ML20211D275

Safety Evaluation Report Related to the Operation of Braidwood Station,Units 1 and 2.Docket Nos. 50-456 and 50-457.(Commonwealth Edison Company)

ML20211D275
Person / Time
Site:	Braidwood
Issue date:	09/30/1986
From:	Office of Nuclear Reactor Regulation
To:
References
NUREG-1002, NUREG-1002-S01, NUREG-1002-S1, NUDOCS 8610220157
Download: ML20211D275 (65)

Text

{{#Wiki_filter:_ NUREG-1002 Supplement No.1 Safety Evaluation Report related to the operation of Braidwood Station, Units 1 arid 2 Docket Nos. 50-456 and 50-457 Commonwealth' Edison Company l l. 1 U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation September 1986 a arooy A. E I sw=ws em E ppc

i f' NOTICE Availability of Reference Materials Cited in NRC Publications Most documents cited in NRC publications will be available from one of the following sources:

1. The NRC Public Document Room,1717 H Street, N.W.

Washington, DC 20555

2. The Superintendent of Documents, U.S. Government Printing Office, Post Office Box 37082, Washington, DC 20013-7082

3. The National Technical Information Service, Springfield, VA 22161 Although the listing that follows represents the majority of documents cited in NRC publications, it is not intended to be exhaustive.

Referenced documents available for inspection and copying for a fee from the NRC Public Docu-ment Room include NRC correspondence and internal NRC memoranda; NRC Office of Inspection and Enforcement bulletins, circulars, information notices, inspection and investigation notices; Licensee Event Reports; vendor reports and correspondence; Commission papers; and applicant and licensee documents and correspondence. The following documents in the NUREG series are available for purchase from the GPC Sales Program: formal NRC staff and contractor reports, NRC-sponsored conference proceedings, and NRC booklets and brochures. Also available are Regulatory Guides, NRC regulations in the Code of Federal Regulations, and Nuclear Regulatory Commission Issuances. Documents available from the National Technical Inforniation Service include NUREG series reports and technical reports prepared by other federal agencies and reports prepared by the Atomic Energy Commission, forerunner agency to the Nuclear Regulatory Commission. Documents available from public and special technical libraries include all open literature items, such as books, journal and periodical articles, and transactions. Federal Register notices, federal and state legislation, and congressional reports can usually be obtained from these libraries. Documents such as theses, dissertations, foreign reports and translations,and non-NRC conference proceedings are available for purchase from the organizat:on sponsoring the publication cited. Single copies of NRC draft reports are available free, to the extent of supply, upon written request to the Division of Technical Information and Document Control, U.S. Nuclear Regulatory Com-mission, Washington, DC 20555. Copies of industry codes and standards used in a substantive manner in the NRC regulatory process are maintained at the NRC Library,7920 Norfolk Avenue, Bethesda, Maryland, and are available there for reference use by the public. Codes and standards are usually copyrighted and may be purchased from the originating organization or, if they are American National Standards, from the American National Standards Institute,1430 Broadway, New York, NY 10018. l e

i NUREG-1002 Supplement No.1 Safety Evaluation Report related to the operation of Braidwood Station, Units 1 and 2 Docket Nos. 50-456 and 50-457 Commonwealth Edison Company U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation September 1986 ,s..., f ', 'i Y t

4 ABSTRACT In November 1983 the staff of the Nuclear Regulatory Commission issued its Safety Evaluation Report (NUREG-1002) regarding the application filed by the Commonwealth Edison Company, as applicant and owner, for a license to operate Braidwood Station, Units 1 and 2 (Docket Nos. STN 50-456 and STN 50-457). The facility is located in Reed Township, Will County, Illinois. This first supple-ment to NUREG-1002 reports the status of certain items-that remained unresolved at the time the Safety Evaluation Report was published. Braidwood SSER 1 iii

TABLE OF CONTENTS Page ABSTRACT..................................... iii 1 INTRODUCTION AND GENERAL DESCRIPTION OF FACILITY................. 1-1 1.1 Introduction................................................ 1-1

1. 7 Summa ry o f Outs ta ndi ng Items................................

1-2 1.8 Confirmatory Issues......................................... 1-3

1. 9 License Conditions..........................................

1-4 2 SITE CHARACTERISTICS............................................. 2-1 2.4 Hydrologic Engineering...................................... 2-1 2.4.3 Flooding Potential................................... 2-1 3 DESIGN CRITERIA FOR STRUCTURES, SYSTEMS, AND COMP 0NENTS.......... 3-1 3.5 Missile Protection.......................................... 3-1 3.5.1 Missile Selection and Description................... 3-1 3.9 Mechanical Systems and Components......................... 3-7 3.9.2 Dynamic Testing and Analysis of Systems, Components, and Equipment............................ 3-7 3.9.3 ASME Ccde Class 1, 2, and 3 Components, Component Supports, and Core Support Structures................ 3-7 4 REACTOR...................................................... 4-1 4.4 Thermal and Hydraulic Design............................... 4-1 4.4.1 Departure From Nucleate Boiling Methodology.......... 4-1 4.4.2 Fuel Rod Bowing.......... 4-3 4.4.7 Inadequate Core Cooling (ICC) Instrumentation........ 4-4 5 REACTOR COOLANT SYSTEM............... 5-1 5.2 Integrity of Reactor Coolant Pressure Boundary............. 5-1 5.2.4 Reactor Coolant Pressure Boundary Inservice Inspection and Testing.......... 5-1 5.3 Reactor Vessel......... 5-4 Braidwood SSER 1 v

TABLE OF CONTENTS (Continued) P.agg 5.3.1 Reactor Vessel and FCPB Materials.................... 5-4 5.3.2 Pressure-Temperature Limits.......................... 5-4 5.3.3 Reactor Vessel Integrity............................. 5-5 5.4 Component and Subsystem Design.............................. 5-5 5.4.2 Steam Generators..................................... 5-5 5.4.3 Residual Heat Removal System......................... 5-7 5.4.6 Seismic and Environmental Qualification of Pressurizer Power-0perated Relief Valves............. 5-8 6 ENGINEERED SAFETY FEATURES....................................... 6-1 6.2 Containment Systems......................................... 6-1 6.2.1 Containment Functional Design........................ 6-1 6.2.2 Containment Heat Removal Systems..................... 6-2 6.2.5 Combustible Gas Control System....................... 6-2 6.2.6 Containment Leakage Testing.......................... 6-3 6.2.7 Fracture Prevention of Containment Pressure Boundary............................................. 6-4 6.3 Emergency Core Cooling System............ 6-5 6.3.2 Evaluation of Single Failures........................ 6-5 6.3.4 Testing.............................................. 6-6 t l l 7 INSTRUMENTATION AND CONTR0L...................................... 7-1 7.2 Reactor Trip System......................................... 7-1 7.2.2 Specific Findings.................................... 7-1 7.3 Engineered Safety Features Systems.......................... 7-1 7.3.2 Specific Findings.................................... 7-1 7.6 Interlock Systems Important to Safety....................... 7-2 7.6.2 Specific Findings.................................... 7-2 8 ELECTRIC POWER SYSTEMS........................................... 8-1 8.2 Offsite Power System........................ 8-1 8.2.4 Adequacy of Station Electric Distribution System Voltages...................................... 8-1 8.4 Other Electrical Features and Requirements for Safety....... 8-1 Braidwood SSER 1 vi

TABLE OF CONTENTS (Continued) Page 8.4.4 Physical Identification and Independence of Redundant Safety-Related Electrical Systems.......... 8-1 9 AUXILIARY SYSTEMS................................................ 9-1 9.1 Fuel Storage and Handling................................... 9-1 9.1.5 Overhead Heavy Load Handling System.................. 9-1 9.2 Water Systems............................................... 9-2 9.2.2 Reactor Auxiliaries Cooling Water Systems............ 9-2 9.3 Process Auxiliaries......................................... 9-2 9.3.2 Process and Postaccident Sampling System............. 9-2 9.3.5 III.D.1.1 Integrity of Systems Outside Containment Likely To Contain Radioactive Material............... 9-3 9.5 Other Auxiliary Systems..................................... 9-7 9.5.4 Emergency Diesel Engine Fuel Oil Stora Tra ns fe r Sys tem....................... ge and 9-7 10 STEAM AND POWER CONVERSION SYSTEM................................ 10-1 10.3 Main Steam Supply System........................ 10-1 10.3.3 Secondary Water Chemistry.......................... 10-1 12 RADIATION PROTECTION........ 12-1 12.5 Operational Radiation Protection Program................... 12-1 12.5.1 Organization................ 12-1 13 CONDUCT OF 0PERATIONS............................................ 13-1 13.3 Emergency Planning........................ 13-1 13.3.1 Background......................................... 13-1 13.3.2 Evaluation of the Emergency (0nsite) Plan.......... 13-1 13.3.3 Offsite Emergency Planning Medical Services........ 13-5 13.3.4 Emergency Preparedness Exercise.................... 13-6 13.3.5 FEMA Finding on Offsite Pre Conclusion.................paredness........... 13-6 13.3.6 13-6 13.4 Review and Audit. 13-7 13.6 Physical Security....................................... 13-9 Braidwood SSER 1 vii

TABLE OF CONTENTS (Continued) Page 13A-1 Appendix 13A........................... 15-1 15 ACCIDENT ANALYSES.................................. 15.2 Normal Operation and Operational Transients................ 15-1 15.2.4 Change in Core Reactivity Transients................ 15-1 15-1 15.3 Design-Basis Accidents.............. 15.3.6 Reacter Coolant Pump Rotor Seizure and Shaft 15-1 Break.................. 15.4 Radiological Consequences of Accidents...................... 15-2 15.4.3 Steam Generator Tube Fail ure........................ 15-2 APPENDIX A CONTINUATION OF CHRON0 LOGY OF NRC STAFF RADIOLOGICAL SAFETY REVIEW 0F BRAIDWOOD STATION, UNITS 1 AND 2 APPENDIX B BIBLIOGRAPHY APPENDIX F NRC STAFF CONTRIBUTORS AND CONSULTANTS APPENDIX H ACRS REPORT ON BRAIDWOOD STATION, UNITS 1 AND 2 APPENDIX I ERRATA TO BRAIDWOOD STATION, UNITS 1 AND 2, SAFETY EVALUATION REPORT APPENDIX J TECHNICAL EVALUATION REPORT ON CONTROL OF HEAVY LOADS AT NUCLEAR POWER PLANTS FOR BRAIDWOOD STATION, UNITS 1 AND 2 Braidwood SSER 1 viii

1 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT 1.1 Introduction In November 1983, the Nuclear Regulatory Commission (NRC) staff issued its Safety Evaluation Report (SER) (NUREG-1002) on the application filed by the Commonwealth Edison Company, as applicant and owner, for a license to operate Braidwood Station, Units 1 and 2 (Docket Nos. STN 50-456 and STN 50-457). At that time, the staff identified items that had not been resolved with the appli-cant. The purpose of this supplement to the SER is to provide the staff evalua-tion 6f the open items that have been resolved, address changes to the SER which resulted from the receipt of additional information from the applicant, and present the comments made by the Advisory Committee on Reactor Safeguards (ACRS) in its letter dated February 11, 1985 (Appendix H). During its 298th meeting on February 7-9, 1985, the Advisory Committee on Reac-tor Safeguards reviewed the Braidwood application. In a February 11, 1985, letter from Chairman David A. Ward to NRC Chairman Nunzio J. Palladino, the Committee concluded that, subject to the resolution of open items identified by the NRC staff and subject to satisfactory completion of construction, staffing, and preoperational testing, there is reasonable assurance that Braidwood Units 1 and 2 can be operated at power levels up to 3425 MWt without undue risk to the health and safety of the public. Each of the following sections or appendices is numbered the same as the corres-ponding SER section or appendix that is being updated. Each section is supple-mentary to and not in lieu of the" discussion in the SER unless otherwise noted. Appendix A continues the chronology of the staff's actions related to the proc-essing of the application for Braidwood Units 1 and 2. Appendix B lists refer-ences cited in this report.* Appendix F lists principal staff members who con- .tributed to this supplement. Appendix H consists of a copy of the letter from the ACRS on Braidwood Units 1 and 2. Appendix I contains errata to the SER. Appendix J contains the Technical Evaluation Report on Control of Heavy Loads at Nuclear Power Plants for Braidwood Station, Units 1 and 2, prepared for the NRC by EG&G Idaho, Inc. Copies of this SER supplement are available for inspection at the NRC Public Document Room, 1717 H Street, N.W., Washington, D.C., and at the Wilmington Township Public Library, 201 South Kankakee Street, Wilmington, Illinteis 60481.

• Availability of all material cited is described on the inside front cover of this report.

Braidwood SSER 1 1-1

The NRC Project Manager for Braidwood Station, Units 1 and 2, is Ms. Janice A. Stevens. Ms. Stevens may be contacted by calling (301) 492-7702 or writing: Janice A. Stevens Division of PWR Licensing-A U.S. Nuclear Regulatory Commissivu Washington, D.C. 20555

1. 7 Summary of Outstanding Items The current status of the outstanding items listed in the original SER follows:

Part A Items Status Section (1) Pump and valve operability Open 3.9.3.2* (2) Seismic and dynamic qualification of equipment Open 3.10* (3) Environmental qualification of electrical and mechanical equipment Open 3.11* (4) Containment pressure boundary components Closed in this 6.2.7 supplement (5) Organizational structure Closed in this 13.1, 13.4 supplement (6) Emergency preparedness plans and facilities Closed in this 13.3* supplement (7) Procedures generation package (PGP) Open 13.5.2 (8) Control room human factors review Open 18.0* Part B Items Status Section (1) Turbine missile evaluation Closed in this

3. 5.1. 3 supplement l

(2) Improved thermal design procedures Closed in this 4.4.1 supplement (3) TMI Action Item II.F.2: Inadequate Core Closed in this 4.4.7 Cooling Instrumentation supplement (4) Steam generator flow-induced vibrations Closed in this 5.4.2 supplement

• This section includes both site-specific-related information and duplicate plant design features.

Braidwood SSER 1 1-2

Part B Items Status Section (5) Conformance of ESF filter system to RG 1.52 Open 6.5.1 (6) Fire protection program Open 9.5.1 (7) Volume reduction system Open 11.1, 11.4.2 1.8 Confirmatory Issues The current status of the confirmatory issues follows: Part A Items Status Section (1) Applicant compliance with the Commission's Open 1.1, 3.l* regulations (2) Site drainage Closed in this 2.4.3.3 supplement (3) Piping vibration test program Closed in this 3.9.2.l* supplement (4) Preservice Inspection Program Open 5.2.4, 6.6* (5) Reactor vessel materials Closed in this 5.3 supplement (6) Electrical distribution system voltage Closed in this 8.2.4* verification supplement (7) Independence of redundant electrical safety Closed in this 8.4.4 equipment supplement (8) RPM qualifications Closed in this 12.5 supplenient (9) Revision to Physical Security Plan Closed in this 13.6 supplement Part B Items Status Section (1) Inservice testing of pumps and valves Open 3.9.6 (2) Steam generator tube surveillance Closed in this 5.4.2.2 supplement (3) Charging pump deadheading Closed in this 6.3.2, 7.3.2 supplement "This section includes both site-specific-related information and duplicate plant design features. Braidwood SSER 1 1-3

Part B Items Status Section (4) Minimum containment pressure analysis for Closed in this 6.2.1.5 performance capabilities of ECCS supplement (5) Containment sump screen Closed in this 6.2.2 supplement (6) Containment leakage testing vent and drain Closed in this 6.2.6 provisions supplement (7) Confirmatory test for sump design Closed in this 6.3.4.1 supplement (8) IE Bulletin 80-06 Closed in this 7.3.2.2 supplement (9) Remote shutdown capability Open 7.4.2.2 (10) TMI Action Plan Item II.D.1 Partial clo-3.9.3.3, sure in this 5.2.2 supplement TMI Action Plan Item II.K.3.1 Closed in this 7.6.2.7 supplement TMI Action Plan Item III.D.1.1 Closed in this 9.3.5 supplement (11) SWS process control program Open 11.4.2 (12) Noble gas monitor Open 11.5.2 (13) RCP rotor seizure an ' shaft break. Closed in this 15.3.6 supplement (14) Anticipated transients without scram (ATWS) Open 15.6 (15) Evaluation of compliance with Opened in this 5.2.4.4 10 CFR 50.55a(a)(3) supplement (16) Steam generator tube failure Opened in this 15.4.3 supplement 1.9 License Conditions The current status of the license conditions follows: Braidwood SSER 1 1-4

Part A Items Status Section (1) Inservice inspection program Open 5.2.4, 6.6* (2) Natural circulation testing Closed in this 5.4.3* supplement (3) Response time testing Closed in this 7.2.2.5* supplement (4) Steam valve inservice inspection Closed in this 10.2* supplement (5) Implementation of secondary water chemistry Closed in this 10.3.3* monitoring and control program as proposed supplement by the Byron /Braidwood FSAR Part B Items Status Section (1) Masonry walls Open 3.8.3 (2) TMI Item II.B.3 postaccident sampling Closed in this 9.3.2 supplement (3) Fire Protection Program Open 9.5.1 (4) Emergency diesel engine auxiliary support Opened in this 9.5.4.1 systems supplement

• This section includes both site-specific-related information and duplicate plant design features.

Braidwood SSER 1 1-5

2 SITE CHARACTERISTICS 2.4 Hydrologic Engineering 2.4.3 Flooding Potential 2.4.3.3 Local Probable Maximum Precipitation in Plant Area The design-basis flood level for local intense precipitation, at elevation 601.35 feet ms1 is 0.35 foot above the grade floor elevation of 601.0 feet ms1. The applicant proposed reinforced concrete curbs to elevation 601.4 feet msl for all exterior grade level doors which was acceptable to the staff.

However, there are two entrances to the auxiliary building from the turbine building at the 601 level where curbs were not feasible.

The provisions for protecting these entrances had not been submitted for staff review so this was left as a confirmatory issue in the SER. The applicant has subsequently proposed, by letter dated February 22, 1984 (T. R. Tramm to H. R. Denton), a resolution that is acceptable to the staff. The applicant has stated that the two doors between the auxiliary and turbine buildings will be normally closed and secured as part n' the security plans. These are metal doors with thresholds and they open tow <rd the turbine building. The staff agrees with the applicant's contention that it is very unlikely that exterior floodwater would reach these doors. However, if floodwater should reach the doors, their physical characteristics and the security requirement to keep them closed will ensure that there would be only minor leakage that could be accommodated by the auxiliary building floor drain system or maintenance crew. The staff thus finds that the applicant's provision for flood protection from local intense precipitation is acceptable and meets the intent of 10 CfR Part 50, Appendix A, GDC 2. Therefore, Confirmatory Issue ti(2) is closed. Braidwood SSER 1 2-1

3 DESIGN CRITERIA FOR STRUCTURES, SYSTEMS, AND COMPONENTS 3.5 Missile Protection 3.5.1 Missile Selection and Description 3.5.1.3 Turbine n ssiles 3.5.1.3.1 Review Basis (1) Introduction During the past several years, the results of turbine inspections at operating nuclear facilities indicate that cracking to various degrees has occurred at the inner radius of turbine disks, particularly those of Westinghouse design. Within this time period, a Westinghouse turbine disk has failed at one facility owned by the Yankee Atomic Electric Company. Furthermore, recent inspections of General Electric turbines have also revealed disk keyway cracks. Stress corrosion has been identified by both manufacturers as the operative cracking mechanism. The staff has followed these developments closely. Thq staff's primary safety objective is to prevent unacceptable doses of radioactivity to the public from releases of radioactive contaminants that could be caused by damage to plant safety-related structures, systems, and components from missile generating tur-bine failures. On tne basis of previous staff reviews and various estimates by others (see Bush, 1973; and Twisdale, Dunn, and Frank, 1982) for a variety of plant layouts, the staff concludes that if a turbine missile is generated the probability of unacceptable damage to safety related structures, systems, and components is in the neighborhood of 10 a or 10 2 per year depending on whether the turbine orientation is favorable or unfavorable. In view of this and opera-ting experience, the staff has shifted the review emphasis to preventing missile-generating turbine failures. In keeping with this shift of emphasis, the staff has recently described turbine missile generation probability guidelines for determining turbine disk ultrasonic inservice inspection frequencies, and tur-bine control and overspeed protection systems maintenance and testing schedules. No change in safety criteria is associated with this change in review emphasis. The major domestic turbine manufacturers are already in the process of estab-lishing models and methods for calculating turbine missile generation probabili-ties for their respective turbine generator systems. This shift of emphasis helps improve turbine generator system reliability by focusing on review and evaluation of the probability of missile generating tur-bine failure, and in the process provides a logically consistent method for establishing inservice inspection and testing schedules. Furthermore, it re-duces considerably the analytical burden placed on licensees by eliminating the need for elaborate and ambiguous analyses of strike and damage probabilities, and at the same time increases the protection of public health and safety by improud maintenance of turbine system integrity. Braidwood SSER 1 3-1

(2) Criteria That Must Be Met To Demonstrate Compliance With Regulations In accordance with General Design Criterion (GDC) 4 of Appendix A to 10 CFR Part 50, nuclear power plant structures, systems, and components important to safety shall be appropriately protected against dynamic effects, including the effects of missiles. Failures of large steam turbines of the main turbine gen-erator have the potential for ejecting large high-energy missiles that can dam-age plant structures, systems, and components. The overall safety objective of the staff is to ensure that structures, systems, and components important to safety are adequately protected from potential turbine missiles. Of those sys-tems important to safety, this topic is primarily concerned with safety related systems; i.e., those structures, systems, or components necessary to perform required safety functions and to ensure (a) the integrity of the reactor coolant pressure boundary (b) the capability to shut down the reactor and maintain it in a safe shutdown condition (c) the capability to prevent accidents that could result in potential offsite exposures that are a significant fraction of the guideline exposures of 10 CFR Part 100, " Reactor Site Criteria" Typical safety related systems are listed in Regulatory Guide (RG) 1.17. The probability of unacceptable damage from turbine missiles (P ) is generally 4 expressed as the product of the probability of (1) turbine failure resulting in the ejection of turbine disk (or internal structure) fragments through the tur-bine casing (P ); (2) ejected missiles perforating intervening barriers and 3 striking safety-related structures, systems, or components (P ); and (3) struck 2 structures, systems, or components failing to perform their safety function (P ). 3 According to NRC guidelines stated in Section 2.2.3 of the Standard Review Plan (SRP) (NUREG-0800) and RG 1.115, the probability of unacceptable damage from turbine missiles should be less than or equal te about 1 chance in 10 million _ 10 7 per year. per year for an individual plant, i.e., P 4 (3) Past Procedures for Demonstrating Compliance With Regulations In the past, analyses for construction perm't (CP) and operating license (OL) reviews assumed the probability of missile generation (P ) to be approximately 1 10 4 per turbine year, based on the historical failure rate (see Bush,1973). The strike probability (P ) was estimated in SRP Section 3.5.1.3 based on postu-2 lated missile sizes, shapes, and energies, and on available plant-specific infor-mation such as turbine placement and orientation, number and type of intervening barriers, target geometry, and potential missile trajectories. The damage prob-ability (Pa) was generally assumed to be 1.0. The overall probability of unac-ceptable damage to safety-related systems (P ), which is the sum over all targets 4 of the product of these probabilities, was then evaluated for compliance with the NRC safety objective. This logic places the regulatory emphasis on the strike probability, i.e., having established an individual plant safety objec-tive of about 10 7 per year, or less, for the probability of unacceptable damage to safety-related systems resulting from turbine missiles, this procedure re-be less than or equal to 10 3 quires that P2 Braidwood SSER 1 3-2

It is well known that nuclear turbine disks crack (see NUREG/CR-1844 and PNO-III-81-04 on Monticello, November 24, 1981), turbine stop and control valves fail (see LER 82-132, Docket No. 50-361, November 19, 1982; and Burns, 1977), and disk ruptures can result in the generation of high-energy missiles (Kalderon, 1972). Furthermore, analyses (NUREG/CR-1884, and Clark, Seth, and Shaffer, 1981) clearly demonstrate the large effects of inservice testing and inspection frequencies on the probabilities of missile generation (P ). It is the staff's view that sufficiently frequent turbine testing and inspection are the best 3 means of ensuring that the criteria on the probability on unacceptable damage to safety related structures, systems, and components (P ) presented in Section 3.5.1.3.1(2) is met. 4 Therefore, it is prudent for turbine manufacturers to perform, and the staff to review, analyses of turbine reliability, which include known and likely failure mechanisms, expressed as a function of time (i.e., inservice inspection or test intervals). While the calculation of strike probability is not difficult in principle, for the most part reducing to a straightforward ballistics analysis, it presents a problem in practice. The problem stems from the fact that numerous modeling approximations and simplifying assumptions are required to make tractable the incorporation into acceptable models of available data on the (1) properties of missiles, (2) interactions of missiles with barriers and obstacles, (3) tra-jectories of missiles as they interact with and perforate (or are deflected by) barriers, and (4) identification and location of safety-related targets. The particular approximations and assumptions made tend to have a large effect on the resulting value of P. Similarly, a reasonably accurate specification of 2 the damage probability (P ) is not a simple matter because it is so difficult 3 to define the missile impact energy required to render given safety-related systems unavailable to perform their safety function, and it is so difficult to postulate sequences of events that would follow a missile producing turbine failure. (4) New Procedure for Demonstrating Compliance With Regulations The new approach places on the applicant the responsibility for demonstrating and maintaining an NRC-specified turbine reliability by appropriate inservice inspection and testing throughout plant life. This shift of emphasis necessi-tates that the applicent show capability to have volumetric (ultrasonic) exami-nations performed which are suitable for inservice inspection of turbine disks and shaf t, and to provide reports for staff review and approval which describe their methods for determining probabilities that turbine missiles will be generated. Westinghouse, on behalf of applicants, has prepared a report (see Appendix C of the Byron /Braidwood FSAR) for staff review and approval that describes methods for determinir.g probabilities that turbine missiles will be generated for their respective turbines. The design-speed missile generation probability is to be related to disk design parameters, material properties, and the inservice volu-metric (ultrasonic) disk inspection interval (for example, see Clark, 1981). The destructive overspeed missile generation probability is to be related to the turbine governor and overspeed protection system's speed sensing and trip-ping characteristics, the design and arrangement of main steam control and stop valves and the reheat steam intercept and stop valves, and the inservice testing and inspection intervals for systems components and valves (for example, see Burns, 1977). The manufacturer has provided applicants and licersees with tables Braidwood SSER 1 3-3

of missile generation probabilities versus time (inservice volumetric disk in-spection interval for design-speed failure, and inservice valve testing interval for destructive overspeed failure) for its particular turbine, which are then to be used to establish inspection and test schedules that meet NRC safety objectives. (see Section 3.5.1.3.1(3) Because of the uncertainties involved in calculating P2 analyses are " ball park" or of this supplement), the staff concludes that P2 " order of magnitude" type calculations only. On the basis of simple estimates f for a variety of plant layouts (for examples see Bush, 1973, and Twisdale, Dunn, and Frank, 1982), the staff further concludes that the strike and damage proba-bility product can be reasonably taken to fall in a characteristic narrow range which is dependent on the gross features of turbine generator orientation, for favorably oriented turbine generators P Pa tends to lie in the range 10 4 to 2 10 3, and for unfavorably oriented turbine generators P P tend to lie in the 2 3 range 10 3 to 10 2 For these reasons (and because of weak data, controversial assumptions, and modeling difficulties), in the evaluation of P, the staff 4 gives credit for the product of the strike and dama0e probabilities of 10 3 for a favorably oriented turbine and 10 2 for an unfavorably oriented turbine, and does not encourage calculations of them. These values represent the staff's Pa lies based on calculations done by the staff and the opinion of where P2 results of calculations done by others. It is the staff's view that the NRC safety objective with regard to turbine missiles is best expressed in terms of two sets of criteria applied to the missile generation probability (see Table 3.1). One set of criteria is to be applied to favorably oriented turbines, and the other is to be applied to un-favorably oriented turbines. Applicants and licensees are expected to meet the set of criteria appropriate to their turbine orientation (as shown in Table 3.1) if the staff has approved the manufacturers' methods and procedures for calcu-lating the probabilities of turbine missile generation. 3.5.1.3.2 Evaluation For Braidwood Station, Units 1 and 2, the steam and power conversion system generates steam in a direct-cycle pressurized-water reactor (PWR) and converts The it to electric power in a turbine generator manufactured by Westinghouse. placement and orientation of the turbine generator is unfavorable with respect to the station reactor buildings; that is, there are safety-related targets in-side the low trajectory missile strike zone. The turbine is a tandem-compound type (single shaft) with one double-flow high pressure turbine, three double-flow low pressure turbines, and a rated rotational speed of 1800 rpm. (1) Destructive Overspeed Failure Prevention In accordance with the requirements of GDC 4, the turbine generator has an overspeed protection system designed to control turbine action under all nor-mal or abnormal conditions and to ensure that a turbine trip from full load will not cause the turbine to overspeed beyond acceptable limits; this mini-mizes the probability of generating turbine missiles. The turbine control and overspeed protection system is, therefore, essential to the safe operation of the plant. Braidwood SSER 1 3-4

The Westinghouse turbine is equipped with a digital electrohydraulic (DEH) con-trol system consisting of a solid-state electronic controller and a highpressure, fire resistant fluid supply used for control of turbine valve operators. The controller compares signals representing turbine speed and firststage pressure with reference values initiated by a load demand signal. The controller then puts out a comparison signal which actuates hydraulic control of the main tur-bine governor and reheat steam interceptor valves, to match generator output to load demand. The turbine governor valve and reheat steam interceptor valves are preceded by main turbine stop-throttle and reheat steam stop valves, respectively. The principal function of these latter valves is to shut off the steam supply to the turbine in the event of a turbine trip. The control system for turbine governor valves includes three separate speed sensors mounted on the turbine as follows: mechanical overspeed trip weight (spring-loaded bolt) electromagnetic pickup for DEH main speed control channel electromagnetic pickup for emergency trip system at turning gear location The following signals act on each of the main turbine valves in case the turbine speed exceeds the specified limit: (a) Main Turbine - Stop Throttle and Reheat Steam Stop Valves Should the turbine exceed approximately 108% of rated speed, these valves will be tripped closed by both (i) the mechanical overspeed trip weight and (ii) a redundant electrical trip from the emergency trip system. (b) Main Steam Control Valves (i) The main speed channel continuously calls for rated speed. When the main power transformer breakers open, the turbine speed tends to rise above rated speed. The DEH system has an anticipatory feature that, upon breaker opening, compares actual speed with rated speed. If cctual speed exceeds rated speed with the breaker open, the speed channel calls for closing of the control valves. (ii) If the unit carried greater than 30% load and the main breaker opens, the control and interceptor valves will be closed by energizing the overspeed protection controller solenoids, which causes high pres-sure fluid associated with the control valves to be dumped. (iii) The overspeed protective controller calls for fully clored control valves at 103% of rated speed. (iv) Should the speed exceed 108% of rated speed, the control valves are tripped closed by both the mechanical overspeed trip weight and a backup electrical trip. Braidwood SSER 1 3-5 m

(c) Reheat Steam Interceptor Valve (i) If the unit carries greater than 30% load and the generator breaker opens, the control and interceptor valves will be closed by dumping emergency high pressure oil associated with the interceptor valve. (ii) The overspeed protective controller calls for fully closed intercep-tor valves at 103% of rated speed. (iii) Should the speed exceed 108% of rated speed, the interceptor valves are tripped closed by both the mechanical overspeed trip weight and redundant electrical trip from the emergency trip system. The trip design philosophy is as follows: Close the turbine governor valves on an overspeed condition (103%) to pre-vent reaching the overspeed trip setting (108%). Actuate the overspeed trip (108%) (one mechanical and one electrical emer-gency trip) to prevent maximum turbine overspeed from exceeding 120%. According to the applicant's inservice inspection and testing program, a sched-ule of valve inspection at periodic intervals for throttle, governor, reheat stop, and interceptor valves will be implemented after initial turbine startup. A functional test of the turbine steam inlet valves is performed periodically while the unit is carrying load. These tests ensure proper operation of throt-tie, governor, reheat stop, and interceptor valves. These valves are observed during the tests for smoothness of movement. (2) Preventing Design-Speed Failure Failures of turbine disks at or below the design speed, nominally 120% of nor-mal operating speed, are caused by a non-ductile raterial failure at nominal stresses lower than the yield stress of the material. Since 1979, the staff has known of the stress corrosion cracking problems in low pressure rotor disks of Westinghouse turbines. Westinghouse has developed and implemented procedures for inservice volumetric inspection of the bore and keyway areas of low pressure turbine disks. Westinghouse has also prepared reports, submitted for NRC review, which describe Westinghouse methods for determining turbine disk inspection in-tervals and relating them to missile generation probabilities that result from stress corrosion cracking. 3.5.1.3.3 Summary The staff has.eviewed the information on Braidwood Station, Units 1 and 2, with regard to the turbine missile issue and concludes that the probability of unac-ceptable damage to safety-related structures, systems, and components from tur-bine missiles is acceptably low (i.e., less than 10 7 per year) provided that the total turbine missile generation probability for each plant is such that conformance with the criteria presented in Table 3.1 is maintained throughout In reaching the life of the plant by acceptable inspection and test programs. this conclusion, the staff has considered the unfavorable orientation of the turbine generators. Braidwood SSER 1 3-6

By letter dated September 26, 1984, the applicant committed to an inspection program based on the manufacturer's recommendations. The staff has reviewed and apprcved the manufacturer's generic turbine integrity methodology which provides procedures for estimating crack growth, missile generation probability and volumetric inspection intervals. Based on the manufacturer's recommenda-tions, the applicant is required to volumetrically inspect all low pressure turbine rotors every third refueling outage. In addition, an acceptable tur-bine valve inspection program has been incorporated into Section 4.3.4.2 of the Technical Specifications. Therefore, License Condition A(4) is no longer necessary. The staff concludes that the turbine missile risLs for the proposed plant design are in compliance with the requirements of GDC 4 and are acceptable. Thus, Out-standing Item B(1) is considered closed. 3.9 Mechanical Systems and Components 3.9.2 Dynamic Testing and Analysis of Systems, Components, and Equipment 3.9.2.1 Preoperational Vibration and Dynamic Effects of Testing on Piping The staff reviewed the applicant's detailed progro for pipe vibration docu-mented in Preoperational Test No. BWPT-EM-12, Revision 1 (see Inspection Report No. 50-456/86-32 dated August 14, 1986). Specified piping was monitored by instrumentation and visually inspected by the applicant during preoperational hot functional testing conducted during March 1986. The following systems were inspected for vibrations: reactor coolant system component cooling system chemical and volume control system (safe shutdown portion) residual heat removal system safety injection system essential service water system contairnient spray system (except spray headers) chilled water (control room) system fuel pool cooling and cleanup system reactor coolant pressurizer system On the basis of a review of the applicant's program', the staff has concluded that the preoperational hot functional tests have demonstrated that piping vibrations in the systems inspected are within acceptable limits and that the piping can expand thermally in a manner consistent with the intent of the de-sign. Therefore, Confirmatory Issue A(3) is considered closed. 3.9.3 ASME Code Class 1, 2, and 3 Components, Component Supports, and Core Support Structures 3.9.3.3 Design and Installation of Pressure Relief Devices As required by TMI Action Plan Item II.D.1, all PWR plant licensees and appli-cants are required to demonstrate that their pressurizer safety valves (SVs), power-operated relief valves (PORVs), PORV block valves, and all associated Braidwood SSER 1 3-7

discharge piping will function adequately under conditions predicted for design-basis transients and accidents. In response to this requirement, the Electric Power Research Institute (EPRI), on behalf of the PWR Owners Group, has com-pleted a full-scale valve testing program and the Owners Group has submitted these test results to the NRC (letter from 0. Kingsley (Alabama Power) to S. Chilk (NRC) dated July 27, 1982, transmitting WCAP-10105, a report performed for the Westinghouse Owners Group). Additionally, each PWR plant applicant for an OL was required to submit a report by fuel load time which would demonstrate the operability of these valves and the associated piping. The applicant responded to this requirement with submittals dated July 1, 1982, and October 26, 1982, that contain information from the EPRI valve test program results which apply to Braidwood Units 1 and 2. A December 30, 1983, submittal also states that the safety and relief valve discharge piping and supports are being modified to ensure functionability. The staff has not completed a detailed review of the applicant's submittals; however, on the basis of a preliminary review, the staff finds that the general approach of using the EPRI test results to demonstrate operability of the safety valves, PORVs, and PORV block valves is acceptable. The applicant's submittals note that Braidwood uses safety valves, PORVs, and PORV block valves similar to valves that performed satisfactorily for test sequences that bound conditions the valve could be exposed to. In summary, on the basis of preliminary review, the staff has concluded that the applicant's general approach to responding to this item is acceptable and provides adequate assurance that reactor coolant system overpressure protection systems at Braidwood can adequately perform their intended functions. If the detailed review reveals that modifications or adjustments to safety valves, PORVs, PORV block valves, or associated piping are needed to ensure that all intended design margins are present, the staff will require that the applicant make appropriate modifications. This is a Confirmatory Issue. Braidwood SSER 1 3-8

g Table 3.1 Reliability criteria { Probability, yr 1 Favorably Unfavorably 5 oriented oriented Required licensee action w A. P < 10 4 P < 10 5 This is the general, minimum reliability re-1 3 quirement for loading the turbine and bringing the system on line. B. 10 4 < P < 10 3 10 5 < P < 10 4 If during operation this condition is reached, 3 t the turbine may be kept in service until the next scheduled outage, at which time the li-censee is to take action to reduce P to meet 1 the appropriate A criterion (above) before returning the turbine to service. C. 10 3 < P < 10 2 10 4 < P < 10 3 If during operation this condition is reached, 3 3 the turbine is to be isolated from the steam supply within 60 days, at which time the licen-see is to take action to reduce P to meet the t appropriate A criterion (above) before return-ing the turbine to service. D. 10 2 < P 10 3 < P If at any time during operation this condition 3 1 is reached, the turbine is to be isolated from the steam supply within 6 days, at which time the licensee is to take action to reduce P 2 to meet the appropriate A criterion (above) before returning the turbine to service.

