LIC-14-0043, Enclosure, Attachment 1 to LIC-14-0043 - Updated Safety Analysis Report (USAR) Page Markups
| ML14143A374 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 05/16/2014 |
| From: | Omaha Public Power District |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| LIC-14-0043 | |
| Download: ML14143A374 (5) | |
Text
LIC-14-0043 Enclosure, Attachment 1 Page 1 Updated Safety Analysis Report (USAR)
Page Markups
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USAR-1.2 Information Use Page 27 of 28 Summary Plant Description Rev. 38 When Fort Calhoun Station was originally licensed, non-RCS safety-related piping was designed and constructed to meet the requirements of USAS B31.7, 1968 (DRAFT) Edition (i.e., the Code of Record for this piping).
Amendment No. XXX to Renewed Facility Operating License No. DPR-40 allows the design and/or analysis of non-RCS safety-related piping to be performed in accordance with ASME Section III, 1980 Edition (no Addenda) as an alternative to USAS B31.7, 1968 (DRAFT) Edition.
The following table lists the principal plant design and operating characteristics. Data relate to 100 percent rated power (1500 MWt) unless specifically stated otherwise.
Table 1.2 Principal Plant Design and Operating Characteristics Plant Net Electrical Power Output, MWe 509.0 Gross Electrical Power Output, MWe 531.9 Electrical Power Output, MWe at 1500 MWt 533.7 Maximum Reactor Core Thermal Output, MWt 1500 Nominal Flow Rate, gpm 214,563 Maximum Core Inlet Temperature, °F (no uncertainties) 545 Maximum Core Outlet Temperature, °F (no uncertainties) 593 Maximum Operating Pressure, psia 2100 Design Pressure, psia 2500 Design Temperature, °F 650 Number of Fuel Assemblies 133 Number of Control Element Assemblies 49 Number of Loops 2 Number of Pumps 4 Steam Generators Number of Units 2 Nominal Total Steam Flow, lb/hr 6.62x106 Feedwater Temperature, Design °F 442.5 Steam Temperature, Design °F 522.6 Steam Quality, expressed as moisture content, max. 0.1 Shell Side Design Pressure, psia 1025 Shell Side Design Temperature, °F 560
USAR-Appendix G Information Use Page 16 of 82 Responses to 70 Criteria Rev. 24 FCS DESIGN CRITERIA*
CRITERION 9 - REACTOR COOLANT PRESSURE BOUNDARY The reactor coolant pressure boundary shall be designed and constructed so as to have an exceedingly low probability of gross rupture or significant leakage throughout its design lifetime.
This criterion is met. Reactor coolant system components are designed for a pressure of 2500 psia and a temperature of 650°F. The nominal operating conditions of 2100 psia and an average reactor coolant system temperature of 572.5°F permit an adequate margin for normal load changes and operating transients. The components are designed and constructed in accordance with the ASME Boiler and Pressure Vessel Code,Section III.
Codes and standards for components of the engineered safeguards systems are delineated in Criterion 1. Reactor coolant loop piping is designed in accordance with ANSI B 31.1 plus nuclear code cases. Other reactor coolant boundary piping is in accordance with the intent of ANSI Draft Code for Nuclear Piping B 31.7 of February 1968. Amendment No. XXX to Renewed Facility Operating License No. DPR-40 allows the design and/or analysis of non-RCS safety-related piping to be performed in accordance with ASME Section III, 1980 Edition (no Addenda) as an alternative to USAS B31.7, 1968 (DRAFT) Edition. Quality Control, inspection, and testing as required by these standards ensure the integrity of the reactor coolant system and are described in Appendix A and Section 4.5 of the USAR.
In addition to the code requirements listed, the reactor coolant loop piping is designed to meet the cyclic loading requirements and transient conditions stated for the reactor pressure vessel in Section 4.2.2 of the USAR. This piping is designed to withstand the dynamic seismic loadings for Class I structures under the rules listed in Section 2.0, Appendix F, of the USAR.
Also, cf. Criteria 33 to 36.
Updated Safety Analysis Report Fort Calhoun Station Classification of Structures and Equipment Appendix F R7 and Seismic Criteria Page 9 of 40 Table F "Loading Combinations and Primary Stress Limits" Primary Stress Limits Loading Combinations Vessels Piping (e) Supports (f)
- 1. Design Loading + PM < SM PM < 1.2Sh Working Design Earthquake Stress PB + PL < 1.5SM PB + PM < 1.2Sh Anchor Bolts F.S.>4.0 (d)
- 2. Normal Operating PM < SD PM < SD Within Loadings + Maximum Yield Hypothetical Earthquake PB < 1.5 [ 1-(PM )2 ] SD PB < 4 SD Cos PM Anchor Bolts
+ (Fluid Transient SD 2 SD F.S.> 2.0 (d)
Loadings (d)) (b) (c)
- 3. Normal Operating PM < SL PM < SL Deflection of supports Loadings + Pipe limited to Rupture + Maximum PB < 1.5 [1-(PM)2 ] SL PB < 4 SL Cos PM maintain supported Hypothetical Earth- SL 2 SL equipment within quake (b) (a), (c) limits shown
Updated Safety Analysis Report Fort Calhoun Station Classification of Structures and Equipment Appendix F R7 and Seismic Criteria Page 10 of 40 Table F-1 (Continued)
NOTES:
(a) These stress criteria are not applied to a piping run within which a pipe break is considered to have occurred.
(b) Loading combinations 2 and 3, stress limits for vessels, are also used in evaluating the effects of local loads imposed on vessels and/or piping, with the symbol PM changed to PL.
(c) The tabulated limits for piping are based on a minimum "shape factor". These limits are modified to incorporate the shape factor of the particular piping being analyzed.
(d) These load cases and limits apply only to the Pressurizer relief valve piping and supports.
(e) As an alternative to USAS B31.7, 1968, non-RCS, safety-related piping analysis may also be performed in accordance with ASME III, 1980 Edition (no Addenda). Material properties shall be from the original code of record (i.e., USAS B31.7, 1968). Associated stress limits shall be in accordance with ASME III, 1980 Edition (no Addenda) for Service Levels as shown below:
Load Combination Service Level A (Eqn. 8): Dead Weight Service Level A (Eqn. 10): Thermal Service Level A (Eqn. 11): Normal Loading: Dead Weight + Thermal Service Level B: Normal + Design Earthquake Service Level C: Normal + (Maximum Hypothetical Earthquake + Fluid Transient Loading) srss (d)
Service Level D: Normal + (Maximum Hypothetical Earthquake + Pipe Rupture) srss (f) Support analysis will continue to be performed in accordance with the existing licensing basis (i.e., Seventh Edition, AISC, American Institute of Steel Construction).