ML20149H177
| ML20149H177 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 11/10/1994 |
| From: | PORTLAND GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20149H176 | List: |
| References | |
| NUDOCS 9411180147 | |
| Download: ML20149H177 (18) | |
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4 h TROJAN NUCLEAR PLANT Defueled Safety Analysis ReDort REVISION 1 ,
The following information is furnished as a guide for the insertion of the new sheets for Revision 1 into the Trojan Nuclear Plant Defueled Safety Analysis Report. This material is denoted by the use of the revision number in the lower outside corner of the page.
Revised pages should be inserted as listed below-t Insert Delete (Front /Back) (Front /Back) l Page xxiii/ xxiv Page xxiii/ xxiv ,
Page xxv/ Blank Page xxv/ Blank Page 3.4-3/Page 3.4-4 Page 3.4-3/Page 3.4-4 Table 3.1-3/ Table 3.1-4 Table 3.1-3/rable 3.1-4 Table 3.2-4/ Table 3.2-5 Table 3.2-4/ Table 3.2-5 Table 3.5-1 Sheets 5/6 Table 3.5-1 Sheet S/6 Table 3.5-1 Sheets 7/8 Table 3.5-1 Sheets 7/8 Figure 3.3-1/ Blank Figure 3.3-1/ Blank Page 5.6-1/Page 5.6-2 Page 5.6-1/Page 5.6-2 O
9411180147 941110 PDR ADDCK 05000344 P PDR
O LIST OF EFFECIIVE PAGES U !
DEFUELED SAFETY ANALYSIS REPORT l l
Section Effective Pages Date Title Page N/A Rev. 0 Table of Contents i - xxii Rev. 0 List of Effective Pages xxiii - xxv Rev. 0 1.0 1.0-1 Rev. 0 1.1 1.1-1 Rev. 0 1.2 1.2-1 through 1.2-2 Rev. 0 13 13-1 Rev. 0 1.4 1.4-1 through 1.4-5 Rev. 0 1.5 1.5-1 Rev. O Figure 1.1-1 N/A Rev. 0 2.0 2.0-1 Rev. 0 2.1 2.1-1 through 2.1-12 Rev. 0 2.2 2.2-1 through 2.2-22 Rev. 0 23 23-1 through 23-12 Rev. 0
/ 2.4 2.4-1 through 2.4-36 Rev.0
- 2.5 2.5-1 through 2.5-47 Rev.0 2.6 2.6-1 through 2.6-10 Rev. O Tables 23-1 through 23-6 N/A Rev. O Figure 23-1 N/A Rev. 0 Figures 2.4-1 and 2.4-2 N/A Rev. O ;
Rev. 0 l 3.0 3.0-1 3.1 3.1-1 through 3.1-28 Rev. 0 3.2 3.2-1 through 3.2-40 Rev. 0 33 33-1 through 33-19 Rev. 0 3.4 3.4-1 through 3.4-3 Rev. 0 3.4 3.4-4 Rev.1 3.4 3.4-5 through 3.4-6 Rev. 0 3.5 3.5-1 Rev. 0 3.6 3.6-1 through 3.6-3 Rev. 0 Table 3.1-1 N/A Rev.1 Tables 3.1-2 through 3.1-7 N/A Rev. O Tables 3.2-1 through 3.2-3 N/A Rev. O Tables 3.2-4 through 3.2-5 N/A Rev.1 4
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LIST OF EFFECTIVE PAGES g DEFUELED SAFETY ANALYSIS REPORT Section Effective Pages Date Table 3.5-1 Sheets 1 through 5 Rev.O Table 3.5-1 Sheets 6 through 8 Rev.1 Table 3.5-1 Sheets 9 through 34 Rev. O Figures 3.1-1 through 3.1-19 N/A Rev.0 Figures 3.2-1 through 3.2-30 N/A Rev. O Figure 33-1 N/A Rev.1 Figures 33-2 through 33-3 N/A Rev.0 4.0 4.0-1 Rev. 0 4.1 4.1-1 through 4.1-4 Rev.O d.2 4.2-1 through 4.2-3 Rev.0 43 43-1 through 43-6 Rev.0 4.4 4.4-1 through 4.4-2 Rev.0 4.5 4.5-1 Rev. 0 Tables 43-1 N/A Rev. O Figures 4.2-1 and 4.2-2 N/A Rev.O g 5.0 5.0-1 