4 REACTOR 4.4 Thermal and Hydraulic Design 4.4.1 Departure From Nucleate Boiling Methodology In the SER, the staff stated that the thermal-hydraulic design methodology using the Westinghouse improved thermal design procedure (ITDP) is acceptable; however, the acceptability of the Braidwood Station design departure from nucle-ate boiling ratio (DNBR) limits required further plant-specific review regard-ing the uncertainties, variances, and distributions of the pertinent parameters used in the ITDP. This SER supplement addresses the staff findings resulting from the Braidwood plant specific review. The applicant, by letter dated May 5, 1982, submi,tted a Westinghouse response to a staff question regarding the ITDP parameter uncertainty distributions and their effects on the design DNBR limit. This response also references a Westing-house generic report which provides a detailed breakdown of the instrumentation error components and a description of the statistical method used in combining these instrumentation errors to determine the measurement uncertainties of the pertinent ITDP parameters such as pressurizer pressure, reactor coolant tempera-ture, reactor coolant system flow rate, and reactor power. The error compo-nents in a measurement channel are combined statistically if they are independ-ent. Error components which are not independent are added arithmetically into groups, and the independent groups are then combined statistically. This method of calculating the measurement uncertainties has been previously found acceptable. The instrument uncertainty values assigned to the Westinghouse report are a conservative bounding set of instrument uncertainties for standard Westinghouse instruments. These values have been reviewed previously for V. C. Summer Nuclear Station, Unit 1 in NUREG-0717, Supplement 4. Attachment I to the May 5, 1982, letter indicated that the Braidwood uncertainties for pressurizer pressure and core coolant temperature are identical to the Westinghouse generic values since the sensors, process racks, and computer and readout devicas are standard Westinghouse-supplied nuclear steam supply system (NSSS) equipment. Therefore, the measurement uncertainties for the pressurizer pressure and reactor coolant temperature are acceptable. The reactor power is determined periodically by a secondary side power calori-metric calculation (i.e., the reactor power is the product of the feedwater flow rate and the enthalpy rise across the steam generator). Only the feedwater pressure and temperature are measured with instruments not supplied by Westing-house. The error allowance for these instruments varies slightly from the West-inghouse generic value. In the feedwater flow measurement, the uncertainty associated with crud buildup in the feedwater venturi is not taken into account. However, since venturi fouling would result in higher measured feedwater flow and higher indicated value of reactor power, neglecting venturi fouling for power measurement is acceptable. The RCS flow is measured periodically with the elbow taps in the cold legs to verify that the RCS flow does not violate the acceptable limit during power Braidwood SSER 1 4-1

operation. The elbow tap flow measurements are normalized against a precision flow calorimetric measurement which will be performed at the beginning of each fuel cycle. Therefore, the overall uncertainty of the RCS flow measurement consists of the uncertainties associated with the precision flow calorimetric and the elbow tap flow measurements. I In the determination of the flow calorimetric uncertainty, several interdepen-dent error components are combined statistically, and thus violate the indepen-dence requirement. For example, the venturi thermal expansion factor, feed-water density, and enthalpy are all dependent on the feedwater temperature; the feedwater density and steam enthalpy are both dependent on steamline pres-sure because the feedwater pressure is calculated from the steamline pressure; and the hot-leg and cold-leg enthalpies are both dependent on the pressurizer However they are treated as independent quantities because the may pressure. nitude of the uncertainties of these interdependent error components ia se sinali compared to the dominant error components, such as the hot-leg taper ato: e st ra-tification uncertainty, that the use of the statistical treatment of these com-ponents has no significant effect on the final result. In a letter Ja+ed Aug-ust 13, 1984, the applicant provided several examples treating these 4rterde-pendent error components both statistically and deterministically to demon-strate the minimal effect on the final results. In addition, the

u. certainty values used in the analysis are the bounding conservative valuas which can offset the small error resulting from the statistical treatmen+ of these in-terdependent error components.

Therefore, the statistical treatment of these error components is acceptable. Drift effect of the measurement instrumentation is not incisded in the analysis except where consideration is necessary because of sensor i;:ation. The appli-cant indicates that the Braidwood plant proceduics 41'l include provisions to ensure that the performance of calorimetric RCS f W measurement will require calibrations within 7 days of the flow measurement. for irstruments used in de-termining the RCS flow. Therefore, neglecting the drift effect in the error analysis is acceptable. The requirement of the calibration of the calorimetric flow measurement instrumentation has been incorporated in the Technical Specifications. The fouling effect of crud buildup in the venturi is not taken into account in the feedwater flow measurement. Since the venturi fouling is a bias which will result in a higher measured feedwater flow as well as RCS flow than the actual values, neglecting venturi fouling effect on RCS flow measurement is not acceptable. The applicant, in a letter dated February 5, 1985, provided a description of the method to be used to detect venturi fouling. These methods are basically visual and augmented, to the extent practical, "by reaching through the inspection port and touching the venturi surface." The inspections will be done on both Units 1 and 2. At each unit, the appli-cant will inspect at least two of the four feedwater ventaris at the first re-fueling. The applicant further committed to continue these inspections until it can be demonstrated to the NRC's satisf action that venturi fouling is not occurring. If all four venturis are cleaned during a refueling outage, no allowance for If all fouling is necessary in the appropriate setpoints and operating limits. four venturis are not cle ned, an allowance of 0.1% will be included to compen-cate for possible undetected fouling. Braidwood SSER 1 4-2 i 1

Excluding the venturf fouling uncertainty, the uncertainty of 2.1% for the RCS flow measured by the elbow tsps, which are normalized with the calorimetric flow measurement, is acceptable. With the inclusion of 0.1% for vectori foul-ing, the overall RCS flow uncertainty of 2.2% is acceptable. The design model DNBR limits for the typical cell and the thimble ct;) are cal-culated using the approved ITDP and WRB-1 critical heat flux (CHF) correlation. Only the application of the plant-specific uncertainty values and the sensi-tivity factors on DNBR of the pertinent ITDP parameters to derive the design limit DNBR values required review. The applicant provided a detailed calcula-tion of the design DNBR limits in the flay 5,1982, letter. It stated that the sensitivity factors and the ranges of applicability of the Braidwood ITDP para-meters are the same as those used in WCAP-9500 with the exception of the range of vessel flow. However, the sensitivity factors of DNBR with respect to core power used for the Braidwood typical cell and thimble cell are in re/erse order compared to the values used in WCAP-9500. The applicant indicated in the Aug-ust 13, 1984, letter that the values applied to the Braidwood units are correct whereas the values given in WCAc-9500 are reversed. However, the use of the incorrect DNBR/ power sensitivity factors in the determination of WCAF-9500 DNBR limits has a negligible effect on the final calculated DNBR limit values. Using the plant-specific uncertainty values of the ITDP parameters with the uncertainty value of 2.2% for the RCS flow, the staff independent calculations agree with the DNBR limits provided in the applicant's response, i.e., the design DNBR limits are 1.336 for the typical cell and 1.311 for the thimble cell. Therefore, the design DNBR limits of 1.34 and 1.32, respectively, for the typical cell and the thimble specified in the Final Safety Analysis Report (FSAR) are acceptable. The Braidwoc1 design safety analysis uses plant-specific safety DNBR limits of 1.49 and 1.47, respectively, for the thimble cells. Therefore, there is about 10.0% margin available for both typical and thimble cells to provide flexibility in the design, operation and analysis of the Braidwood units. The staff concludes that the DNBR calculation for the Braid-wood units is acceptable. Therefore, Outstanding Item B(2) is considered closed. 4.4.2 Fuel Rod Bowing Subsequent to issuance of the original SER, the Westinghouse topical report WCAP-8691, Revision 1, " Fuel Rod Bow Evaluation," was approved by the staff. This rod bow penalty evaluation method applies statistical convolution of the critical heat flux test data and inter-fuel rod gap closure data to derive the rod bow penalty on DNBR. The use of this method results in a significantly lower rod bow penalty compared to the interim method previously used. The ap-plicant has submitted a table of rod bow penalty as a function of fuel burnup calculated with the approved Westinghouse method. Since rod bow and gap closure increase with fuel burnup, the rod ~ bow penalty on DNBR increases with burnup. However, even though the plant mty be operated at higher burnup, the maximun fuel burnup used for the rod bow penalty calcu-lation is 33,000 MWD /MTU, The reason for using 33,000 MWD /MTV as a cutoff point is because the physical burndown effect of the high peaking fuel rod will exceed the rod bow effects at higher bernup. By the time the fuel exceeds a Braidwood.MER 1 4-3

burnup of 33,000 MWD /MTV, it is not capable of achieving limiting peaking fac-tors because the fissionable isotopes have decreased and the fission product inventory has built up. Therefore, the rod bow penalty value of less than 3% DNBR at 33,000 MWD /MTU represents the maximum rod bow penalty for Braidwood which has 17 x 17 optimized fuel assemblies. Since the use of the plant-specific design limit DNBR of 1.49 and 1.47 for the typical and thimble cells, respectively, has an inherent thermal margin of about 10.0%, the rod bow penalty can be accommodated by the available thermal margin. Therefore, no rod bow penalty is required for the Braidwood plant. A description of the available thermal margin and the rod bow penalty that is compensated by the thermal mar-gin is included in the " Bases" of the Technical Specifications to avoid multiple usage of the available margin. 4.4.7 Inadequate Core Cooling Instrumentation 4.4.7.1 Clarification of Requirements A clarification of requirements for inadequate core cooling (ICC) instrumenta-tion which is to be installed and operational before fuel load was provided in TMI Action Plan Item II.F.2 of NUREG-0737, " Clarification of TMI Action Plan Requirements." On November 4,1982, the Commission determined that an instru-mentation system for detection of ICC consisting of an upgraded subcooling mar-gin monitor, core exit thermocouples, and a reactor coolant inventory tracking system is required for the operation of pressurized water reactor fa::ilities. 4.4.7.2 Inadequate Core Cooling Detection System Design In response to TMI Action Plan requirements, the applicant has transmitted letters dated June 7, 1982; August 13, 1982; February 8, 1983; December 27, 1983; July 6, 1984; August 20, 1984; September 25, 1984; October 15, 1984; and February 22, 1985. The applicant has selected an ICC instrumentation package for use in Braidwood consisting of three instrumentation subsystems: (1) subcooled margin monitor to measure saturation /superheat margin, (2) heated junction thermocouples to monitor level / temperature in the upper region of the reactor vessel, and (3) incore instrumentation thermocouples to measure temperature at the core exit. The processing and display hardware includes two subsystems of hardware: a qualified, safety-related subsystem of ICC instrumentation; and an unquali-fied, non-safety subsystem of ICC instrumentation. The backup displays for reactor lesel and core exit temperature are safety grade; the primary displays are non-safety grade. Human factors engineering reviews have been applied to both types of displays. Plant-specific procedures vill be prepared based on the NRC-approved Westing-house Owner's Group guidelines. Minor departures from these guidelines will be necessary because the Braidwood heated junction thermocouple instrunientation covers a level range different from the standard dh ferential pressure instruments. The Braidwood Unit 1 ICC instrumentation system will be operational by fuel load. Braidwood SSER 1 4-4

4.4.7.2.1 Subcooled Margin Monitor The primary display for the subcooling margin monitor (SMM) is the plant com-puter's cathode ray tube (CRT) which is mounted on the main control board. The plant process computer will compute the degrees of subcooling for saturation and output this number in digital form on the safety parameter display system (SPDS) iconic display CRT. Subcooling is determined from 2 wide-range reactor coolant pressure instrument channels and 65 core exit thermocouples (CETs). The display range is 0-3000 psig for pressure channels and 200-2300 F for CETs. *The SMM is displayed on a control board indicator and can also be displayed on any of the various process computer output devices. Two digital CET monitors are pro-vided as backup displays. The two thermocouple monitors are each powered from a separate engineered safety feature (ESF) bus. There are 33 thermocouples on one monitor and 32 on the other. The thermocouples have been grouped so that either monitor can display representative temperatures across the entire core cross-section. The process computer system is a highly reliable system with four separate central process units (CPUs). One CPU serves as a backup for any other CPU which may fail. The computer system is powered from either of the two inde-pendent ac sources with an automatic dc battery backup capability. The avail-ability of the process computer is expected to be at least 99%. All signals to the computer are isolated from safety-related instrument channels by the qualified isolators. The backup method to determine the subcooling margin is from two separate safety-related wide range reactor coolant pressure indicators and two separate safety-related core exit thermocouple monitors (which provide the average of the 10 highest CET temperature inputs). The operator reads the wide-range pressure. indicator to determine the appropriate saturation curve and compares the derived temperature from this curve with the core exit thermocouple monitor reading (i.e., the average of the ten highest thermocouple temperatures) to determine the subcooling margin. The operator also monitors the four contain-ment pressure indicators and the two radiation monitors to determine whether the containment is at normal or adverse conditions (for use of the correct sat-uration curves). The adverse containment condition is defined as: (1) con-tainment pressure greater than 5 psig; or (2) containment radiation greater than 104 R/hr. The staff concludes that the proposed SMM display system satisfies the TMI Action Plan Item II.F.2 requirements. 4.4.7.2.2 Core Exit Thermocouple System The design of the in core instrumentation system includes 65 Type K (chromel-alumel) thermocouples. The thermocouples are installed into guide tubes which penetrate the reactor vessel head and terminate at the exit flow end of selected fuel assemblies. The CET system is seismically and environmentally qualified to the requirements of IEEE 344-1975 and IEEE 323-1974, respectively. The iso-lation devices in the CET processors are accessible for maintenance following i an accident. Braidwood SSER 1 4-5 e

The processing equipment for the CET will perform the following functions: l l (1) Process all core exit thermocouple inputs. Processing of 33 CET inputs will be performe6 Dy Channel A and 32 CET inputs by Channel B. (2) Provide 33 Channel A,and 32 Channel B thermocouple outputs (8 per quad-rant), respectively, to the backup displays. I (3) Provide data link outputs to the process computer for all 65 thermocouple inputs. These outputs are isolated signals. The primary displays for ICC detection are generated by the plant process com-puter using isolated outputs from the heated junction thermocouple system (HJTC) and CET processor cabinets and nuclear steam supply (NSS) protection system cabinets (for reactor coolant loop pressures). The main control room primary displays for ICC detection are part of the safety parametcr display system (SPOS). The primary display for CET is on the SPDS. Additional displays include a spa-tially oriented core map indicating the temperature at each of the CET locations, a core exit temperature representative of the CET inputs, and trends of core exit temperature. Both The backup displays for HJTC and CET are driven by a two-channel system. the HJTC and CET systems use microprocessor-based designs for the signal pro-cessing function in conjunction with main control room indication, digital and analog, respectively. Each channel will accejt and process ICC input signals and provide outputs to thc channel related indicator and the plant process com-puter. Selectable temperatures from 65 core exit thermocouples, 33 for Chan-nel A and 32 for Channel B are available on the backup displays. 4.4.7.2.3 Heated Junction Thermocouple System Two identical HJTC probe assemblies are installed in each of the Braidwood units. These probe assemblies are identical to System 80 probe assemblies. There are eight heated / unheated thermocouple pairs (sensors) in each probe as-sembly. The HJTC sensor arrangement for the Braidwood units has two sensors located in the upper head and six sensors in the upper plenum. Only two sen-sors are placed in the upper head because once the water level falls below the top of the rod cluster control assemblies (RCCA) guide thimbles, the upper head liquid inventory no longer exists to communicate with the sensors in the upper plenum and reactor core. The processing equipment for the HJTC performs the following functions: (1) determines if liquid inventory exists at the HJTC position (2) processes all inputs and calculated outputs for dispicy (3) provides an alarm output to the plant annunciator system when any of the HJTC detects the absence of liquid level (4) provides control of heater power for proper HJTC output signal level (5) provides an input to the process computer for percent liquid inventory above the fuel alignment plate Braidwood SSER 1 4-6

i 1 The primary display provides liquid level inventory above the fuel alignment i plate and trends of liquid level inventory. The backup HJTC display provides percent liquid inventory level above the fuel alignment plate derived from the eight discrete HJTC positions; i.e., the unheated junction temperature at eight positions and the heated junction temperature at eight positions. 4.4.7.3 Evaluation 2 1 The staff has reviewed the applicant's submittals and concludes that the ICC detection system design is acceptable. Therefore, Outstanding Item B(3) is 4 a considered closed. i l r l i f l i h i 4 Braidwood SSER 1 4-7

5 REACTOR COOLANT SYSTEM

5. 2 Integrity of Reactor Coolant Pressure Boundary 5.2.4 Reactor Coolant Pressure Boundary Inservice Inspection and Testing 5.2.4.4 Evaluation of Compliance With 10 CFR 50.55a(a)(3) for Braidwood Unit 1 For nuclear power facilities whose construction permit was issued on or after July 1,1974,10 CFR 50.55(g)(3) specifies that components shall meet the pre-service examination requirements set forth in editions and addenda of Section XI of the ASME Boiler and Pressure Vessel Code applied to the construction of the particular component.

However, 10 CFR 50.55a(a)(3) permits alternative require-ments to 10 CFR 50.55(g)(3) when authorized by NRC's Director of the Office of Nuclear Reactor Regulation. This regulation requires that the applicant demon-strates that (1) the proposed alternatives would provide an acceptable level of - quality and safety or (2) compliance with the specified requirements of this section would result in hardship or unusual difficulties without a compensating increase in the level of quality and safety. The Braidwnod Unit 1 steam generators and pressurizer were ultrasonically examined to the preservice inspection requirements of Section XI of the ASME Code,'1977 Edition and Addenda through Summer 1978. In a letter dated Feb-ruary 28, 1986, the applicant requested relief from these requirements for these components. Article IWB-3112 of this Code Edition and Addenda indicates that flaws exceeding the standards of Table IWB 3410-1 shall be unacceptable for service unless such flaws are removed or repaired to the extent necessary to meet the flaw indication standards before placement of the component in service. In its February 28, 1986, letter, the applicant stated that two indi-cations exceeded the standards of Table IWB 3410-1. One unacceptable indica-tion was found in the upper shell-to-transition cone circumferential weld of the loop 1 steam generator. Another unacceptable indication was identified in the upper middle shell-to-lower. middle shell circumferential weld of the pres-surizer. Instead of excavating and weld repairing these flaws, the applicant proposes to leave the flaws in the vessels. The proposal is based on fracture mechanics analyses that indicate that these flaws will not grow to an unaccept-able size'during the life of the plant. The fracture mechanics analyses are documented in Westinghouse Reports WCAP-11063, " Background and Technical Basis for the Handbook on Flaw Evaluation for Byron Units 1 and 2 Stean Generators i and Pressurizers," and WCAP-11064, " Handbook on Flaw Evaluation for Byron Units 1 and 2 Steam Generators and Pressurizers." These reports were submitted in a letter from A. D. Miosi to H. R. Denton dated June 11, 1986. This letter indicates that the flaw evaluation is applicable to the Braidwood Units 1 and 2 steam generators and pressurizers. The applicant concluded that the flaws in the Braidwood Unit 1 steam generator and pressurizer were small subsurface slag inclusions formed during vessel fabrication and not cracks or lack of fusion. This conclusion was based on volumetric and destructive examination of flaws in the Byron Units 1 and 2 steam Braidwood SSER 1 5-1

i generators and volumetric examination of the flaws in the Braidwood Unit 1 steam generator and pressurizer. Volumetric examination of welds in Braidwood Unit 1 and Byron Units 1 and 2 steam generators and pressurizers was performed using ultrasonic inspection techniques. The ultrasonic procedures, equipment, and calibration blocks used for the Braidwood Unit 1 inspection meet the require-ments of the Section XI, 1977 Edition with Addenda through Summer 1978, which were the applicable requirements for inspection of Byron Units 1 and 2. The calibration blocks are of the same' material specification with the same size machined reflectors. The transducers used at both plants were 2.25 megahertz (MHz) with 0.5-inch X 1.0-inch element sizes. Destructive examination of the flaws in the Byron Units 1 and 2-steam generators was performed by metallo-graphic examination of core samples and by visual examination during flaw ex-cavation. As a result of these examinations, the flaws in the Byron Units 1 and 2 steam generators were determined to be (1) embedded slag inclusions re-sulting from the welding process, (2) smaller in size than the size estimated by the ultrasonic sizing method, and (3) subsurface flaws that did not open to the surface. The ultrasonic characteristics of the flaws in the Braidwood Unit 1 steam generator and pressurizer were similar to the ultrasonic charac-teristics of the flaws removed and evaluated in the Byron Units 1 and 2 steam On the basis of similarity in the ultrasonic characteristics, the generators. flaws in the Braidwood steam generator and the pressurizer are believed to be small subsurface slag inclusions. The applicant indicates that the location of the subsurface slag inclusions at Braidwood makes their removal very difficult and their removal would not guar-antee an increase in the vessel integrity. Core sampling of the steam generator flaw is impractical because of the components' geometry and impractical for the pressurizer because of its cladding. The indications can De excavated, but would require weld repair. Weld repair could result in additional slag inclu-sions and increased internal residual weld stress. The applicant has provided flaw evaluation charts in Appendix A to WCAP-11064 for circumferential and longitudinal oriented welds in the Braidwood pressurizers and steam generators. These charts were constructed using fracture mechanics analyses. The method and criteria used in the fracture mechanics analyses are documented in WCAP-11063. The fracture mechanics analyses that were performed to develop the flaw evaluation charts were in accordance with the methodology and criteria specified in Paragraph IWB-3600 and Appendix A of the ASME Code, Section XI, except that stresses were not linearized and stress intensity fac-tors were not calculated in accordance with the recommendations in Appendix A. Instead of linearizing the stresses, the proposed method represented the actual stress profile by a cubic polynomial. Stress intensity factors were calculated using the expressions discussed by McGowan and Raymund (1979), Newman and Raju (1980), Buchalet and Banford (1976), and Shah and Kobayashi (1981). These stress profiles and stress intensity factor expressions are believed to pro-vide a more accurate determination of the critical flaw size, and are particu-larly important during the evaluation of emergency and faulted conditions in which the stress profile is generally nonlinear and often very steep. Important parameters in a fracture mechanics analyses are the materials' brittle fracture resistance and the projected flaw growth rate during operation of the component. The standard measurement of brittle fracture resistance for the vessel materials in the Braidwood pressurizer and steam generator is their 'Braidwood SSER 1 5-2

crack initiation and arrest fracture toughness. These values of fracture tough-ness are used to determine a critical flaw size. Westinghouse indicates that the critical flaw size calculation used the crack initiation and arrest frac-ture toughness for vessel materials that are recommended in Appendix A of the ASME Code, Section XI. The critical flaw size for each weld location was determined using a reference temperature, RT f 10 F and an upper shelf

NDT, toughness of 200 ksi/in.

These values are acceptable for the steam generator and pressurizer welds because the weld material in these components is not sub-ject to neutron irradiation damage. The amount of projected flaw growth was calculated using the transients reported in Table 2-1 and 2-2 of WCAP-11063 and the rate of fatigue growth recommended in Appendix A of Section XI of the ASME Code. The rates of fatigue growth docu-mented in Appendix A apply to surface flaws in a water reactor environment and subsurface flaws in an air environment, but do not apply to accelerated growth resulting from stress corrosion mechanisms. The staff believes that stress corrosion should not be a problem for the flaw in the pressurizer, because it is located 2.2 inches form the inside of the vessel. Hence, it will not be in contact with a corrosive environment and will not be subjected to a stress corrosion mechanism. Stress corrosion of the upper shell-to-transition cone circumferential weld was observed in the Indian Point Unit 3 steam generator. The operating characteristics and weld configuration are similar for the Indian Point Unit 3 and Braidwood Unit 1 steam generators. Stress corrosion of the Indian Point Unit 3 steam generator vessels was studied by Brookhaven National Laboratory (NUREG/CR-3281). Major factors in the accelerated growth of the stress corrosion observed in the Indian Point Unit 3 vessel were high' residual weld stresses resulting from stress relief at 1000 F and copper cations in solu-tion. The copper cations resulted from corrosion of the copper condenser tube material. The stress corrosion growth of the flaw in the Braidwood steam gen-erator is not considered likely because its upper-shell-to-cone weld was stress relieved at 1125 1 25 F, its feedwater system heat exchangers and main conden-sers are.made of Type 304 stainless steel, and the defect is located 0.55 inch from the inside surface. However, to account for uncertainties in flaw growth rate, the staff requires that flaws evaluated in accordance with the criteria in IWB-3600 be re-examined in intervals listed in IWB-2420 of Section XI of the ASME Code. On the basis of the location and depth determined by ultrasonic examination, the flaw evaluation charts in Appendix A of WCAP-11064 indicate that the' flaw in the steam generator weld is acceptable in accordance with the analytical criteria of IWB-3600 of Section XI of the.ASME Code and the steam generator secondary side hydrotest and leak test must be performed at temperatures greater than 165 F and 150 F, respectively. On the basis of the location and depth determined by ultrasonic examination, the flaw evaluation charts in Appendix A of WCAP-11064 indicate that the flaw in the pressurizer weld is acceptable in accordance with the analytical criteria of IWB-3600 of Section XI of the ASME Code and the pressurizer primary side hydrotest and leak test must be performed at temperatures greater than 120 F. The staff concludes the following: Braidwood SSER 1 5-3

(1) The flaws that exceed the standards of Table IWB-3410-1 of Section XI of the ASME Code in the steam generator and pressurizer are most likely slag inclusions resulting from weld fabrication. (2) Excavation and weld repair'of these flaws would not increase the level of quality and safety of the components. (3) The fracture mechanics _ evaluations illustrated in charts in WCAP-11063 and WCAP-11064 demonstrate that the flaws will not grow during the life of the plant to a size that will affect the integrity of the vessels. (4) The proposed alternative provides an acceptable level of quality and safety, subject to the following examination and test requirements: (a) The areas containing the flaws must be inspected in accordance with the inspection interval requirements of IWB-2420 of Section XI of the ASME Code. (b) The loop 1 steam generator secondary side hydrotest and leak tests must be performed-at temperatures greater than 165 F and 150 F, respectively. (c) The pressurizer primary side hydrotest and leak test must be per-formed at temperatures greater than 120 F. (5) After the applicant commits to the inservice examination and test require-ments of item (4), the applicant will have demonstrated compliance with the criteria in 10 CFR 50.55a(a)(3). Hence, the applicant may be per-mitted to place the pressurizer and loop 1 steam generator into service without weld repair of the two flaws that exceed the preservice acceptance standard of Section XI of the ASME Code. This is a Confirmatory Issue. 5.3 Reactor Vessel 5.3.1 Reactor Vessel Materials and RCPB Materials 5.3.1.2 Evaluation of Compliance to 10 CFR 50, Appendix H The SER indicated that the reactor vessel beltline surveillance program had met all the requirements of 10 CFR 50, Appendix H, and American Society for Testing Materials (ASTM) Standard E-185 except that the applicant had not indicated the surveillance capsule removal schedule for Units 1 and 2 and the materials in the Unit 2 surveillance capsules. The applicant has subsequently provided the required information. On the basis of this information, the staff has deter-mined that the Braidwood reactor vessel beltline surveillance program complies ~ with the requirements of 10 CFR 50, Appendix H, and ASTM E-185. 5.3.2 Pressure-Temperature Limits Pressure-temperature limits must meet the requirements of Appendix G, 10 CFR 50. A revision to 10 CFR 50, Appendix G, was published in the Federal Register on Braidwood SSER 1 5-4

May 27, 1983, and became effective on July 26, 1983. The amended Appendix G of 10 CFR 50 states that when pressure exceeds 20% of the preservice system hydrostatic test pressure, the temperature of the closure flange regions that are highly stressed by bolt preload must exceed the reference temperature of the materials in those regions by at least 120aF for normal operation and by 90 F for hydrostatic pressure tests and leak tests, unless a lower temperature can be justified by showing that the margins of safety for those regions, when they are controlling, are equivalent to those required for the beltline, when it-is controlling. By letters dated January 3,1984, and August 13, 1986, the applicant provided 'l heatup and cooldown curves for the Braidwood reactor vessels. The applicant stated that the heatup and cooldown curves for the Braidwood reactor vessels were reviewed against the amended Appendix G requirements. Only the cooldown curve for Braidwood Unit 2 had to be revised to incorporate the new closure flange limitations. The staff has reviewed the' revised curves submitted by let-ters dated January 3, 1984, and August 13, 1986, and has found them acceptable. 5.3.3 Reactor Vessel Integrity The SER indicated that the review would be completed when the staff received the reactor vessel material data for Braidwood Units 1 and 2. The staff has reviewed the site-specific data concerning reactor vessel materials for Braid-wood Units 1 and 2, which confirm the bases of staff conclusions included in the SER. Therefore, Confirmatory Issue A(5) is considered closed.

5. 4 Component and Subsystem Design 5.4.2 Steam Generators

5. 4. 2. 2 Steam Generator Tube Inservice Inspection 5.4.2.2.1 Compliance With the Standard Review Plan The staff has completed its review of applicant's commitments to conform with the Westinghouse Standard Technical Specifications (NUREG-0452) in the area of steam generator tube inservice inspection and finds them acceptable.

Thus, Confirmatory Issue B(2) is considered closed.

