evaluation are for a s nyice load condition and therefore does not restrict ;
the elevated temperature condition to a, one time only, limited duration incident. !
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t D Fuel Handlinc Incident
[V For the fuel handling incident, no credit is taken for systems or components to mitigate the incident such as fuel building ventilation 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 storage incidents described above.
Cask Droo incident Movement of the cask has been carefully examined and in no case does the cask pass over systems or equipment important to safety. An analysis has revealed that a 100-ton,6-foot diameter cask dropped 42 feet straight down into the fuel 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 O the 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 equipment from this leakage would occur.
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 over the fuel pool.
Loss of Heat Sink Loss of PCCW cooling to the SFP heat exchanger (E-25) will result in a heat-up of the pool. PCCW normally flows through the shell side of the fuel pool heat exchanger and cools the tube side pool water, if PCCW is not available, temporary hoses may be connected to flanged connections l provided on the shell side of the heat exchanger to allow the alignment of alternate cooling water i supplies. Loss of PCCW cooling is bounded by the " Loss of Spent Fuel Pool Cooling" incidents
'V DSAR 3-73 Rev.15
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3.3:7 Electrical Systems
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The station electrical systems include the equipment and facilities which provide power to desired plant equipment, instrumentation and controls. The station electrical system has a preferred and standby power supply. The system is designed to provide reliable power for the permanently defueled condition. Individual power systems are designed with sufficient sources, relay protection, control, and necessary switching. l I
3.3.7.1 Offsite Power !
Four 345 kV transmission lines connect the plant to the power grid at the following locations: !
= Central Maine Power Company's Mason generating station
- Central Maine Power Company's Surowiec substation at Pownal
. Central Maine Power Company's Buxton substation !
- Maine Electric Power Company's Maxcys substation at Windsor g< The bus arrangement and outgoing lines are shown in Figure 3.3-23. The 345 KV switchyard utilizes a ring bus arrangement which has the capability of being developed into a breaker and a !
half design.
Station power is supplied by transformer X14. The transformer provides power to the four 4.16 kV buses. The 115 kV source for the station transformer is the '115 kV switchyard.
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115 kV line undervoltage is annunciated in the Main Control Room. The alarm ensures that the Control Room personnel will be alerted to a loss of reserve station service.
l : Each 345 kV and 115 kV breaker is monitored and protected by two completely independent i relaying systems which provide primary and backup schemes for phase and ground fault clearing.
Equipment is included for stuck breaker protection. There are dual trip coils on each circuit breaker. A separate battery for each protective relay system provides operating and tripping power. Necessary control facilities are located in either the relay house or in the station main i control room. Any single circuit breaker can be switched and isolated without affecting another circuit, or the power, or the protection to any other circuit.
1 The station control room is provided with controls, instrumentation, indicators, and trouble annunciators. Electrical conditions which affect the switchyard status, and any breaker operation, l
are annunciated in the control room. Direct communication links between the control room and the lead dispatch offices of the interconnected power systems facilitate continuous verification of the status of the transmission lines. Remote metering at the Control Room indicate the status of the 115 kV bus at Surowiec substation.
The offsite power system is a reliable source of electrical power for plant equipment. In the defueled condition, no active systems meet the criteria for safety related systems or components
. as the consequences of accidents are significantly lower than the limits of 10CFR100.11. In the event of an interruption in power, the robust design of the passive systems assure the continued safe storage of fuel.
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. ,m b - SECTION 5.0 ACCIDENT ANALYSIS 5.1 Introduction Earlier sections of this report de::cribe the major systems and components of the plant from the perspective of safe spent fuel handling, spent fuel storage, and other decommissioning activities as would be appropriate to a permanently defueled plant.
This section uses the previous information and examines the potential consequences of accidents and incidents, notwithstanding the precautions taken to prevent their happening, to assess the adequacy of the plant design in minimizing or mitigating potential consequences of such occurrences. Additionally, the accident analyses presented in this section provide assurance that the health and safety of the public is protected from the consequences of even the most severe of the hypothetical incidents analyzed.
