RS-17-082, Dresden Nuclear Power Station, Units 1, 2 & 3, Revision 12 to Updated Final Safety Analysis Report, Chapter 01, Introduction and General Description of Plant
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DRESDEN - UFSAR Rev 6 June 2005 1.3-1 1.3 COMPARISON TABLES Certain original design features of Dresden Units 2 and 3 are similar to those of other BWRs designed in the same time frame as Dresden, especially Quad Cities and other GE BWR/3-type plants. These similarities, in addition to subtle plant differences, are documented in the original FSAR and Amendments. A discussion of features developed by GE for use in the Dresden Station original design is provided in Section 1.2.5.
DRESDEN - UFSAR 1.4-1 1.4 IDENTIFICATION OF AGENTS AND CONTRACTORS
As owner, CECo engaged, or approved the engagement of, the contractors identified below in the construction of both units. However, irrespective of the explanation of contractual arrangements offered below, CECo was the sole applicant for the construction permit and operating license for both units, and as owner and applicant, is responsible for the design, construction, and operation of them.
Dresden Units 2 and 3 were designed and built by GE as prime contractor for CECo. General Electric Company engaged the architect-engineering services of Sargent and Lundy, Incorporated (S&L), Chicago, Illinois, to provide the design of the nonnuclear portions of the units and to prepare specifications for the purchase and construction thereof. Commonwealth Edison Company reviewed the designs and construction and purchase specifications prepared by S&L and GE to assure that the general plant arrangements, equipment, and operating provisions were satisfactory to it. The units were constructed under the general direction of GE throug h a construction management organization at the site, United Engineers and Constructors, Inc., utilizing appropriate construction, erection, and equipment subcontracts.
Preoperational testing of equipment and systems and initial operation were performed by CECo personnel under the technical direction of GE. Personnel provided by CECo for operation were drawn from the experienced operating staff of Dresden Unit 1, trained and qualified in the startup of this boiling water reactor, and had several years of operational experience. Startup testing is described in Chapter 14.
The units were turned over to CECo after a demonstration of unit operational capability at a specified output. CECo then assumed responsibility for their subsequent operation.
DRESDEN - UFSAR 1.4-2 ENDNOTES 1.4-1 FSAR Section 1.7.
1.4-2 FSAR Section 1.7; UFSAR Section 1.7.
DRESDEN - UFSAR Rev. 4 1.5-1 1.5 REQUIREMENTS FOR FURTHER TECHNICAL INFORMATION For a licensed operating facility such as Dresden Station, requirements for further technical information are regularly promulgated by the NRC at both the plant-specific and generic levels. Responses to these requests are documented in docketed correspondence to the NRC. The NRC-requested or EGC-initiated studies or analyses, to the extent they impact the plant design or safety analysis, are reflected in plant modifications, changes to procedures, and changes to the Technical Specifications, as appropriate. These results are documented in special or periodic submittals to the NRC and updates of the UFSAR.
DRESDEN - UFSAR 1.6-1 1.6 MATERIAL INCORPORATED BY REFERENCE Incorporated into the design of these units are features to improve both operational performance and overall safety which have been presented in special topical reports. These reports which have been provided to the NRC for review include those listed below:
A. APED 5286 - Design Basis for Critical Heat Flux in Boiling Wa ter Reactors (S eptember 1966)
B. APED 5446 - Control Rod Velo city Limiter (March 1967)
C. APED 5449 - Control Rod Worth Minimizer (March 1967)
D. APED 5450 - Design Provisions for In-Service Inspection (April 1967)
E. APED 5453 - Vibration Analysis and Testing of Reactor Internals (April 1967)
F. APED 5555 - Impact Testing on Collet Assembly for Control Rod Drive Mechanism 7 RDB144A (November 1967)
G. TR67SL211 An Analysis of Turbine Missiles Resulting from Last Stage Wheel Failure (October 1967)
H. APED 5608 - General Electric Company Analytical and Experiment al Program for Resolution of ACRS Safety Concerns (April 1968)
I. APED 5455 - The Mechanical Effects of Reactivity Transients (January 1968)
J. APED 5528 - Nuclea r Excursion Technology (August 1967)
K. APED 5448 - Analysis Methods of Hypothetical Super-Prompt Critical Reactivity Transients in Large Power Reactors (April 1968)
L. APED 5458 - Effectiveness of Core Standby Cooling Systems for General Electric Boiling Water Reactors (March 1968)
M. APED 5640 - Xenon Considerations in Design of Large Boiling Water Reactors (June 1968)
N. APED 5454 - Metal Water Reactions - Effects on Core Cooling and Containment (March 1968)
O. APED 5460 - Design and Performance of General Electric Boiling Water Reactor Jet Pumps (September 1968)
DRESDEN - UFSAR Rev. 4 1.7-1 1.7 DRAWINGS AND OTHER DETAILED INFORMATION A list of drawings provided to the AEC as part of the license application was not included in the FSAR and, therefore, has not been developed for this updated report.