5.4.2.2.2 Evaluation of the Inspection Program In the SER, the staff identified a generic problem concerning a potential for tube degradation caused by flow-induced vibration in the preheater section of Westinghouse Model D steam generators. The staff evaluation of the information submitted by the applicant in its letter dated December 16, 1983, relative to the changes being made to the Braidwood steam generators to minimize tube vibration, follows. r The potential for tube well degradation resulting from flow-induced vibration in Westinghouse Model D4 and'05 steam generators has been thoroughly evaluated and documented in NUREG-1014, " Safety Evaluation Report Related to Model D4/05 Steam Generator Design Modification." l Braidwood SSER 1-5-5

The primary cause of tube vibration in heat exchangers is hydrodynamic excita-tion caused by secondary fluid flow on the outside of the tubes. In the range of normal steam generator operating conditions, the effects of primary fluid flow inside the tubes and mechan.ically induced tube vibration are considered to be negligible. To evaluate flow-induced tube vibration in the preheater region of the tube bundle, Westinghouse undertook an extensive program using data from operat-ing plants, full-and partial-scale model tests, and analytical tube vibration Operating plant data consisted of tube wear data from tubes removed models. from steam generators, eddy current tests, and tube motion data from accelerom-eters installed inside selected tubes. Model testing generated tube wear data, flow velocity distributions, tube motion parameters, and flow-induced tube vibration forcing functions. The tube vibration analyses applied the forcing functions to produce tube motion data. The results of these evaluations were consistent with the early operating experience of preheat steam generators. On the basis of the above extensive model test and -analysis program, Westing-house designed, verified, and implemented a modification to the steam generator to reduce tube vibratory response to preheater inlet flow excitation. Addition-ally, the magnitude of the flow-forcing function was reduced through implementa-The vibra-tion of a preheater flow bypass arrangement in the feedwater system. tion of the performance of the modifications in reducing tube excitation and response was achieved with input from a full-scale test under simulated conser-vative flow and tube support conditions. The above design modifications developed by Westinghouse for the preheater sec-tion of Model 04 and D5 steam generators provide a substantial reduction in tube vibration. As a result, the potential for tube wear has been reduced to within acceptable levels. In the Model 04/05 steam generators in Braidwood Units 1 and 2, the modifica-tions consist of expanding selected tubes into the baffle plates in the pre-heater, and splitting the feedwater flow through the auxiliary feedwater nozzle. The close support condition, resulting from tube expansion at the supports, significantly changes the response frequency and also the G-Delta value (pro-duct of the peak to peak acceleration and root-mean-square displacement). The G-Delta parameter provides a measure of tube wear resulting from vibration. A reduced value of G-Delta is indicative of diminished potential for tube wear. The split feedwater flow reduces the mass flow and velocity of the fluid in the preheater section. Both modifications combine to provide a substantial im- ~ provement by reducing the potential for tube wear. The design modifications and their consequences'for steam generators and plant performance were reviewed extensively by the NRC staff'and an independent panel In NUREG-1014, the staff concluded that the proposed modification of experts. ensures substantial improvement by reducing the potential for tube wear to with-in acceptable levels. This conclusion was reached after a thorough review of the test models and testing results as well as evaluation of analytical models and analytical results. Fatigue of the tubes in the preheater region that are subject to flow-induced excitation is not a concern because the maximum resultant stresses in the tube are below the endurance limit of the material. Braidwood SSER 1 5-6

For areas of the tube bundle other than the preheater, parallel flow analyses were performed to determine the vibratory deflections. These analyses indicate that the flow velocities are sufficiently low that they result in negligible fatigue and vibratory amplitudes. The support system, therefore, is considered adequate with regard to parallel flow excitation. To evaluate crossflow at the exit'of the downcomer and flow to the tube bundle and at the top of the bundle of the U-bend area, Westinghouse performed an experimental research program of crossflow in tube arrays with the specific parameters of the Model D4/D5 steam generator using air and water model tests. The results of this research indicate that these regions of the bundle are not subject to the vortex shedding mechanisms of tube excitation. Vortex shedding was found not to be a significant mechanism in these two regions for the follow-ing reasons: (1) Flow turbulence in the downcomer and tube bundle inlet region inhibit the formation of Von Karman vorticies. (2) Both axial and crossflow velocity component exist on the tubes. The axial flow component disrupts the Von Karman vorticies. This research program was also the basis for evaluating the fluid elastic mechanism associated with crossflow at the tube sheet. The evaluation showed the adequacy of the tube support arrangement. Flow turbulence can result in some tube excitation in these regions. This exci-tation is of little concern, however, because (1) Maximum stresses in the tubes are at least an order of magnitude below the fatigue endurance limit of the tube material, and (2) Tube support arrangements preclude significant vibratory motion. In summary, tube vibration has been thoroughly evaluated. Mechanical and pri-mary flow excitation are considered negligible. Secondary flow excitation has been evaluated. From this evaluation, the staff has concluded that the pro-posed expansion of selected tubes and splitting the feedwater flow through the auxiliary feedwater nozzle provides a reduction in tube vibration and in the potential for tube wear to within acceptable levels. Any tube wear resulting from the tube vibration would be limited and would progress slowly. This allows use of a periodic tube inservice inspection program for detection and followup of tube wear. This inservice inspection program, in conjunction with tube plug- .ging criteria, provides for safe operation of the Model D4/05 steam generators. Thus, Outstanding Item B(4) is considered closed. 5.4.3 Residual ~ Heat Removal System The SER stated that if the Diablo Canyon natural circulation tests are not com-pleted or do not provide satisfactory results, the applicant will perform such tests at Byron Station before startup after the first refueling outage. In either case, the resulting test data will then be applied to Braidwood Units 1 and 2. The applicant's commitments are documented in FSAR Amendment 39, Question 212.154. Therefore, License Condition A(2) is no longer required. Braidwood SSER 1 5-7

5.4.6 Seismic and Environmental Qualification of Pressurizer Power-0perated Relief Valves In a letter dated October 7,1982, the staff required the applicant to justify that the seismic and environmental design of the pressurizer power-operated relief valves (PORVs) will be appropriate for safety-related functions. Spe-cifically, the staff questioned the use of the PORVs as high point vents in achieving cold shutdown and in mitigating steam generator tube ruptures. The applicant provided additional information concerning each of these functions in Amendment 43 in response to Question 212.160, dated September 1983. The applicant stated that the PORV at Braidwood would be upgraded to operate fol-lowing a safe shutdown earthquake (SSE). In addition, the applicant justified' that the current environmental design of the PORV (service in a mild environ-ment) is adequate. The staff has concluded that the above issues are resolved. High Point Vents High point vents are required by 10 CFR 50.44(c)(3)(iii) for noncondensible gas removal. Specifically, venting capability is required for the reactor coolant system, the reactor vessel head, and other systems required to maintain adequate core cooling if the accumulation of noncondensible gases would cause the loss of function of these systems. High point vents are not required, however, for the tubes in U-tube steam generators. Because severe accidents involving nonconden-sible gas generation might exceed the environmental design of the.PORVs, the applicant does not rely on the PORVs as high point vents but instead relies on the reactor vessel head vents, which are provided with redundant valving in parallel paths. The staff's acceptance of the head vent design is discussed in Section 5.4.5 of the SER. Because the head vent design is adequate to re-move noncondensible gas from the reactor system and because noncondensible gas accumulation in the pressurizer will not affect core cooling, the staff has concluded that the high point vent design for Braidwood Station is adequate without reliance on the PORVs. Cold Shutdown Caoability Branch Technical Position (BTP) RSB 5-1, attached to SRP Section 5.4.7, re-quires that the reactor be designed to be taken from normal operating condi-tions to cold shutdown using 'only safety-related equipment. Because the PORVs -may be used to depressurize the reactor system during the approach to cold shutdown, the staff required that they be designed to operate following an SSE or that an alternative safety-related method for reactor system depressuriza-tion be provided. The PORVs will be upgraded to operate following an SSE; therefore, this matter is resolved. Environmental consideration for cold shutdown is the same as for a steam generator tube rupture (SGTR) as discussed below. Steam Generator Tube Rupture Use of the PORVs may be required to depressurize the reactor coolant system during SGTR events when the pressurizer level is maintained by the emergency core cooling system (ECCS). The staff required the applicant to demonstrate Braidwood SSER 1 5-8

4 that the PORVs are seismically and environmentally qualified to mitigate this event or to provide other appropriately qualified safety-related equipment. The applicant calculated that the containment environment produced by an SGTR would be mild with a maximum containment temperature of 160 F. The tempera-ture increase would be pruduced by opening the PORV assuming failure of the nonsafety-related pressurtier relief tank. The PORV block valves, which are environmentally qualified for harsh environments,'ould be relied on to isolate c a stuck open PORV. The staff concurs with the applicant's assessment that the environmental qualification of the PORVs is adequate. The applicant has com-mitted to upgrade the seismic qualification of the PORVs to SSE; therefore, this matter is resolved. i 1 i 1 l i I f Braidwood SSER 1 5-9

6 ENGINEERED SAFETY FEATURES 6.2 Containment Systems 6.2.1 Containment Functional Design 6.2.1.5 Minimum Containment Pressure Analysis for Performance Capability Studies on the ECCS Appendix K to 10 CFR 50 requires that the value of containment pressure in eval-uating the cooling effectiveness of the emergency core cooling water during core reflood shall not exceed a pressure calculated conservatively for that purpose. Also, the calculation must include the effect on operation of all installed containment pressure-reducing systems and processes. The staff stated in the SER that the applicant's analysis was acceptably conser-vative and in conformance with the provisions of BTP CSB 6-1, " Minimum Contain-eent Pressure Model for PWR ECCS Performance Evaluation," with three exceptions. These exceptions have, subsequent to the publication of the SER, been resolved, as described below: (1) The applicant assumed an essential service water (ESW) temperature of 100 F for calculating the heat removal capacity of the reactor containment fan coolers (RCFCs). BTP CSB 6-1 states that the minimum cooling water tem-perature be used to maximize the heat removal capacity of the RCFCs. The applicant has redone the analysis (see response to Question 022.24, FSAR Amendment 43) assuming an ESW temperature of 45 F, which is acceptable. (2) The applicant assumed a 40 second delay in RCFC initiation, whereas other information provided by the applicant indicated that the delay time could be as short as 17 seconds. For the reanalysis performed in response to Question 022.24, FSAR Amendment 43, the applicant assumed an initiation time of 15 seconds, which is acceptable. (3) The applicant did not consider.the effect of miniflow purge system opera-tion at the onset of a LOCA on the containment pressure response. The applicant has provided.an analysis of this effect (see response to Ques-tion 022.23, FSAR Amendment 39). The applicant assumed the system isola-tion valves would be fully closed 7.62 seconds after onset of the accident. This time is based on the time required to reach the isolation set point, the signal delay time, and a 5-second valve closure time. Discharge of containment atmosphere through the supply and exhaust lines was assumed for the entire 7.62 seconds; flow resistance effects and the reduction in flow area caused by the closing of the valves were conservatively ignored in calculating the containment atmosphere release. The applicant calculated a containment pressure drop of less than 0.05 psi, and estimated that the peak fuel clad temperature would increase by approximately 1*F. Therefore, the staff concludes that the effect of operation of the miniflow purge system on the minimum containment pressure analysis is insignificant and need not be considered. Braidwocd SSER 1 6-1

In conclusion, the applicant has provided a reanalysis of the minimum contain-ment pressure transient that is in accordance with the provisions of BTP CSB 6-1 and resolves previous staff concerns. The FSAR has been revisec to reflect the effect that the revised minimum containment pressure transient has on the ECCS performance capability. Therefore, this resolves Confirmatory Issue B(4). 6.2.2 Containment Heat Removal Systems The SER indicated that the staff would confirm that the applicant's sump design conforms to the guidelines in-RG 1.82. The latest revisions of drawings S-904, S-905, S-995, S-996 and S-1065, which detail the addition of an outer screen l that encompasses both of the existing outer containment recirculation pump screens, have been reviewed and approved for Byron Unit 1 (see Inspection Report No. 50-454/84-19). Because the sump design is identical for all four Byron /Braidwood units, the staff considers Confirmatory Issue B(5) closed for Braidwood Units 1 and 2. 6.2.5 Combustible Gas Control System The staff stated in the SER that the two hydrogen recombiners at the Braidwood site, including their associated piping and valves, will perform the intended hydrogen control function assuming any single active component failure coinci-dent with loss of offsite power. The acceptability of this statement, however, was contingent on the incorporation of a design modification whereby the suction and discharge valves associated with a recombiner would receive electrical power from the same Class 1E power supply that serves the recombiner. This will pre-clude the loss of both reco.nbiner trains in the event of the loss of one of the two Class 1E power supply divisions, assuming the suction and discharge valves are normally closed. The SER also stated that the applicant committed to make the design change described above before initial fuel loading. By letter dated February 22, 1984, the applicant informed the staff that the present recombiner system design differs from that described in the SER with respect to the power supplies serving the valves; the applicant also provided justification for the design, as discussed below. There are two hydrogen recombiners permanently installed at the Braidwood Sta-tion. Through the use of cross-tie piping, either recombiner may be used on either unit. The suction and discharge valve operators are powered from oppo-site division Class 1E power supplies. Specifically, recombiner 1 and the suc-tion line valve are powered from Division Ell, and the discharge line valve is powered from Division E12. Recombiner 2 and the suction line valvt are powered from Division E12, and the~ discharge line valve is powered from Division Ell. Also, the discharge line valves are normally kept open. The suction and dis-charge valves are not powered from a common power supply because certain single failures in that configuration could compromise recombiner system effectiveness. The present electrical division assignments prevent backflow through a failed recombiner. As presently arranged, both hydrogen recombiners operate in par-allel, using the same suction and exhaust piping from each containment. With a Class IE power supply failure, sufficient redundancy exists to operate at least one recombiner at design capacity. If each recombiner and its suction and discharge valves were powered from the same supply, a single failure of either Class 1E power supply could prevent isolation of the associated recom-biner piping circuit. Flow from the redundant recombiner would follow the path Braidwood SSER 1 6-2 .=

of least resistance and backflow through the failed recombiner, the design flow rate of air from the containment would not be achieved. T5e applicant's hydrogen recombiner system design eliminates the potential for backflow as discussed above, and is, therefore, an improvement on the design described in the SER. This is, however, contingent on the recombiner discharge valves being kept open during normal operation; the applicant must ensure that appropriate administrative controls are instituted to maintain the discharge valves open. The applicant has committed to satisfy this requirement. On the basis of the above discussion, the staff concludes that the applicant's recombiner system design is acceptable. 6.2.6 Containment Leakage Testing Fluid systems penetrating containment that may be opened to the containment atmosphere under post-accident conditions must, in general, be vented and drained during containment integrated leakage rate (Type A) testing to ensure exposure of the system containment isolation valves to the test medium (contain-ment air) and test differential pressure. In so doing, potential containment atmosphere leak paths will be included in the Type A test. Certain exceptions are allowed, as noted in paragraph III. A.1(d) of Appendix J to 10 CFR 50. As stated in the SER, the applicant had not finished preparing the Type A' test procedures concerning venting and draining, but did commit to comply with the appropriate requirements of Appendix J. The staff stated that once the appli-cant's ~ procedures had been completed, the staff would confirm that proper vent-ing and draining provisions would be employed. By letter dated April 19, 1983, the applicant submitted information describing the venting and draining provisions. Fluid systems penetrating containment will be vented and drained during Type A tests, with the following exceptions: (1) Paragraph III.A.1(d)-of Appendix J states, in part, that those portions of the fluid systems that are part of the reactor coolant pressure boundary and are open directly to the containment atmosphere under post-accident conditions and become' an extension of the boundary of the containment shall be vented to the containment atmosphere during Type A tests. It further states that portions of closed systems inside containment that penetrate containment and rupture as a result of a LOCA shall be vented to the con-tainment atmosphere. The applicant submits that the main steamlines, the main feedwater and auxiliary feedwater lines, and the component cooling system excess letdown heat exchanger inlet and outlet lines do not fall into either category. These lines are not part of the reactor coolant pressure boundary, and are not postulated to rupture during a LOCA, because they satisfy the design provisions of Section 11.0. of Standard Review Plan (SRP) (NUREG-0800) Section 6.2.4, " Containment Isolation System," for closed systems inside containment. Therefore, the staff concludes that these lines need not be vented and drained during Type A tests. (2) Paragraph III. A.1(d) of Appendix J also states that systems required to maintain the plant in a safe condition during the test shall be operable in their normal mode and need not be vented, and systems normally filled Braidwood SSER 1 6-3

with water and operating under post-accident conditions need not be vented. The applicant submits that the residual heat removal system suction lines, the fire protection system lines, and the chemical and volume control sys-tem (CVCS) charging and loop fill headers are required to maintain the plant in a safe condition during the test. Also, the containment isola-tion valves in the CVCS charging and loop fill headers will be isolated post-accident and sealed by high pressure water from the centrifugal charg-ing pumps, and therefore do not constitute a potential leak path for con-tainment atmosphere. Therefore, the staff concludes that the above lines need not be vented and drained during Type A tests. (3) The applicant submits that the essential service water lines to the reactor containment fan coolers, the safety injection system injection lines, and the chemical volume and control system seal injection lines are normally filled with water and operating under post-accident conditions. The staff concurs, and concludes that these lines need not be vented and drained during Type A tests. In summary, the staff concludes that the lines specified in the applicant's letter dated April 19, 1983, and discussed above, need not be vented and drained during the performance of Type A tests. This resolves Confirmatory Issue B(6). 6.2.7 Fracture Prevention of Containment Pressure Boundary ~The ferritic materials in the Braidwood Units 1.and 2 containment system that constitute the containment pressure boundary have been assessed to determine if the material fracture toughness is in compliance with the requirements of Gen-eral Design Criterion (GDC) 51, " Fracture Prevention of Containment Pressure Boundary." GDC 51 requires that under operating, maintanance, testing, and postulated acci-dent conditions the (1) ferritic materials of the containment pressure boundary l behave in a nonbrittle manner and (2) probability of rapidly propagating frac-1 ture is minimized. The containment for Braidwood Units 1 and 2 is a reinforced concrete structure with a thin steel liner on the inside surface that serves as a leaktight mem-brane. The ferritic' materials of the containment pressure boundary that were considered in the staff's assessment are those that have been applied in the fabrication of the equipment hatch, personnel lock, penetrations and fluid These system components, including the valves required to isolate the system. components are the parts of the containment system that are not backed by con-crete and must sustain loads during the performance of the containment function. .The staff has determined that the fracture toughness requirements contained in ASME Code editions and addenda typical'of those used in the design of the Braidwood containment may not ensure compliance with GDC 51 for all areas of ~ he containment pressure boundary. t As a result, the staff elected to apply in the licensing reviews of ferritic l containment pressure boundary materials, the criteria for Class 2 components identified in the Summer 1977 Addenda of Section III of the ASME Code. Because the fracture toughness criteria that have been applied in construction typically Braidwood SSER 1 6-4

differ in Code classification and Code edition and addenda, the staff chose the criteria in the Summer 1977 Addenda of Section III of the Code to provide a uni-form review, consistent with the safety function of the containment pressure boundary materials. Therefore, the staff has reviewed the components of the Braidwood Units 1 and 2 containment pressure boundary according to the fracture toughness requirements of the Summer 1977 Addenda of ASME Section III for Class 2 components. Considered in this review were components of the containment system that'are load bearing and provide a pressure boundary in the performance of the contain-ment function under operating, maintenance, testing, and postulated accident conditions as addressed in GDC 51. These components are the equipment hatch, personnel airlock, penetrations, and elements of specific containment penetrat-ing systems. The staff assessment of the fracture toughness of materials was based on the metallurgical characterization of these materials and fracture toughness data presented in NUREG-0577, " Potential for Low Fracture Toughness and Lamellar Tearing on PWR Steam Generator and Reactor Coolant Pump Supports," and ASME Code Section III, Summer 1977 Addenda, Subsection NC. The metallurgical characterization of these materials, with respect to their fracture toughness, was developed from a review of how these materials were fabricated and what thermal history they experienced during fabrication. The metallurgical characterization of these materials, when correlated with the data presented in NUREG-0577 and the Summer 1977 Addenda of the ASME Code Sec-tion III, provided the technical basis the staff's for evaluation of compliance with GDC 51. On the basis of its review of the available fracture toughness data and material ~ fabrication histories, and the use of correlations between metallurgical charac-teristics and material fracture toughness, the staff concludes that the ferritic components in the Braidwood Units 1 and 2 containment pressure boundary meet the fracture toughness requirements that are specified for Class 2 components by the 1977 Addenda of Section III of the ASME Code. Compliance with these Code requirements provides reasonable assurance-that the Braidwood Units 1 and 2 con-tainment pressure b_oundary will behave in a nonbrittle manner, the probability of rapidly propagating fracture will be minimized, and the requirements of GDC 51 are satisfied. Therefore, Outstanding Item A(4) is closed. 6.3 Emergency Core Cooling System 6.3.2 Evaluation of Single Failures The SER documented a commitment by the applicant to install an automatic system to ensure adequate minimum centrifugal charging pump flow to prevent deadheading that could damage the pump in the ECCS mode of operation. The staff accepted the commitment subject to confirmation of the applicant's design. The details of the modification were submitted in a letter dated September 16, 1983. The modification is currently scheduled to be completed before startup at Braidwood. In the original design, the centrifugal charging pump bypass lines that are designed to protect the pumps from overheating at high pressures and low flows are automatically isolated by a safety injection (SI) signal. The bypass would Braidwood SSER 1 6-5

4 1 not normally be needed for pump protection during LOCA events because reactor system pressures would be relatively low. Isolating the bypass lines provides an additional 8 pounds per second of fluid for core cooling. For certain event sequences, such as a stuck-open pressurizer relief valve that is subsequently isolated, charging pump deadheading might result with the orig-inal design. To protect the charging pumps from possible damage, the applicant will remove the SI closure signal from the bypass line valves, relocate the valves from a series to parallel arrangement so that single valve failure will not result in loss of bypass flow, and provide for automatic closure of the bypass lines on low reactor system pressure (1400 psig) and SI. The valves 4 ~ i will be automatically opened if the reactor system pressure increases above 2000 psig. Closure of additional isolation valves in the bypass lines will occur before switching to the' recirculation mode of ECCS operation. Protec-tion from deadheading would not be required in the recirculation mode because either the course of the accident or operator action will provide for reactor system depressurization before the refueling water storage tank could be emptied. 1 l The design details of the permanent modification have been reviewed and approved by the staff (see Section 7.3.2.13 of this SSER). Therefore, Confirmatory Issue B(3) is considered closed. 6.3.4 Testing 6.3.4.1 Preoperational Tests The original SER indicated that the staff would confirm the sump design by re-l viewing the results of ECCS testing. Staff review of the test results for Byron Station is documented in Inspection Report No. 50-454/84-24 and No. 50-455/84-17, transmitted by letter dated June 12, 1984. Because the sump design is identical for the Byron and Braidwood Stations, the staff considers Confirmatory Issue B(7) closed for Braidwood Station, Units 1 and 2. Braidwood SSER 1 6-6

4 7 INSTRUMENTATION AND CONTROL 7.2 Reactor Trip System 7.2.2 Specific Findings 7.2.2.5 Response Time Testing By letter dated September 28, 1983, the applicant indicated that the report entitled "The Use of. Process Noise Measurement to Determine Response Character-istics of Protection Sensors in U.S. Plants" submitted by E. P. Rahe, Westing-house, to D. G. Eisenhut, NRC, on August 15, 1983, provides justification for the use of this technique for response time testing. The staff has reviewed the Westinghouse report that describes the test method and provides the results of tests conducted at operating reactors from 1977 through 1982 using this technique. The staff concludes that the use of process noise measurements will provide an acceptable means to fulfill the requirements for response time testing as specified in the plant Technical Specifications. Therefore, License Condition A(3) is considered closed. 7.3 Engineered-Safety Features Systems 7.3.2 Specific Findings 7.3.2.2 Compliance With IE Bulletin 80-06 The SER stated that a test will be conducted to verify that the actual installed engineered safety features (ESF) instrumentation and controls ane in compliance with the requirements of IE Bulletin 80-06. Staff confirmation of test comple-tion is documented in Byron Inspection Report Nos. 50-454/84-70(DRP) and 50-455/ 84-48(DRP), transmitted by letter dated November 26, 1984. Because the equip-ment design is identical for Byron and Braidwood, Confirmatory Issue B(8) is considered closed. 7.3.2.13 Charging Pump 00adheading In the SER, the staff stated that the applicant had provided a conceptual out-line for a design change that would provide open and close signals for a solenoid-operated valve in each charging pump miniflow line to prevent pump deadheading. These signals would be developed from four redundant reactor coolant. system wide range pressure transmitters combined in two-out-of-four logic for both high (open) and low (close) pressure signals taken in coinci-dence with the safety injection signal. The staff reviewed the conceptual information and found it acceptable with a confirmatory detailed design review pending. The licensee has submitted detailed design information (logic diagrams and electrical schematics) for the staff's review and has provided appropriate revisions covering the modifications in Amendment 45 to the FSAR. The staff Braidwood SSER 1 7-1

has reviewed the detailed information provided and finds the design modifica-tion acceptable. Therefore,. Confirmatory Issue B(3) is considered closed. 7.6 Interlock Systems Important to Safety 7.6.2 Specific Findings 7.6.2.7 TMI Action Plan Item II.K.3.1, Installation and Testing of Automatic Power-Operated Relief Valve Isolation System By letter dated December 27, 1983, the applicant referred to a Westinghouse generic report (WCAP-9804) in addressing TMI Action Plan Item II.K.3.2, " Report on Overall Safety Effect of Power-Operated Relief Valve (PORV) Isolation System." The applicant asserted that the generic report was applicable to Braidwood Units 1 and 2. The staff has made an independent assessment of the frequency of a small-break loss-of-coolant accident (SBLOCA) as a result of a stuck-open PORV or safety valve (SV). The similarity of the safety valves (Crosby Model HP-BP-86) that are used by many other Westinghouse plants, and based.on the EPRI test results (see EPRI-2628-SR), the staff estimates the failure rate of the Braidwood SVs to be similar to that of other Westinghouse plants--I x 10-2 per demand. Because the design and operation of Braidwood is similar to that of Zion Nuclear Plant, the staff expects a similar SV challenge. frequency at Braid-wood similar to that at Zion. Therefore, the staff estimate of SBLOCA frequency resulting from a stuck-open SV is 3 x 10 4 per reactor year, the same as for Zion. On the basis of the similarity of the Braidwood PORVs to those used by Zion, and the similarity.in design and operation of the two plants, the staff estimates an SBLOCA frequency resulting from a stuck-open PORV of 1.5 x 10 3 per reactor-The SBLOCA frequency is within the range of the SBLOCA frequency of 10 4 year. to 10-2 per reactor year given in WASH-1400. The staff has, therefore, deter-mined that the requirements of TMI Action Plan Item II.K.3.2 are met with the existing PORV, SV, and high pressure reactor trip setpoints. According to the criteria set forth in the clarification of TMI Action Plan Item II.K.3.2, there is no need for an automatic PORV isolation system. Therefore, Confirmatory Issue B(10) dealing with TMI Action Plan Item II.K.3.1 is considered closed. Braidwood SSER 1 7-2

8 ELECTRIC POWER SYSTEMS 8.2 Offsite Power System 8.2.4 Adequacy of Station Electric Distribution System Voltages The staff reviewed Pre-operational Test No. BWPT-AP-16, Sections 9.1 and 9.2 I (see Inspection Report No. 50-456/86024). This test measured and documented the loaded voltage levels of the safety-related buses from the 4-kV level down to the 120-V level. The test evaluation compared measured engineered safety feature (ESF) bus voltages with the ESF. bus voltages predicted by a computer model. Acceptance criteria were met in that measured bus voltages fell within i 3% of the predicted values. Therefore, Confirmatory Issue A(6) is considered closed. 8.4 Other Electrical Features and Requirements for Safety 8.4.4 Physical Identification and Independence of Redundant Safety-Related Electrical Systems A site visit was conducted on May 21-23, 1985, at Braidwood Station, Units 1 and 2 to view the installation and arrangement of electrical equipment cables. During this visit, specific issues identified during the Braidwood Station Construction Appraisal Team (CAT) inspection.were discussed. The CAT inspec-tion revealed several items regarding physical separation, particularly between Class 1E and non-Class 1E cables. Commonwealth Edison Company has established the separation criteria between redundant Class 1E raceways in accordance with RG 1.75 for Braidwood Station. However, the applicant has established separation criteria of non-Class 1E from Class 1E raceways that deviates from the specific separation distances detailed in RG 1.75. Acceptability of the applicant's lesser separation. distance with its bases and justification were not specifically addressed in the Braidwood SER. Therefore, the staff has performed the following evaluation of the applicant's separation criteria of Class IE cables from non-Class 1E cables to eliminate any further differences in interpretation of separation requirements in this area. The applicant instituted a test program conducted by Wy'le Laboratories and per-formed calculations and analyses to justify lesser separation distances. By letter dated August 6, 1985 (A. D. Miosi, Ceco, to H. R. Denton, NRC), the applicant submitted the test results with its associated information, and the analysis on the separation criteria. The purpose of these tests was to establish a basis of analysis that could be applied in justifying a lesser physical separation distance. Any lesser separation distance than the separation criteria specified in RG 1.75 must be Gstablished by the test results. Braidwood SSER 1 8-1 i

To perform a test program to verify the adequacy of the raceway separation criteria, it was necessary to define the worst-case electrical failure that could be postulated to occur in a raceway. The Braidwood raceway separation test program was based on the following failure assumptions: (1) The cable or equipment in the circuit develops an electrical fault that is not cleared as a result of the postulated failure of the primary over-current protective device. (2) The fault current used was the radiation monitoring system (RMS) value that produces the maximum possible credible heating effect without trip-ping the breaker by magnetic force. (3) Load current effects from other loads on the same circuit was not consid-ered to cause the next higher level overcurrent device to trip. The worst-case failure of a cable for which the electrical separat, ion criteria must protect cables in an adjacent raceway is a sustained overload condition in which the magnitude of the current is such that the. cable would be able to sustain the overload for a significant length of time. This condition would allow the cable to generate the greatest amount of heat over a period of time and, therefore, has the greatest potential for causing damage to nearby cir-cuits. On the other hand, if the cables were exposed to the maximum short circuit current available at the bus, the higher fault current would lead to rapid clearing of the fault by a breaker. This condition causes less energy to be generated to the ambient temperature and hence results in less temperature rise in the adjacent raceway. For the purpose of the test, the cables were subjected to the overload currents for the length of time it took to open the circuit through failure of the cable conductors. This is considered to be a very conservative test since no credit was taken for any current interrupting devices operating in the circuit. The purpose of these tests was to establish an analytical basis for demonstrat-ing the minimum acceptable separation distance. Any separation distance less than the separation criterion specified in l'G 1.75 shall also be established by the test program. In selecting the test configuration, the primary concern was to ensure that the quantity and types of raceway and cable arrangements tested would satisfac-torily represent actual plant configurations and provide a basis for applying the results of the testing to similar configurations that were not. tested. Using this criterion, the following representative configurations were selected for the staff's evaluation. i (1) Separation distances of 1 foot (12 inches) vertical and 3 inches horizontal between safety-related and nonsafety-related raceways The staff's analysis concludes that fire or failure resulting from elec-trical faults induced in nonsafety-related cables in a raceway would not cause electrical failure of safety-related cables in a raceway located 12 inches directly above or below or 3 inches horizontally away from the nonsafety-related raceway. The analysis was based on actual results of tests performed to establish electrical separation distance. The cable Braidwood SSER 1 8-2 l I -

failures addressed in the establishment of separation distance in this analysis are those that were induced by an electrical fault within the nonsafety-related cable only. The raceway configuration chosen'for the test was one in which an open top cable tray-containing nonsafety related power cables was located 2 inches below a cable tray containing safety-related cables. The con-figuration also included a 2-inch flexible steel conduit containing safety-related cables running vertically, separated 2 inches horizontally from the nonsafety-related cable tray. The value of overload current that was selected for the test was approxi-mately 6.5 times.the rated current overload value for the given cable size. This value is based on the fact that a stalled motor draws about 6.5 times rated current. The current of a stalled motor was selected because it was considered a credible overload current that may occur during normal operat-ing conditions. The target cables in the upper cable tray and vertical flex. conduit were continually energized during the test with their rated current. The actual value of overload current that the faulted cables were exposed to during the test were 462A for 3/C #2AWG-(American wire gauge), 737A for ( 3/C 1/0, and 2070A for 3/C 350 MCM (thousand circular mils). These values were based on 6.5 times rated current overcurrent test. The length of time for which each of the faulted cables were energized with the over-load was very conservative. As stated previously, the overload current value was selected because it was representative of the test current a stalled motor may draw. This was evaluated as the most credible cause of a sustained overload current. In reality, the motor windings would even-tually short together and result in~a full short circuit that would be of a magnitude high enough to trip upstream circuit breakers even if a feeder breaker were to fail. Calculated fault currents.were 4600 amperes for the size 2 AWG cable, 5400 amperes for the size 1/0 AWG cables, and 6700 am-peres for the size 350 MCM cables short circuit test. The test results demonstrated that these fault current values caused relatively minor damage to the~ fault cable insulation, particularly when compared with the extreme degradation incurred with the lower (6.5 times rated current) overcurrent tests.. The major reason for the decreased insulation system i damage was the fact that the' conductor circuits open much faster at higher current values. At the completion of each cable test, functional tests for the target l cables (consisting of the insulation resistance test, high potential test, l overcurrent test, and post-test functional test) were performed. The target cables passed these tests in accordance with the acceptance criteria and cable manufacturer's specification. The results of Wyle Laboratories' investigation (Test Report No. 46511-3) discussed in the August 6, 1986, submittal (A. D. Miosi, Ceco, to H. R. Denton, NRR) demonstrate that all of the target cables in upper cable tray-(located 12 inches above the cable tray containing the faulted cable) and in the vertical conduit (located 2 inches horizontally away from faulted i cable tray) maintained their integrity to conduct specified current and voltage before, during, and after the fault specimens were subjected to l Braidwood SSER 1 8-3

the overload currents. The target cables passed the post-functional insu-lation resistance tests at 500. volts dc and the high potential withstand tests at 22M volts ac. The temperature measured on the target caDies in the upper ~caole tray and in the flex conduit was much less than the tem-perature for which the cables are continuously rated and significantly less than the emergency temperature rating of 130*C of the power, control, and instrument cables. The staff has reviewed the results of tests conducted by Wyle Laboratories and the applicant's analysis of the resulting report (letter from A. D. On the basis Miosi, Ceco, to H. R. Denton, NRR, dated August 6, 1986). of its review, the staff concludes that the separation distance of 12 inches vertical and 3 inches horizontal between safety-related and nonsafety-related raceway is adequate to prevent a fault in nonsafety-related cable causing failure of safety-related cables and is, therefore, ) 1 acceptable. Separation of a safety-related cable in free air in contact with a raceway (2) containing a nonsafety-related cable and of a nonsafety-related cable in free air in contact with a raceway containing a safety-related cable The purpose of this analysis and test was to demonstrate that fire or failure resulting from electrical faults induced in nonsafety-related cables in free air or in raceway would not cause electrical functional failure of safety-related cables in raceway or in free air respectively. This configuration consisted of a test between two horizontal, rigid steel conduits, and various free-air instrumentation cables. The faulted cable was a 3/C 500 MCM routed in a rigid steel conduit. Three target cables were located in a 1-inch rigid steel conduit in contact with the conduit containing the faulted cable. Three other target cables, respectively, were mounted in free air in contact with the conduit of the faulted cable. This configuration test demonstrated the adequacy of separation design that (1) two horizontal, rigid conduits are physically ceparated by zero inches vertically when a worst-case electrical fault occurs in the lower conduit; or (2) free-air cables are physically separated from a horizontal rigid steel conduit by zero inches horizontally when a vorst-case electrical fault occurs in the conduit. All instrumentation cables used in both safety-related and nonsafety-related applications were rated for 600 volts with insulation tested to a minimum of 1500 volts with an overall Jacket and were applied in circuits with a system voltage less than 30 volts. I Control cables were applied in circuits with a system voltage of either 120 volts ac or 125 volts dc. Low voltage power cables were applied in circuits with a system voltage of 480 volts ac. Control and low power cables had insulation rated at 600 volts. The cable was also tested to show that it could withstand voltage transients up to 1500 volts. Medium voltage power cables were applied in circuits with system voltages of l 4160 or 6900 volts. These were required to have insulation rated at 5 and 8 kilovolts respectively. The cable was also tested to show that it could withstand voltage transients of up to 16 and 22 kilovolts respectively. Therefore, there was a conservative design margin in the cable to ensure adequate isolation from voltage transients in the nonsafety-related l circuit from adversely affecting a safety-related circuit. l Braidwood SSER 1 8-4

For the purpose of the verification test, it was assumed that the circuit breaker feeding the overloaded cable fails to trip and the overcurrent would persist in the cable. The fault current considered the most credible i severe overload condition that the cable may see during plant operation was that resulting from a motoc failing to start but continuing to draw locked rotor current as described above. The actual test current values were selected from the largest motor that was fed with a 500 MCM 600-volt cable at either Byron or Braidwood. This motor was a 250 horsepower motor that had a locked rotor current of approximately 1700 amperes. If the voltage drop is taken into consideration, the actual current that would be seen by the cable is approximately 1300 amperes. The overcurrent test, therefore, consisted of energizing the 500 MCM size to 1300 amperes for 1 hour and 1700 amperes until the cable open circuited. The two-step over-current test was selected to cimulate a worst-case condition by energizing the cables with a fault current that causes the cable to generate consider-able heat but not cause an open circuit, and then jump the fault curr3nt to a value that would eventually open circuit the cable. The 1-hour time limit on the 1300-ampere portion of the test was considered conservative because a stalled motor would be alarmed and deenergized long before 1 hour had elapsed. Alternatively, the motor winding would short together l and result in a full short circuit that would be interrupted by the up-l stream breakers. l The target cables were energized continually during the test. The target l cables passed pre-and post-functional tests that consisted of both insula-l tion resistance and high potential withstand tests. l l As previously stated, the primary objective in the selection of the test i configuration was to ensure that the quantity of raceways and cable l arrangements tested would satisfactorily represent actual plant configu-rations and provide a basis for applying the results of the testing to similar configurations that were not tested. l The results of these tests performed (Test Report No. 17769-1 by Wyle Laboratories) indicate that all of the target cables maintained integrity to conduit specified current and voltage before, during, and after the fault specimen was subjected to the overload current. At the completion of each cable test, the functional tests were performed for the target cables. The target cables passed these tests in accordance with the acceptance criteria and the cable manufacturer's specification. l The staff has reviewed the results of Wyle Laboratories Test Report No. 17769-1 dated August 23, 1985, and the applicant's analysis of these l configuration tests. On the basis of its review, the staff concludes that I it is acceptable for (1) safety-related cables in free air to come in con-tact with a raceway containing nonsafety-related cables and (2) nonsafety-related cable in free air to come in contact with a raceway containing t safety-related cables. This analysis has demonstrated that safety-related cable will not be degraded below an acceptable level as a result of the reduced separation as specified in the FSAR. Confirmatory Issue A(7) is considered closed. Braidwood SSER 1 8-5 l

9 AUXILIARY SYSTEMS 9.1 Fuel Handling and Storage 9.1.5 Overhead Heavy Load Handling System As a result of Generic Task A-36, " Control of Heavy Loads Near Spent Fuel," NUREG-0612, " Control of Heavy Loads at Nuclear Plants" was developed. Follow-ing the issuance of NUREG-0612, a generic letter dated December 22, 1980, was sent by D. G. Eisenhut to all licensees of of all operating reactors, applicants for operating licenses, and holders of construction permits requesting that re-sponses be prepared to indicate the degree of compliance with the guidelines of NUREG-0612. The responses were to be made in two stages. The first response (Phase I, Section 5.1.1 of NUREG-0612) was to identify the load handling equip-ment within the scope of NUREG-0612 and to describe the associated general load handling operations such as safe load paths; procedures; operator training; special and general purpose lifting devices; the maintenance, testing, and re-pair of equipment; and the handling equipment specifications. The second re-sponse (Phase II) was intended to show that either single-failure proof handling equipment was not needed or that single-failure proof equipment had been pro-vided. This supplement to the SER contains the staff's evaluation of both Phase I and Phase II. By letter dated December 22, 1980, the applicant was requested to review its provisions for handling and control of heavy loads at Braidwood to determine the extent to which the guidelines of NUREG-0612 were satisfied and to commit to mutually agreeable changes and modifications that would be required to fully satisfy these guidelines. The staff and its consultant, Idaho National Engineering Laboratory (INEL) have reviewed the applicant's submittals for Byron and Braidwood Stations. As a result of its review, INEL has issued a techr.ical evaluation report (TER) (Appendix J). The staff reviewed the TER and concurs with its findings that the guidelines of NUREG-0612, Section 5.1.1 have been satisfied. The staff concludes that Phase I of NUREG-0612 for Braidwood is acceptable. The staff has eliminated the need for further effort regarding compliance with the criteria of Phase II (Sections 5.1.2 through 5.1.6) of NUREG-0612 on the basis of the Phase I (Section 5.1.1) compliance and Phase II reviews to date. These Phase II reviews consisted of an evaluation of the responses for_ twelve randomly selected operating plants that formed a pilot program. The staff de-termined from these reviews that the majority of risk associated with the han-dling of heavy loads has been resolved by implementation of Phase I, and, in addition, no further heavy loads handling concerns were identified from the pilot progran reviews. It is therefore concluded that the objective identified in NUREG-0612 for providing " maximum practical defense in depth" is satisfied without the need for further action regarding Phase II. Braidwood SSER 1 9-1