With the permanent defueling of the Maine Yankee facility and the certification of the cessation of operations, the postulated accidents associated with reactor operation are no longer applicable and Q need not be considered. Likewise, the unirradiated nuclear fuel has been removed from the Maine Yankee site and therefore accidents involving new fuel assemblies are also no longer applicable.
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However, those accidents associated with the storage or handling of irradiated fuel or radioactive waste storage or processing remain applicable and are discussed within this section.
The general classification of accidents for the permanently defueled condition are limited. These groupings are listed as follows:
- 1. Inadvertent criticality of the stored spent fuel,
- 2. Fuel assembly handling accident,
- 3. Spent fuel shipping cask drop in the spent fuel pool,
- 4. Loss of spent fuel decay heat removal capability,
- 5. Loss of spent fuel poolinventory,
- 6. Radioactive release from a subsystem or component, or
- 7. Low level waste storage accident.
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C 5.3 Fuel Handlina Accident The purpose of this cection is to assess anticipated spent fuel pool fuel handling operations in order to arrive at the accident which would result in appropriately conservative off-site and control room radioactive release effects. Fuel handling incidents originating in the containment are not applicable to the pemlanently defueled condition. Fuel handling operation associated with the use of a spent fuel cask are addressed in section 5.4.
The likelihood of a fuel handling incident in the spent fuel pool is minimized by implementation of appropriate and long standing administrative controls and physical limitations imposed on fuel hand!!ng operations. All fuel handling operations cre conducted in accordance with prescribed procedures under the direct surveillance and supervision of qualified personnel.
The fuel handling equipment and facility are designed for the transfer and handling of a single fuel assembly at any time, and movement of equipment when handling the fuel is administratively restricted. The fuel handling manipulators and hoists are designed so that fuel cannot be raised above a position which provides adequate shield water depth for the safety of operating personnel.
in the spent fuel pool, the design of fuel storage racks and manipulator equipment, in conjunction with appropriate administrative controls, is such that:
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- 1. Fuel is always maintained by mechanical restraint. Fuel at rest is positioned by '
positive restraints in a suberitical geometrical array, with no credit for boric acid in the water.
- 2. Fuel can be manipulated only one assembly at a time.
- 3. Violation of procedures by placing one fuel assembly in juxtaposition with any group of assemblies in racks will not result in criticality.
- 4. The spent fuel shipping container does not pass over spent fuel during transfer ]
operations.
The fuel assembly is immersed continuously while in the spent fuel pool. Adequate cooling of fuel during underwater handling is provided by convective heat transfer to the surrounding water. The fuel handling equipment and spent fuel pool are described in detail in Section 3.
Inadvertent disengagement of the fuel assembly from the fuel handling equipment is prevented by design, mechanical, and procedural interlocks. Consequently, the possibility of dropping and (h,
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4 No fuel overheating occurs, and thus, no significant solid fission products are released.
- 5. No credit is taken for the isolation of the spent fuel building. The release is
. modeled as an instantaneous puff release at accident (95 percentile) ground l l
level meteorological conditions.
- 6. The free air volume of the CR is 38,000 cubic feet.
- 7. Normal CR ventilation is either 120 or 900 cubic feet per minute for the duration of the accident. ]
- 8. No credit is assumed for CR isolation or air filtration.
The radiological analyses was performed using the code ELISA, Reference 1.
The results of this analysis for the EAB are presented in Table 5.3.2. These results show that the -
projected doses from the fuel handing accident are insignificant in comparison to the 10 CFR 100 limits and far less than the Environmental Protection Agency Protective Action Guidelines (PAGs).
l The calculated doses for the CR portion of the analysis at the CR ventilation intakes and inside the CR are presented in Table 5.3.3. These results show that the anticipated doses are within the 10
!(p) CFR 50, General Design Criteria 19 dose limits.
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- 54. Soent Fuel Cask Droo Spent fuel shipping casks are designed, as per the requirements of 10 CFR 71, to withstand a free I fall of 30 feet onto an unyielding surface. For this reason; radiological consequences of a postulated spent fuel cask drop accident outside the spent fuel pool are not required if potential drop distances are less than 30 feet.