Applicable drawings, pictures, plot and building plans, sketches, electrical diagrams and piping diagrams are included at the end of the sections in which they are referenced or at the end of the related sections in the case of duplicate drawing references. An equipment symbol chart which provides an explanation of the symbols used on the station piping and instrumentation drawings (P&IDs) is shown on Drawing M-11, Sheet 2. A complete P&ID index is provided in Drawing M-11, Sheet 1.
References on the figures contained in the UFSAR to ComEd, CECo, and Commonwealth Edison will be revised to reflect the change in facility ownership to EGC when other changes to that figure are needed.
DRESDEN - UFSAR Rev. 8 June 2009 REGULATORY GUIDE REFERENCE SECTIONS 1.8-1 1.8 CONFORMANCE TO NRC REGULATORY GUIDES Dresden was designed and partially constructed before the issuance of the first Regulatory Guides in 1970. During this time frame the NRC issued Safety Guides for utility guidance. Therefore, Dresden was not designed specifically to conform to Regulatory Guides. Conformance to the provisions of Regulatory Guides is generally indicated under two general categories, full compliance or compliance with intent or objectives of the Regulatory Guide via an alternate approach. Full compliance indicates that the provisions of the Regulatory Guides are met by direct conformance or by the assessed capability of the design.
In certain cases, CECo/EGC has assessed the design against a particular Regulatory Guide or specifically committed to the NRC to conform in part or in whole to a particular Regulatory Guide. Where appropriate, these Regulatory Guides are disc ussed in the applicable sections of the UFSAR. Table 1.8-1 provides a list of the Regulatory Guides and Safety Guides discussed and the sections in which they are discussed. This table is not a listing of Regulatory Guides that have been committed to by EGC.
DRESDEN - UFSAR Rev. 8 June 2009 (Sheet 1 of 4)
Table 1.8-1 REGULATORY GUIDE REFERENCE SECTIONS Commitment to or conformance with the identified Regulatory or Safety Guides is to the extent identified in the referenced UFSAR sections.