9.2 Water Systems 9.2.2 Reactor Auxiliaries Cooling Water Systems The SER indicated that during the limiting mode of plant operation, a simulta-neous loss-of-coolant accident (LOCA) in one unit and safe shutdown of the other, the component cooling water system (CCWS) is,s,plit on receipt of an engineered safety features actuation signal (ESFAS). For purposes of clarifi-cation, this splitting is performed manually by the operator some time after receipt of an ESF S. TN staff's original approval was based on this design. 9.3 Process Auxiliaries 9.3.2 Process and Postaccident Sampling System TMI Action Plan Item II.B.3, Postaccident Sampling Capability By the submitta) of Amendment 38 and letters dated August 26 and October 26, 1982, the applicant provided a description of systems, equipment, and proce-dures to be used for sampling the reactor coolant and the containment atmosphere following an accident resulting in core degradation. The applicant has also providei! Information on methods of transporting samples for offsite analyses. The postaccident sampling syste.a (PASS) provides the capability to obtain and analyze samples within 3 hours of the time a decision is made to sample. Sam-pies can be obtained from the reactor coolant system, containment sumps, and containment atmosphere under accident conditions. Provisions are incorporated to obtain grab samples for offsite analyses. The applicant also provided_a description of radiochemical analyses capabilities including provisions to identify and quantify radioactive isotopes (noble gases, iodine, and cesium isotopes, and nonvolatile isotopes). Analyses capabilities are also incorporated for dissolved gases, chloride, and boron concentrations in liquid samples. The PASS also provides the capability to measure hydrogen con-centration in the containment atmosphere. Sample lines are routed to an access-ible area and shielded to protect operators. An isolated auxiliary system is not required to be operational in order to use the PASS. Furthermore, all valves in the high radiation sampling system are designed for the environment in which they need to operate. The PASS is also capable of performing inline analysis of hydrogen, dissolved oxygen, and chloride in the primary coolant during r.ormal and accident conditions. Provisions will also be made to purge the sample lines for reducing plateout, for minimizing sample loss or distortion, for preventing blockage of sample lines for disposal of samples, and for passive flow restric-tions. Sufficient shielding will be provided to meet the requirements of GDC 19 in Appendix A to 10 CFR Part 50, assuming the source term defined in RG 1.4, " Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss-of-Coolant Accident for Pressurized Water Reactors." The applicant pro-vided an alternative backup power source to the PASS for use during loss of offsite power. In addition, inline monitoring of dissolved oxygen at less than 0.1 ppm (parts per million) can be made if the reactor coolant chlorides concen-tration is greater than 0.15 ppm. On the basis of its review of Amendment 38 and letters dated August 26 and October 26, 1982, the staff concludes that the postat.cident sampling system is acceptable, except for Clarification Items (2), (10), and (11) identified in TMI Action Plan Item II.B.3. Braidwood SSER 1 9-2

By letters dated December 11, 1983, and January 5 and May 4, 1984, the appli-cant provided additional information regarding Clarification Items (2), (10)', and (11). The applicant provided a procedure for estimating the degree of reactor core damage based on the Westinghouse Owners Group (WOG) generic methodology (" Post-Accident Core Damage Assessment Methodology," Revision 1, March 1984), using measured and predicted postaccident radionuclide concentration from failed fuels. T'he procedure takes into consideration other physical parameters such as reac-tor core temperature data, reactor water level, sample location, containment radiation levels, and hydrogen concentrations. The staff concludes that these provisions meet Clarification Item (2) and are, therefore, acceptable. The range, accuracy, and sensitivity of the radiological and chemical analysis was given by the applicant. Information was provided on the applicability of these procedures and instruments in the postaccident water chemistry and radia-tion environment. Containment atmosphere will be sampled and analyzed for hydrogen, oxygen, and gamma spectrum every 6 months. The postaccident sampling system is the same as the one used for routine sampling operations. The appli-cant has developed a special retraining program for all radiation chemistry technicians that will be administered every 6 months. The staff concludes that these provisions meet Clarification Item (10) and are, therefore, acceptable. The licensee has addressed provisions for purging to ensure samples are repre-sentative, size of sample line to limit reactor coolant loss from a rupture of the sample line, and ventilation exhaust from PASS filtered through charcoal absorbers and HEPA filters. The containment atmosphere sample line is heat traced to aid in obtaining representative samples. The staff concludes that these provisions meet Clarification Item (11) of Item II.B.3 and are, there-fore, acceptable. The staff concludes that all eleven Clarification Items are met and the post-accident sampling system is acceptable. The applicant committed to have the postaccident sampling system operational before initial criticality. There-fore, License Condition B(2) is no longer applicable. 9.3.5 III.D.1.1 Integrity of Systems Outside Containment Likely To Contain Radioactive Material 9.3.5.1 Summary Description The applicant has provided a description of the program designed to reduce leakage from systems outside containment that would or could contain primary coolant or highly radioactive fluids during a serious transient or accident. The applicant's program will be initiated during the preoperational test phase. The applicant has indicated that, before loading fuel, all systems or portion. of systems constructed in accordance with ASME Section III will be hydrostati-cally tested to 125% of the system's design pressure. In the case of gaseous systems, a pneumatic type of pressure decay test at 125% of system's design pressure will be performed. All systems in the leak reduction program will be tested before initial plant startup durin0 the Preoperational Test Program. Braidwood SSER 1 9-3

During these tests, system walkdowns will be conducted by the System Test Engi-neer and deficiency reports will be generated for leakage and defective compo-nents. In addition to the individual system tests, integrated types of tests such as Integrated Hot Functional (IHF) and Emergency Core Cooling System (ECCS) Full Flow Tests will be conducted. During these integrated tests, additional system walkdowns will be conducted for vibrational testing and inspection of piping thermal expansion. Deficiency reports will also be generated during these walkdowns. The applicant has indicated that, at the time Braidwood Unit I reaches full-power operation, a report will be submitted to the NRC staff detailing all re-corded leakage and, as the direct result of the evaluation of this leakage, all preventive maintenance will be performed. The report will also identify general leakage criteria to be applied during the first fuel cycle as the basis for instituting corrective action in the form of preventive maintenance. Levels of leakage will be kept as low as practicable through this program's.:ommitment to generate work requests for all practicable repairable leakage problems. The applicant has committed to reviewing leakage problems presenting as low as reasonably achievable (ALARA) concerns and to refine the leakage criteria over time as more information is accumulated through inspections. Thus, the criteria can be revised to incorporate new modifications and techniques designed to keep leakage as low as practicable. Before the start of the second fuel cycle, the applicant has committed to revising the general criteria based on the experience gained during.the first operating cycle on Braidwood Unit 1. These revised criteria will be used as the basis for the long-term leakage monitoring program on Braidwood Units 1 and 2. The applicant has committed to leak testing the following systems or portions of a system that could contain highly radioactive gases or fluids: (1) chemical and volume control (2) containment spray (3) radioactive waste gas (4) offgas, including hydrogen recombiners (5) residual heat removal (6) safety injection (7) fuel pool cooling and cleanup (8) process sampling (9) auxiliary building equipment drains and floor drains The applicant has committed to performing integrated leak tests at least during each refueling outage on each system, or portions of systems, that could potentially contain highly radioactive fluids or gases. Station surveillance and procedures will be used to (1) monitor the leak testing of piping so that the appropriate lines are examined at the required intervals (2) direct leak tests examinations so that systems are tested at approxi-mately operating pressures or higher (3) align systems so that all piping tested is properly pressurized Braidwood SSER 1 9-4

(4) identify lines that contain gases requiring pressure decay and/or metered makeup testing (5) quantify results of leakage examinations (6) initiate corrective action Systems or portions of systems that will be excluded from-the leak. reduction program include (1) chemical and volume control (a) chemical mixing tank and associated piping (b) boric acid addition portion of the system (c) resin fill tank and associated piping (2) containment spray (a) spray additive tank and associated piping (t) 5-inch recirculation line from containment spray pumps back to esfueling water storage tank (3) radioactNe waste gas (a) drain lines from gas decay tanks (b) relief lines from gas decay tanks ( (c) Unit 2 tie-ins until these lines come into service l (4) off gas (a) calibration lines to hydrogen analyzers (b) steam system portion (c) Unit 2 tie-ins until these lines come into service (5) safety injection (a) refueling water storage tanks and associated piping (b) accumulator fill lines (c) leakoff lines from recirculation line isolation valve caps (6) -boric acid processing (7) boron thermal regeneration Braidwood SSER 1 9-5

(8) Unit 2 fuel pool cooling and cleanup tie-ins until the time they become operational (9) process radiation monitoring (10) process sampling except for (a) pressurizer steam and liquid sample lines, reactor coolant sample lines, and residual heat removal heat exchanger sample lines (b) chemical and volume control system demineralizer outlet sample line (c) letdown heat exchanger sample line (11) reactor building equipment drains except for a line that extends from the reactor coolant drain tank to the waste gas compressor (12) reactor building floor drains (13) primary containment purge line (14) auxiliary building equipment drains and floor drains, except for casing drain lines from (a) containment spray pumps (b) safety injection pumps (c) residual heat removal pumps (d) chemical and volume control pumps (15) solid radwaste disposal (16) chemical radwaste disposal In addition to the leak reduction program, the applicant has stated that all Class 1, 2, and 3 systems will be leak tested at prescribed intervals, in ac-cordance with the requirements of the 1980 Edition, with addenda through the Winter of 1981 Addenda, of Section XI of the ASME Code, " Rules for Inservice Inspection of Nuclear Power Plant Components," as described by Braidwood Sta-Portions of Class 1, 2, tion's Inservice Inspection and Testing Program Plan. and 3 systems excluded from this leakage program will be leak tested through the inservice inspection program. The applicant has committed to document leakage observed during the performance of inservice tests and to preparing a work request to repair this leakage. Work requests of this type will be assigned a high priority by the applicant and designated as an ALARA concern. A review for possible modification to re-duce leakage in the future will be initiated. l Braidwood SSER 1 9-6

i Piping and components that make up the containment penetrations will be tested during every outage as part of the 10 CFR 50, Appendix J, leakage testing pro-gram for Type A and Type B testing. Type C testing will be' performed in accor-dance with the Technical Specifications. 9.3.5.2 Evaluation and Findings The staff has reviewed the applicant's leak reduction program. The' program as presented meets the requirements of TMI Action Plan Item III.D.1.1 except that initial-leak-test results have not been provided. With their submittal, the program will conform to III.D.1.1. Therefore, Confirmatory Issue B(10) concern-ing TMI Action Plan Item III.D.1.1 is considered closed. 9.5 Other Auxiliary Systems 9.5.4. Emergency Diesel Engine Fuel Oil Storage and Transfer System 9.5.4.1 Emergency Diesel Engine Auxiliary Support Systems (General) I The SER stated that the applicant had committed to implement certain procedures for no-load operation of the diesel generators. Specifically, if diesel troubleshooting continues for 3.to 4 hours, the diesel would be loaded to at least 25% of full load for 1 hour. By letter of April 23, 1984 (T. R. Tramm to H. R. Denton), the applicant clari-fied its intentions regarding load testing of the diesels. The applicant has ascertained from its diesel manufacturer, Cooper-Bessemer, that troubleshooting involving repeated starts is no worse than continuous no-load operation. There-fore, the applicant intends to load the diesel to 25% for 1 hour after 8 hours of no-load operation, regardless of the number of starts. The staff finds this procedure acceptable. l The SER stated that the controls and monitoring instrumentation are installed ( on a freestanding floar-mounted panel separate from the engine skids, and i located in a vibration-free floor area. The response to NRC Question 040.94 l states that the design of the. floor slab is such that the slab mass has been j l proportioned to the equipment mass to minimize vibration and impact loads. The staff does not consider this to be a vibration-free floor area. By letter dated October 16, 1984, the applicant has provided an evaluation to resolve this concern, which is being reviewed by the staff. Until the review is completed, a license condition has been added that requires the applicant, before startup after the first refueling outage, to dynamically qualify the controls and monitoring instrumentation for their present location, or install them on a freestanding floor-mounted panel in such a manner (including the use of vibration-isolation mounts if necessary) that any induced vibrations will not result in a cyclic fatigue failure for the expected life of the instrument. The staff has determined that exemptions to GDC 13 and 17 are required. GDC 13 requires that instrumentation and controls shall be provided to monitor vari-ables and systems over their anticipated ranges for normal operations, antici-pated operational occurrences, and accident conditions. GDC 17 requires that provisions be included to minimize the probabili_ty of losing electric power Braidwood SSER 1 9-7 = _

from any of the remaining supplies as a result of, or coincident with, the loss of power generated by the nuclear power unit, loss of power from the transmis-sion network, or loss of power from the onsite electric power supplies. The staff does not expect that there will be enough induced vibrations before the first refueling outage to cause cycle fatigue failure of the instruments. Therefore, the staff concludes that the exemptions to GDC 13 and 17 for the first cycle of operation will not endanger life or property or the common defense and security and is otherwise in the public interest. Braidwood SSER 1 9-8

i l 10 STEAM AND POWER CONVERSION SYSTEM 10.3 Main Steam Supply System 10.3.3 Secondary Water Chemistry In the SER, the staff found the applicant's secondary water chemistry monitor-ing and control program acceptable, and stated that the license would be con-ditioned to require that this program be carried out. The staff has since decided to incorporate this program into Section 6 (Administrative Controls) of the Technical Specifications which will be issued with the licenst

Thus, License Condition A(5) is no longer necessary.

l i l Braidwood SSER 1 10-1

12 RADIATION PROTECTION 12.5 Operational Radiation Protection Program 12.5.1 Organization The original SER contained a conditional acceptance of the Station Health Physi-cist as the plant's Radiation Protection Manager (RPM), subject to him comple-ting a formal training program. By letter dated June 7, 1985, the applicant documented that'the training program had been completed by the-Station Health-Physicist except for attending the first refueling outage at Byron Station, Unit 1. The applicant has committed to having the Braidwood RPM attend either the first refueling _cutage at Byron Unit 1 or a previously approved training meeting between Byron and Braidwood health physics personnel to discuss both the positive and negative health physics aspects of the Byron refueling outage. The staff finds this approach acceptable. The staff finds that the Braidwood Station Health Physicist meets the require-ments of Regulatory Guide (RG) 1.8 and is acceptable as'the Braidwood RPM. Accordingly, Confirmatory Issue A(8) is closed. j Braidwood SSER 1 12-1

l l-l 13 CONOUCT OF OPERATIONS 13.3 Emeraency Planning 13.3.1 Background The staff's previous evaluation of the Generating Station Emergency Plan (GSEP) and the Braidwood Annex to the GSEP was published in a letter from the NRC to Commonwealth Edison Company, dated March 4, 1986. As this evaluation was never -issued in a supplement to the Safety Evaluation Report for Braidwood (NUREG-1002), it is included here in Appendix 13A. Subsequent to our initial review, the applicant submitted Revision 5 to the Generating Stations Emergency Plan (GSEP) dated July 1985, with an effective date of November 22, 1985. [The GSEP is a generic emergency plan applicable to all nuclear generating stations operated by Commonwealth Edison Company (Ceco). i The GSEP includes site specific annexes that contain more specific information relevant to each nuclear generating station.] In February 1986, the applicant submitted Revision 1 of the Braidwood Annex to the GSEP, dated March 1986. In addition, by correspondence dated May 20, 1986~, the applicant responded to staff concerns on the Braidwood Station emergency preparedness program that were docu- .mented in the March 4, 1986 letter. This section provides the staff's evaluation of the aforementioned additional information. r 13.3.2 Evaluation of the Emergency (Onsite) Plan l l 13.3.2.1 Assignment of Responsibility (Organization Control) Item requiring resolution: A formal letter of agreement should be executed between the applicant and the U.S. Coast Guard, which would specify the support to be provided by that l organization in response to an emergency condition at Braidwood Station. Revision 5 to the Generating Station's Emergency Plan (GSEP)~ indicates that the F Station Manager or his alternate is. responsible for promptly notifying the U.S. Coast Guard in the event of any oil or hazardous substance discharge into a l lake or waterway (Kankakee River, in the case of Braidwood Station), or radio-l active contanination of a lake or waterway under Coast Guard jurisdiction at levels requiring assistance to effect~ protective actions. The plan describes the manner in which the Coast Guard would be notified by the applicant, and states that the Illinois Department of Conservation would also contact the Coast Guard. The Coast Guard would then be responsible for officially closing the ~ affected waterway (s) to traffic. The staff has concluded that the applicant's GSEP adequately addresses the rea-l sons for and means of notifying the U.S. Coast Guard. Also, since the support Braidwood SSER~1 13-1 --- - - - - - - -=-

requested of the Coast Guard would not be beyond the realm of emergency assis-tance normally provided by that agency, a formal letter of agreement between the applicant and the U.S. Coast Guard is not required. This item is resolved. 13.3.2.2 Onsite Emergency Organization Item requiring resolution: The emergency plan should specify those individuals in the onsite and offsite emergency organizations who have the undelegatable authority to authorize emergency worker exposures in excess of normal regulatory limits. In correspondence dated May 20, 1986 the applicant has committed to provide the required clarification in Revision 6 of the GSEP. This plan revision will in-dicate that the individual in command and control of emergency response activi-ties will have the undelegatable. authority to authorize emergency worker expo-sures in excess of regulatory limits. It will also indicate that-the individ-ual should still attempt to consult with corporate medical staff or the sta-tion's Radiation Protection Superprior to authorizing the exposure. The staff has concluded that the applicant's clarification'is adequate. This item is resolved. The title of the person responsible for direct management of the station in the normal station organization has been changed to the Station Manager from Station Superintendent. The emergency duties and responsibilities of that individual .are unchanged. The onsite emergency organization is known as the Station Group. The position of Onsite Environs Director has been formally added to the Station ~ Group. The duties and responsibilities associated with that position are-ade-quately described in the GSEP. The applicant's provisions for onshift augmenta-tion, described in Section 4.2 of the GSEP, have been upgraded. All ten-Station Group directors are now required to report to their emergency duty stations following any Alert declaration, as well as following any Site Area or General Emergency declaration. To better ensure that onsite and offsite emergency re-sponse facilities become fully operational in a timely manner, the applicant has described a minimum staffing concept in Revision 5 to the GSEP. Essential activities associated with each facility are specified along with the appropriate minimum staffs needed to perform these activities. The applicant's minimum shift staffing levels continue to meet the objectives of Table B-1 in NUREG-0654, Revision 1. Therefore, these changes are acceptable. i 13.3.2.3 Emergency Classification System Items requiring resolution: The applicant must justify classifying emergencies based on the same level of equipment degradation, but differing in cause of degradation as either an l Alert or a Site Area Emergency. The plan must indicate that unmonitored gaseous radiation releases will be measured by environmental survey teams so that unmonitored releases will be appropriately classified. I l Braidwood SSER 1 13-2 l

__=.__.__ The applicant states that the damage to equipment resulting from a fire, un-planned explosion, or aircraft / missile impact potentially can be more severe than the damage caused by other means. Therefore, the applicant considers a Site Area Emergency to be more appropriate and conservative in the event of equipment degradation due to a fire, unplanned explosion, or aircraft / missile impact, while an Alert classification is sufficient for equipment degradation l resulting from any other occurrence. l The applicant has deleted listings from the GSEP of general conditions asso-l ciated with the four emergency classes (Notification of. an Unusual Event through General Emergency) that are also stated in Appendix 1 of NUREG-0654, Revision 1. The Braidwood Annex to the GSEP contains a table of station-specific Emergency l Action Levels (EAls) that are based on various plant sensors, onsite and offsite radiation monitoring information, and various natural phenomena. These station specific EALs -would be used for rapid classification of emergency situations at Braidwood Station. This item is resolved. l The second concern addressed the need to indicate the possibility of an unmor.i-i tored radioactive release. Such a release would have to be estimated from field measurements taken by environmental survey teams so that the emergency condition would be appropriately classified. The applicant's GSEP contains a description 'of its environmental survey teams, whose duties during a radioactive release from the plant include: dose rate surveys; air sampling; and soil, water,.and vegetation sampling. The applicant's procedures for the Environmental Director (ED-series) include methodologies for estimating radioactive release rates based on survey results from the environmental survey teams. The estimated release rates can be compared to the appropriate EALs. Thus, there is sufficient assur-- ance that abnormal situations involving unmonitored radioactive releases can be properly classified through the use of measurements taken by environmental sur-vey teams. This item is resolved. 13.3.2.4 Notification Methods and Procedures Items requiring resolution: A followup message form and provisions for periodically transmitting adequate followup messages to state and local authorities should be ' developed. Revision 5 to the GSEP specifies that after any Alert, Site Area Emergency, or General Emergency declaration, f.ilowup messages to State and local authorities will be transniitted at least every sixty minutes, which is an acceptable fre-quency. Rather than develop a standardized followup message form for use by persons at all its nuclear generating stations and the associated State and local emergency response centers, the applicant has implemented the acceptable alternative of procedurally specifying, by position title, those persons in the Technical Support Center (TSC), Corporate Command Center (CCC), and Emergency Operations Facility (EOF) who are responsible for followup message preparation, approval, and transmittal. Messages would be transmitted from that facility where command and control of emergency response activities is located. Follow-up messages would be documented in the same manner as would other telephone con-l versations. Appropriate procedures adequately address the content of periodic followup messages in a manner consistent with the GSEP. This item is resolved. l l Braidwood SSER 1 13-3 l l

13.3.2.5 Protective Response Item requiring resolution: The applicant's evacuation time estimate study for Braidwood Station's plume exposure pathway EPZ must be completed. The applicant submitted Revision 1 to this study, dated January 1986, in Febru-ary 1986. A review ~of the applicant's evacuation time estimate study has been completed by the staff cnd by Dr. Thomas Urbanik, an NRC consultant from the . Texas Transportation Institute of Texas A&M University. It has been concluded that the applicant's revised study is in accordance with the guidance of NUREG-0654, Revision 1, and is acceptable. This item is resolved. 13.3.2.6 Drills Item requiring resolution: The Plan should indicate that communications drills between Braidwood Station and the appropriate emergency response organization (s) in Indiana will take place quarterly. As indicated in March 15, 1985 correspondence from the applicant, there is an agreement between the Illinois Emergency Services and Disaster Agency (ESDA) and the Indiana Department of Civil Defense and Emergency Management that states that ESDA is responsible for notifying Indiana in the event that the ingestion pathway EPZ is impacted by a radioactive release from Braidwood Station. In addition, the Illinois Plan for Radiological Accidents - Volume 1 indicates that the State of Illinois is responsible for conducting quarterly communica-tions drills with other states located within the ingestion exposure pathway EPZ of an Illinois nuclear generating plant. This item is resolved. 13.3.2.7 Radiological Emergency Response Training Item requiring resolution: The applicant needs to clearly indicate that construction and contractor per-sonnel who would be onsite would receive an initial orientation and annual retraining on the plan and relevant procedures to ensure that they are aware of what actions they should take during an emergency. The applicant's GSEP states that station personnel who are not members of the onsite emergency organization are provided with an annual review of the emer-gency plan by Training Department personnel. This review is provided as part of the training program given to all persons who are granted unescorted access privileges at Braidwood Station, regardless of whether they are station or cor-porate office employees, construction or contractor personnel, or other persons who can justify their need for unescorted access to onsite areas in order to perform assigned tasks. In correspondence dated May 20, 1986 the applicant stated that all contractor and construction personnel at Braidwood Station will receive written instructions on what to do in an emergency. The initial distri-bution of these instructions will coincide with the implementation of the Sta-tion's security program. Subsequent distributions will be done annually. Re-cords of the distributions will be maintained by the applicant. Braidwood SSER 1 13-4

(.. I I In additi.on, copies of the Braidwood Emergency Information Booklet (letter dated November 19, 1985, from D. H. Smith, CECO, to H. R. Denton, NRR), which are also sent to public gathering places and residences within the 10-mile EPZ, will be made available on request at the Station's security gatehouse. Sufficient commitments have been made.in the GSEP and in the May 20, 1986 cor-respondence to indicate that all persons granted unescorted access privileges, as well as all onsite construction and contractor personnel, will receive in-I .structions on what to do in an emergency. This item is resolved. 13.3.2.8 Responsibility for the Planning Effort: Development, Periodic Review, and Distribution of Emergency Plans Item to be resolved: The applicant's emergency plan must indicate that the scope of independent j audits of Braidwood Station's emergency preparedness program will be adequate l and that relevant audit results will be made available to representatives of state and local governments, per the requirements of 10'CFR 50.54(t). Revision 5 to the GSEP states that the scope of these audits will meet the re-quirements of 10 CFR 50.54(t). In correspondence dated May 20, 1986, the appli-cant has stated that Revision 6 to the GSEP will contain additional information regarding how audit results relating to the interface between the applicant and state and local governmental organizations will be made available to those i organizations. This item is resolved based on Revision 5 to the GSEP and the applicant's commitment to include additional information in Revision 6 to the GSEP. 13.3.3 Offsite Emergency Planning Medical Services In a recent decision, GUARD v. NRC, 753 F.2d 1144 (D.C. Cir. 1985), the U.S. Court of Appeals vacated the Commission's interpretation of 10 CFR S50.47(b)(12) to the extent that a list of facilities was found to constitute adequate arrange-ments for medical services for members of the public offsite exposed to danger-l ous levels of radiation. The Commission has now provided guidance to be fol-lowed in determining compliance with this regulation pending its determination of how it will proceed in response to the Court's remand. In particular, the Commission directed that Licensing Boards, and in uncontested cases, the staff, should consider the uncertainty attendant to the Commission's interpretation of this regulation, especially in regard to its interpretation of the term " con-l taminated injured individuals." In GUARD, the Court left open to the Commis-l sion the discretion to reconsider whether that term should include members of the offsite public exposed to dangerous levels of radiation and, thus, whether arrangements for this population of individuals are required at all. For this reason, the Commission observed that it may reasonably be concluded that "no additional actions should be taken now on the strength of the present interpretation of that term." Accordingly, the Commission observed that it can be found "that any deficiency which may be found in complying with a finalized post GUARD planning standard (b)(12) is insignificant for the purposes of 10 CFR SSO.47(c)(1)." In this regard, the Commission, as a generic matter, noted the low probability of accidents which might result in exposure of mem-bers of the offsite public to dangerous levels of radiation as well as the slow l Braidwood SSER 1 13-5 I i 6 ~..-

development of adverse reactions to overexposure. See Emergency Planning, Statement of Policy 50 FR 10892, May 21, 1985. Consistent with the foregoing Statement of Policy, on May 20, 1986 the applicant committed to fully comply with the Commission's response to the Court remand. By memorandum dated March 3, 1985. FEMA provided confirmation that a list of medical facilities has been identified for use in case of an emergency at Braid-wood. FEMA also identified other related and supporting arrangements and services in the same memorandum. Accordingly, on the basis of the factors identified by the Commission in its Statement of Policy, the staff has determined that the requirements of 10 CFR S50.47(c)(1) have been satisfied so as to warrant issuance of the operat - ing license pending further action by the Commission with respect to the re-quirements of 10 CFR S50.47(b)(12). 13.3.4 Emergency Preparedness Exercise The Inspection Report Nos. 50-456/85037'(DRSS) and 50-457/85036 (DRSS) cover-ing the joint full participation exercise identified no violations of NRC re-quirements. The results of the FEMA evaluation of the exercise are addressed below in Section 13.3.5. 13.3.5 FEMA Finding on Offsite Preparedness The interim finding by FEMA on the offsite radiological preparedness plans for Braidwood Station was dated April 21, 1986 and was forwarded to the NRC by memo-randum dated April 30, 1986. The interim finding was based on FEMA's Region V review of the Illinois Plan for Radiological Accidents State General Plan (Vol-ume 1) and the preliminary site-specific plan for Braidwood (Volume VII), an analysis of the state's schedule of corrective actions for the planning inade-quacies identified in the Regional Assistance Committee review and an evaluation of the initial joint full participation exercise conducted on November 6, 1985. FEMA found reasonable assurance that the health and safety of the public can be protected in the event of an incident at Braidwood Station. 13.3.6 Conclusion Based on our review against the criteria in " Criteria for Preparation and Eval-uation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants," NUREG-0654/ FEMA-REP-1, Revision 1, November, 1980, we conclude that the Emergency Plan for Braidwood provides for'an acceptable state of emergency preparedness and meets the requirements of 10 CFR Part 50 and Appendix E thereto. The Federal Emergency Management Agency (FEMA) has provided interim findings on the State and local emergency response plans. FEMA concludas that there is rea-sonable assurance that the health and safety of the public can be protected in the event of an incident at Braidwood Station. Based upon our review of the applicant's plans and procedures, the NRC and FEMA evaluation of the joint exercise, and our review of the FEMA findings, we find that the state of onsite and offsite emergency preparedness provides reasonable Braidwood SSER 1 13-6

. _ ~. _ assurance that adequate protective measures can and will be taken in the event of a radiological emergency. Therefore, Outstanding Item A(6) is considered closed. 13.4 Review and Audit The SER noted that the staff had not completed its review of all the documents submitted relating to the applicant's organizational structure and the review and audit functions. Since that time, the staff has reviewed the Braidwood FSAR through Amendment 47, Commonwealth Edison Company's topical report CE-1A through Revision 31,'and Section 6, " Administrative Controls," of the Techni-cal Specifications (NUREG-1097) that formed Appendix A to the Byron operating license NPF-23 that was issued October 31, 1984. The Byron Technical Specifications include organization charts of the offsite .and plant organizations as constituted at the time of issuance of NPF-23 and i descriptive specifications for the review and audit functions. These charts and specifications will also be in the Braidwood Technical Specifications when 1 the Braidwood operating license is issued. Therefore, the staff concludes that, i unless the applicant requests a change to the applicable charts or specifica-tions, the applicant's organizational structure and review and audit functions are acceptable. By letter dated October 5, 1981, the applicant committed to provide, as required by TMI Action Plan Item I.B.1.2 (NUREG-0737), an Independent Safety Engineering Group (for both Braidwood and Byron Stations) consisting of four dedicated full-time engineers located on site, reporting to the Supervisor, Safety Engineering Groups, Office of Nuclear Safety. By Amendment 47, the applicant advised that j the Supervisor, Safety Engineering Groups, now reports to the Manager of Nuclear Safety rather than to the Director of Nuclear Safety. By Amendment 47, the applicant also redefined the functions of the Independent Safety Engineering Group to read as foi?ows: The principal fuqction of the ISEG is to examine plant operating 4 characteristics, NRC issuances, industry advisories, and other appropriate sources of plant design and operating experience in-formation that may indicate areas.for improving plant safety. The ISEG is to perform independent review and audits of plant activities including maintenance, modifications, operational problems, and operational analysis, and aid in the establishment of programmatic requirements for plant activities. Where use-ful improvements can be achieved, it is expected that this group will develop and present detailed recommendations to corporate management for such things as revised procedures or equipment modifications. The ISEG also functions to maintain surveillance of plant opera-tions and maintenance activities to provide independent verifica-tion that these activities are performed correctly and tat human errors are reduced as far as practicable. ISEG will then be in a position to advise utility management on the overall quality and safety of operations. ISEG need not perform detailed audits of plant operations and shall not be responsible for sign-off func-tions such that it becomes involved in the operating organization. Braidwood SSER 1 13-7 I

This description is consistent with the description of the function as given in NUREG-0737. The applicant proposed to adopt the staff's definition "per INPO recommendations." The staff finds that the applicant's Independent Safety Engineering review group meets the requirement of TMI Action Plan Item 1.B.1.2 and is accept ale. By letter dated August 7, 1984, the applicant described a change in the organi-zational arrangements for conducting the Braidwood startup testing. The appli-cant has found, based on experience gained at the applicant's LaSalle and Byron plants, that, as the time of fuel load approaches, preparation for operation, including such things as training, emergency planning and operating procedures preparation, places burdens on the station personnel that interfe 'e with the testing schedule. Therefore, the Braidwood preoperational testing responsibili-ties will be divided between the station personnel and a newly formed Project Startup Organization. At Byron Station, the operating and technical staffs manage and execute the initial test program. The station technical staffs are responsible for writing and conducting the initial test program. The Test Review Board is responsible for the onsite review and approval of the test procedures and test results. At Braidwood, two Commonwealth Edison organizations have been established to execute the initial testing program: the Station Organization and the Project Startup Organization. For initial startup testing, Braidwocd Station's efforts will parallel those of Byron Station's; that is, Braidwood Station's operating and technical staffs will manage and execute the initial startup test program. The station technical staffs are responsible for writing and conducting the initial startup test program. The Test Review Board is responsible for the review and the approval of the startup test procedures and startup test results. After April 1984, responsibility for preoperational testing at Braidwood Station rests with the Project Startup Organization, which is headed up by the Project Startup Superintendent. The Project Startup Organization's Testing Staff will be responsible for writing and conducting the preoperational test program. A Test Review Board will be responsible for the onsite review and approval of the preoperational test procedurn and test results. The staff concurs with efforts to relieve burdens on operating personnel and concludes that the manner in which the applicant proposes to accomplish this is acceptable. Generic Letter 84-16 specifies that each operating shif t should have at least one individual who has had at least 6 months of previous experience on a hot, operating nuclear power plant of a similar type (either PWR or BWR), including startup and shutdown experience and at least 6 weeks of experience above 20% power. By letter dated June 18, 1986 (A. D. Miosi to H. R. Denton), the appli-cant provided information regarding the operators it plans to use at Braidwood Unit 1 to meet the guidelines of Generic Letter 84-16. This was supplemented by additional information submitted by letter dated July 30, 1986 (A. D. Miosi to H. R. Denton). The staff's refiew of the information provided in the two letters finds the applicant now has seven individuals who will be fully qualified to provide the Braidwood SSER 1 13-8

on shift hot operating experience specified by Generic Letter 84-16 after they each receive their Senior Operators License for Braidwood. These seven individ-uals will be able to provide coverage on the operating shifts. An eighth in-dividual will be available very shortly after initial fuel loading. Thus, the staff concludes that, subject to-their successfully obtaining Senior Operators Licen3es at Braidwood Station, the applicant will have a sufficient number of experienced senior operators to provide on-shift operating experience meeting the guidelines of Generic Letter 84-16. Therefore, Outstanding Item A(5) is considered closed. 13.6 Physical Security The applicant has filed with the Nuclear Regulatory Commission a Physical Security Plan, a Security Personr.el Training and Qualification Plan, and a Safeguards Contingency Plan for the Braidwood Nuclear Power Station. This SER summarizes how the applicant has provided for meeting the requirements of 10 CFR 73. The SER is composed of a basic analysis that is available for public review, and a protected Appendix. Based on a review of the subject docu-ments and visits to the site, the staff has concluded that the protection pro-vided by the applicant against radiological sabotage at the Braidwood Station meets the requirements of 10 CFR 73. Accordingly, the protection will ensure that the health and safety of the public will not be endangered. Therefore, Conformatory Issue A(9) is considered closed. Physical Security Organization To satisfy the requirements of 10 CFR 73.55(b), the applicant has provided a physical security organization that includes a Security Shift Supervisor who is on site at all times with the authority to direct the physical protection activities. To implement the commitments made in the physical security plan, training and qualification plan, and the safeguards contingency plan, written security procedures specifying the duties of the security organization members have been developed and are available for inspection. The training program and critical security tasks and duties for the security organization personnel are defined in the Braidwood Security Personnel Training and Qualification Plan, which meets the requirements of 10 CFR 73, Appendix B, for the training, equip-ping, and qualification of the security organization members. The physical security plan and the training program provide commitments that preclude the assignment of any individual to a security-related duty or task before the individual being trained, equipped, and qualified to perform the assigned duty in accordance with the approved guard training and qualification plan. Physical Barriers In meeting the requirements of 10 CFR 73.55(c), the applicant has provided a protected area barrier that meets the definition in 10 CFR 73.2(f)(1). This barrier consists of two parallel fences with the inside parallel fence desig-nated as the protected area barrier. An isolation zone, which permits obser-vation of activities at the perimeter, is provided along both sides of the protected area barrier (except for the locations listed in the Appendix). The staff has reviewed those locations and determined that the security meas-ures in place are satisfactory and continue to meet the requirements of Braidwood SSER 1 13-9

l l 10 CFR 73.55(c). Illumination of 0.2 foot-candles is maintained for the iso-lation zones, protected area barriers, and external portions of the protected I area. Identification of Vital Areas The Appendix contains a discussion of the applicant's vital area program and identifies those areas and items of equipment determined to be vital for pro-tection purposes. Vital equipment is located within vital areas that are located within the protected area and require passage through at least two barriers, as defined in 10 CFR 73.2(f)(1) and (2), to gain access to the vital equipment. Vital area barriers are separated from the protected area barrier. The control. room and central alarm station are provided with bullet-resistant walls, doors, ceilings, floors, and windows. On the basis of these findings and the analysis set forth in paragraph C of the Appendix, the staff has concluded that the applicant's program for identification and protection of vital equipment satisfies the regulatory. intent. However, this program is subject to onsite validation by the staff in the future, and to subsequent changes if found to be necessary. Access Requirements In accordance with 10 CFR 73.55(d), all points of personnel and vehicle access to the protected area are controlled. The individual responsible for controll-ing the final point of access into the protected area is located in a bullet-resistant structure. As part of the access control program, vehicles (except under emergency conditions), personnel, packages, and materials entering the protected area are searched for explosives, firearms, and incendiary devices by electronic search equipment and/or physical search. Vehicles admitted to the protected area, except licensee-designated vehicles, are controlled by escorts when in operation. Licensee-designated vehicles are limited to onsite station functions and remain in the protected area except for operational maintenance, repair, security, and emergency purposes. Posi-tive control over the vehicles is maintained by personnel authorized to use the vehicles or by the escort personnel. A picture badge / key card system, using encoded information, identifies individuals that are authorized unescorted access to protected and vital areas, and is used to control access to these Individuals not authorized unescorted accers are issued non picture areas. badges that indicate an escort is required. Access authorizations are limited to those individuals who have a need for access to perform their duties. Unoccupied vital areas are locked and alarmed. During periods of refueling or major maintenance, access to the reactor containment (s) is positively controlled by a member of the security organization to assure that only authorized individ-uals and materials are permitted to enter. In addition, all doors and personnel / equipment hatenes into the reactor containment (s) are locked and alarmed.