Rev. 0 5.1 5.1-1 Rev. 0 5.2 5.2-1 through 5.2-7 Rev.0 53 53-1 through 53-8 Rev.0 5.4 5.41 through 5.4-4 Rev.0 5.5 5.5-1 through 5.5-10 Rev. 0 5.6 5.6-1 through 5.6-2 Rev.1 5.6 5.6-3 through 5.6-9 Rev.0 5.7 5.7-1 Rev.0 5.8 5.8-1 Rev.0 5.9 5.9-1 through 5.9-2 Rev.O Tables 5.5-1 N/A Rev. O Figures 5.2-1 through 5.2-6 N/A Rev. O Figures 53-1 through 53-3 N/A Rev. 0 Figure 5.4-1 N/A Rev. 0 6.0 6.0-1 through 6.0-8 Rev. 0 6.1 6.1-1 through 6.1-4 Rev. 0 6.2 6.2-1 through 6.2-3 Rev. 0 63 63-1 through 63-7 Rev.0 Revision 1 xxiv 0
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f LIST OF EFFECTIVE PAGES DEFUELED SAFETY ANALYSIS REPORT ,
l Section Effective Pages Date l 6.4 6.4-1 through 6.4-2 Rev. 0 l I
Tables 6.0-1 through 6.0-5 N/A Rev. O Tables 6.2-1 and 6.2-2 N/A Rev. 0 :
Tables 63-1 and 63-3 N/A Rev. 0 Figure 6.1-1 N/A Rev. 0 Figures 6.3-1 through 6.3-6 N/A Rev. 0 7.0 7.0-1 Rev. 0 7.1 7.1-1 through 7.1-6 Rev. 0 7.2 7.2-1 through 7.2-6 Rev. 0 73 7.3-1 through 73-3 Rev. 0 i
7.4 7.4-1 Rev. O 7.5 7.5-1 Rev. 0 7.6 7.6-1 Rev. 0 8.0 8.0-1 Rev. 0 9.0 9.0-1 Rev. 0 l
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O connected to bus Al and the other connected to bus A2, are provided to supply standby power to the buses. Further description of the diesel generators is provided in Section 3.4.2.1.5. The SFP cooling pumps are energized from motor control centers (MCCs) fed from bus A6. Emergency power can be supplied to A6 from EDG "B" via i bus A2 or from EDG "A" via buses Al and A5 using the bus A5 to A6 cross-tie breaker.
For defueled conditions, train separation is no longer required. The ability to cross-tie buses provides flexibility for maintaining power to station loads.
3.4.2.1.3 480-V System The 480-V system is designed to provide sufficient electric power for operation of Plant loads from load centers and MCC buses. The system consists ofload centers, MCCs, loads fed from these load centers and MCCs, interconnecting cables, and associated instrumentation and control circuits.
For detueled conditions, train separation is no longer required. The ability to cross-tie buses provides flexibility for maintaining power to station loads.
l 3.4.2.1.4 120-V A-C System l
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The 120-V instrument a-c and the preferred instrument a-c systems are designed to provide reliable electric power for control and instrumentation.
The 120-V instrument a-c system was designed to supply 208/120-V control and instrumentation power for equipment not required for the safe shutdown of the reactor.
This system supplies power to the normally open service water isolation valves at the Seismic Category I interface with the Seismic Category II portion of the system. In the defueled condition this automatic isolation is no longer needed.