Regulatory Guide Title UFSAR Section(s) 1.3 Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss-of-Coolant Accident for Boiling Water Reactors 15.6 1.7 Control of Combustible Gas Concentrations in Containment Following a Loss-of-Coolant Accident 6.2 1.8 (Safety Guide 8, March 1971) Qualification and Training of Personnel for Nuclear Power Plants T.S. 5.3.1 (1) 1.21 Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes, Releases of Radioactive Materials in Liquid, and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants 11.2 1.23 Onsite Meteorological Programs 2.3 1.26 Quality Group Classifications and Standards for Water, Steam, and Radioactive Waste Containing Components of Nuclear Power Plants (for Comment) 5.2, 6.6 1.28, Rev. 3, August 1985 Quality Assurance Program Requirements - Design and Construction (1)
DRESDEN - UFSAR Rev. 8
June 2009 Table 1.8-1 (Continued)
REGULATORY GUIDE REFERENCE SECTIONS (Sheet 2 of 4)
Regulatory Guide Title UFSAR Section(s) 1.30 (Safety Guide 30, August 1972) Quality Assurance Program Requirements for the
Installation, Inspection, and Testing of Instrumentation and Electrical Equipment (1) 1.33 (Safety Guide 33, November 1972) Quality Assurance Program Requirements - Operation 13.5 1.34 Control of Electroslag Weld Properties 5.2, 5.3 1.36 Nonmetallic Thermal Insulation for Austenitic Stainless Steel 6.1 1.37, March 1973 Quality Assurance Requirements for Cleaning Fluid Systems and Associated Components of Water-Cooled Nuclear Power Plants (1) 1.38, March 1973 Quality Assurance Re quirements for Pack aging, Shipping, Receiving, Storage, and Handling of Items for Water-Cooled Nuclear Power Plants (1) 1.39, March 1973 Housekeeping Requirements for Water-Cooled Nuclear Power Plants (1) 1.44 Control of the Use of Sensitized Stainless Steel 5.3 1.45 Reactor Coolant Pressure Boundary Leakage Detection Systems 5.2 1.49 Power Levels of Nuclear Power Plants T.S. 1.1/2.1 bases DRESDEN - UFSAR Rev. 8 June 2009 Table 1.8-1 (Continued)
REGULATORY GUIDE REFERENCE SECTIONS (Sheet 3 of 4)
Regulatory Guide Title UFSAR Section(s) 1.50 Control of Preheat Temperature for Welding of Low-Allow Steel 5.3 1.52 Design, Testing, and Maintenance Criteria for Post-Accident Engineered Safety Feature Atmosphere Cleanup System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants 6.5 6.4 1.54, June 1973 Quality Assurance Requirements for Protective Coatings Applied to Water-Cooled Nuclear Power Plants (1) 1.61 Damping Values for Seismic Design of Nuclear Power Plants 3.9, 3.7 1.70, Rev 3, November 1978 Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, LWR Edition 1.1, 5.3, 12.2 1.75 Physical Independence of Electric Systems 7.5 1.77, May 1974 Assumptions Used for Evaluating a Cont rol Rod Ejection Accident for Pressurized Water Reactors 3.2, 4.3 1.78 Assumptions for Evaluating the Habitability of a Nuclear Power Plant Control Room During a Postulated Hazardous Chemical Release 6.4, 2.2 1.91 Evaluation of Explosions Postulated to Occur on Transportation Routes Near Nuclear Power Plants 2.2 1.97 Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident 7.1, 7.5, 9.1, 3.11
DRESDEN - UFSAR Rev. 8 June 2009 Table 1.8-1 (Continued)
REGULATORY GUIDE REFERENCE SECTIONS (Sheet 4 of 4) Regulatory Guide Title UFSAR Section(s) 1.99, Rev. 2, May 1988 Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials 5.2, 5.3 1.100 Seismic Qualification of Electric Equipment for Nuclear Power Plants 3.10 1.101, Rev. 2, October 1981 Emergency Planning and Preparedness for Nuclear Power Reactors 13.3 1.109 Calculation of Annual Doses to Man from Routine Re leases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I 11.3 ODCM 1.111 Methods for Estimating Atmosp heric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors 11.3 ODCM 1.113 Estimating Aquatic Dispersion of Effluent from Accidental and Routine Reactor Releases for the Purpose of Implementing, Appendix I ODCM 1.181 Content of the Updated Final Safety Analysis Report in Accordance with 10 CFR 50.71(e) 1.1 1.183 Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors 15.4.10, 15.6.4, 15.6.5, 15.7.3 1.190 Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence (Drafts were DG-1053 and DG-1025 (9/93))
5.3 4.8 Table 1, December 1975 Environmental Technical Specifications for Nuclear Power Plants T.S. 5.5 Notes: 1. These items are committed to in Topical Report NO-AA-10 for Dres den Station, but not specifically referenced in the text of the rebaselined UFSAR. Exceptions or alternatives identified in the UFSAR take precedence over commitments in the Topical Report.