Keys, locks, combinations and related equipment are changed on an annual basis.

In addition, when an individual's access authorization has been terminated because of the lack of reliability or trustworthiness or poor work performance, the keys, locks, combinations and related equipment to which that person had access are changed. Braidwood SSER 1 13-10

Detection Aids In satisfying the requirements of 10 CFR 73.55(e), the applicant has installed intrusion detection systems at the protected area barrier, at entrances to vital areas, and at all emergency exits. Alarms from the intrusion detection system annunciate within the continuously manned central alarm station and a secondary alarm station located within the protected area. The central alarm station is located so that the interior of the station is not visible from outside the perimeter of the protected area. In addition, the central station is constructed so that walls, floors, ceilings', doors, and windows are bullet-resistant. The alarm stations are located and designed in such a manner that a single act cannot interdict the capability of calling for assistance or respond-ing to alarms. The central alarm station centains no other functions or duties that would interfere with its alarm response function. The intrusion detection system transmission lines and associated alarm annunciation hardware are self-checking and tamper-indicating. Alarm annunciators indicate the type of alarm and its location when activated. An automatic indication of when the alarm system is on standby power is provided in the central alarm station. Communications As required in 10 CFR 73.55(f), the applicant has provided for the capability of continuous communications between the central and secondary alarm station operators, guards, watchmen, and armed response personnel through the use of a conventional telephone system and a security radio system. In addition, direct communication with the local law enforcement authorities is maintained through the use uf a conventional telephone system and two-way FM radio links. All non portable communication links, except the conventional telephone system, are provided with an uninterruptible emergency power source. Test and Maintenance Requirements In meeting the requirements of 10 CFR 73.55(g), the applicant has established a program for the testing and maintenance of all intrusion alarms, emergency alarms, communication equipment, physical barriers, and other security-related devices and equipment. Equipment or devices that do not meet the design per-formance criteria or have failed to otherwise operate will be compensated for by appropriate compensatory measures as defined in the Braidwood Nuclear Power Station Security Plan and in site procedures. The compensatory measures defined in these plans will ensure that the effectiveness of the security system is not reduced by failures or other contingencies affecting the operation of the secu-rity related equipment or structures. Intrusion detection systems are tested for proper performance at the beginning and end of any period that they are used for security. Such testing will be conducted at least once every 7 days. Communication systems for onsite communications are tested at the beginning of I each security shift. Offsite communications are tested at least once each day. Audits of the security program are conducted once every 12 mor.ths by personnel independent of site security management and supervision. The audits, focusing on the effectiveness of the physical protection provided by the onsite security organization implementing the approved security program plans, include, but are not limited to, a review of the security procedurcs and practices, system Braidwood SSER 1 13-11

testing and maintenance programs, and local law enforcement assistance agree-ments. A report is prepared documenting audit findings and recommendations and j is submitted to the plant management. Response Requirements In meeting the requirements of 10 CFR 73.55(h), the applicant has provided for armed responders to be immediately available for response duties on all shifts consistent with the requirements of the regulations. Considerations used in support of this number are attached (see Appendix). In addition, liaison with local law enforcement authorities to provide additional response support in the event of security events has been established and documented. The applicant's safeguards contingency plan for dealing with thefts, threats and radiological sabotage events satisfies the requirements of 10 CFR 73, Appendix C. The plan identifies appropriate security events that could ini-tiate a radiological sabotage event and identifies the applicant's preplanning, response resources, safeguards contingency participants, and coordination activities for each identified event. Through this plan, upon the detection response activities and objectives include the neutralization of the existing threat by requiring the response force members to interpose themselves between the adversary and their objective, instructions to use force commensurate with that used by the adversary, and authority to request sufficient assistance from the local law enforcement authorities to maintain control over the situation. To assist in the assessment / response activities, a closed-circuit television system, providing the capability to observe the entire protected area perimeter, isolation zones, and a majority of the protected area, is provided to the security organization. Employee Screening Program In meeting the requirements of 10 CFR 73.55(a) to protect against the design basis threat as stated in 10 CFR 73.1 (a)(1)(ii), the applicant has provided an employee screening program. Personnel who successfully complete the employee screening program or its equivalent may be granted unescorted access to pro-tected and vital areas at the Braidwood site. All other personnel requiring access to the site are escorted by persons authorized and trained for escort duties and who have successfully completed the employee screening program. The employee screening program is based on accepted industry standards and includes a background investigation, a psychological evaluation, and a continuing obser-vation program. In addition, the applicant may recognize the screening program of other nuclear utilities or contractors based on a comparability review con-ducted by the applicant. The plan also provides for a " grandfather clause" exclusion that allows recognition of a certain period of trustworthy service with the utility or contractor as being equivalent to the overall employee screening program. The staff has reviewed the applicant's screening program against the accepted industry standards (ANSI N18.17-1973) and has determined that the program is acceptable. Braidwood SSER 1 13-12

APPENDIX 13A ORIGINAL STAFF EVALUATION OF EMERGENCY PLANNING INPUT FOR BRAIDWOOD STATION, UNITS 1 AND 2 13.3 Emergency Planning The Commonwealth Edison Company filed with the Nuclear Regulatory Commission revisions to the Commonwealth Edison Generating Stations Emergency Plan (GSEP) dated January 3, 1980; April 24, 1980; June 4, 1980; July 30, 1980; December 31, 1980; March 27, 1981; October 20, 1981; May, 1984; and October, 1984. The GSEP (filed in May and October,.1984 and dated April and July, 1984) is a generic emergency plan applicable to all nuclear generating stations operated by CECO. The GSEP includes site-specific annexes that contain additional infor-mation and guidance ~ specific to each nuclear generating station. The Braidwood Annex to the GSEP, dated October, 1984, was submitted to the Nuclear Regulatory Commission in January,1985..The plan is for the Braidwood onsite and corporate activities only. By correspondence. dated March 15, 1985, CECO submitted clari-fications to this annex and committed to incorporate these clarifications to an annex revision scheduled to be issued around October, 1985. The Commission staff has conducted a review of the GSEP and the Braidwood Annex (all herein-after referred to as the Plan) as part of the overall emergency preparedness planning evaluation for the Braidwood Station. Evaluation of the state of emergency preparedness for Braidwood also involves the review of State and local radiological emergency response plans by the Federal Emergency Management Agency (FEMA). NUREG-0800 states that the FEMA findings on offsite plans are reviewed by the NRC and a full participation exercise is conducted at the facility. In accordance with the revised rule on emergency planning (47 [R 30232), no NRC or FEMA findings and determinations concerning the state or adequacy of offsite emergency preparedness are reauired before issuance of an operating license authorizing only fuel loading and low-power operations up to 5% of rated power. The findings and determinations of FEMA on the adequacy of the State and local emeryncy response plans and the overall. conclusion of the NRC on the state of ewegency preparedness for Braidwood will be presented in a future supplement to the SER. ~ The Plan was reviewed against the requirements of 50.47(b) and Appendix E of 10 CFR 50, and the criteria of the 16 planning standards in Part II of the " Criteria for Preparation and Evaluation of Radiological Emergency Response-i Plans and Preparedness in Support of Nuclear Power Plants," NUREG-0654/ l FEMA-REP-1, Revision 1, dated November 1980. The criteria of NUREG-0654 have l been endorsed in Regulatory Guide 1.101, Revision 2, " Emergency Planning and Preparedness for Nuclear Power Reactors," dated October, 1981, and thus have the same status as a regulatory guide. Section 13.3 of the July 1981 revision of the Standard Review Plan (NUREG-0800) was also used during the staff review. of Braidwood Station Emergency Preparedness. Braidwood SSER 1 13A-1 Appendix 13A

Section 13.3.2 of this supplement lists each planning standard of 10 CFR 50.47(b) in order followed by an evaluation of the applicable portions of the Emergency (Onsite) Plan that relate principally to that particular planning standard. Section 13.3.3 of this supplement provides the staff review results and conclusions. 13.3.2 Evaluation of the Emergency (Onsite) Plan 13.3.2.1 Assignment of Responsibility (Organization Control) The applicant's GSEP organization is divided into onsite and offsite functional The onsite GSEP organization consists of a Station Group, which functions areas. under a Station Director who is responsible for organizing and coordinating emergency response activities at and within the immediate vicinity of the sta-tion. The offsite GSEP organization is comprised of a Corporate Command Center (CCC) Group and a Recovery Group. During emergencies of limited extent, such as a transportation accident involving radioactive or other hazardous materials or a condition that would be classified as a Notification of an Unusual Event or as an Alert, the CCC Group's Director and staff may be activated to support the Stati u Group in evaluating, coordinating, and directing the overall company response tc the emergency. For more serious emergencies, such as any Site Area or General Cmergency, CECO has developed a prioritized Nuclear Duty Officer /CCC Director / Recovery Group notification list. This list enables the corporate Nuclear Duty Officer to initiate activation of the Recovery Group, which func-tions under the Recovery Manager at the nearsite Emergency Operations Facility (EOF). The CCC Group becomes a support staff to the Recovery Group upon the Recovery Manager's decision that the Recovery Group is ready to assume its respc.. ibilities at the EOF. The Recovery Manager's responsibilities include the evaluation, coordination, and direction of the overall company and industry The CCC response and management of the nuclear plant recovery operations. ~ Director is responsible for initial coordination with governmental agencies, with the Recovery Manager assuming this responsibility after the E0F is functional. Figures 1 and 2 illustrate the CCC Group (limited response offsite GSEP organi-zation) and Recovery Group (full response offsite GSEP organization), respec-tively. The Plan describes the duties, responsibilities, and interfaces of 4 l each position in these organizations, as well as analogous information for each position in the Station Group. The applicant has a sufficient pool of trained personnel available from its nuclear generating stations and corporate staff to ensure that the GSEP organizations are capable of continuous operations for a protracted period of time, as evident from information provided in the GSEP Telephone Directory. The Administrative Director, Administration / Logistics Manager, and the Manpower / Logistics Director are responsible for ensuring that adequate numbers of trained personnel are available for protracted emergency i organizations' operations. The Shift Engineer, who is on duty 24 hours a day, becomes the initial Station Director upon detection of an abnormal condition, and is responsible for classi-fying the emergency and implementing the Plan as required. Multiple communica-4 tions links are described in the Plan that will enable the Station to communi-cate 24 hours a day with Federal, State, and local emergency rest.onse organiza-tions to ensure the rapid transmittal of accurate notification information and emergency assessment data. Braidwood SSER 1 13A-2 Appendix 13A

Formal agreements with appropriate agencies and organizations includ-ing law enforcement, ambulance services, medical and hospital support, l fire departments, and state and local authorities responsible for implementation of protective measures for the public are maintained separate from the Plan. With the exception of a letter of agreement with the U.S. Coast Guard, the staff determined that adequate letters of agreement have been executed. The staff will examine the agree-ment with the U.S. Coast Guard and report its finding in a future supplement to the SER. 13.3.2.2 Onsite Emergency Organization The normal station organization is shown in Figure 3. The Braidwood Station is managed by a Station Superintendent who is responsible for direct management of the station. During an emergency situation (normal working hours) the Sta-tion Superintendent is the Station Group Director. During periods when the Station Superintendent is unavailable, his responsibilities are delegated to alternates who satisfy the requirements of ANSI N18.1-1971, " Experience Require-ments for Plant Manager." The Shift Engineer, on duty 24 hours per day, is the initial Station Director and, as such, has the duties and responsibilities specified in Table 4.2-1 of the generic GSEP. Should the Shift Engineer become incapacitated, a line of succession is provided in the GSEP. The Braidwood Annex _to the GSEP specifies responsibilities of the Station Director that cannot be delegated: the deci-sion to declare an emergency condition and the decision to notify and recommend protective actions to offsite authorities in the event that the Recovery Manager or CCC Director have not been contacted or are not prepared to make an informed decision. The GSEP addresses the need for limiting personnel exposures under emergency conditions, and the need to obtain the prior approval of the Station Superintendent, the CECO Medical Director, and/or the Station's Radiation Pro-tection Supervisor before potentially exposing volunteer emergency workers to doses which will potentially exceed 10 CFR 20 limits for life saving actions or actions needed to protect facilities, terminate radioactive releases, or to control fires. However, the Plan does not specify which individuals in the onsite and offsite emergency organizations have the ultimate responsibility for authorizing excessive exposured to volunteers under the aforementioned circum-stances. This is an Open Item. As described previously, the onsite emergency organization.. the Station Group. Figure 4 shows this organization. The major duties and responsibilities of each Station Group director are defined in the Plan. The interfaces between the Station Group, the CCC Group, and the Recovery Group are indicated in l Figures 1 and 2. Interfaces between and among the Station Group, CCC Group, and Recovery Group staffs and the staffs of governmental and private sector organizations and technical and engineering contractors have been specified in the Plan. The onsite emergency organization for nonnormal working hours, backshifts, and holidays is described in the Plan. Emergency assignments have been made, and the relationship between this emergency organizatiun and the normal staff com-plement is shown in the Plan. Positions and/or titles and qualifications of plant personnel who are assigned major emergency functional duties are listed. Braidwood SSER 1 13A-3 Appendix 13A

The minimum shift staffing levels provided in Figure 4.2-2 of the Plan meet the objectives of Table B-1 in NUREG-0654, Revision 1. This onshift staff includes the following areas of expertise: one Shift Engineer (Senior Reactor Operator (SRO)); one Shift Foreman (SRO); two Nuclear Station Operators (three, if both units are operating); three equipment operators / attendants (four, if both units are operating); two Radiation / Chemistry Technicians; and one Shift Technical Advisor (also referred to as the Station Control Room Engineer (SCRE)). The applicant's provisions for onshift augmentation within the first hour after an emergency declaration is described in Section 4.2 of the GSEP. CECO has established a 24-hour Station Duty Officer who would be notified first after a Plan activation. This individual would initiate a prioritized notification (call-list) procedure whereby individuals are contacted who are capable of per-forming the specific response functions identified in Table B-1 of NUREG-0654. The call-list is prioritized according to the least amount of travel time of the applicant's staff members to the site. This is done to ensure that the desired functions can be properly covered by qualified people within the time goals of Table B-1 of NUREG-0654. Further, unannounced offshift notification drills will be conducted at least every 6 months to demonstrate that the goals af Table B-1 are maintained. Drill records will be maintained for inspection. rigure 4.2-3 of the Plan 1; dicates which personnel must respond, or may respond depending upon the nature of the emergency, following the declaration of a spe-cific emergency classification. Within about 60 minutes of any Alert declara-tion, for example, a minimum of eight and a maximum of 25 individuals would be required to augment onshift personnel. Following any Site Area or General Emergency declaration, twenty-nine personnel would augment the onsite emergency orga1ization within about 60 minutes. These 29 persons would function in the following emergency roles: all nine Station Group directors; an Environs Director; seven radiation / chemistry personnel for inplant, onsite, and offsite surveys; four radiation / chemistry personnel for protective actions; three engineers for plant system engineering; two electrical / mechanical personnel; one Instrument and Control Technician; one Radwaste Operator for equipment repair and correction; and one dedicated communicator. 13.3.2.3 Emergency Response Support and Resources Arrangements for requesting and utilizing outside resources have been made, including authority to request implementation of the Department of Energy (00E) Radiological Assistance Plan and the Federal Radiological Emergency Response Plan. Either the Station Director, Recovery Manager, or CCC Director may request DOE assistance, or other Federal response, if deemed necessary or desirable. The applicant also retains contractors to provide various emergency support services to the Braidwood Station, including technical experts for the following: accident analyses; environmental radiological monitoring, bioassay, and radiochemical analyses; medical and health physics support; meteorological monitoring and forecasting support; and personnel dosimetry services. The Plan describes the radiological laboratories at each CFCo nuclear generating station. Because each station's management and resources are similar, each station can make available some of its equipment, manpower, and counting facili-ties to a station affected by an emergency situation. Braidwood SSER 1 13A-4 Appendix 13A

The CECO organization provides for dispatching licensee representatives to governmental emergency operations centers, if requested. Working space will be available for Federal, State, and local offsite representatives as well as for contractor and other support groups in the applicant's E0F located near Mazon, Illinois. This facility presently serves as the EOF for the applicant's LaSalle County and Dresden Nuclear Generating Stations. This facility is the central point for providing information needed by primary response agencies for implementation of protective actions. 13.3.2.4 Emergency Classification System The Plan provides for classification of emergency conditions into one of six categories. The first category, Transportation Accident, concerns an emergency involving the transportation of radioactive or other hazardous material from a nuclear generating station. The next four classes, in order of increasing severity, are Notification of an Unusual Event, Alert, Site Area Emergency, and General Emergency. The applicant's definitions of these classes are consistent with the descriptions found in Appendix I of NUREG 0654, Revision 1. The sixth category is Recovery, which the applicant has defined as that period when the emergency pnase is over and activities are underway to return the station to normal operation. The Plan contains guidance for determining when the individ-ual in charge of the applicant's emergency response activities can declare the beginning of the Recovery phase of operations, as well as guidance for down-grading of an emergency classification, and for terminating an emergency condi-tion. The Plan also summarizes major CECO emergency response actions for the first five er.ergency classes (Transportation Accident through General Emergency). The applicant's actions to be taken in response to each of the four emergency classes described in Appendix I of NUREG-0654, Revision 1, are consistent with the guidance provided in that document. The GSEP contains general descriptions of initiating conditions for the appli-cant's first five emergency classes. The descriptions for Notification of an Unusual Event through General Emergency classes include most of the examples given in Appendix I of NUREG-0654, Revision 1. The Braidwood Annex contains station-specific Emergency Action Levels (EALs) that are based on various plant sensors, onsite and offsite radiation monitoring information, and various natural phenomena. Tnese station-specific EALs will be used for rapid classi-fication of emergency situations. The staff has evaluated the adequacy of the site-specific EALs, as contained in the Braidwood Annex to the generic GSEP, versus the guidance in Appendix 1 of NUREG-0654, Revision 1. The EALs were determined to be adequate, with the following two exceptions: (1) Four EALs addressed the situation in which equipment described in the station's Technical Specifications had been degraded beyond the limiting condition for operation that requires a reactor shutdown, or had been degraded such that a Technical Specification safety limit had been exceeded. Such equipment degradation would be classified as an Alert, per EAL Condi-tion 14; however, the same resulting degradation would be classified as a Site Area Emergency, per EAL Conditions 1, 4, or 5, if it were caused by l aircraft or missile crash, unplanned explosion, or fire, respectively. The applicant must justify the rationale for classifying the same level of Braidwood SSER 1 13A-5 Appendix 13A

equipment degradation either as an Alert (Condition 14) or as a Site Area Emergency (Conditions 1, 4, and 5), depending on the cause of the degrada-tion. This is an Open Item. (2) The wording of EALs for Gaseous Radiation Releases implied that releases would always be monitored and quantified by plant instrumentation. The plan must contain guidance in appropriate EALs for Condition 27 to indicate that unmonitored gaseous radiation releases will be estimated from field peasurements taken by environmental survey teams, so that unmonitored re-leases will be approximately classified. This EAL is in the Byron Annex, but missing from the Braidwood Annex. This is an Open Item. 13.3.2.5 Notification Methods and Procedures The applicant has made provisions for initial and followup notifications of State and local authorities in case of an emergency at the Braidwood Station. The Shift Engineer, as acting Station Director, has the authority and responsi-bility for initiating emergency notifications to these authorities. The initial notification scheme is shown in Figures 5A, SB, and SC, and was set up by the State of Illinois and agreed to by the applicant. The Nuclear Accident Report-3 ing System (NARS) is the dedicated voice communications system utilized for notifying State and local emergency operations centers of declared emergencies. l l Its capability to enable timely notification of offsite authorities has been demonstrated at Ceco's operating nuclear stations. Initial messages to State and local representatives of the Illinois Department of Nuclear Safety (IDNS) and the Illinois Emergency Services and Disaster Agency (IESDA) are reported in the format of the current NARS Form, a copy of which is included in the Plan. The format and content of the NARS form has been agreed to by CECO and the Directors of IDNS and IESDA. The NARS form is standard for all Ceco nuclear generating stations and is not subject to alteration by CECO onsite or offsite review groups. The current NARS Form includes: information about the emergency ,^ class; whether a release of radioactive material has taken place; potentially affected population and areas; and whether offsite protective actions may be necessary. Provisions exist for verification of messages. The plan has estab-lished procedures which describe mutually agreeable bases for notification of offsite response organizations consistent with the standard emergency classifi-f cation and action scheme set forth in Appendix 1 of NUREG-0654, Revision 1. The plan contains guidance for the formulation of followup messages to State and local authorities that is consistent with the guidance found in Criterion 4, Planning Standard E, of NUREG-0654, Revision 1. However, the Plan neither addresses the frequency of followup message transmittal nor contains a copy of a standardized followup message form, which would provide greater assurance that followup message content would be developed in accordance with regulatory I guidance. The lack of an adequate followup message and Plan provisions addressing the frequency of followup message transmittal to State and local authorities is an Open Item. The plan has established procedures for notifying, alerting, and mobilizing applicant emergency response personnel. These procedures include both station and corporate personnel. Ceco and the State have developed predetermined written messages intended for the public and consistent with the emergency classification scheme. These Braidwood SSER 1 13A-6 Appendix 13A _ _ ~ - \\

messages are part of the State Emergency Plan and are not included in the applicant's plan. The Plan describes the Prompt Public Notification System (PPNS) as consisting of a permanently installed outdoor notification system within the zero (0) to ten (10) mile radius around the station. The zero (0) to ten (10) mile radius around the. station is primarily an agricultural area with a population density below 2000 persons per square mile. The prompt notification system to be installed consists of mechanical and electronic sirens which will cover this entire area with a minimum sound level of 60 db. Additionally, for the heavily populated areas within the zero (0) to ten (10) mile radius around the station, the prompt notification system will cover these communities with a minimum sound level of 70 db to ensure complete coverage. Responsibility for activating the PPNS is given to local offsite authorities. Per instructions to be provided in the emergency information brochures to be distributed within the plume exposure EPZ, persons hearing a siren are to tune their radios to predesignated Emergency Broadcast System (EBS) radio stations to obtain detailed instructional messages from local authorities. State and local procedures provide the contents of these messages. By letter dated March 3, 1986, the applicant committed to providing information in a subsequent Annex which would address how members of the transient popula-tion, such as b' oaters and fishermen who may be at recreation areas, will receive an instructional message on what actions they would take upon hearing the siren. Although the siren system will be operated by local government agencies, it will be maintained by CECO. CECO will seek agreement with these agencies to test the system monthly and to report inoperable equipment to designated main-tenance personnel for timely repair. In addition to this nonroutine repair program, CECO will provide for a periodic, preventive maintenance program. 13.3.2.6 Emergency Communications The Plan describes multiple communications systems, which include the use of normal and dedicated telephone lines on land lines and microwave voice channels, mobile radio units, handi-talkies, and computer peripherals, thus providing both a primary and several backup means for communications. A dedicated telephone communications system, called the Nuclear Accident Report-ing System (NARS), is described and provides for the notification of State and local authorities in the event of an emergency. This system links the Control Room, Corporate Command Center (CCC), Technical Support Center (TSC), System Power Supply Office, Emergency Operations Facility (EOF), and State and local Emergency Operations Centers. Initial contact points are manned 24 hours per day. Additional dedicated communications systems include: A microwave voice channel between the CCC and Control Room, TSC, and the EOF; A dedicated telephone link that enables communication between the CCC, TSC, and the E0F; A dedicated telephone link that enables communication between the Control Room and the TSC; Braidwood SSER 1 13A-7 Appendix 13A

I A dedicated telephone link that enables communication between the Control Room and the Operational Support Center (OSC); and A dedicated telephone link that enables communication between the TSC and OSC. Two separate communication system have been installed to allow coordinated environmental monitoring and assessment during an emergency. The first system consists of the necessary hardware to allow communications between the Corporate Command Center, the Control Room, the TSC, the EOF mobile units in Commonwealth vehicles, and handi-talkies held by environmental monitoring teams in the field. The second system consists of a dedicated telephone which allows continuous communication between the Corporate Command Center and the Illinois Department of Nuclear Safety REAC in Springfield. The Braidwood Station will also have other intraplant and plant-to-offsite communications equipment, including: a public address system; a commercial phone system; Security / Operations radioconsoles and handi-talkies; and sound-powered phones. The NRC will install dedicated telephones linking the station's control room, TSC, and E0F with the NRC Incident Response Center in Washington, D.C. and the regional NRC office in Glen Ellyn, IL. Also there will be a separate dedicated Health Physics Network (HPN) telephone between the NRC and the Braidwood Radia-tion Protection Office, TSC, and E0F. Operation of these systems will be under the direction of the NRC. 13.3.2.7 Public Education and Information The Plan provides for the annual distribution of informational brochures to the public which address how they will be notified and what their actions should be during an emergency. The public information brochure for the Braidwood Station included the following information: what to do if a take-shelter request is given; what to do if an evacuation request is given; methods for notification during an emergency; educational information concerning radiation; a map of major evacuation routes; a list of communities likely to serve as host shelter areas; and instructions on how to obtain additional information, especially for the disabled or their caretakers or those without transportation. The public information brochure described above was mailed to all residents in the plume exposure EPZ of the Braidwood Station and was also provided to appropriate locations (city halls, State parks, campgrounds) and other areas where a tran-sient population may obtain a copy. The E0F provides support and interface between CECO, State, and local agencies, and the news media. The plan provides for dispatching the Emergency News Center Director to the EOF. An Emergency News Center functions under the direction of this person and is the single point of contact for disseminating information to the public. The Emergency News Center Director's responsibilities include coordinating information releases with Federal, State, and local agencies, and participating in rumor control activities managed by State agencies. A tech-nical spokesperson, knowledgeable about the affected station and its operations, will be available to brief the press at the Emergency News Center. Braidwood SSER 1 13A-8 Appendix 13A

CECO will offer programs at least annually to acquaint news media with the GSEP, information concerning radiation, and points of contact for release of public information in an emergency. 13.3.2.8 Emergency Facilities and Equipment Emergency facilities needed to support an emergency response effort have been described, including an onsite Technical Support Center (TSC), onsite Opera-tional Support Center (OSC), nearsite Emergency Operations Facility (EOF),'and a Corporate Command Center (CCC) located at the applicant's offices in Chicago, Illinois. The TSC and OSC will be activated for any Alert or higher emergency classification. The EOF will'be activated for any Site Area or General Emer-gency classification, and the CCC may be activated for any class. The TSC will be located at the south end of the turbine building and is sized for at least 25 persons and. supporting equipment. Personnel in the TSC will be protected from radiological hazards, including direct radiation and airborne contaminants under accident conditions to the same degree as Control Room per-sonnel. To ensure adequate radiological protection, permanent radiation moni-toring systems will be inscalled. These systems continuously indicate radia-tion dose rates and airborne radioactivity inside the TSC while in use. In addition, protective breathing apparatus and thyroid blocking agents will be available for use as needed. The TSC will have access to a complete set of as-built drawings and other records, including general arrangement diagrams, P& ids, piping system isometrics, and electrical schematics. The TSC will have the capability to record and display vital plant data in real time. The station's primary OSC will be located in Meeting Room Number 1 of the Ser-vice Building. The Shift Engineer's office in the Auxiliary Building will serve as a backup OSC. Operations, Radwaste, and Rad / Chem personnel will report to the OSC for assignment. A limited inventory of supplies will be kept in the OSC. This inventory will include portable lighting, respirators, protective clott..W, and portable survey instruments. Communications and management controls from the OSC to the TSC and Control Room will be provided. Braidwood Station has an EOF located at Mazon, Illinois, approximately 14 miles WNW of the station. This facility presently serves as the EOF for Ceco's-LaSalle County and Dresden Nuclear Generating Stations. Since this EOF is be-yond 10 miles from the Braidwood Station, it conforms to the NUREG-0696 habit-ability guidelines for EOF locations between 10 and 20 miles from a nuclear power plant. The EOF will be utilized to evaluate and coordinate the emergency reentry / recovery operations on a continuing basis. Liaison with Federal, State, and local officials will be maintained at this center, which will also be used for' receipt and analysis of field monitoring data submitted by field teams. 'In December 1983, the NRC issued Supplement 1 to NU3EG-0737 which contained final requirements and guidance for Emergency Response Facilities (ERFs) that superseded the previous requirements of NUREG-0737. A determination of adequacy of the applicant's final ERFs will be performed during a post-implementation appraisal. That appraisal will be conducted against the pro-visions of Supplement 1 to NUREG-0737 at a future date. i Braidwood SSER 1 13A-9 Appendix 13A

The CCC is located on the 12th floor of the Edison Building in downtown Chicago, and is the location from which the CCC Director will normally direct cverall company activities involved in coping with an emergency, if he has assumed command. If the Recovery Group is activated at the E0F, the CCC will be the location for a support staff reporting to the Recovery Group. In addition to the above functions, the CCC will serve as the corporate environmental center where environmental monitoring will be coordinated and offsite dose projections performed under the direction of the CCC Environmental Director. The CCC has dedicated communication with the Control Room, TSC, E0F, State of Illinois Emergency Services and Disaster Agency, the Illinois Department of Nuclear Safety REAC, company cars, and field radios. A CECO submittal dated January 12, 1982 not only described the Braidwood Sta-tion's ERFs but also contained commitments and preliminary descriptions of the Safety Parameter Display System (SPDS), indicating that SPDS consoles would be located in the TSC and E0F. The staff finds these locations acceptable. The consoles will be installed to ensure that personnel in the TSC and EOF have in-dications of direct and derived plant variables as necessary to assess plant status. Supplement 1 to NUREG-0737 contained final requirements and guidance for SPDS that superseded the previous requirements of NUREG-0737. A determina-tion of the adequacy of the Braidwood Station's SPDS will be performed during a post-implementation appraisal, to be conducted against the provisions of Sup-plement 1 at a future date. This item does not require resolution prior to licensing. Onsite monitoring systems and instrumentation used to initiate emergency measures and/or provide continuing assessment are identified. These systems include the following: a meteorological monitoring system with wind speed, direction, and temperature capability; seismic instrumentation to measure ground acceleration levels; radiation monitors in process lines that actually or potentially con-tain radioactive ef fluents; area radiation monitors to measure upward devia-tions in radiation levels in specific locations in the station; fire and smoke detection instruments placed in strategic plant locations; portable dose rate and radiation detection instruments; and laboratory counting and analysis facilities. The plan also indicates the nonradiological process monitors that will be used under accident conditions (such as, reactor coolant system pres-sure and temperature, liquid levels, containment pressure and temperature, flow rates, and so forth). Such process monitors are referenced in the Station's EALs. Hydrological monitors have not been installed because of the plant site's hydrological characteristics, as described in the FSAR and the Plan. EALs are, however, specified for both flood and low water conditions, based on probable maximum precipitation amounts and cooling pond dike failure scenarios, respectively. Provisions for offsite monitoring equipment have been made. Seismic data, respiratory protection equipment, portable detection instruments, and counting room equipment can be obtained from the Dresden, LaSalle County, Quad Cities, Zion, and Byron Stations. The Illinois Department of Nuclear Safety maintains a mobile laboratory equipped with radioassay capability to respond to radiation emergencies. The Environmental / Emergency Coordinator is responsible for the receipt and analysis of all field monitoring data and the determination of where environ-mental sample media will be taken for analysis. Braidwood SSER 1 13A-10 Appendix 13A

The meteorological monitoring equipment installed at the Braidwood Station meets the criteria of Regulatory Guide 1.23, "0nsite Meteorological Programs," dated 1972. The applicant submitted plans (letter dated January 19, 1981) for upgrading the meteorological program as per NUREG-0737, " Clarification of TMI Action Plan Requirements," Task Item III.A.2. This material also ir.cluded a . description of the Offsite Dose Calculation System (0DCS). The plan describes the ODCS and its objectives. These objectives include: (1) meet the meteoro-logical criteria of NUREG-0654, Revision 1; (2) provide, where possible, redun-dant independent pathways of data transmission and redundant data processing computers for use in an emergency situation; (3) provide quick and reasonably accurate estimates of radiation dose to persons living offsite, including preparation of procedures and training of users required to accomplish this assessment; and (4) provide a method for meteorological contractors to secure 4 i meteorological data for assessment of routine releases and to detect equipment failure quickly. The Emergency Plan describes the Braidwood Station's onsite meteorological monitoring system, provisions for the system's routine and emer-gency maintenance, and data review and quality control. Offsite meteorological information can also be obtained from CECO's LaSalle. County and Dresden Stations, i the National Weather Service station in Joliet Illinois, or from CECO's meteor-i ological services contractor. The latter source of data can provide both cur-rent and forecast meteorological information. The Plan indicates that EAL alarms based on offsite dose rates will, in accord-ance with Appendix 1 of NUREG-0654, Revision 1, be factored into the Class A model. The station process computers will process this information and will produce initial transport and diffusion estimates within 15 minutes following. initiation of the calculational procedure. This information will be immediately available to the Control Room operators. In December 1983, the NRC issued Supplement 1 to NUREG-0737 which contained final requirements and guidance for ERFs that superseded the previous require-ments of item III.A.1.2 (Upgraded Emergency Support Facilities) and III.A.2.2 (Meteorological Data) of NUREG-0737. The adequacy of CECO's final ODCS and upgraded meteorological monitoring program will be determined during a future 4 post-implementation appraisal to be conducted against the provisions of NUREG 0737, Supplement 1. l Procedures will be developed for emergency preparedness including quarterly inventory and operational readiness of emergency equipment and supplies. Suf-ficient equipment for emergency kits exists to ensure a minimum inventory in case of replacement delay. The station will maintain portable survey instru-mentation to assess inplant, onsite, and offsite contamination levels, exposure rates, and airborne gaseous, radioiodine, and particulate concentrations. Addi-tionally, during emergency situations, emergency equipment and supplies can be obtained from other CECO nuclear generating facilities. The staff finds that the applicant's Emergency Response Facilities (ERFs) and equipment meet the requirements of 10 CFR Part 50, Appendix E, and the guidance criteria of NUREG-0654 on an interim basis for licensing. As noted above, the staff will confirm the adequacy of the applicant's final ERFs during a post implementation inspection in accordance with the requirements of Supplement 1 of NUREG-0737 on a schedule to be developed between the applicant and the NRC. Braidwood SSER 1 13A-11 Appendix 13A 1

i 13.3.2.9 Accident Assessment The Plan describes several system and radiological effluent parameter values characteristic of a spectrum of offnormal conditions and accidents. Parameter values and other reliable information are tabulated to cross-reference initia-ting conditions for each of the emergency classes. Specific alarm setpoints, both visual and audio, will be in the Control Room to alert the operators. The onsite radiation monitoring capability includes installed process, effluent, airborne and area radiation monitoring systems; portable survey instrumentation; counting equipment for radiochemical analysis; and a personnel dosimetry pro-gram to record integrated exposure. The radiation monitoring system is designed to continuously monitor the containment atmosphere, plant effluents, and various inplant locations. The system includes Control Room readouts and recorders for each parameter that is monitored and an audible Control Room alarm when pre-determined setpoints are exceeded. The system can be subdivided into process / effluent instrumentation and an area monitoring system. The process / effluent instrumentation will consist of pumps, filter samplers, detectors, and associated electronics to determine noble gas, iodine, and par-ticulate concentrations in plant cubicles or in liquid and gaseous effluents. Several monitored effluent pathways have control functions which will terminate the release at predetermined setpoints. These setpoints are premised on compli-ance with federal regulations. The area monitoring system provides information of existing radiation levels in various areas of the plant to ensure safe occu-pancy. It is equipped with Control Room and local readout and audible alarms to warn personnel of an increased radiation level. Two General Atomic Company wide-range monitors will be installed on the auxil-iary building vent stacks (final release points), one monitor per stack. The monitor has a range for radioactive gas concentration 1 x 10 7 pCi/cc to 1 x 10s pCi/cc. Each monitor system has a microprocessor which utilizes digital processing techniques to analyze data and control monitor functions. Control Room readouts include a. chart recorder and an RM-23 remote display module for all monitored parameters. Two General Atomic Company P0-12 detectors will be provided for each of the four main steamlines upstreem of the safety and relief valves. The range of the monitor is 10 1 mR/hr to 104 mR/hr. The monitors will be mounted external to the main steamline piping, and corrections made for the loss of low energy gammas. The General Atomic Company wide range gas monitor includes a sampling rack for collection of the auxiliary building vent stack particulate and radioiodine samples. Filter holders and valves are provided to allow grab sample collec-tion for isotopic analyses in the station's counting rooms. The sampling rack is shielded to minimize personnel exposure. The sampling media will be analyzed by a gamma ray spectrometer which utilizes a Ge(Li) detector. The iodine car-tridges are reverse blown for at least ten minutes to reduce the level of entrapped noble gases. In addition, silver zeolite cartridges are available to further reduce the interference of noble gases. Braidwood SSER 1 13A-12 Appendix 13A