3.4-3 l
The 120-V preferred instrument a-c system consists of four 120-V bus sections which are supplied from four power sources consisting of a static inverter and a 120-V instrument a-c bus. For defueled conditions, the system no longer pedorms any safety functions.
3.4.2.1.5 Standby Power Supply System Two emergency diesel generator units are provided to generate and supply standby power to the redundant 4.16-kV buses Al and A2. Diesel generator unit No. I supplies power to bus Al and diesel generator unit No. 2 supp!ies power to bus A2.
With the teactor defueled, the loads on the diesel generators are substantially lower than the original design load requirementsm. Most of the loads have been disabled and isolated from die system. This allows manualloading of selected loads onto the buses to restore SFP cooling.
The standby power supply system consists of two identical diesel generator units, accessories, fuel storage and transfer systems, interconnecting cables and instrumentation.
Each diesel generator set is a complete package unit including auxiliaries required to make it a self-sufficient power source capable of continuous operation at rated fullload, and rated voltage and frequency until either manually or automatically stopped. Each )
l generator is driven by two fast-starting, tandem-mounted, diesel engines, with complete '
air-starting systems, closed-loop jacket cooling water system, diesel generator lube oil system, and engine and generator control panels.
Diesel fuel oil is supplied to each engine from the diesel generator fuel oil day tanks.
These tanks are replenished by transfer pumps from remotely located fuel oil storage tanks.
Power to start each diesel engine is provided by two air-starting systems. Each tandem diesel ge; ntor is equipped with four paired air-starting motor units, two per diesel; any g Revisice ' 3.4_4 m
l O TABLE 3.1-3 l
LOCATIONS OF GAS STORAGE TANKS )
Equip. No. Service Location Dimensions psig T-112 EDG air EDG Room in Turbine 2-ft 6-in x 300 A,B,C,D, tank B!dg. near Column 45 8-ft high max E,F,G,H line wall, Elevation 45 ft.
T-118 Compressed Turbine Bldg. Elevation 3-ft dia x 165 A,B,C air receivers 45 ft. outside EDG 8-ft high max rooms between column 51 -55 T-151 N2bulk gas Roof of Control Bldg. 24-in dia x 2450 A,B,C,D, tanks 21-ft long E, F T-159 Liquid CO 2 Turbine Bldg. between 4-ft dia x 300 storage tank column 77 - 83, 10-ft long Elevation 45 ft.
T-157 Air Main steam support 183 in' each 125 O A,B,C,D accumulator structure, Elevation 64 ft.
T-125 Hydro- Turbine Bldg. south end, 11-ft dia x 125 pneumatic Elevation 45 ft., between 19-ft long pressure column U and T accumulator (air)
T-204 A, B CCWS surge Elevation 77 ft., column 7-ft dia x 150 tank (N 2) line E and 61, between 8-ft 3-in high ,
Reactor and Auxilag Building O
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TABLE 3.1-4 ANALYZED SFP LOAD DROPS AND MISSILES LOAD DROPS Max. Max.
Dry Droo Impact Impact Item Weicht Hei Enercy Surface No. Item Descriotion (lb1 in.) ght( (in.-lb~s) (in.)