DRESDEN - UFSAR 1.9-1 1.9 UNIT 2 SYSTEMATIC EVALUATION PROGRAM
1.9.1 Summary
The Systematic Evaluation Program (SEP) was initiated by the NRC to review the designs of older operating nuclear reactor plants to reconfirm and document their safety. The review provided an assessment of the significance of differences between current technical positions on safety issues and those that existed when a particular plant was licensed, a basis for deciding on how these differences should be resolved in an integrated plant review, and a documented evaluation of plant safety.
The results of the initial review were published as NUREG-0823, entitled, "Integrated Plant Safety Assessment Systematic Evaluation Program for Dresden Nuclear Power Station, Unit 2." This report was issued in February of 1983, and Supplement 1 to NUREG-0823 was issued in October of 1989.
The review compared the as-built design with current review criteria in 137 different areas defined as "topics." The "definition" and other information for each of these topics appear in Appendix A of NUREG-0823. During the review, 49 of the topics were deleted from consid eration by the SEP because a review was being made under other programs (Unresolved Safety Issue [USI] or Three Mile Island [TMI] Action Plan Tasks) or the topic was not applicable to the plant; that is, the topic was applicable to pressurized water reactors rather than to BWRs. The topics deleted because they were being reviewed under either the USI or TMI programs are listed in Appendix B of NUREG-0823, and the topics deleted because they did not apply to the plant are listed in Appendix C of NUREG-0823. The status of the USI or TMI tasks are addressed in a provisional operating license conversion safety evaluation report, NUREG-1403. That report was issued following completion of the SEP Integrated Plant Safety A ssessment Re port (IPSAR) and together with the IPSAR was considered during the conversion of the Dresden Unit 2 provisional operating license to a full-term operating license.
Of the original 137 topics, 88 were, therefore, reviewed for Dresden Unit 2; of those, 54 met current criteria or were acceptable on another defined basis. No modifications were made by CECo during topic review. References for correspondence pertaining to safety evaluation reports (SERs) for each of the 88 topics appear in Appendix E of NUREG-0823.
The review of the remaining 34 topics found that certain aspects of plant design differed from current criteria. The topics that differed from current licensing criteria consisted of 73 individual issues. These issues were considered in the integrated assessment of the plant, which consisted of evaluating the safety significance and other factors of the identified differences from current design criteria to arrive at decisions on whether backfitting was necessary from an overall plant safety viewpoint. To arrive at these decisions, engineering judgement was used as well as the results of a limited probabilistic risk assessment study. This study and staff comments are in Appendix D of NUREG-0823.
Table 4.1 of NUREG-0823 summarizes the staff's backfitting positions reached in the integrated assessment. In general, backfit requirements fell into one or more of the following categories:
DRESDEN - UFSAR 1.9-2 A. Equipment modification or addition; B. Procedure development or Technical Specification changes; C. Refined engineering analysis or continuation of ongoing evaluation; and
D. No backfit modifications necessary.
Eight issues required primarily equipment modification or addition, 17 issues required primarily procedure development or changes, and 23 issues required primarily refined engineering analysis or continuation of an ongoing evaluation. Twenty-five issues did not require any backfitting.
Safety improvements are being planned as a result of the integrated assessment and are listed below. Some safety improvements have already been implemented by the licensee. The following descriptions summarize the backfit actions addressed by the integrated assessment. The NUREG-0823 sections relating to the issue are given in parentheses.
1.9.2 Safety Improvements Agreed To and To Be Implemented by the Licensee As a Result of SEP
The safety improvements identified by SEP fall into three categories. The first category comprises hardware modifications or additions that CECo agreed to make and that are required by the NRC. The second category comprises procedural or Technical Specification changes that become part of the operating license. The third category comprises additional engineering analysis followed by corrective measures where required. These three categories are listed below, and the issues are discussed in the NUREG-0823 sections given in parentheses.