In cases where the instrumentation used for effluent assessment is inoperable or offscale, actual releases will be determined periodically by collecting grab samples, counting the samples, and calculating the releases. Two high range containment radiation monitors will be installed for each oper-ating reactor. The monitors will detect and measure the radiation level within the reactor containment during and following an accident. The range of the monitors is 1 rad /hr to 108 rads /hr (beta gamma) or alternatively,1 R/hr to 107 R/hr for gamma only. Plots of activity in containment (Ci) versus contain-ment radiation reading (R/hr) for each reactor are developed to aid the Control Room operator in an assessment of core damage. These values are related to EALs for rapid classification of an emergency conditions. Monitoring of increasing iodine levels in buildings under accident conditions ~ will include the use of portable instruments using silver zeolite as a sample media. The Braidwood Station will have a Transportable Data Acquisition and Analysis System for analyzing samples that cannot be counted and analyzed in the normal Station counting room because of background problems. Auxiliary counting room locations have been identified within the Turbine Building. The applicant estimates that a sample can be obtained, purged, and analyzed for iodine content within a two-hour time frame. The Station will maintain portable survey instrumentation to assess contamina-tion levels, exposure rates, and gaseous, iodine, and particulate airborne radioactivity concentrations. This equipment will include G-M's, ion chambers, and air samplers. The equipment will be operated and calibrated by Station personnel. The Station counting room contains Ge(Li) gamma spectrometer systems, gas-flow proportional counters for alpha and beta / gamma analysis, and liquid scintillation counters for tritium analysis. The Station will use film badges, TLDs, direct reading pocket ion chambers, and electronic dosimeters to monitor personnel exposures. In addition, a whole body counting system for bioassay determinations will be located onsite. The Plan and Appendix E of the applicant's FSAR describe an extensive post-accident primary coolant and containment atmosphere sampling system. The post-accident primary coolant sampling system wili provide samples from the reactor loops, pressurizer, and the residual heat removal (RHR) system. The system will allow sample collection and analysis within the exposure guidelines given in NUREG-Q737. This system provides analytical capabilities for boron and isotopic analysis for diluted samples (1000 to 1), and online analysis of pH, dissolved oxygen, specific conductivity, chloride, and hydrogen. The contain-ment atmosphere sampling system will provide representative grab samples at the time of an accident and fixed intervals thereafter. The Plan describes means by which operator exposure will be limited when utilizing either the coolant or containment air sampling system. The Plan describes an Offsite Dose Calculation System (ODCS) which meets the design objectives of the NRC Class A model. The system will be computerized and used to predict offsite doses on a real-time basis using effluent and meteorological monitors. The ODCS provides access to meteorological informa-tion at any CECO facility on a real-time basis, and can be accessed from the Control Room, TSC, EOF, and CCC. The ODCS can determine the magnitude of a release or potential release by using any of the following: (1) evaluation of Braidwood SSER 1 13A-13 Appendix 13A

plant conditions, (2) offsite radiological measurements, and (3) dose projec-tions offsite. The ODCS will achieve operational status at the Braidwood Station per the schedule provided in CECO's response to NUREG-0737, Supple-ment 1. The Plan describes the offsite radiological environmental monitoring program, including fixed continuous air samplers and a fixed thermoluminescent dosimeter (TLD) monitoring network which meets the NRC Radiological Assessment Branch Technical Position for Environmental Radiological Monitoring Programs. Maps are provided in the Plan that depict the TLD and air sampler locations. The Plan describes the capabilities and resources for field monitoring within the plume exposure EPZ. Teams will have adequate monitoring equipment to locate and characterize the plume, and make airborne measurements of radiciodine to levels of 1 E-7 pCi/cc under field conditions. Adequate communications systems for the field teams will be provided. 13.3.2.10 Protective Response The Plan describes protective actions to be taken by onsite personnel. The station will have a siren to signal all persons to assemble in predesignated Assembly areas for onsite emergency response personnel are specified in 1 areas. the Plan. Persons not having an emergency response assignment, including visi-tors and contractors, will be required to assemble in other predesignated areas also identified in the Plan. Onsite accountability is the responsibility of the Security Director, who will account for all individuals within the protected area at the time the assembly is announced and will be able to ascertain the names of missing individuals within about 30 minutes, utilizing the computerized security control system. If site evacuation is necessary (such as for a Site Area or General Emergency), personnel will be relocated and monitored at one or more of the following locations: (1) the Dresden Nuclear Generating Station near Morris, Illinois; (2) the LaSalle County Nuclear Generating Station near Marseilles, Illinois; and (3) the Joliet Generating Station in Joliet, Illinois. The Plan indicates the evacuation routes to these relocation areas. Traffic 7 control for onsite areas during a site evacuation will be the responsibility of i the Braidwood Station security force. The Plan describes how radiological moni-toring and decontamination (if necessary) will be provided for evacuees at the offsite relocation site (s). The plan makes provisions for respiratory protec-tion, use of protective clothing, and use of radioprotective drugs for onsite i emergency workers. The criteria for issuance of these protective measures are described in CECO Radiation Protection Standards and radiation / chemistry i procedures. The Plan provides the bases for recommendations for protective actions for the public. These protective action recommendations are consistent with the guid-ance set for in Table 5.1 of the Manual of Protective Action Guides and Protec-l tive Actions for Nuclear Incidents (EPA-520/1-75-001, September 1975 as revised February 1980), and the guidance of the U.S. Food and Drug Administration cover-ing contamination of human food and animal feed (Federal Register, Vol. 43, No. 242, December 15, 1978). The plan summarizes possible recommended protec-tive actions to be made to State and local agencies during an emergency. The applicant's protective action formulation guidance, as stated in the Plan, is consistent with current NRC guidance. The plan clearly indicates that prompt Braidwood SSER 1 13A-14 Appendix 13A

notification will be made directly to offsite authorities responsible for imple-menting protective measures within the plume exposure pathway and ingestion exposure pathway EPZs. Population distribution data by sector and distance within a 50-mile radius have been compiled and are included in the plan. Maps indicating major evacuation routes for the public and station personnel are provided in the plan. Detailed evacuation routes (maps) for the general publi will be contained in the State and local emergency plans. The Plan contains evacuation time data for normal and adverse weather conditions from the August 1985 " Preliminary Evacuatior Time Estimates Within the Plume Ex-posure Pathway Emergency Planning Zone for the Braidwood Nuclear Generating Sta-tion." A review of these estimates was made by the staff and by Dr. Thomas Urbanik, an NRC consultant from the Texas Transportation Institute of Texas A&M University. A final conclusion as to the adequacy of the estimates can not be made until the following three areas are completed: (1) review and approval by local officials; (2) evacuation time estimates for specific special facilities; and (3) consideration of the transport dependent population. No other problems were noted in the general methodology used in the study. The applicant submit-ted the complete evacuation time estimate information for the plume exposure EPZ on February 5, 1986. Evaluation of this information by the staff is an Open Item; the results of the staff review will be reported in a future supple-ment to the SER. The evacuation time estimates may be used by the Environmental / Emergency Coordi-l nator as an aid in determining the recommended protection action for the offsite public (that is, sheltering or evacuation). The applicant's procedures will describe the bases for the choice of recommended protective actions including such factors as evacuation time estimates and local protection afforded in resi-dential units. l 13.3.2.11 Radiological Exposure Control Emergency response personnel may receive radiation exposures in excess of the limits imposed by 10 CFR 20. The Plan indicates that whenever possible, prior authorization of the Station Superintendent, CECO Medical Director, and/or the Station's Radiation Protection Supervisor should be obtained before potentially exposing volunteer emergency workers to doses exceeding these regulatory limits. However, as indicated in Section 13.3.2.2 of this SER, the Plan does not specify which individuals in the onsite and offsite emergency organizations have the ultimate responsibility for authorizing these "once in a lifetime" exposures under the appropriate circumstances. The Plan does, however, contain emergency exposure guidelines for whole-body and thyroid doses which are consistent with the EPA Emergency Worker and Lifesaving Activity Protective Action Guides. The station will provide and distribute self-reading and accumulative type dosi-metry to personnel involved in onsite emergency response, regardless of company ) affiliation. Accumulated exposure records will be maintained and checked daily during en emergency condition. l Provisions for minimizing the affects of radiological exposures or contamina-tion problems include the distribution of respirators, use of protective x Braidwood SSER 1 13A-15 Appendix 13A

clothing, and use of thyroid blocking agents. The Station Director is respon-sible for preventing or minimizing direct or subsequent inhalation exposures due to radioactive materials deposited on the ground or other surfaces. CECO Radiation Protection Standards, used by the Station Director as general methods to be used in contamination control, include criteria for issuance of respira-tory protection and protective clothing. The CECO Medical Director is respon-sible for maintaining adequate supplies of thyroid blocking agents and for establishing the specific policy for its use. Onsite contamination control procedures for personnel, equipment, and access control are in place. Decontamination of personnel and equipment is required when contamination levels exceed predetermined values. Criteria for permitting return of contaminated areas and their contents to normal use are stated in the appropriate contamination control procedures. The station will supply clothing and decontamination materials to onsite per-sonnel required to relocate and found to be contaminated. In addition, the station will provide for monitoring, decontamination, and bioassay capabilities at the relocation sites. 13.3.2.12 Medical and Public Health Support Although there are no resident physicians, nurses, or industrial hygienists on the staffs of CECO nuclear generating stations, select station radiation protec-tion and supervisory personnel will be trained and qualified to administer first aid. The Plan indicates that such personnel will be annually retrained in first aid techniques and that at least one of these individuals will be available onshift at all times. The Braidwood Station will have an inplant first aid / decontamination room. First aid kits, stretchers, sinks, eyewashes, and emer-l gency showers will be placed in various locations throughout the station. Because of the specialized nature of the diagnosis and treatment of radiation injuries, CECO's Corporate Medical Office maintains a roster of physicians especially competent in this area of medicine and available for the care of persons with these special problems. These specialists may be in direct charge of the care of these patients or may serve as consultants to other physicians in charge of their care. In addition, Radiation Management Corporation (RMC) provides health physics and medical support, including bioassay result interpretation. The applicant had made arrangements with the Braidwood Fire Department, con-firmed in writing, for ambulance service for transporting contaminated, injured personnel from the Braidwood Station to a local hospital. This service will be available 24 hours per day. Radiation monitoring will be provided by the sta-tion whenever it becomes necessary to use the ambulance service to transport a contaminated person. The applicant had made arrangements, confirmed in writing, with St. Joseph's Hospital in Joliet, IL,.for receiving and treating contaminated or exposed persons. This hospital is also the supporting hospital for the applicant's Dresden Station, and has been involved in the handling of actual patients from Dresden. This hospital will be utilized for decontamination and initial treat-ment of persons with injuries involving radioactivity and requiring immediate Braidwood SSER 1 13A-16 Appendix 13A

hospital care. CECO will provide medical consultants to aid in any special care necessary at this hospital. Radiation monitoring will be provided by the station whenever it becomes necessary to use the hospital to treat a contami-nated injured patient. Backup medical support, confirmed in writing, is also maintained with North-western Memorial Hospital in Chicago, which is equipped and staffed for dealing with persons having radiation injuries. The Plan states that, whenever neces-sary, contaminated, injured personnel would be transferred to this major facility for extended specialized treatment. By letter dated October 23, 1985, the NRC staff requested that the applicant commit to complying with the Interim Guidance contained in the Commission .Stitement of Policy on Emergency Planning Standard 10 CFR 50.47(b)(12), pub-lished in the Federal Register (50 FR 20892) May 21, 1985. This issue is an Open Item for authorization above 5% of rated power pending a satisfactory response by the applicant to the Commission's policy statement. i l 13.3.2.13 Recovery and Reentry Planning and Post accident Operations The Plan describes an extensive Recovery Organization (Figure 2) which follows the recommendations of the Atomic Industrial Forum and the Institute for Nuclear Power Operations. The Recovery Organization will be activated upon activation of the EOF, which will automatically take place for any Site Area or General Emergency declaration. Designated CECO personnel will assemble at the EOF and assume additional responsibilities for assigned positions. These responsibili-ties are described in the Plan. There will be three major groups of emergency control personnel functioning at the EOF. They are (1) Recovery personnel; (2) Environmental Control personnel; and (3) Emergency News personnel. Recovery personnel function under the direction of the Recovery Manager and serve as the command post for direction of all recovery operations. Environmental Control personnel are under the direction of the Environmertal/ Emergency Coordinator and function to evaluate emergency situations that, affect the public. Emer-gency News personnel operate from the Emergency News Center, which functions as the single point contact to interface with Federal, State, and local authorities who are responsible for dissemina. ting information to the public. A technical spokesperson will be chosen by the Recovery Manager and will have the authority and responsibility to discuss technical problems associated with the emergency. The spokesperson shall be available to brief the press at the Joint Public Information Center. The Recovery Manager is responsible for determining that a Recovery mode may be entered. The Recovery Manager is the designated CECO individual given the requisite authority, management ability, and technical knowledge to manage recovery operations. The primary Recovery Manager is the Division Vice-President and General Manager, Nuclear Stations. The Plan includes guidelines for determining when a Recovery phase can be declared, how to downgrade an emer-gency classification, and when an emergency condition can be declared terminated. Provisions will be made for informing members of appropriate emergency response organizations of any change in emergency classification. The Plan contains provisions which address entry to previously evacuated onsite areas for the purpose of saving lives, search and rescue of missing and injured persons, or manipulation, repair, or recovery of critical equipment or systems. Braidwood SSER 1 13A-17 Appendix 13A

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l The Plan also contains guidelines for recommending re-entry into previously evacuated offsite areas. The Plan describes the Offsite Oose Calculation Sys-tem (0DCS) as the method used for estimating the environmental impact of an unplanned airborne release of a radioactive release from the station, including estimating total population exposures. 13.3.2.14 Exercises and Drills The Plan ensures that an annual exercise will'be conducted at Braidwood to test the adequacy of timing and content of implementing procedures and methods, to test emergency equipment and communications networks, and to ensure that emer-gency personnel are familiar with their duties. Both full participation and partial participation exercises will be conducted. The plan states that once every 6 years, an exercise should be scheduled between the hours of 6.00 p.m. and midnight, and another between midnight and 6:00 a.m. Full participation exercises which test as much of the appropriate plans (licensee, State, or local) as is reasonably achievable without mandatory public participation will be scheduled in order to permit agencies to fulfill their full-scale exercise frequency requirements as delineated in 10 CFR Part 50, Appendix E, Sections IV.F.1 and IV.F.2. This exercise is observed during the NRC's pre-operational inspection program, and has been scheduled for the late fall of 1985. Partial participation exercises will be conducted at the frequency specified in Appendix E of 10 CFR Part 50. This exercise means that appropriate offsite authorities will actively take part in the exercise sufficient to test direc-tion and control functions; i.e., protective action decisionmaking related to EALs and communication capabilities among affected State and local authorities and the licensee. A written scenario will be prepared for each annual exercise. This scenario will include: (1) the basic objectives of the exercise; (2) the date, time period, places, and participating organizations; (3) the simulated events; (4) the time schedule of real and simulated initiating events; (5) arrangements for qualified observers; and (6) a narrative summary describing the conduct of the exercise to include such things as simulated casualties, rescue of personnel, deployment of radiological monitoring teams, and public information activities. A critique will be conducted as soon as practical after each exercise. The critique will discuss the exercise results as to the ability of the GSEP organi-zation to respond to a simulated emergency situation as called for in the GSEP. The Supervisor of Emergency Planning will ensure that when deficiencies in the plan, corresponding implementing procedures, or training program are discovered during exercises and/or drills, necessary corrective actions will be completed. Medical emergency drills, involving a simulated contaminated individual, will contain provisions for participation by local support services agencies (i.e., ambulance and offsite support hospital) and will be conducted annually. Health physics drills will be conducted semiannually. These drills will include re-sponse to, and analysis of, simulated airborne and liquid samples within the plant. At least annually, these drills will include a test of post-accident sampling systems. Plant environs and radiological monitoring drills will be conducted annually. These drills will include collection and analysis of sample media such as water, grass, soil, and air. An assembly / accountability drill Braidwood SSER 1 13A-18 Appendix 13A

will be conducted annually, and will include identifying the locations of all persons within the protected area af ter an assembly is ordered. The station will also conduct unannounced offshift notification drills at least every six i months. These drills shall involve implementation of the station's notifica-i tion procedure and documentation of the times that persons are contacted. These i drills shall serve to demonstrate the capability to augment the onshift staff in a timely manner. Fire drills will be conducted in accordance with station i Technical Specifications. The current revision to the GSEP deleted reference to an annual operator response drill, which was to be conducted in the event that the' annual exercise requirement would have been deleted. This drill is no i longer required since an annual exercise remains a requirement. Reference to an annual operator response drill should, therefore, be deleted from the Braid- ~ 1 wood Annex to remain consistent with the GSEP. J I The capability to notify NRC Region III, FEMA Region V, and other federal emer-l gency response organizations listed in the GSEP Telephone Directory will be demonstrated from the CECO corporate office at least quarterly. The capability to notify the NRC from the Braidwood Station's Control Room, TSC, and E0F will be demonstrated at least monthly, following installation of dedicated communi-cations equipment to be arranged thro'gh the NRC. The capability to notify the u Illinois Emergency Services and Disaster Agency and Illinois Department of Nuclear Safety will be demonstrated, using the NARS, at least monthly. The communications systems described in Section F of this Appendix will be tested annually. These systems include communications between the station, State, and local emergency operations centers, and field assessment teams. The Plan does not indicate whether communications drills with Indiana (the only State other than Illinois within the ingestion pathway EPZ) will be tested quarterly. j Inclusion of provisions for quarterly communications tests with appropriate Indiana emergency response organization (s) is an Open Item. l 13.3.2.15 Radiological Emergency Response Training ) Appropriate initial and annual retraining will be given to all Ceco emergency response personnel. CECO's Production Training Center is responsible for ensuring that necessary training is given; however, the station's Training Department will'actually perform the training of onsite response emergency personnel, under the guidance of the Production Training Center. Station per-1 sonnel who are assigned positions in the offsite emergency organization will receive appropriate additional training from staff members of the Division Vice-President and General Manager, Nuclear Stations. The Supervisor of Emer-l gency Planning will notify the Production Training Center whenever new personnel 1 are assigned emergency organization positions. The Production Training Center will ensure that appropriate initial training, and retraining sessions are scheduled and given. They will also maintain records of all emergency personnel trained. Station personnel not specifically assigned to emergency organization ) positions will be provided with an annual review of the Plan by the Station training staff. i The training program for emergency response personnel allows each member to meet the following objectives: know the objectives of the Plan; understand the i emergency classification system; display an adequate knowledge of personal { responsibilities and duties as listed in the Plan and its implementing proce-dures; know the persons with whom they may interface while performing emergency J i j Braidwood SSER 1 13A-19 Appendix 13A

duties; and display a functional knowledge of the documents necessary to.ful-fill these duties. The proficiency of the applicant's emergency response personnel is ensured by the following means: Assigning persons to emergency duties which are similar to those performed as a part of their regular work assignment; The initial and annual retraining of emergency personnel on applicable generic and site specific portions of the Plan and corresponding Emergency Plan Implementing Procedures; and Participation in exercises and drills designed to sharpen those skills which they are expected to use during a radiological emergency. The Station Superintendent is responsible for making an annual written offer to train those non-Commonwealth organizations, referenced in the Plan and Letters of Agreement, which may provide specialized services during a radio-logical emergency (e.g., firefighting, medical services, transport of injured). This training will acquaint the participants with the special problems poten-tially encountered during a radiological emergency, notification procedures, and their expected roles. Those organizations who must enter the site will also receive onsite training. They will also be instructed in site access procedures and the identity (by position and title) of those persons in the onsite organization who will control their support activities. CECO will offer programs (at least annually) to acquaint news media with the Plan, information concerning radiation, and points of contact for release of public information in an emergency. The Plan does not clearly indicate that Unit 2 construction and other contractor personnel who would be onsite but outside of the temporary protected area bound-ary while Unit I was operational, would receive an initial orientation and annual retraining on the Plan and relevant implementing procedures sufficient to ensure that they are aware of actions they should take during an emergency. This is an Open Item. i 13.3.2.16 Responsibility for the Planning Effort: Development, Periodic Review, and Distribution of Emergency Plans i The Division Vice-President and General Manager, Nuclear Stations, has overall responsibility for radiological emergency response planning within CECO. A staff assigned to this individual has the responsibility for developing and updating the Plan and coordinating the Plan with other response organizations. This staff is headed by the Supervisor of Emergency Planning. Training of emergency planning staff will be performed as a matter of practice. Actual training received is subject to the availabii.ity of appropriate courses and the l availability of individuals to be scheduled for those courses. l To ensure that the Plan and the corresponding implementing procedures are kept current and updated, the Supervisor of Emergency Planning will ensure the fol-lowing: (1) each copy of the Plan will be assigned a serial number; (2) an assignment record will be maintained of all copies; (3) the Plan will be dis-l tributed on a controlled basis to all individuals requiring them; (4) the Plan Braidwood SSER 1 13A-20 Appendix 13A

I will be reviewed and certified current on an annual basis and updated as needed; ~ i (5) all changes to the Plan will be approved by onsite and offsite review com-mittees; (6) all persons in possession of the Plan will receive authorized changes, which will be marked and dated to show where changes have been made; and (7) names and phone numbers of emergency organizations and support person-nel will be reviewed and updated at least quarterly. The generic GSEP also i indicates that Emergency Plan Implementing Procedures (EPIPs) will be developed i and will be reviewed every two years. Per a commitment in the Braidwood FSAR, i Braidwood Station EPIPs will be reviewed annually. t The Plan contains a detailed listing of supporting plans and their sources. A section is the Plan outlines the required contents of implementing procedures, and lists the subjects of procedures required to implement the Plan. The Plan contains a specific table of contents. An independent audit of the Emergency Plan and its implementing procedures will i 4 be conducted on an annual basis by the Ceco Quality Assurance Department. The i Plan states that actions shall be taken for evaluation and correction of all i audit findings. However, it does not state that the scope of these audits will + be in accordance with the requirements of 10 CFR 50.54(t); nor does the Plan indicate that portions of the audits that address the adequacy of the Station's interface with State and local governments will be made available to those i organizations. The Plan must indicate that the scope of independent audits of j the emergency preparedness program will be adequate, and that appropriate por-tions of audit results will be made available to representatives of State and 1 local governments, per the requirements of 10 CFR 50.54(t). This is an Open j Item. i 13.3.3 Conclusion Based on its review, the staff concludes that Commonwealth Edison Company's ) Generating Stations Emergency Plan and the Braidwood site-specific Annex, upon j satisfactory correction of the items listed below, will meet the planning standards of 10 CFR 50.47(b) and the requirements of 10 CFR 50, Appendix E, and conform with the guidance in NUREG-0654, Revision 1. These items are as follows: ) (1) A formal letter of agreement should be executed between CECO and the U.S. Coast Guard which would specify the support to be provided by that organi-zation in response to an emergency condition at the Braidwood Station. 4 (Section 13.3.2.1) 1 (2) The Plan should specify those individuals in the onsite and offsite emer-gency organizations who have the undelegatable authority to authorize emergency worker exposures in excess of regulatory limits. (Sec- { tions 13.3.2.2 and 13.3.2.11) { (3) The Station's Emergency Action Levels (EALs) need to be modified as indicated in this SER. (Section 13.3.2.4) I J (4) A followup message form and provisions for periodically transmitting ade-1 quate followup messages to State and local authorities should be developed. 1 (Section 13.3.2.5) i Braidwood SSER 1 13A-21 Appe.idix 13A 4

(5) The evacuation time estimate study for the Braidwood Station's plume expo-sure EPZ should be modified as indicated in this SER. (Section 13.3.2.10) A letter of commitment to compliance with the Interim Guidance contained (6) in the Commission Statement of Policy on Emergency Planning Standard 10 CFR 50.47(b)(12) is required for licensing above 5% rated power. (Sec-tion 13.3.2.12) Indicate in the Plan that communications drills between the Braidwood (7) Station and the appropriate emergency response organization (s) in Ir. diana will take place quarterly. (Section 13.3.2.14) Indicate in the Plan that construction and other contractor personnel, who (8) would not have duties in the onsite emergency organization, would receive an initial orientation and annual retraining on the Plan and relevant pro-cedures sufficient to ensure that they are aware of actions they should take during an emergency. (Section 13.3.2.15) (9) Indicate in the Plan that the scope of independent audits of the station's emergency preparedness program will be in accordance with the requirements of 10 CFR 50.54(t). Also, indicate in the Plan that portions of such audits that address the adequacy of the Station's interface with State and local governmental organizations will be made available for review by these organizations, per the aforementioned regulatory requirement. (Sec-tion 13.3.2.16) Upon correction of the above items and after receiving the findings and deter-minations of the Federal Emergency Management Agency (FEMA) on the State and local emergency response plans, a supplement to this report will provide the staf f's overall conclusions on the status of emergency preparedness for the Braidwood Station and related emergency planning zones. The staff will con-duct an Emergency Preparedness Implementation Appraisal of the Braidwood Sta-tion as part of its routine pre-licensing program in accordance with 10 CFR 50.47(a)(2). 1 Braidwood SSER 1 13A-22 Appendix 13A

REFERENCES American National Standards Institute, ANSI N18.1-1971, " Experience Require-ments for Plant Manager" Commoraealth letter to NRC describing Braidwood emergency facilities compliance with NUREG-0737 requirements --, Letter to NRC regarding upgrading of meteorological program to meet NUREG-0737 requirements --, letter to NRC regarding response to meet NUREG-0737, Supplement 1 l requirements --, letter to NRC agreeing to provide letters of agreement with NUREG-0654; Revision 1, requirements regarding notification system U.S. Environmental Protection Agency, EPA report EPA-520/1-75-001, " Manual of Protective Action Guides and Protective Actions for Nuclear Incidents," Septem-ber 1975, as revised February 1980. U.S. Nuclear Regulatory Commission, NUREG report NUREG-0654, " Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Pre-paredness in Support of Nuclear Power Plants," Revision 1, October 1981 --, NUREG-0696, " Functional Criteria for Emergency Response Facilities," February 1981 --, NUREG-0737, " Clarification of TMI Action Plan Requirements," November 1981 --, NUREG-0737, Supplement 1, " Clarification of TMI Action Plan Requirements," January 1983 --, NUREG-0800, " Standard Review Plan for Review of Safety Analysis Reports for Nuclear Power Plants," July 1981 l Braidwood SSER 1 13A-23 Appendix 13A -.~.

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15 ACCIDENT ANALYSES 15.2 Normal Operation and Operational Transients 15.2.4 Change in Core Reactivity Transients 15.2.4.3 Rod Cluster Control Assembly Malfunctions The SER indicated that a potential controller problem existed for the dropped control rod event which could lead to the imposition of operating restrictions. It also indicated that a detailed analysis would probably show that even if the problem did occur, thermal limits would not be exceeded. Since then Westing-house has developed a solution for the problem via a new methodology for ana-lyzing the event and has documented it in a topical report (WCAP-10297P). This report and its methodology have been approved by the staff. The solution requires a reactor-cycle-specific analysis showing that departure from nucleate boiling (DNB) limits will not be exceeded. Final Safety Analysis Report (FSAR) Amendment 44 includes a discussion of this analysis, and the results for cycle one operation indicate that DNB limits will be met for this cycle. Thus opera-ting limits will not be necessary for cycle one. Each future reload cycle will require similar cycle-specific analysis as part of the normal reload analysis. 15.3 Desian-Basis Accidents 15.3.6 Reactor Coolant Pump Rotor Seizure and Shaft Break In the SER, the staff requested that the postulated reactor coolant pump locked rotor accident be reevaluated assuming turbine trip and consequential loss of of fsite power and assuming single failure of safety systems. The staff also required the applicant to provide an analysis of a postulated sheared reactor coolant pump shaft accident to verify that the consequences were no more severe than those for a locked rotor. The applicant provided additional information in letters dated June 7, 1982, September 22, 1982, May 2, 1984, and September 26, 1984. The applicant evaluated the time required for offsite power to be lost following the reactor trip / turbine trip which would result from reactor coolant pump shaft failure. The applicant determined that offsite power could not be lost until after the departure from nucleate boiling ratio (DNBR) went through its minimum value during the event. Therefore, loss of offsite power would not affect the amount of fuel failure. The applicant also evaluated the effect of a single active component failure on the event consequences, including single failure of the emergency core cooling system (ECCS), auxiliary feedwater, pres-surizer PORV, and secondary system isolation valves. The thermal-hydraulic transient was determined to be terminated before any single active component failures of these systems could affect the results. The applicant evaluated the consequences of a postulated sheared shaft accident and determined that the consequences were not significantly different from those for a locked rotor. The staff agrees with the appilcant's above conclusions. Braidwood SSER 1 15-1

The staff requested additional analyses of the consequences of a stuck-open secondary relief valve on the offsite dose consequences. Nine percent of the fuel was originally calculated to experience DNB in the FSAR analysis and was assumed to fail. This large amount of fuel failure could have a significant effect on the offsite dose consequences assuming a stuck-open secondary relief valve and operator action to isolate feedwater from the affected steam generator in accordance with Westinghouse Emergency Response Guidelines. If the steam generator with the stuck-open valve were allowed to dry out, a direct path would exist for fission products from failed fuel to pass through any steam generator tube leaks directly into the atmosphere. The applicant's calculation which resulted in 9% of the fuel in DNB assumed an initial power peaking factor of 3.0 for which DNB was assumed to occur as an initial condition for the transient. As discussed in Section 4.3 of the FSAR, the power distribution at Braidwood will be limited so that the maximum peaking factor will be no greater than 2.32. Using the less limiting power peaking as-sumptions, the applicant determined that DNB will not occur for a locked-rotor / sheared-shaft accident. This result was confirmed by the Argonne National Laboratory under contract to the NRC staff (Bollinger). In the absence of DNB, j fuel failure is not predicted to occur. In the event that a secondary relief valve stuck open, the offsite dose conse-quences would be bounded by those of a postulated main steamline break outside containment. These results have already been found to be acceptable by the staff in SER section 15.4.2. The staff concludes that Confirmatory Issue B(13) is closed. 15.4 Radiological Consequences of Accidents 15.4.3 Steam Generator Tube Failure In the Byron /Braidwood FSAR, the applicant made general, unverified assumptions concerning system performance following a complete severance of a single steam generator tube. In addition, the FSAR assumed that the break flow was termi-nated within 30 minutes of the event by operator actions to equalize the primary and secondary pressures. In the original SER, the staff addressed the accident, including the sequence of events and the radiological consequences, and found them acceptable. However, the actual steam generator tube rupture (SGTR) that occurred at the Ginna Nuclear Power Plant indicated that more than 30 minutes could be required for pressure equalization, implying that the Byron /Braidwood analysis may ba nonconservative with respect to assumed operator actions. As a result, the staff requested additional information including an evaluation of operator action times, as to whether liquid can enter the steamlines, and what the effects were on the integrity of steam piping and supports. The staff also requested that a reactor systems analysis be performed for natural circu-lation cooldown with a steam generator tube rupture (SGTR) including the effect of the worst single failure of a system that is either required or expected to operate during the event. By letter dated October 12, 1984, the applicant com-mitted to work with Westinghouse and the NRC staff to find resolution consist-ent with other Westinghouse owners. The staff requires that the applicant sub-t mit all plant-specific information at least 6 months before the first Braidwood refueling outage. This includes plant-specific radiological consequence analyses and steamline static load analyses in the event of overfill. Braidwood SSER 1 15-2

i i To justify safe operation until the SGTR issue is satisfactorily resolved, the i staff notes the following: (1) components necessary for mitigation of the design-basis SGTR are generally safety-related, (2) the emergency procedures for SGTR are based on generic guidelines which have been reviewed and approved for implementation by the NRC staff, and (3) there is a low probability of a design-basis SGTR during initial operation. The staff concludes that there is sufficient assurance that the plant can operate safely until the issue is com-pletely resolved. This is a Confirmatory Issue. I l i l 1 1 l i l i 4 i l i } l 1 i I l ) I l Braidwood SSER 1 15-3 i 1 --- - - -.=

APPENDIX A CONTINUATION OF NRC STAFF RADIOLOGICAL SAFETY REVIEW 0F BRAIDWOOD STATION, UNITS 1 AND 2 October 24, 1983 Letter from applicant concerning rod swap analysis method. October 28, 1983 Letter from applicant concerning the Braidwood Station Security Plan. October 31, 1983 Letter to applicant concerning request for additional in-formation about revised licensed operator requalification topical report. November 1, 1983 Letter to applicant concerning staff evaluation of the l modification to Westinghouse 04/05/E steam generators. l November 3, 1983 Letter to applicant concerning the Emergency Operation l Facilities at Byron and Braidwood. l November 5, 1983 Letter from applicant responding to Generic Letter No. 83-28. November 7, 1983 Letter from applicant concerning a supplemental response to i unresolved issues contained in the draft Safety l Evaluation Report (SER) (NUREG-1002). November 7, 1983 Letter from applicant concerning the control of heavy loads. November 17, 1983 Letter from applicant concerning Security Plan revisions. December 1, 1983 Letter to applicant transmitting two copies of the Braid-wood SER. December 5, 1983 Letter from applicant concerning masonry walls. December 6, 1983 Letter from applicant concerning revised commitment regard-i ing NUREG-0737, Supplement 1 (Generic Letter No. 82-33). I December 6, 1983 Letter from applicant concerning dropped rod reanalysis. December 12, 1983 Letter to applicant transmitting 20 copies of the Braidwood SER. December 15, 1983 Letter from applicant concerning Environmental Report, OL l Stage. December 15, 1983 Letter from applicant concerning revised commitment regard-ing NUREG-0737, Supplement 1 (Generic Letter No. 82-33). ( Braidwood SSER 1 1 Appendix A {