1 Spent fuel assembly 1,624 12 19,488 8.42 x 8.42 with reactor contro'l rod 2 Spent fuel assembly 356 172 61,200 9.0 x 9.0 handline tool 3 Spent fuel assembly 1,976 12 23,700 8.42 x 8.42 with handline tool 4 BPRA handline tool 800 148 118.400 18.0 x 18.0 5 Thimble plug removal 290 192 55,700 8.69 x 8.69 tool 6 Fuel assembly channel 350 254 88,900 8.42 x 8.42 spacine tool g
7 Radiation specimen transfer basket 600 12 7,200 8.42 x 8.42 W 8 New fuel tool 80 480 38.400 9.0 x 9.0 9 Electric chain hoist 90 530 47.700 8 diameter 10 Portable RCCA 900 148 133,200 8.42 x 8.42 chance tool 11 Miscellaneous hand SC) 300 150,000 36 x 48 tooling / portable ;
vacuum system -
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TORNADO MISSILES _ l Vertical Velocity at Item Dry Weight SFP Surface No. Item Descriotion ~ (lb) (moh) 1 4-in. x 4-in. x 12-ft lone wood plank 108 140
, 2 3-in. diameter x 10-ft tone steel cine 76 52.5 O
IAllLE 3.2-4 cal CUl ATED PEstfl.TS - CONTROL. Bull DillC
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Earthquebe Description LeaJ 8trese Calculated Alloweble location of Hesber Combinetton Combination Sttese Streen Rese Le of itemi er u 24 o 76 peo= Floor of control room on Col. Line O D4 8. 19.8 bei 24.0 bel t.etween 46 and 51 - El. 93'-0*
l'-O" Concrete Floor of conteel room on Col. I.Ine 0 1. 5 0 e 1. 8 B. 7.5 k-ft 8.6 k-ft Allowable le UltI=ste Sieb between 46 and St - El . 9 3'-0" Homent Ca reel t y W 18 e 45 Been Floor of cable spreedlag room on Col. Line O Dt 1. 16.0 bei 24.0 kel ,
bet we en 4 6 a nd 51 - El . 7 7'-0" 6" Concrete Floor of cable spreading room on Col. Line 0 1.5 D + l . 8 l. l.8 k-ft 5.6 k-ft Allowable is Ultimate Slob between 4 6 and 51 - El . 7 7'-0* lb.ent cepecity del 0.942 1.0 alloweblo le by Column W 14 e 64 Co l e.en Column I.ine 0 El. 6t'-0"
-t eract len Fossut e Co l umn 1.t ne 0 - 4 6 - 31. 6 l '-O" D .t 1. 0.979 1.0 Allevable le by Colven W 12 m 142 Celemn Internetten Fosmule tee t e s From t i.e Indiested toevate it is apparent t lie t stralne are not celticall tlierefore, strain values are not given.
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l TA.BLE 3.2 5 h SIGNIFICANT ELEVATIONS IN THE SPENT FUEL POOL l l
Descriotion Elevation Top of the spent fuel pool 93' 0" Normal water level approximately 91' Water level after loss of gates to the fuel 84' transfer canal and cask loading pit with the cask loading pit filled Siphon breakers 83'11" Bottom of gate to the fuel transfer canal 68' 4" Top of fuel racks 67' 4" Top of fuel assemblies 66' 7" Wetted side of the bottom of the spent 52' 6" <
fuel pool I
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r TABLE 3.5-1 Sheet 5 of 34 systems that are intended to store or delay the release of gaseous radioactive waste, including portions of structures housing the system, to be Seismic Category I. Whereas Trojan's seismic classification guidelines for radioactive waste management are not as specific as those given in Regulatory Guide 1.143, the seismic classification of Trojan's solid, liquid, and gaseous radwaste systems, and the structures housing the systems, are in full compliance with the above seismic requirements of Regulatory Guide 1.143 for such systems.
(c) Regulatory Position C.1.p of R6 vision 3 to Regulatory Guide 1.29 also requires that any other systems (other than radwaste systems) not covered specifically in the Regulatory Guide that contain or may contain radioactive material, and whose postulated failure would result in conservatively calculated potential offsite doses greater than 0.5 rem (whole body or its equivalent to any part of the body), be designated Seismic Categon I. The Trojan classification method does not contain this additional requirement for the classification of structures, systems, and components outside of the Control, Auxiliary, and Fuel Building Complex. These items were classified in accordance with Revision 0 of Regulatory Guide 1.29.
(d) Regulatory Position C.2 of this Regulatory Guide requires that portions of non-safety-related systems whose failure could reduce the functioning of a Seismic Category I plant feature to an unacceptable level be designed and constructed so that the SSE would tot cause such failure. The origmal Plant design included system interaction considerations as well as failure modes and effects analyses primarily in determming system and equipment locations. Portions of non-safety-related Seismic Category II systems (e.g., pipe supports) were not originally designed to Seismic Categon I requirements.