1.9.2.1 Category 1, Equipment Modifications or Additions Required by NRC A. Modify roof parapets to ensure ponded water is within roof load capacity (4.1.3);
B. Provide locking devices for manual isolation valves (4.18.3);
C. Provide second isolation valve on containment penetration branch lines (4.18.6);
D. Modify existing dc power system monitoring for breaker or fuse position and battery availability (4.23.3 and 4.28);
E. Install Class 1E protection at interface of reactor protection system and its power supply (4.24.3);
F. Modify diesel-generator annunciators (4.26.1); and
G. Provide for bypassing the diesel-generator underfrequency protective trip during accident conditions (4.26.2).
DRESDEN - UFSAR 1.9-3 1.9.2.2 Category 2, Technical Specification Changes and Procedure Development A. Modify existing flood emergency plan to provide ability to cope with design basis flood (4.1.2 and 4.1.4);
B. Modify the water control structures inspection program to ensure it is overseen by qualified personnel and that special inspections are conducted following extreme events (4.4.3);
C. Develop procedures for achieving cold shutdown from outside the control room (4.15 and 4.25.1);
D. Provide procedures for testing the shutdown cooling system temperature interlocks (4.17 and 4.25.4); E. Provide mechanical locking devices and administrative procedures to ensure valve closure (4.18.1);
F. Modify procedures for post-accident engineered safety features leakage (4.18.2);
G. Provide procedures to ensure disconnect links between redundant electrical divisions are open (4.21.2);
H. Provide assurance that tie breakers are not used during power operations (4.21.3);
I. Limit allowable time for obtaining DG 2/3 control power from Unit 3 (4.21.4);
J. Prohibit paralleling of shared dc systems during power operations (4.23.1);
K. Prohibit placing DG 2/3 switch in "bypass" during normal operation (4.23.2);
L. Revise procedures to achieve cold shutdown using safety-grade systems (4.25.2); and
M. Modify plant Technical Specification limits for primary coolant and iodine activity (4.31 and 4.32).
1.9.2.3 Category 3, Additional Engineering Evaluation A. Identify radiography requirements of vessels and pump casing (4.2.1);
B. Demonstrate fracture toughness for various components or that failure consequence is acceptable (4.2.2);
C. Ensure failure of ventilation stack does not affect safe shutdown (4.3.2);
DRESDEN - UFSAR 1.9-4 D. Identify and ensure components outside qualified structures can withstand tornado loading or that their loss does not affect safe shutdown (4.3.3);
E. Demonstrate failure of roof decks does not affect plant safety (4.3.4);
F. Demonstrate struct ural capability of plant to withstand load combinations (4.3.5 and 4.10);
G. Ensure operability of DG 2 and DG 2/3 following loss of ventilation systems resulting from tornado missiles (4.5.3);
H. Ensure capability to achi eve safe shutdown using tornado-missile-protected systems (4.5.4);
I. Provide schedule and basis for reinspection of low-pressure turbines (4.6);
J. Address effects of jet impingement on target pipe (4.7.1);
K. Demonstrate deformation of pipe associated with glob al strain does not affect functionability (4.7.2);
L. Ensure detectability for th rough-wall cracks in high-energy fluid systems piping (4.7.3);
M. Provide criteria and results of pipe whip load formulation and ensure pipe whip and jet impingement do not affect the containment liner (4.7.4);
N. Determine seismic capability of mechanical equipment (4.9.2);
O. Provide analysis of structural integrity of cable trays (4.9.3);
P. Ensure adequate setpoints for thermal overload protection of motor-operated valves or bypass thermal overloads (4.12.1);
Q. Provide leakage detection capability in conjunction with pipe breaks inside containment (4.13.1);
R. Provide seismically qualified leakage detection system (4.13.2);
S. Ensure adequacy of protective relaying (4.21.1);
T. Demonstrate adequate isolation of Class 1E sources from non-Class 1E loads (4.21.5);
U. Ensure common-mode electrical faults do not disable the neutron flux monitoring systems (4.24.1); and
V. Ensure the reactor protection system is protected from faults generated in process computer (4.24.2).