December 16,'1983 Letter from applicant concerning steam generator tube vibration. December 19, 1983 Letter from applicant concerning charging pump deadheading. December 20, 1983 Letter from applicant concerning FSAR Amendment 44. 3 December 21, 1983 Letter from applicant concerning environmental review. December 23, 1983 Letter from applicant concerning implementation of 10 CFR 61 and 10 CFR 20.311. December 27, 1983 Letter from applicant concerning automatic power-operated relief valve (PORV) isolation. December 27, 1983 Letter from applicant concerning instrumentation for the detection of inadequate core cooling. December 28, 1983 Letter from applicant concerning 40 year operating license. December 29, 1983 Letter from applicant concerning NUREG-0737, Supplement 1, safety parameter display system (SPDS) safety analysis. December 30, 1983 Letter from applicant concerning system leakage monitoring. December 30, 1983 Letter from applicant concerning noble gas monitors. December 30, 1983 Letter from applicant concerning pressurizer safety and relief valve. January 3, 1984 Letter from applicant concerning reactor vessel temperature limits. . January 3, 1984 Letter from applicant concerning diesel generator controls. January 4, 1984 Letter from applicant concerning supplemental response to Generic Letter No. 83-10c and d. l January 5, 1984 Letter from applicant concerning post-accident sampling system. January 5, 1984 Letter from applicant transmitting Amendment No. 5 to the Braidwood Environmental Report, OL Stage. January 6, 1984 Letter to applicant transmitting two copies of the Draft Environmental Statement (DES) (NUREG-1026) and advising that 20 additional copies will be forwarded when they are returned from the printer-contractor. January 6, 1984 Letter from applicant concerning control room preliminary design approval (PDA). January 12, 1984 Letter from applicant concerning schedule for Advisory Committee on Reactor Safeguards (ACRS) subcommittee and full' committee meetings. Braidwood SSER 1 2 Appendix A

January 16, 1984 Letter to applicant transmitting 20 copies of the DES for review and comment. February 10, 1984 Letter from applicant concerning supplemental information to the resolution of control of heavy loads at nuclear power plants. February 21, 1984 Letter to applicant concerning deletion of home telephone numbers, unlisted utility numbers, etc., from emergency plans. February 21, 1984 Letter from applicant transmitting the 1983 Annual Report. February 22, 1984. Letter from applicant concerning masonry walls. February 22, 1984 Letter from applicant concerning system leakage monitoring. February 22, 1984 Letter from applicant concerning effects of local intense precipitation. February 22, 1984 Letter from applicant concerning General Design Criterion (GDC) 51 compliance review. February 22, 1984 Letter from applicant concerning hydrogen recombiners. February 28, 1984 Letter from applicant concerning preservice testing of snubbers. March 21, 1984 Letter from applicant concerning response to Generic Letter 83-35, Clarification of TMI Item Plan II.K.3.31 (small-break loss of-coolant accident (SBLOCA) analyses). March 22, 1984 Letter to applicant ~ requesting additional information on fire protection and masonry walls. March 26, 1984 Letter to applicant concerning' supplemental response to Generic Letter 83-10c and d, " Automatic Trip of Reactor Coolant Pumps." March 30, 1984 Letter to applicant concerning mechanical equipment environ-mental qualification program. April 2, 1984 Letter to applicant concerning control room habitability plant visit. April 11, 1984 Letter from applicant concerning supplemental information to the resolution of control of heavy loads at nuclear power plants. April 13, 1984 Letter from applicant concerning operating shift experience. April 19, 1984 Letter to applicant requesting additional information on steam generator tube rupture. Braidwood S3ER 1 3 Appendix A

April 20, 1984 Letter to applicant concerning instrumentation for the detection of inadequate core cooling. April 23, 1984 Letter from applicant concerning diesel generator load testing. April 26, 1984 Letter from applicant concerning circulating water pump trip. April 27, 1984 Meeting of representatives from NRC and Commonwealth Edison Company (CECO) in Bethesda, Md., to discuss alternate means of determining subcooling margin at Byron /Braidwood. (Sum-mary issued May 2, 1984.) May 2, 1984 Letter from applicant concerning Technical Specification changes. May 2, 1984 Letter from applicant concerning reactor coolant pump -transients. May 3, 1984 Letter from applicant concerning testing personnel quali-fications. May 3, 1984 Letter from applicant concerning penetrameter placement during radiography. May 4, 1984 Letter from applicant concerning core damage assessment procedure. May 15, 1984 Letter from applicant concerning Draft Annex to GSEP. May 16, 1984 Letter from applicant concerning response to various fed-eral, state and local comments on the DES. May 17, 1984 Letter from applicant concerning site drainage. May 18, 1984 Letter from applicant concerning diesel generator vibrations. May 24, 1984 Letter from applicant concerning steam generator baseline-inspection. June 1, 1984 _ Letter from applicant concerning Environmental Report, OL Stage. June 4, 1984 Letter to applicant requesting additional information on Braidwood Station Physical Security Plan. June 11, 1984 Letter to applicant requesting additional information on safety-related d.c. system. June 12, 1984 Letter from applicant concerning FSAR Amendment 45. June.18, 1984 Letter from applicant concerning piping design criteria. Braidwood SSER 1 4 Appendix A

June 30, 1984 Letter from applicant concerning supplemental response.to Generic Letter No. 83-28, " Required Actions Based on Generic Implications of Salem ATWS Events." July 6, 1984 Letter from applicant submitting additional information on safety-related d.c. system. July 6, 1984 Letter from applicant concerning subcooling margin monitor. July 10, 1984 Letter to applicant transmitting 20 copies of the Final Environaental Statement (FES) for Braidwood (NUREG-1026). July 16, 1984 Letter from applicant concerning masonry walls. July 19, 1984 Letter to applicant requesting additional information on volume reduction system. July 31, 1984 Representatives from NRC and CECO meet in NRC's Silver Spring, Md., office to discuss safeguards information - NOT RELEASABLE OR OPEN TO PUBLIC. (Summary issued August 14, 1984.) August 7, 1984 . Letter from applicant concerning FSAR change concerning project startup trganization. August 9, 1984 Letter from applicant concerning GDC 51 compliance review. l August 11, 1984 Letter from applicant concerning system leakage monitoring. August 13, 1984 Letter from applicant concerning post-accident sampling system. August 13, 1984 Letter from applicant concerning improved thermal design procedure. August 13, 1984 Letter from applicant concerning inservice inspection progrTm. August 16, 1984 Letter from applicant transmitting Amendment Nos. 44 and 4S FSAR Amendments to the application for construction permits and operating licenses. August-20, 1984 Letter from applicant concerning instrumentation for the detection of inadequate core cooling. August 21, 1984 Letter from applicant concerning diesel generator 2A availability. August 24, 1984 Letter from applicant concerning revision to Braidwood Station Security Plan. August 24, 1984 Letter from applicant concerning volume reduction system. Braidwood SSER 1 5 Appendix A

September 7, 1984 Letter from applicant concerning pipe whip restraints uti-lizing crushable energy absorbing material. September 12, 1984 Letter to applicant requesting additional information on quality assurance (QA) for startup at Braidwood Station. September 17, 1984 Letter from applicant.concerning elimination of postulated pipe breaks in the reactor coolant system (RCS) primary loop. September 18, 1984 Letter from applicant concerning environmental qualifica-tion of equipment. September 20, 1984 Letter from applicant concerning pipe whip restraints uti-lizing crushable energy absorbing material. September 24, 1984 Letter from applicant transmitting the complete Braidwood Station Security Plan. September 25, 1984 Letter from applicant transmitting Drawings 5-15 and 5-16 of the Braidwood Station Security Plan. September 25, 1984 Letter from applicant concerning pipe whip restraints using crushable energy absorbing material. September 25, 1984-Letter from applicant concerning instrumentation for the detection of inadequate core cooling. September 26, 1984 Letter from applicant concerning turbine missiles. September 26, 1984 Letter from applicant concerning pump und valve operability. September 26, 1984 Letter from applicant concerning iodine sampling of stack effluent. September 26, 1984 Letter from applicant concerning reactor coolant pump transients. October 4, 1984 Letter from applicant concerning potential deferral of ser-vice dates. October 4, 1984 Letter from applicant concerning ventilation system filtra-tion. October 11, 1984 Letter from applicant concerning fire protection. October 11, 1984 Letter from applicant concerning Technical Specifications. October 12, 1984 Letter from applicant concerning steam generator tube rupture. October 15, 1984 Letter from applit. ant concerning polar crane speeds. October 15, 1984 Letter from applicant concerning post-accident sampling system. Braidwood SSER 1 6 Appendix A

October 15, 1984 Letter from applicant concerning inadequate core cooling. October 15, 1984 Letter from applicant responding to NRC request for addi-tional information for Question No. 420.24. October 16, 1984 Letter from applicant concerning diesel generator vibration. October 16, 1984 Letter from applicant concerning radwaste solidification system. October 17, 1984 Letter from applicant concerning radiological monitoring. October 18, 1984 Letter from applicant concerning heating, ventilating and air conditioning (HVAC) systems. October 19, 1984 Letter from applicant concerning mechanical equipment qualification. October 23, 1984 Letter from applicant concerning Technical Specifications. October 23, 1984 Letter from applicanti concerning post-accident sampling system. October 24, 1984 Letter from applicant concerning iodine sampling of stack effluent. November 2, 1984 Letter from applicant concerning ASME Code Case N-340. November 5, 1984 Letter from applicant concerning the Braidwood Station Security Plan. November 9, 1984 Letter to applicant requesting additional information on the Byron /Braidwood SPDS. November 15, 1984 Letter from applicant concerning elimination of arbitrary intennediate pipe breaks. November 29, 1984 Letter from applicant concerning startup tests. December 7, 1984 Letter from applicant concerning GDC 51 compliance review. December 11, 1984 Letter to applicant concerning historic site near the Braidwood Station. December 14, 1984-Letter to applicant requesting additional information - Byron /Braidwood Volume Reduction System. December 24, 1984 Letter from applicant concerning revision to the expected fuel load dates for Byron and Braidwood Stations. January 7,1985 Letter to applicant concerning elimination of arbitrary intermediate pipe breaks. 'anuary 9, 1985 Letter from applicant concerning volume reduction system. J Braidwood SSER 1 7 Appendix A

January 9, 1985 Letter from applicant concerning startup test program. January 9, 1985 Letter from applicant concerning ventilation system. January 14, 1985 Letter to applicant concerning design verification activities at Byron and Braidwood Stations. January 14, 1985 Letter from applicant transmitting an antitrust analysis update for Byron Unit 2 and Braidwood Units 1 and 2. Letter from applicant concerning pipe whip restraints uti-January 16, 1985 lizing crushable energy absorbing material. January 17, 1985 Letter to applicant concerning piping design criteria. January 17, 1985 Letter from applicant concerning volume reduction system. January 21, 1985 Letter from applicant concerning startup tests. January 21, 1985 Letter from applicant concerning Annex to GSEP. January 22, 1985 Letter from applicant transmitting advance copies of FSAR revisions to Amendment 46. January 23, 1985 Letter from applicant concerning improved thermal design procedures. January 31, 1985 Letter from applicant concerning startup tests. February 1, 1985 Letter from applicant transmitting FSAR Amendment 46. February 1, 1985 Letter to applicant concerning initial test program. February 5, 1985 Letter from applicant concerning volume reduction system. February 5, 1985 Letter from applicant concerning improved thermal design procedure. February 6, 1985 Letter from applicant concerning FSAR changes. February 6, 1985 Letter from applicant concerning volume reduction system. February 8, 1985 Letter from applicant concerning environmental qualifica-tion of equipment. 'ebruary 11, 1985 Letter from applicant concerning initial test program. February 11, 1985 Letter from applicant concerning source range neutron r monitors. February *11, 1985 Advisory Committee on Reactor Safeguards Report on the Braidwood Station Units 1 and 2 issued. Braidwood SSER 1 8 Appendix A \\

February 15, 1985 Letter from applicant transmitting supplemental response to Generic Letter No. 83-28, " Required Actions Based on Generic Implications of Salem ATWS. Events." I February 15, 1985 Letter to applicant transmitting the ACRS Report on the Braidwood Station, Units 1 and 2. February 20, 1985 Letter to applicant concerning receipt of updated antitrust information on Byron Station, Unit 2, and Braidwood Station, Units 1 and 2. Federal Register notice included to be published on February 28, 1985, with 30-day intervention period. February 20, 1985 Letter from applicant transmitting an affidavit stating that distribution of FSAR Amendment No. 46 for Byron Units 1 and 2 and Braidwood Units 1 and 2 has been made. February 22, 1985 Letter from applicant transmitting a copy of the 1984' Annual Report for Commonwealth Edison Company. February 22, 1985 Letter from applicant concerning instrumentation for the detection.of inadequate core cooling. February 25, 1985 Letter from applicant transmitting the Revisions to Braid-wood and Byron security plans. Revision 10 was transmitted for Braidwood and Revision 13 for Byron. February 26, 1985 Letter to applicant concerning limiting condition for oper-ation relaxation program. February 26, 1985 Letter from applicant requesting an exemption from obser-vation training per 10 CFR 55.7. March 1, 1985 Letter from applicant transmitting Amendment 6 to the Fire Protection Report. March 1, 1985 Letter from applicant concerning archeological site 11Ka179, determination of effect. March ~8, 1985 Letter from applicant concerning FSAR changes. March 12, 1985 Letter to applicant requesting additional information fol-lowing Preliminary Staff Review of Licensee Responses to Generic Letter 83-28. March 13, 1985 Letter to applicant concerning automatic shunt trip for reactor trip breakers. March 15, 1985 Letter from applicant concerning Annex to GSEP. March 27, 1985 Meeting of representatives from NRC, Brookhaven National Laboratory, CECO, and Westinghouse in Bethesda, Md., to discuss Byron /Braidwood Limiting Condition for Operation Relaxation Program. (Summary issued March 29, 1985.) Braidwood SSER 1 9 Appendix A

April 10, 1985 Meeting of representatives from NRC and CECO in Bethesda, Md., to present applicant's design verification activities for Byron Unit 2 and Braidwood Units 1 and 2. (Summary issued April 29, 1985.) April 16, 1985 Letter from applicant concerning updated FSAR. April 24, 1985 Letter from applicant concerning design verification activities. April 29, 1985 Letter to applicant transmitting copies of the legal ad-vertisements placed in the approved trade journals for Byron and Braidwood Stations. May 6, 1985 Letter from applicant concerning reactor vessel materials. May 14, 1985 Letter to applicant concerning site visit to Braidwood Station. May 17, 1985 Letter to applicant concerning Generic Letter 83-28, Item 1.1, " Post-Trip Review Braidwood Station." May 24, 1985 Letter to applicant concerning TMI Action Plan Item II.K.3.30. May 29, 1985 Letter from applicant concerning Initial Test Program. June 7, 1985 Letter from applicant concerning radiation protection manager training. June 10, 1985 Letter to applicant concerning environmental assessment and finding of no significant impact, exemption from submittal of an updated FSAR. June 10, 1985 Letter to Westinghouse withholding from public disclosure CAW-34-44, WCAP-10553 and WCAP-10554, " Technical Bases for Eliminating Large Primary Loop Pipe Rupture as the Struc-tural Design Basis for Byron and Braidwood." June 10, 1985 Letter from applicant transmitting an amendment to the Braidwood Environmental Report. Enclosed were 3 originals and 5 copies; the remaining 36 copies to follow by slower mail. June 10, 1985 Letter from applicant transmitting Amendment No. 7 to the Environmental Report, OL Stage for the Braidwood Station. June 17, 1985 Letter from applicant concerning a response to Generic Letter 85-02, " Steam Generator Tube Integrity." June 18, 1985 Letter from applicant concerning ASME Code Cases N-403 and N-413. June 19, 1985 Letter from applicant concerning fees for Technical Speci-fication revisions. Braidwood SSER 1 10 Appendix A

June 19, 1985 Letter from applicant concerning piping design criteria. June 20, 1985' Letter ~from applicant transmitting SER Section 3.9.6, " Pre-service / Inservice Testing of Pumps and Valves." June 28, 1985 Letter from applicant concerning elimination of postulated pipe breaks _in the RCS primary loops. July 9, 1985 Letter from applicant responding to a request for additional ~ information to FSAR Question 423.45 on diesel generator testing. July 11, 1985 Letter from applicant transmitting an affidavit for Amend-ment 7 of the Braidwood Environmental Report, OL Stage. July 11, 1985 Letter to applicant concerning Generic Letter 83-28, Item 1.1, " Post-Trip Review for Byron /Braidwood." July 15, 1985 Letter to applicant concerning supplemental request for additionel information on Class 1E cables at Byron Unit 2 and Braidsood Units 1 and 2. July 17, 1985-Letter from applicant concerning preservice inspection non-destructivt testing program plan. July 17, 1985 Letter from applicant concerning Technical Specifications for Braidwood Units 1 and 2. July 18, 1985 Letter to applicant concerning use of ASME Code Case N-411 for Byron Units 1 and 2 and Braidwood Units 1 and 2. August 2, 1985 Letter from applicant concerning environmental effects of high-energy-line breaks. August 6, 1985 Letter from applicant concerning supplemental request for additional information on separation criteria. August 20, 1985 Letter from applicant concerning SPDS. August 21, 1985 Letter from applicant concerning supplemental request to Generic Letter No. 83-28, " Required Actions Based on Generic Implications of Salem ATWS Events." August 22, 1985 Letter from applicant responding to Generic Letter 85-12, " Implementation of TMI Action Item II.K.3.5 - Automatic Trip of Reactor Coolant Pumps." August 22, 1985 Letter from applicant concerning supplemental response to Generic Letter 85-02. August 23, 1985 Letter to applicant concerning use of ASME Code Cases N-403 and N-413 for Byron and Braidwood Stations. Braidwood SSER 1 11 Appendix A

August 23, 1985 Letter to applicant requesting additional information on vibration of diesel generator instrumentation at Byron and Braidwood Stations. August 27, 1985 Letter to applicant transmitting an Exemption from Submit-tal Date for Updated' Final Safety Analysis Report (FSAR) for Byron /Braidwood. September 5, 1985 Letter from applicant concerning fluid jet impingement analyses. September 11, 1985 Letter from applicant transmitting supplemental informa. tion for SER Open Items 3.9.3.2, 3.10, and 3.11. September 17, 1985 Letter to applicant concerning Generic Letter 83-28, Items 3.1.1 and 3.1.2, " Post Maintenance Testing Verifi-cation (RTS Components)." September 25, 1985 Letter from applicant requesting amendment to construc-tion permit to reflect.an exemption from GDC 4. October 7, 1985 Letter to applicant concerning initial review of the First Draft Technical Specifications for Braidwood Units 1 and 2. October 7, 1985 Letter to applicant concerning visual weld inspection requirements. October 7, 1985 Letter from applicant concerning startup test program. October 8, 1985 Letter to applicant concerning acceptance of criteria for firecode CT gypsum fire stops at Byron and Braidwood Stations. October 10, 1985 Letter from applicant concerning schedule for complying with 10 CFR 50.62, dealing with ATWS. October 23, 1985 Letter to applicant concerning interim guidance on emer-gency planning standard 10 CFR 50.47(b)(12) regarding Byron Unit 2 and Braidwood Units 1 and 2. October 24, 1985 Letter to applicant concerning Draft Technical Evaluation Report (TER) for Generic Letter 83-28, Item 1.2. November 5, 1985 Letter to applicant concerning Safety Evaluation Report for Generic Letter 83-28 Items 3.1.3 and 3.2.3, " Post-Maintenance Testing," for Byron and Braidwood Stations. November 15, 1985 Letter from applicant concerning a model implementation schedule. November 18, 1985 Letter to applicant concerning revision to the expected fuel load dates for Byron Unit 2 and Braidwood Units 1 and 2. Braidwood SSER 1 12 Appendix A

November 19, 1985 . Letter to applicant concerning Emergency Planning Public Information Booklet. November 20, 1985 Letter from applicant responding to First Draft Technical Specifications. i December 11, 1985 Letter from applicant concerning environmental effects of high energy-line breaks. December 16, 1985 . Letter from applicant concerning NF Jurisdictional Boun-dary for Component Supports. January 13, 1986 Letter to applicant transmi.tting the Braidwood Units 1 and 2 Technical Specifications, Second Oraft. i January 17, 1986 Letter from opplicant concerning pressurized thermal shock. January 28, 1986 Letter from applicant concerning containment coatings. February 4,1986 Meeting of representatives from NRC, CECO, Sargent and Lundy (S&L), Carb. Co., and Region III in Bethesda, Md., to discuss the Braidwood containment liner coating system. (Summary issued April.4, 1986.) February 5, 1986 Letter from applicant concerning emergency evacuation time estimates. February 10, 1986 Letter from applicant concerning containment coatings. February 13, 1986 Letter to applicant concerning startup test program. February 25, 1986 Letter to applicant concerning Byron /Braidwood Supplemen-tal Safety Evaluation Report for Physical Identification and Independence of Redundant Safety-Related Electrical Systems. February 28, 1986 Letter from applicant concerning preservice inspection (PSI) disposition of indications for loop 1 steam gener-ator and the pressurizer. March 3, 1986 Letter from applicant concerning Braidwood Annex to GSEP. March 3, 1986 Letter from applicant concerning Environmental Protection Plan (Nonradiological), Appendix B of Operating License. March 4, 1986 Letter to applicant concerning Supplemental Safety Evalu-ation Report concerning emergency planning for Braidwood Station, Units 1 and 2. l March 12, 1986 Letter from applicant concerning inspection of cast stain-less steel component welds. March 18, 1986 Letter from applicant concerning change in CECO Attorney of Record. Michael I. Miller, Esq., is the legal contact. ,Braidwood SSER 1 13 Appendix A

April 1, 1986 Letter from applicant transmitting revised Emergency Oper-ations Facility Procedures. April 1, 1986 Letter from applicant concerning post-accident control room habitability analysis for Braidwood Station. April 2, 1986 Letter from applicant concerning reactor vessel nozzle analysis for Braidwood Unit 2. April 7, 1986 Letter from applicant responding to Second Draft of Tech-nical Specifications. April 9, 1986 Letter to applicant concerning chloride content for the Category I concrete structures at Braidwood Units 1 and 2. April 21, 1986 Letter to applicant concerning Generic Letter 83-28, Items a.2.1 and 4.2.2. April 22, 1986 Letter from applicant concerning Supplemental Safety Eval-uation Report for Physical Identification and Independence of Redundant Safety-Related Electrical Systems FSAR Changes and S&L Design Manual Change for Byron and Braidwood Units 1 and 2. April 23, 1986 Letter from applicant concerning FSAR Amendment No. 47. April 25, 1986 Letter from applicant concerning TMI Action Plan Item I. A.1.1, "Shif t Technical Advisor." April 29, 1986 Letter from applicant concerning environmental effects of high-energy-line breaks. April 29, 1986 Letter to applicant transmitting Amendment No. 1 to Con-struction Permits CPPR-131 (Byron Unit 2) and CPPR-132 and CPPR-133 (Braidwood Units 1 and 2) to include the exemption from GDC 4 issued October 25, 1985. May 5, 1986 Letter to applicant concerning safety evaluation of com-pliance with Item 1.2 of Generic Letter 83-28. May 9, 1986 Letter from applicant concerning TMI Action Plan Item I. A.1.1, "Shif t Technical Advisor." May 27, 1986 Letter from applicant concerning Security Plan for Braid-wood Units 1 and 2. May 27, 1986 Letter from applicant concerning Braidwood Unit 1 supple-mental summary of changes to Amendment No. 7 to the Fire Protection Report. June 2, 1986 Letter from applicant concerning NEPA Code deviations for Braidwood Unit 1. Braidwood SSER 1 14 Appendix A

June 4, 1986 Letter from applicant concerning a model implementation schedule. June 4, 1986 Letter from applicant concerning Braidwood Unit 1 PDA of Braidwood Unit 1 control room. June 10, 1986 Letter to applicant concerning proof and review copy of Technical Specifications for Braidwood Units 1 and 2. June 11, 1986 Letter from applicant concerning Braidwood Unit 1 preser-vice inspection steam generators and pressurizer. June 11, 1986 Letter from applicant concerning Fire Protection Report, Amendment 8. June 18, 1986 Letter from applicant concerning Braidwood licensed oper-ator hot participation experience resumes. June 23, 1986 Letter from applicant concerning Startup Test Program. June 25, 1986 Letter to applicant concerning Safety Evaluation Report for Generic Letter 83-28, Item 2.1 (Part 1), for Braidwood Station Units 1 and 2. June 25, 1986 Letter to applicant concerning Safety Evaluation Report for Generic Letter 83-28, Items 3.2.1 and 3.2.2, for Braidwood Units 1 and 2. June 25, 1986 Letter from applicant concerning inspection of cast stain-less steel component welds. July 1, 1986 Letter from applicant commenting on proof and review copy of Technical Specifications for Braidwood Units 1 and 2. July 1, 1986 Letter from applicant concerning Seismic Equipment Quali-fication and Electrical and Mechanical Environmental Equipment Qualification Supplemental Information. July 1, 1986 Letter from applicant concerning Federal Emergency Manage-ment Agency (FEMA) Interim Finding on Offsite Emergency. July 2, 1986 Letter from applicant concerning confirmation of design adequacy for jet impingement effects. July 2, 1986 Letter from applicant concerning exceptions to FSAR Appen-dix A Regulatory Guides 1.52 and 1.140. July 17,1986 Letter from applicant commenting on proof and review copy of Technical Specifications. July 22, 1986 Letter from applicant concerning evaluation of environ-mental effects of main steam line break outside contain-ment (IE Information Notice 84-90). Braidwood SSER 1 15 Appendix A

July 30, 1986 Letter from applicant concerning Braidwood licensed oper-ator hot participation experience resumds. July 30, 1986 Letter from applicant concerning IE Information Notices 86-02 and 86-03. July 31, 1986 Letter from applicant concerning preservice inspection program notes and relief requests. August'1, 19'86 Letter from applicant concerning reactor vessel nozzle analysis. August 4, 1986 Letter from applicant transmitting supplemental informa-tion to Amendment 7 of the Fire Protection Report and National Fire Protection Association (NFPA) Code Deviations. August 6, 1986 Letter from applicant concerning Braidwood Units 1 and 2 Technical Specifications. l August 6, 1986 Letter to applicant concerning revisions to the Braidwood security plans. t J i 16 Appendix A Braidwood SSER 1 .)

APPENDIX B BIBLIOGRAPHY Bollinger, J. S., "RELAPS/ MOD 1.5/FRAP-T6 Analysis of a Siezed Shaft / Sheared Shaft Transient," Report No. ANL/ LWR /NRC 83-11, September 1983. I Buchalet, C. B., and W. H. Bamford, " Stress Intensity Factor Solutions for Continuous Surface Flaws in Reactor Pressure Vessels," in Mechanics of Crack Growth, ASTM STP 590, 1976, pp. 385-402. Burns, J. J., Jr., " Reliability of Nuclear Power Plant Steam Turbine Overspeed Control Systems, 1977 ASME Failure Prevention and Reliability Conference, Chicago, IL, September 1977, p. 27. Bush, S. H., " Probability of Damage to Nuclear Components Due to Turbine Failure," Nuclear Safety, Vol. 14, No. 3, p. 187 (May-June) 1973. Clark, W. G., Jr., B. B. reth, and D. H. Shaffer, " Procedures for Estimating the .the Probability of Steam lurbine Disc Rupture From Stress Corrosion Cracking," ASME/IEEE Power Generation Conference, October 4-8, 1981, St. Louis, M0. Czajkowski, C. J., " Investigation of Shell Cracking on the Steam Generators at Indian Point No. 3," Brookhaven National Laboratory, NUREG/CR-3281, June 1983. Electric Power Research Institute (EPRI), Special Report NP-2628-SR, "EPRI PWR Safety and Relief Valve Test Program - Safety and Relief Valve Test Report," December 1982. Kalderon, D., " Steam Turbine Failure at Hinkley Point A," Proceedings of the Institute of Mechanical Engineers, Vol. 186, No. 31/72, p. 341, 1972. Letter from O. Kingsley, Alabama Power and Light Co., to S. Chilk, NRC, dated July 27, 1982, transmitting WCAP-10105, " Review of Pressurizer Safety Valve Performance as Observed in the EPRI Safety and Relief Valve Test Program." McGowan, J. J., and M. Raymund, " Stress Intensity Factor Solutions for Internal Longitudinal Semi-elliptic Surface Flaw in a Cylinder Under Arbitrary Loading," ASTM STP 677, 1979, pp. 365-380. Newman, J. C., Jr., and I. S. Raju, " Stress Intensity Factors for Internal Sur-face Cracks in Cylindrical Pressure Vessels," ASME Trans., Jour. of Pressure Vessel Technology, Vol. 102, 1980, pp. 342-346. } Shah, R. C., and A. S. Kobayashi, " Stress Intensity Factor for an Elliptical Crack Under Arbitrary Loading," Engineering Fracture Machanics, Vol. 3, 1981, pp. 71-96. l Braidwood SSER 1 1 Appendix B t

1 1 Twisdale, L. A., W. L. Dunn, and R. A. Frank, " Turbine Missile Risk Methodology and Computer Code," EPRI Seminar on Turbine Missile Effects in Nuclear Power Plants, Palo Alto, CA, October 25-26, 1982. U.S. Atomic Energy Commission, WASH-1400 (NUREG/75-014), " Reactor Safety Study: An Assessment of Risks in U.S. Commercial Nuclear Power Plants," December 1975. U.S. Court of Appeals, Guard v. NRC, 753 F.2d 1144 (D.C. Cir. 1985). U.S. Environmental Protection Agency, EPA report EPA-520/1-75-001, " Manual of Protective Action Guides and Protective Actions for Nuclear Incidents," September 1975, revised February 1980. U.S. Nuclear Regulatory Commission, Generic Letter 84-16, " Adequacy of On-Shift Operating Experience for Near Term Operating License Applicants," June 27, 1984. --, NUREG-0452, " Standard Technical Specifications for Westinghouse Pressurized Water Reactors," Revisions of May 1978 and June 1978; Revision 2, July 1980; Rev. 4, November 1981. --, NUREG-0577, " Potential for Low Fracture Toughness and Lamellar Tearing on PWR Steam Generator and Reactor Coolant Pump Supports," October 1979. --, NUREG-0612, " Control of Heavy Loads at Nuclear Plants," July 1980. --, NUREG-0654/ FEMA-REP-1, Rev. 1, " Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants," November 1980. --, NUREG-0717, Supp. 4, " Safety Evaluation related to the operation of Virgil C. Summer Nuclear Station," Docket No. 50-395, August 1982. --, NUREG-0696, " Functional Criteria for Emergency Response Facilities," February 1981. --, NUREG-0737, " Clarification of TMI Action Plan Requirements," November 1980; Supplement 1, December 1982 (Generic Letter 82-33). --, NUREG-0800 (formerly NUREG-75/087), " Standard Review Plan for Review of Safety Analysis Reports for Nuclear Power Plants--LWR Edition," July 1981. --, NUREG-1014, " Safety Evaluation Report Related to Model D4/05 Steam Gen-erator Design Modification," October 1983. --, NUREG-1097, " Technical Specifications for Byron Station Unit 1," Docket No. 50-454, October 1984. --. NUREG/CR-1884, Lawrence Livermore Laboratory, " Observations and Comments on the Turbine Failure at Yankee Atomic Electric Company, Rowe, Massachusetts," March 1981. U.S. Nuclear Regulatory Commission, Office of Inspection and Enforcement (IE) Bulletin 80-06, " Engineered Safety Feature (ESF) Reset Controls," March 13, 1980. Braidwood SSER 1 2 Appendix 8

--, Inspection Report No. 50-455/84-7, June 12, 1984. --,' Inspection Report No. 50-454/84-19, May 18, 1984. --, Inspection Report No. 50-454/84-24, June 12, 1984. --, Inspection Report No. 50-455/84-48 (DRP), November 26, 1984. --, Inspection Report No. 50-454/84-70 (ORP), November 26, 1984. --, Inspection Report No. 50-457/85036 (ORSS), November 27, 1985. --, Inspection Report No. 50-456/85037 (DRSS), November 27, 1985. --, Inspection Report No. 50-456/86024, July 18, 1986. --, Inspection Report No. 50-456/86-32, August 14, 1986. W:stinghouse Electric Corporation, WCAP-9500, " Reference Core Report 17x17 Optimized Fuel Assembly," July 1979. --, WCAP-9804, "Probabilistic Analysis and Operation Data in Response to NUREG-0737 Item II.K.3.2 for Westinghouse NSSS Plants," February 1981. --, WCAP-10297P, " Dropped Rod Methodology for Negative Flux Rate Trip Plants," January 1983. --, WCAP-11063, " Background and Technical Basis for the Handbook on Flaw Evaluation for Byron Units 1 and 2 Steam Generators and Pressurizers," June 1986. --, WCAP-11064, " Handbook on Flaw Evaluation for Byron Units 1 and 2 Steam Generators and Pressurizers,'? June 1986. W:stinghouse Owners Group (WOG), " Post-Accident Core Damage Assessment Method-odology," Revision 1, March 1984. Wyle Laboratories, Test Report No. 17769-1, " Test Report on Electrical Cable Siparation Verification Testing for Sargent and Lundy Engineers and Commonwealth ' Edison Company for Use in Illinois Power Company's Clinton Nuclear Generating Station and Commonwealth Edison Company's Byron and Braidwood Nuclear Generating Stations," August 23, 1985. --, Test Report No. 46511-3, " Test Report on Verification Testing Between Class IE and Non-Class 1E Power Cables in Raceways," March 28, 1983. Industry Codes and Standards American National Standards Institute (ANSI) ANSI N18.1-1971, " Experience Requirements for Plant Manager." --, ANSI N18.17-1973, Industrial Security for Nuclear Power Plants." Braidwood SSER 1 3 Appendix B

-. _ - _ = 'American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section III. --, Section VIII. l --, Section XI,1977 Edition and Addenda Through Summer 1978. American Society for Testing and Materials (ASTM), E-185-73, " Standard Recom-mended Practice for Surveillance Test for Nuclear Reactor Vessel." Institute of Electrical and Electronic Engineers (IEEE), 323-1974, " Standard for Qualifying Class.1E Equipment for Nuclear Power Generating Stations," February 28, 1974. --, 344-1975, " Recommended Practices for Seismic Qualification of Class 1E-Equipment for Nuclear Power Generating Stations," January 31, 1975. l l l l 1 Braidwood SSER 1 4 Appendix B

APPENDIX F NRC STAFF CONTRIBUTORS AND CONSULTANTS Name Title Review Branch *. Frederick H. Burrows Reactor Engineer Electrical /Instrumenta-tion and Control Systems (PWR-A) Frank C. Cherney Section Leader Engineering Issues Barry J. Elliot Materials Engineer Engineering (PWR-A) Ronald N. Gardner Section Chief Region III Robert J. Giardina Reactor Systems Plant Systems (PWR-A) Engineer, Mechanical C. Gary Hammer Mechanical Engineer Engineering (PWR-A) Marthe E. Harwell Section Leader Editioral Section (TIDC) John J. Hayes, Jr. Nuclear Engineer Plant Systems (PWR-A) Yi-Hsiung Hsii Nuclear Engineer Reactor Systems (PWR-A) Tai L. Huang Nuclear Engineer Reactor Systems (BWR) Walton L. Jensen Nuclear Engineer Reactor Systems (PWR-A) l Rudy O. Karsch Nuclear Engineer Reactor Systems (PWR-A) Michael A. Lamastra Radiation Engineer Plant Systems (BWR) Robert A. Meck Emergency Preparedness Emergency Preparedness Specialist Leonard N. Olshan Byron Project Project Directorate #3 Manager (PWR-A) Robert L. Perch Project Manager, Facility Operations l Technical Specifi-cations l

• Reflects reorganization since SER was issued.