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The design of the Maine Yankee spent fuel cask transfer system is such that the cask drop distance is less than 30 feet whenever the cask is not directly over the spent fuel pool. It is I concluded, therefore, that an evaluation of the spent fuel cask drop accident outside of the spent fuel poolis not necessary.
-_ Operations with tne spent fuel shipping container are designed act to pass ever spent fuel storage L racks or spent fuel assemblies during cask loading or fuel transfer operations. ,
L At the current time, Maine Yankee is prohibited from lifting a spent fuel shipping cask over the spent fuel pool. Therefore, an accident analysis of a spent fuel cask drop in the spent fuel pool is l not required and does not supply safety analysis limits. ]
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A MYAPC heat calculations were conservatively determined assuming prior plant operation at 2700 MWt for the most recently discharged spent fuel. Actual power operation was limited to 2440 MWt for the ]
last batch of fuel used for power production at Maine Yankee. ]
In' late October and early November 1997, Maine Yankee performed passive cooling tests in the spent fuel pool to obtain data in assessing the pool heatup rate, heat losses due to evaporation at
' elevated temperatures, and accuracy of the calculation of spent fuel decay heat. Conservative assumptions were used for inclusion of the fuel building heat losses in this assessment. The results of these tests indicate that passive (i.e.: no forced flow through the cooling system) coo!!ng of the spent fuel pool will result in a steady bulk water temperature of between 190 and 200*F with water surface wind speeds between 3 and 5 miles per hour. When compared to the analytical predictions of the spent fuel decay heat using the NRC Branch Technical Position ASB 9-2, these results demonstrate substantial margin to the decay heat analysis assumptions presented in this
. section.
5.5.1.1 Blocked / Improper Cell Flow The design of the spent fuel racks (Reference 1) is such that the top of the rack cells may be blocked and sufficient cooling of the fuel assembly is still assured. Analyses of various types of
' flow blockage of the cell exit have been performed to demonstrate satisfactory preservation of the stored fuel in a coolable geometry under these conditions.
Two types of cell blockage were considered. The first assumed that a fuel assembly was laying horizontally across the limiting spent fuel storage cell. Conservatively neglecting flow upwards ,
through the horizontal pins, this event results in a net blockage of the storage cell exit area of 79%
The second flow blockage scenario involves the placement of a 8 by 10 foot section of masonry wall from the south end of the spent fuel pool onto the top of the storage cells. This event blocks 100% of multiple cell exit openings and requires the cooling of the spent fuel through the 1 inch !
diameter flow holes located near the top of the cells, j j
' Specific analyses were performed using the RETRAN code (Reference 2) in analyzing the blockage of cells to conservatively predict the limiting fuel cell coolant conditions in verifying that localized boiling or that the onset of Critical Heat Flux (CHF) will not occur in the individual fuel storage cell.
The limiting cell was then evaluated to determine the peak fuel pin surface cladding and fuel pellet temperatures. These analyses were performed as part of Reference 1.
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i p MYAPC 5.6 ' Low Level Waste Releare incidents 5.6.1 Radioactive Waste Gas System Leaks and Failures i
Radioactive waste gas decay tanks permit decay of accumulated radioactive gases prior to their l release as a means of reducing the normal release of radioactive materials to the atmosphere. The radioactive contents were principally the noble gases krypton and xenon, the particulate daughters of some of the krypton and xenon isotopes, and trace quantities of halogens. Since these noble gases are generated from the fissioning process during power operation, there will be no generation of fission gases and no more gases sent to the waste gas decay tanks from the Reactor Coolant System with the reoctor permanently defueled. The inventory of the waste gas decay i tanks resulting from the last period of power operations has been released and the tanks contain !
only residual amounts of radioactive waste gases.
The projected radioactive gas release information and projected doses that are applicable to the permanently defueled condition are defined in sections 5.3,5.5, and 5.6.
fS 5.6.2 Radioactive Liquid Waste System Leaks and Failures U
Because of the continued processing of radioactive fluids at Maine Yankee during the permanently defueled condition, the analysis of radioactive liquid waste system leaks end failurce remains essentially the same as that used as the licensing basis for the plant during power operations. The use of the full power transient is used as representative of the maximum dose consequences associated with the offsite receptor due to a liquid and gaseous release. This transient is not a credible accident scenario in the defueled condition. It is not intended to be the sole basis upon which the retention of systems or components is required.