The system interaction requirements in this Regulatory Guide are now implemented in '
Seismic II/I provisions. A system interaction review identifying potentir.1 II/I items has been completed for safety-related systems and equipment in the Control, Auxilian, and Fuel Buildings. Design and analysis of Seismic Categon II/I items are for SSE seismic loadings (operating basis earthquake loads are not required to be analyzed). The design and construction of Seismic Category II/I items wn! be considered quality-related but not safety-related.
The original Plant design meets Regulatory Position C.2 that is given in Revision 0 to this Regulatory Guide. The position to design and construct Seismic Category II/I items for SSE loads complies with Revision 3 to this Regulatory Guide.
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TABLE 3.5-1 Sheet 6 of 34 h
Regulatorv Guide 1.30 - Quality Assurance Requirements for the Installation, Inspection, and Testing of Instrumentation and Electrical Equipment (8/72), Rev. O Compliance Comply.
Reculatorv Guide 1.33 - Quality Assurance Program Requirements (2/78), Rev. 2 Compliance Comply with exception.
(a) Section 5.1 of ANSI N18.7-1976 requires a summary document of those sources ,
providing administrative controls and quality assurance during the operational I l
phase to index such source documents to the criteria of ANSI N18.7-1976. PGE has not compiled such a summary document that indexes procedures and ;
instructions to this standard since the manpower required to perform such a task is not justified.
(b) Paragraph 5.2.1.1 of ANSI N18.7-1976 discusses the RO authority and responsibility. Paragraph 5.2.1.3 discusses the SRO responsibilities. No personnel 1re required to hold RO or SRO licenses per the Technical Specifications.
(c) Section 5.2.2 of ANSI N18.7-1976 requires temporaq changes which do not change the intent of an approved procedure to be approved by the supervisor in charge of the shift who holds a SRO license on the unit affected. Temporary procedure changes which do not change the intent shall be approved by two members of Plant Management, at least one of whom is a Certified Fuel Handler.
This approved temporary change is then reviewed and approved within 14 days by the responsible manager designated in writing in accordance with Technical Specifications.
(d) Not all quality-related procedures are reviewed biennially.
The following non-routine procedures whose usage is dictated by a facility event shall be reviewed at least even 2 years to determine if changes are necessary or desirable: Off-Normal Instructions, Alarm Response Procedures, Emergency h
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TABLE 3.5-1 Sheet 7 of 34 Procedures for the Radiological Emergency Plan and Nuclear Security Safeguards Contingency Procedures.
Other Plant procedures shall be reviewed and revised appropriately when affected by a corrective action, a change in the license or Technical Specifications, an update to the Safety Analysis Report, a Topical Report, A Plant Setpoint Change a Vendor Equipment Technical Manual change, or a similar action required by Section 5.2 of ANSI N18.7-1976, i.e., any modification to a system, an unusual incident (accident), unexpected transient, significant operator error or equipment malfunction. These procedures will be reviewed and audited as specified in procedures, but not necessarily every two years. 'Ibese measures will ensure t procedures are being revised when appropriate and that the procedure rev revision program is effectively implemented.
Section 5.3 of ANSI N18.7-1976 requires procedures for starting up the reactor, (e) steady-state power operation and load changing, shutting down and tripping the reactor, and changing modes. Procedures are not required for these actions in a g defueled mode.
(f) ANSI N18.7-1976 states that certain PGE other complies ANSI with these standards standards will be utilize and their compliance with ANSI N18.7-1976.
associated Regulatory Guides only as described in each Regulatory Guide position in the DSAR.