Actions leading to transfer of a radioactive fluid from a system to the environment or to another system require positive operator control and monitoring.
All operator actions required are performed in accordance with written operating procedures which ]
include instructions, checklists and allowable release information. Following final radioactivity analysis, effluent is released to the environment through a process radiation monitor interlocked ]
with the discharge valve such that a high radiation alarm will close the valve and terminate the release.
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,m (j With these considerations in mind, the probability of the accidental release of radioactivity from the radwaste systems as a result of operator error is minimal.
The purpose of this section, however, is to consider postulated liquid waste system single component leaks or failures in order to arrive at the postulated incident which could have maximum off-site effects.
Method of Analysis:
A. Releases to the Atmosohere (Gaseous Releases)
For the purpose of establishing an upper limit on the activity released from a single component failure in the liquid waste system, it is assumed that the primary drain tank fails, releasing its total inventory. The primary drain tank failure has been selected since this tank has the highest inventory of dissolved noble gases and halogens during operational periods. The release takes place ] 1 as a liquid spill on the floor of the compartment in the waste processing building where the tank is located. Radioactivity is released to the atmosphere from noble gases and halogens evolved from the spilled liquid.
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i ) The following conservative assumptions are used in conjunction with the meteorological and dose assumptions given in Appendices 5A and 5B.
- 1. Eighty percent of the primary drain tank's 8,150 gallon capacity is filled with undecayed, un-degasified primary reactor coolant (with activity concentrations at Technical Specification limits of 1.0 pCl/g Dose Equivalent 1-131 and 100/E pCi/g).
- 2. One hundred percent of the tank's inventory is spilled and all of the noble gases and 1 percent of the radiciodines are available for direct release through the building ventilation system to the environment.
- 3. Duration of the release is 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
B. Potential Releases to the Groundwater Table Postulated liquid spills escaping concrete structures may be released to the site groundwater table. The groundwater table at the Maine Yankee site, however, flows towards Back River and Montsweag Bay, both of which are tidal saltwater
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v 6.5.4 Fitness For Duty Program l
As a result of the permanently shutdown condition of the plant and the 10 CFR 50.82(a)(1) certifications, the NRC has concluded that the Fitness for Duty Program rule,10 CFR 26, no longer j applies to Maine Yankee, Reference 1. l 6.5.5 Offsite Dose Calculation Manual The Maine Yankee Offsite Dose Calculation Manual (ODCM) is defined by Technical Specifications to conta:n the methodology and parameters used in the calculation of off-site doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm / Trip Setpoints and in the conduct of the Environmental Radiological Monitoring Program.
The ODCM shall also contain the Radioactive Effluent Control and Radiological Environmental Monitoring Programs and descriptions of the information that should be included in the Annual Radiological Environmental Operating and Annual Radioactive Effluent Release Reports.
Changes to the ODCM shall be made in accordance with Maine Yankee Technical Specifications
() and may be made if the change will maintain the level of radioactive effluent control required by applicable regulations and not adversely impact the accuracy or reliability of effluent, dose, or l
setpoint calculations.
6.5.6 Quality Assurance Program The Maine Yankee Quality Assurance Program is docketed as a separate document and is !
required by 10 CFR 50.54(a). Changes to the Quality Assurance Plan are evaluated under 10 CFR 50.54 (a) which allows changes to be made without NRC approval if these changes do not reduce the commitments in the program description previously accepted by the NRC. These changes
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' must be submitted to the NRC in accordance with the requirements of 50.71(e), FSAR update l requirements.
l 6.5.7 Process Control Program l
The Process Control Program (PCP) contains the current formulas, sampling analyses, tests and {
determinations to be made to ensure that processing and packaging of solid radioactive wastes O
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References:
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- 1. Letter, USNRC to Maine Yankee," Fitness for Duty Programs'(10 CFR 26) for Maine l Yankee Atomic Power Station", dated January 12,1998. l 4
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