(g) Section 5.2.6 of ANSI N18.71976 requires procedurcs " . for control of ;
equipment, as necessary, to maintain personnel and reactor safety and avoid i unauthorized operation of equipment. The procedures shall require independent i verification, where appropriate, to ensure that necessary measures, such as tagging i equipment, have been implemented correctly." \
PGE procedures do not require independent verification for clearances for equipment removed from service or system lineups except for designated locke valves. Trojan will maintain personnel safety by performing clearance operations in accordance with Occupational Safety and Health Admuustration (OSHA) regulations.
PGE procedures do not require independent verification for system alignments performed in conjunction with operating procedures. Problems resulting from component mispositioning can be identified and corrected within sufficient time to O prevent fuel damage. Due to plant closure, reactor safety is no longer a concern.
Reculatorv Guide Revision 1
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- TABLE 3.5-1 Sheet 8 of 34 h 137 - Quality Assurance Requirements for Cleaning of Fluid Systems and Assn.ated Components of Water-Cooled Nuclear Power Plants (3/73), Rev. O fomoliance Comply.
Reculatorv Guide 1.38 - Quality Assurance Requirements for Packaging, Shipping, Receiving, Storage, and Handling of Items for Water-Cooled Nuclear Power Plants (5/77), Rev. 2 Comoliance Rev. 2 - Comply with exception.
(a) Quality Asstuance personnel auditing inspection activities will be qualified per ANSI N45.2.23 in lieu of ANSI N45.2.6 (b) The requirements of ANSI N45.2.2-1972 for classification levels will be applied to O
quality-related materials, parts, and components only when specific protection measures are required. In lieu of the detailed requirements of ANSI N45.2.2-1972, quality-related items will receive a thorough engineering evaluation to assure that adequate protective measures are specified for packaging, shipping, receiving, storage, and handling of items for nuclear power plants. These protective measures will be consistent with standard / commercial engineering practices and manufacturers' recommendations.
(c) Marking may be applied to bare austenitic stainless steel and nickel alloy metal surfaces provided that it has been established that the marking is not deleterious to the item rather than as stated in Paragraph A3.9 of ANSI N45 M Proper chemical controls will be employed to ensure there is no adverse metallurgical impact on the steel or nickel alloy. This position has been adopted in the document ASME NOA-2, Addenda 2a.
Reculatorv Guide 1.39 - Housekeeping Requirements for Water-Cooled Nuclear Power Plants (9/77), Rev. 2 Compliance Comply.
O Revision 1
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5.6 RADIATION PROTECTION PROGRAM 5.6.1 RADIATION PROTECTION DESIGN FEATURES 5.6.1.1 Shielding. Radiation Zoninc and ' Access Control The shielding, radiation zoning and access control design features are based on potential radiation sources during normal Plant operation, shutdown, and emergency operations .
which conservatively bound the current defueled statem . With the reactor shutdown and the fuel stored in the SFP, the number and magnitude of the potential radiation sources have been reduced substantially from the original design bases source terms.
5.6.1.2 Plant Ventilation Systems O
V The design features of the Plant ventilation systems are described in Section 5.2.
5.6.1.3 Area Radiation Monitoring Instrumentation The area radiation monitoring system (ARMS) is provided to supplement the personnel and area radiation monitoring of the Plant radiation protection program. Radiation detectors provide local and/or remote indication and alarm of direct radiation dose rate.
The area radiation monitoring system does not perform any safety functions or perform any quantitative measurements of releases of radioactive material to the environment.
7 The ARMS measure radiation levels over the range of 104 to 10 mR/hr. Each monitor generally consists of a detector, remote and local alarms, remote and local indicators and a remote power supply. The remote indicators are located in the ARMS control panelin the control room.
O 5.6-1 Revision 1
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The two monitors located at Elevation 93 feet of the Fuel Building (ARM-12 and h ARM-13) serve as criticality alarms in the new fuel and spent fuel handling and storage areas.
5.6.2 EOUTPMENT. INSTRUMENTATION AND FACILITIES The various facilities provided at Trojan for radiation protection activities, and the Procedures and equipment to be employed for measunng and minimi7ing personnel exposure are described in this section.
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