Braidwood SSER 1 1 Appendix F

Name Title Review Branch

• Thomas J. Ploski Emergency Preparedness Region III Specialist James C. Pulsipher Containment Systems Engineering Branch Engineer (PWR-A)

Jai R. N. Rajan Mechanical Engineer Engineering (PWR-8) Sang Chil Rhow Reactor Systems Electrical /Instrumenta-Engineer, Electrical tion and Control Systems (PWR-B) Howard J. Richings Reactor Physicist Reactor Systems (BWR) Madelyn M. Rushbrook Licensing Assistant Project Directorate #5 (PWR-A) John 0. Schiffgens Materials Engineer Facility Operations (PWR-A) Gary B. Staley Hydraulic Engineer Engineering (PWR-B) Thomas M. Tongue Senior Resident Region III Inspector Jared.S. Wermiel Section Leader Plant, Electrical, Instrumentation and Control Systems (PWR-8) Paul C. Wu Chemical Engineer Reactor Construction Programs I i

• Reflects reorganization since SER was issued.

Braidwood SSER 1 2 Appendix F

l APPENDIX H ACRS REPORT ON BRAIDWOOD STATION, UNITS 1 AND 2 Braidwood SSER 1 Appendix H

l l l l / 'g UNITED STATES / NUCLEAR REGULATORY COMMISSION c { ,l ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON, D. C. 30666 S,

• 1. g e...e February 11, 1985 Honorable Nunzio J. Palladino Chairman U. S. Nuclear Regulatory Comission Washington, D. C. 20555

SUBJECT:

ACRS REPORT ON THE BRAIDWOOD STATION, UNITS 1 AND 2 During its 298th meeting, February 7-9, 1985, the Advisory Comittee on Reactor Safeguards reviewed the appifcation of the Comonwealth Edison Company (the Applicant) for a license to operate the Braidwood Station. Units 1 and 2. The ACRS comented on the construction permit applica-tion for this Station (and the sister Byron Station) 1.1 its report dated May 13, 1975. A tour of the facility was made by members of a Subcom-mittee on March 8, 1984, and Subcomittee meetings to consider this application were held in Joliet, Illinois on March 8 and 9,1984, and Washington, D. C. on January 29, 1985. During our review, we had the benefit of discussions with the NRC Staff and representatives and consultants of the Applicant, Westinghouse Electric Corporation, and Sargent & Lundy Corporation. We also had the benefit of the documents listed. l The Braidwood Station is located in Will County, Illinois, about 20 i miles south-southwest of Joliet, Illinois. The Station ~ uses two Westinghouse four-loop pressurized water reactors, each having a rated power level of 3425 MWt. Each is housed in a steel-lined, reinforced concrete containment building with a design pressure of 50 psig. With the exception of site-related matters (Braidwood uses an artificial cooling pond for heat rejection, whereas Byron uses cooling towers), the l Braidwood plant design is identical to that of Byron, which was granted an operating license on October 31, 1984. The safe shutdown earthquake for the Braidwood site (as for Byron) has been set at 0.209, the operat-l ing basis earthquake at 0.09g. l Construction of Unit 1 is about GO percent complete, and Unit 2 is about 54 percent complete. The Applicant currently estimates the fuel load date to be April 1, 1986 for Unit 1, and for Unit 2 to be July 1, 1987. At the time the Subcomittee considered the Braidwood Station in March 1984, the Applicant was engaged in a major reorganization and strength-ening of provisions for quality assurance / quality control matters. This was a consequence of problems encountered in the area of quality assurance / quality control, particularly on its Byron project, but to some degree also in connection with its Braidwood project. Since that l Braidwood SSER 1 1 Appendix H

Honorable Nunzio J. Palladino February 11,1985 time, increased attention has been given to this area at all levels in the Applicant's organization. Work at the Braidwood Station has been reviewed as part of the continuing Systematic Assessment of Licensee Performance (SALP) program being carried out by NRC Region III personnel, by a larger than normal Construction Appraisal Team (CAT) inspection conducted by the Office of Inspection and Enforcement, and by an Institute of Nuclear Power Operations (INPO) evaluation. Though problems have, of course, been identified as a result of these inspections and reviews, the deficiencies found have either been i remedied or are in the course of being corrected. In addition to the ~ above. the Applicant has undertaken an extensive internal review, the Braidwood Construction Assessment Program (BCAP), wherein extensive j reinspection of completed work and records is being conducted by qualified teams not connected with those engaged in the original work. ) From these various independent and wide-ranging inspections and reviews ] (along with the associated corrective actions) there is a good basis for ) confidence that the quality of the construction and of the managing i organization for the Braidwood Station will be fully adequate. During our meeting, the NRC Staff identified a small number of open items that must be resolved prior to the granting of an operating i license. It is expected that these will be resolved in a manner satis-factory to the NRC Staff. We believe that, subject to the resolution of open items identified by the NRC Staff, and subject to satisfactory completion of construction, staffing, and preoperational testing, there is reasonable assurance that the Braidwood Station, Units 1 and 2 can be operated at power levels up to 3425 MWt without undue risk to the health and safety of the public. Sincerely, David A. Ward Chairman

References:

1. Comonwealth Edison Company. " Final Safety Analysis Report. Byron /Braidwood Stations," Volumes 1-14 and Amendments 1-45 2. U. S. Nuclear Regulatory Comission, " Safety Evaluation Report Related to the Operation of Braidwood Station, Units 1 and 2 " USNRC Report NUREG-1002, dated November 1983 3. U. S. Nuclear Regulatory Comission ASLB Initial Decision in the Matter of Comonwealth Edison Company (Byron Station Units 1 and 2), LBP-84-2, dated January 13, 1984 4. U. S. Nuclear Regulatory Commission, " Final Environmental Statement Related to the Operation of Braidwood Station Units 1 and 2 " Chapter 4. Section 4.3.1.1.5, Effects of Cooling Pond Dike Failure, USNRC Report NUREG-1026, dated June 1984 Braidwood SSER 1 2 Appendix H

_=. _. Honorable Nunzio J. Palladino February 11,1985 5. U. S. Nuclear Regulatory Commission 'ASLB Supplemental Initial Decision in the Matter of Comonwealth Edison Company (Byron Station, Units.I and 2), LBP-84-41, dated October 16, 1984 6. Handouts received from Stanley E. Campbell on behalf of Sinnissippi Alliance for the Environment at the March 8-9, 1984 Subcomittee meeting at Joliet, Illinois i I l Braidwood SSER 1 3 Appendix H

i APPENDIX I ERRATA TO BRAIDWOOD STATION, UNITS 1 AND 2, SAFETY EVALUATION R'LIORT Braidwood SSER 1 Appendix I

l APPENDIX I ERRATA TO BRAIDWOOD STATION, UNITS 1 AND 2 SAFETY EVALUATION REPORT l Page Line Change 1-12 18 Delete "(8) control room human factors review 18.0". I i l l l l Braidwood SSER 1 1 Appendix I l

= ~. ~ - - I i i f APPENDIX J TECHNICAL EVALUATION REPORT ON CONTROL OF HEAVY LOADS AT NUCLEAR POWER PLANTS FOR BRAIDWOOD STATION, UNITS 1 AND 2 / i 1 ) 1 l ) J i i i j i -. i i I l i l Braidwood SSER 1 Appendix J l

I TECHNICAL EVALUATION REPORT CONTROL OF HEAVY LOADS AT NUCLEAR POWER PLANTS BYRON UNITS.1 AND 2 BRAIDWOOD UNITS 1 AND 2 (PHASE I Final) Docket tios. 50-454, 50-455, 50-456, and 50-457 Author S. A. Jensen Principal Technical Investigator T. H. Stickley Revised April 1984 EG&G Idaho, Inc. Idaho Falls, Idaho 83415 Prepared for the U.S. Nuclear Regulatory Commission Under DOE Contract No. DE-AC07-761001570 Fin. No. A6457 Braidwood SSER 1 Appendix J

ABSTRACT The Nuclear Regulatory Commission (NRC) has requested that all nuclear plants either operating or under construction submit a response of compliancy with NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants." EG&G Idaho, Inc., has contracted with the NRC to evaluate the responses of those plants presently under construction. This report contains EG&G's evaluation and recommendations for Byron /Braidwood for the requirements of Section 5.1.1 of NUREG-0612. Braidwood SSER 1 ii Appendix J

EXECUTIVE

SUMMARY

Byron /Braidwood is consistent with the guidelines of NUREG-0612. Braidwood SSER 1 111 Appendix J

CONTENTS 11 ABSTRACT.. EXECUTIVE

SUMMARY

iii 1. INTRODUCTION 1 1.1 Purpose of Review.......... I 1.2 Generic Background.................................... 1 1.3 Plant-Specific Background........................ 3 2. EVALUATION AND RECOMMENDATIONS........ 4 2.1 Overview...... 4 2.2 Heavy Lead Overhead Handling Systems........... 4 2.3 General Guidelines..................................... 7 3. CONCLUDING

SUMMARY

19 3.1 Applicable Le 4-Harr. ; <us'.e.ns.. 19 19 3.2 Guideline Reco.c .2to:ns 4. REFERENCES............................................. 25 TABLES 2.1 Crane / Hoist Systems Considered as Potential Sources for Damage of Safety Comoonents................ 6 3.1 NUREG Compliance Matrix....... 20 Braidwood SSER 1 iv Appendix J

CONTROL OF HEAVY LOADS AT NUCLEAR POWER PLANTS BYRON /BRAIDWOOD (PHASE I) 1. INTRODUCTION 1.1 Purpose of Review This technical evaluation report documents the EG&G Idaho, Inc., review of general load-handling policy and procedures at Byron /Braidwood. This evaluation was performed with the objective of assessing conformance to the general load-handling guidelines of NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants" [1], Section 5.1.1. This constitutes Phase I of a two phase evaluation. 1.2 Generic Background Generic Technical Activity Task A-36 was established by the U.S. Nuclear Regulatory Commission (NRC) staff to systematically examine staff licensing criteria'and the adequacy of measures in effect at operating nuclear power plants to assure the safe handling of heavy loads and to recommend necessary changes to izase measures. This activity was initiated by a letter issued by the NRC staff on May 17, 1978 [2], to all power reactor applicants, requesting information concerning the control of heavy loads near spent fuel. The results of Task A-36 were reported in NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants." The staff's conclusion from this evaluation was that existing measures to control the handling of heavy loads at operating plants, although providing protection from certain potential problems, do not adequately cover the major causes of load-handling accidents and should be upgraded. Braidwood SSER 1 1 Appendix J

In order to upgrade measures for the control of heavy loads, the staff developed a series of guidelines designed to achieve a two phase objective using an accepted approach or protection philosophy,. The first portion of the objective, achieved through a set of general guidelines identified in NUREG-0612, Article 5.1.1, is to ensure that all load-handling systems at nuclear power plants are designed and operated such that their probability of failure is uniformly small and appropriate for the critical tasks in which they are employed. The second portion of the staff's objective, achieved through guidelines identified in NUREG-0612, Articles 5.1.2 through 5.1.5, is to ensure that, for load-handling systems in areas where their failure might result in significant consequences, either (a) features are provided, in addition to those required for all load-handling systems, to ensure that the potential for a load drop is extremely small (e.g., a single-failure proof crane) or (b) conservative evaluations of load-handling accidents indicate that the potential consequences of any load drop are acceptably small. Acceptability of accident consequences is qu:.ified in NUREG-0612 into four accident analysis evalur en crit - e. cos.ch used to develop the staff guidelines for minimizing the i 41 for a load drop was based on defense in depth and is .e r. summar uea as follows: Provide sufficient operator training, handling system o design, load-handling instructions, and equipment inspection to assure reliable operation of the handling system o Define safe load travel paths through procedures and operator training so that, to the extent practical, heavy loads are not carried over or near irradiated fuel or safe shutdown equipment 1 Braidwood SSER 1 2 Appendix J

o Provide mechanical stops or electrical interlocks to prevent movement of heavy loads over irradiated fuel or in proximity to equipment associated with redundant. shutdown paths. Staff guidelines resulting from'the foregoing are tabulated in Section 5 of NUREG-0612. 1.3 plant-Specific Background On December 22, 1980, the NRC issued a letter [3] to Commonwealth Edison, the applicant for Byron /Braidwood requesting that the applicant review provisions for handling and control of heavy loads at Byron /Braidwood, evaluate these provisions with respect to the guidelines of NUREG-0612, and provide certain additional information to be used for an independent determination of conformance to these guidelines. On April 7, 1982, Commonwealth Edison provided the initial response [4] to this request. On October 25, 1982, Commonwealth Edison provided additional information in response [4] to a preliminary draft of this report. Further information was provid:c in submi't:13 [10], [11] dated February 10, 1984, and April 11, 1984. 4 Braidwood SSER 1 3 Appendix J

2. EVALUATION AND RECOMMENDATIONS 2.1 Overview The following sections summarize Commonwealth Edison's review of heavy load handling at Byron /Braidwood accompanied by EG&G's evaluation, conclusions, and recommendations to the applicant for bringing the facilities more completely into compliance with the intent of NUREG-0612. Commonwealth Edison's review of the facilities does not differentiate between the units. EG&G has evaluated the submittals as though all units are of identical design. The applicant has indicated the weight of a heavy load for this facility (as defined in NUREG-0612, Article 1.2) as 2000 lbs. 2.2 Heavy Load Overhead Handling Systems This section reviews the applicant's list of overhead handling systems which are subject to the criteria of NUREG-0612 and a review of the justification for excluding overhead nandling systems from the above-mentioned list. 2.2.1 Scope " Report the results of your review of plant arrangements to identify all overhead handling systems fror which a load drop may result in damage to any system required fev 71 ant sh"tdown or decay heat removal (taking no credit for , interlocks, technical specications, operating procedures, or detailed structural ar. sis) and justify the exclusion of any overhead handling system from your list by verifying that there is su ficient physical separation from any load-impact point and any d safety-related component to permit a determination by inspection that no heavy load drop can n uit in damage to any system or component required for plant shutdown or decay heat removal." Braidwood SSER 1 4 Appendix J

~ A. Summary of Applicant's Statements The applicant's review of overhead handling systems identified the cranes and hoists shown in Table 2.1 as those which handle heavy loads in the vicinity of irradiated fuel or safe shutdown equipment. The applicant has also identified other cranes that have been excluded from satisifying the criteria of the general guidelines of NUREG-0612. B. EG&G Evaluation The applicant appears to have included all applicable handling systems in their tables showing handling for which a load drop could damage equipment. C. EG&G Conclusions and Recommendations Based on the information provided, EG&G concludes that the applicant has included all applicable hoists and cranes in their list of handling systems which must comply with the requirements of the general guidelines of NUREG-0612. I i t l' L 1 3 Braidwood SSER 1 5 Appendix J

. = - I CRANE / HOIST SY'TEMS CONSIDERED AS POTENTIAL SOURCES FOR DAMAGE S TABLE 2.1 OF SAFETY COMPONENTS. l System Designations 4, i Polar Crane Cable Tray Drawbridge Winch 4 i ' Stud Tensioner Hoists (3) i Fuel Building Crane Spent-Fuel Pit Bridge Crane Trolley Beam 24 i Trolley Beam 25 Trolley Beam 53 j Trolley Beam 54 Trolley Beam 23 Trolley Beam 42 (Braidwood only) Turbine Building Cranes PTS-2 l PTS-3 PTS-4 PTS-5 PTS-8 (Braidwood only) i PTS-9 (Braidwood only) SG-1 P SG-2 SG-3 l SG-4 i l l l-f i Braidwood SSER 1 6 Appendix J

2.3 General Guidelines This section addresses the extent to which the applicable han,dling systems comply with the general guidelines of NUREG-0612 Article 5.1.1. EG&G's conclusions and recommendations are provided in summaries for each guideline. The NRC has established ~seven general guidelines which must be met in order to provide the defense-in-depth approach for the handling of heavy loads. These guidelines consist of the following criteria from l Section 5.1.1 of NUREG-0612: o Guideline 1--Safe Load Paths o Guiduline 2--Load-Handling Procedures o Guideline 3--Crane Operator Training o Guideline 4--Special Lifting Devices o Guideline 5--Lifting Devices (not specially designed) o Guideline 6--Cranes (Inspection, Testing, and Maintenance) o Guideline 7--Crane Design. These seven guidelines should be satisfied for all overhead handling systems and programs in order to handle heavy loads in the vicinity of the reactor vessel, near spent fuel in the spent-fuel pool, or in other areas where a load drop may damage. safe shutdown systems. The succeeding paragraphs address the guidelines individually. Braidwood SSER 1 7 Appendix J

2.3.1 Safe Load Paths [ Guideline 1, NUREG-0612, Article 5.1.1(1)] " Safe load paths should be defined for the movement of heavy loads to minimize the potential for heavy loads, if dropped, to impact irradiated fuel in the reactor vessel and in the spent-fuel pool, or to impact safe shutdown equipment. The path should follow, to the extent practical, structural floor members, beams, etc., such that if the load is dropped, the structure is .more likely to withstand the impact. These load paths should be defined in procedures, shown on equipment layout drawings, and clearly marked on the floor in the area where the load is to be handled. Deviations from defined load paths should require written alternative procedures approved by the plant safety review committee." A. Summary of~ Applicant's Statements The applicant has evaluated load path locations for ' Byron /Braidwood. The applicant states that load movement follows the safest and shortest route with the load as close to the floor as possible. Byron /Braidwood will incorporate into Maintenance / Equipment Removal Procedures references to the applicable M-517 and M-27 prints to identify safe load paths. To the extent necessary, these procedures will be available prior to fuel load. Procedures for heavy lead movement inside Containment will incorporate .111ty Control or Quality Assurance hold points as necessary and provide independent erification of proper load paths. Crane operators at Byron /Braidwood Stations will move heavy loads under the direction of a maintenance foreman or mechanic. The Byron /Braidwood Stations are presently writing an administrative procedure to describe the job responsibility of the person directing the heavy load movement. Furthermore, the existing maintenance and fuel handling procedures are being revised to reflect the administrative procedure. Braidwood SSER 1 8 Appendix J

B. EG&G Evaluation The applicant response and drawings submitted indicates that the intent of Guideline 1 criteria have been satisfied at Byron /Braidwood. Load paths have been develcoed for all heavy loads which have been identified. The applicant's position on the unfeasibility of marking load paths on the floor is acceptable since a supervisor is present to ensure that the best load path is followed. C. EG&G Conclusions and Recommendations EG&G concludes from the applicant's response that the Byron /Braidwood Stations are consistent with the intent of Guideline 1. 2.3.2 Load-Handling' Procedures [ Guideline 2, NUREG-0612, Article 5.1.1(2)]' " Procedures should be developed to cover lead-handling operations for heavy loads that are or could be handled over or in proximity to irradiated fuel or safe shutdown equipment. At a minimum, procedures should cover handling of those loads listed in Tacle 3-1 of NUREG-0612. These procecures snould include: identification of required equipment; inspections and acceptance criteria requirec before movement of load; tne steos and precer secuence to be followed in handling tne loaa; defining the safe oath; and other soecial precautions." A. Summary of Aeplicant's Statements The applicant states that procedures will be developed to cover load-handling operations for the heavy loads identified in Table 3.1-1 of NUREG-0612. These procedures will identify the required equipment, the inspection and acceptance criteria prior to load movement, the steps and Braidwood SSER 1 9 Appendix J

i i l sequence in handling the lead, and define the safe load pa:n 'and'other special precautions. They also state that to the extent necessary approved procecures will be in effect prior j to fuel loading. B. EG&G Evaluation l The applicant has stated that load-handling procedures will be developed which will comply with the requirements of Guideline 2. These guidelines should be available for possible review by the NRC prior to' fuel loading. i I C.- EG&G Conclusions and Recommetdations i l The Byron /Braidwood Stations are consistent with Guideline 2. l 1 2 '. 3. 3 Crane Ocerator Trainino (Guideline 3, NURE3-0612, Article 5.1.1(3)] 1 " Crane operators should be trained, qualified, and conduct 1 themselves in accordance with Chapter 2-3 of ANSI B30.2-1976, ' Overhead and Gantry. Cranes' [5]." 4 I A. Summary of Applicant's Statements l The applicant states that Byron /Braidwood will comply with ANSI 830.2-1976 with respect to operator training, qualification, and conduct. Training records will be i available for inspection and review prior to fuel load. i i B. EG1G Evaluation Byron /Braidwood Stations are consistent with the intent of Guideline 3. I i i i i 1 l Braidwood SSER 1 10 Appendix J -.----.--.-_m.-----_..,_-,__.-..---._.._,--_-,--_-_.__ - ~

C. EG&G Conclusion and Recommendations Based on the applicant's statement, Byron /Braidwood,is consistent with Guideline 3. 2.3.4 Soecial Lifting Devices [ Guideline 4, NUREG-0612, Article 5.1.1(4)] "Special lif ting devices should satisfy the guidelines of ANSI N14.6-1978, ' Standard for Special Lif ting Devices for Shipping Containers Weighing 10,000 Pounds (4500 kg) or More for Nuclear Materials' [6]. This standard should apply to all special lifting devices which carry heavy loads in areas as defined above. For operating plants, certain inspections and load tests may be accepted in lieu of certain material requirements in the s ta nda rd. In addition, the stress design factor stated in Section 3.2.1.1 of ANSI N14.6 should be based on the combined maximum static and dynamic loads that could~be imparted on the handling device based on characteristics of the crane which will be used. This is in lieu of the guideline in Section 3.2.1.1 of ANSI N14.6 which bases the stress design factor on only the weight (static load) or the load and of the intervening components of the special handling device." A. Summary of Applicant's Statements The applicant has stated that (a) the lifting devices have been or will be designed in accordance with industrial standards using good engineering practices; (b) special lifting devices for the reactor vessel head ard uoper internals have been provided by Westinghouse; and (c) Westinghouse used standard quality control procecures in the fabrication of the l'ifting devices. Both lifting rigs have been designed for 200*4 of the dead load using AISC allowables and load tested to 125% of their rated load. In regard to acceptance testing, and maintenance the applicant has made the following statements: "The Byron and Braidwood Station procedures will comply with the intent of Section 5, Acceptance esting, Maintenance, Braidwood SSER 1 11 Appendix J

and Assurance of Continued Compliance with some exceptions. In Commonwealth Edison's judgment, the periodic load testing of the special lifting devices to 150% of the maximum load is not practical nor warranted, and may invalidste any vendor product guarantees. As stated in our April 1982 Heavy Load Movement Report, the special lifting devices were load tested to 125%, which is in accordance with the proof-load test indicated on the vendor drawings. Additionally, the logistics of moving heavy test loads into the Reactor Containment Building to accommodate such periodic load testing are dif ficult. " Prior to use of specially designed lifting assemblies, visual inspection will be performed and certain critical and accessible parts or members such as hooks and pins will be non-cestructively examined at appropriate time intervals. In our judgment, the visual inspection and limited NDE are acequate to detect potential failures. "However, should an incident occur in which a special lif ting device is overloaded, damaged or distorted, an engineering assessment will be performed. This assessment will address ANSI N14.6 and include consideration of the loac test up to tne original procurement load test value or 150% wnichever is less. The recuirement to perform this assesstent will be incorcorated into olant procedures." B. EG&G Evaluation Information provided by the applicant indicates that stress design factors are consistent with the intent of ANSI N14.6-1978 for the two special lifting devices identified. The lifting devices mentioned nave been load tested to weights substantially in excess of the maximum load currently lifted and, therefore, meet the intent of ANSI N14.6-1978 guidelines for acceptance load testirg. Braidwood SSER 1 12 Appendix J

The applicant also states that current procecures meet the intent of Section 5 with some exceptions. The applicant takes exception to periodic performance of load testing. -It is noted that ANSI N14.6-1978 provides acceptable alternatives to periodic load-tests if an initial acceptance load test has been satisfactorily performed; the owner may opt to perform an annual (or prior-to-use, depending on frequence of use) series of inspections in accordance with Section 5.3.1(2) of the ANSI standard. This testing shall include " dimensional testing, visual inspection, and nondestructive testing of major load carrying welds and critical areas." In this regard, the applicant has proposed performance of visual inspections and limited NDE prior to each use of the lifting devices and is of the opinion that such an inspection program is ade.uate to detect potential failure. Based upon the applicant's statement that the current inspection program is adequate, the degree of load-handling reliability necessary to satisfy periodic inspection requirements of Section 5 has been satisfied. It is recommended that the applicant review requirements for dimensional and nondestructive testing of these lifting devices and modify their inspection program accordingly. For the remaining exception, the commitment to assess potential damage to the lifting device and determine the need for an overload test to a' weight substantially in excess of the rated capacity is consistent with the intent of the ANSI standard. Since the lifting devices were designed and fabricated by Westinghouse, and were fabricated using Westinghouse's quality control procedures EG&G feels that quality control practices were consi:stnt with the intent of ANSI N14.6. Braidwood SSER 1 13 Appendix J

Based on the above discussion the two special lifting devices mentioned by the applicant meet the intent of ANSI N14.6. C. EG&G Conclusions and Recommendations Byron /Braidwood Stations are consistent with the intent of Guideline 4. 2.3.5 Lif ting Devices (Not Specially Designed) [ Guideline 5, NUREG-0612, Article 5.1.1(5)] " Lifting devices that -are not specially designed should be installed and used in accordance with the guidelines of ANSI B30.9-1971, ' Slings' [7]. However, in selecting the proper sling, the load used should be the sum of the static and maximum dynamic load. The rating identified on the sling should be in terms of the ' static load' which produces the maximum static and dynamic load. Where this restricts slings to use on only certain cranes, the slings should be clearly marked as to the cranes with which they may be used." A. Summary of Acolicant's Statements The applicant states that all lifting devices were designed according to industrial standards using good engineering practices. Byron /Braidwood procures ano insoects slings to ANSI B30.9-1971. Ir:cections of slings are conducted annually and slings are examined visually prior to use. Slings are installed and used in accordance with ANSI B30.9-1971. Sling selection is based on the sum of the static and maximum dynamic loads. Slings are not restricted to special cranes. Braidwood SSER 1 14 Appendix J

The Turbine Building Crane 25 ton Auxiliary Hoist is the only hoist capable of operating faster than 30 fpm. The maximum operating speed is 33.8 fpm. The only safety related component located in the Turbine Bldg. is th'e SX piping. The SM piping is' embedded a minimum of 6 ft under the surface of the basemat and is located three floors below the main turbine floor. Any auxiliary hoist load drop will not effect the SX piping. Additionally, the SX piping is redundant and is separated by a distance of 49 ft-6 in. 'The 33.8 fpm does not result in any significant addition of dynamic load and therefore any modification of the hoist speed is not warrsnted. B. EG&G Evaluation The a'plicant is consistent with this guideline on the basis p of the above statements. The dynamic loads generated by cranes and hoists at Byron /Braidwood are reasonably small percentages of the overall static load and, therefore, may be disregarded when selecting slings. C. EG&G Conclusions and Recommendations Byron /Braidwood Stations are consistent with the Guideline 5 of NUREG-0612, based on the previous evaluation. 2.3.6 Cranes (Inspection, Testing, and Maintenance) [ Guideline 6, NUREG-0612. Article 5.1.1(6)) "The crane should be inspected, tested, and maintained in accordance with Chapter 2-2 of ANSI B30.2-1976, ' Overhead and Gantry Cranes,' with the exception that tests and inspections Braidwood SSER 1 15 Appendix J

.~ E should be performed prior to use where it is not practical to j meet the frequencies of ANSI B30.2 for periodic inspection and i test, or where frequency of crane use is less than the specified inspection and test frequency (e.g., the polar crane inside a PWR containment may only be used every 12 to 18 months durin) refueling operations, and is generally not accessible during power operation. ANSI B30.2, however, calls for certain l inspections to be performed daily or monthly. For such cranes having. limited usage, the inspections, test, and maintenance should be performed prior to their use)." A. Summary of Applicant's Statements 4 Cranes will be inspected, tested, and maintained in accordance with Chapter 2-2 of ANSI B30.2-1975, with the i I' exception that tests and inspections _should be performed prior to use where it is not practical to meet the frequencies of ANSI B30.2. For cranes having limited usage, I the inspections and tests will be performed prior to their use. Approved procedures will be in effect prior to fuel load. B EG&G Evaluat[on i The applicanc states that crane inspection, testing, and maintenance programs will be in accordance with ANSI B30.2-1976, with exceptions as allowed by GJideline 6. C. EG&G Conclusions and Recommendations Byron /Braidwood Station's are consistent with Guideline 6 on I the basis of the applicant's statement. 2.3.7 Crane Design [ Guideline 7, NUREG-0612, Article 5.1.1(7)1 "The crane should be designed to meet the aoplicable criteria and guidelines of Chapter 2-1 of ANSI B30.2-1976, ' Overhead and I i i Braidwood SSER 1 16 Appendix J

Gantry Cranes,' and of CMAA-70, Specifications for Electric Overhead Traveling Cranes' [8].. An alternative to a specification in ~ ANSI B30.2 or CMAA-70 may be accepted in lieu of specific compliance if the intent of the specification is satisfied." A. Summary of Aoplicant's Statements The polar cranes, turbine building cranes, and fuel-handling building cranes were designed in accordance with the 1975 Revision of CMAA-70 and ANSI B30.2-1976. Welding was performed in accordance with AWS D.1.1. The turbine building cranes were also designed and fabricated in accordance with CMAA-70, 1975 Revision and ANSI B30.2 1976. The polar crane is provided with limit switches for bridge overtravel, plus two upper and one lower limit switch for each hoist. Mechanical end stops are also provided on the bridge. The Fuel-Handling Building Crane is provided with end stops on the runways and bridge, plus upper and lower limit switches on both hoists. The trolley beams were designed and fabricated in accordance with AISC-1978 standards. The PTS and single girder systems (SG) were designed in accordance witn MMA and AISC-1978 standards. The jib cranes were designed and fabricated according to AISC-1978 standards. B. EG&G Evaluation The cranes mentioned by the applicant in their response are consistent with the intent of Guideline 7 based on the applicant's statements. Braidwood SSER 1 17 Appendix J

C. EG&G Conclusions and Recommendations Byron /Braidwood Stations are consistent with the intent of Guideline 7 on the basis of the applicant's statements. Braidwood SSER 1 18 Appendix J

3. CONCLUDING

SUMMARY

3.1 Applicable Load-Handling Systems The list of cranes and hoists supplied by the applicant as.be.ing subject to the provisions of NUREG-0612 is adequate (see Section 2.2.1). 3.2 Guideline Recommendations Compliance with the seven NRC guidelines for heavy load handling (Section 2.3) are partially satisfied at Byron /Braidwood. This conclusion is represented in tabular form as Table 3.1. Specific recommendations to aid in compliance with the intent of these guidelines are provided as follows: Guideline Recommendation 1. (Section 2.3.1) a. Define the duties of supervisors with respect to heavy load handling. 2. (Section 2.3.2) a. Byron /Braidwood are consistent with this guideline. 3. (Section 2.3.3) a. Byron /Braidwood are consistent with this guideline. 4. (Section 2.3.4) a. Byron /Braidwood are consistent with the intent of this guideline. 5. (Section 2.3.5) a. Byron /Braidwood are consistent with the intent of this guideline. 6. (Section 2.3.6) a. Byron /Braidwood are consistent with this guideline 7. (Section 2.3.7) a. Byron /Braidwood are consistent with this guideline l l Braidwood SSER 1 19 Appendix J

. _ _ _ _.. _, _ _ -.. _ ~ _ _ _ _. -. I i I i 03 as IAntf 3.1, BYROII/BRAIDWOOD IeURf C Cre4Pt.l AleCf MAIRIX o I Guidelines a E 1 2 3 3 5 6 I rei 2 Weight or C rane Special Crane = Capacity Safe load Ope ra tor I.ifting Test and Crane w j [sittigenent Desionagfo!) _ Ileavy_ load s _ Llons L Paths Procedures Trainir!g Devige3 11Jng1 1nlesqtion Design j Polar Crane Heactor Vessel Head, 230/40 C .C C C C C C 411.150 lb I Reactor upper i Internals, i 145,000 lb ] Reactor Lower Inte rna l s, j 269,600 Ib Reactor Enolant Pump flotors, i l 77,500 lb O Reactor Core j Ba rrel Assembly, 217,300 lb i I i plain Hook tower l_oad f Block, 6,183 lb j j Auxilia ry Itook t ower 1 toad Block, 1.T70 lb Cable Trey Cable Irny 10 C C C N/A C C C j prawbridge niinch Drawbridge, 9,tMWs Ib i j Stud Tensioner Reactor Vessel Hetad 2 C C C N/A C C C Holsts (3) Stud TensieHeer, N/A Reactor Vossel Skrad Studs, 806 lb fuel Basilding Crane Spent-Fuel cash, 125 C C C N/A C C C 218,000 lb (IN-12) E Iuel Asseehty, 4 Q 1,467 lb 4 a (Main Holst lower X Load Block, c 5,600 lb ft 4

l 3 L J 4 txt

l I AlltI 3.1.

(continued l l Guidelines c2. I m 1 2 3 3 6 I I m I Q Wight or Crane Special Crane l Capacity Safe t saad Ope ra to r 8.ifting Test and Crane H [quipment Desegnatiore __ Ileavy loads LT ogt}__ Path 1_ Procedures Training De_yice_3 Ellegg 1stssteqt ioet Desiert 4 Aasailiary lloist j tower load illsack, I (3,500 lb est.) j ) I f ailed-f asel l 1 Cannister, 940 lbs { 1 Control Rod l Cluster, 158 lbs Spent fuel Pit IIew Fuel Assembly, 2 C C C C C C C 1 Bridge Crano 1,467 Ib 1 spent-fasel Asseehty, 1,467 lb l y fuel-Handling tools, 375 lb ma x imase failed-finet Cannister, 940 lb Control Rod j Cluster, 158 lb Irelley Seam 24 RieR Heat Exchangers 12 C C C N/A C C C Tube Bundle, 14,500 lb j Concrete Plasgs, 15,000 lb } Irelsey Bessa 2S RitR teeat Excleange rs 12 C C C M/A C C C l fube Sundle, r 14,500 Ils V ] concrete Plesqs, l i = ik,Ooo ib C2. I E 1rolley Been 53 Cha rg i ng Pimp, 8 C C C N/A C C C F,500 lb f I ? l

c3 -s 1 o.n I Allt f 3.1. (continued) f CieideIinos $a 1 2 J 9 3 6 I m C ra ne Special Crane W Weight or Q Capacity Sat'e load Ope ra tor 1.irting Tost and Crane W (qu_ipmen LDes_i3nat gn awavy_tnads _ Llonsj _ _Pa g!Ls___ P.rocedures T ra in i ng Deyices S tjpg s jnspectiori pe_.11RD m i (.lsa v y rHI Piep

Motor,

4. 3% lb 8

C C C N/A C C C 1rolley Beam 54 Cha rg ing Piump, 7,500 lb Charging Pump -sed Moto r, 4,37*> Ib SG-1 Diesel Cenorator 2 C C C N/A C C C Cylinder Itcad Covers, 830 lb SG-2 Diesel Gene ra to r 2 C C C N/A C C C y N Cylinder llead Covers, 83ts th 1 2 C C C N/A C C C i SC-3 Dieset Cencrator Cylinder Itcad Covers, 81(I Ib 2 C C C N/A C C C Diesel Generator SG-ti Cy l i nde r lica d Covers, 831: Ib PIS-? and Concrete PItegs, 6 C C C N/A C C C 18,100 I t. PIS-3 (Unit 2) Con ta i simen t Spray Pump / Motor, /,3rtl Ib Cha rging Ptop, 7,500 lb p V V Safety injsection Pump, 5, 7 641 th sa Clea rg e og hop { Mo t o r, ti, 31 *> Ih 4 C.,

) a,s l E. Tanaf i.1. (continued) . Q i 8 Cuidelines CL 1 2 3 5 6 I m w m Weight nr Crane. Special Crane Capacity Safe Load Ope ra to r L 4 f t irig . Test and Crsne [9'dPeer11_Deliena don _tleayy,! oads _ _tTongi Pa dis Procedures Trainine Devices Elipes Irispec t ion Desien Nesety t ri jec t inn Pump stotor, ,345 sb i PIS-4 and Safety injection 6 C C C 88/A C C C l a Pump, 5,760 Ib i i PTS-5 (Unit 23 Safety injection j Pump 100 tor, 3,100 lb 1 Irolley Beam 23 Cha rg ing Pump 10 C C C N/A C C C l 7,500 lb 4 i l Containment spray l Pump /ptotor, 7,307 lb g w i RHft Pump /900 tor, j 6,200 lb i Safety leijection Pump, 5,26n Ib l l' Safety Iejection footor, 4,345 lb i l ESW Pump, 9,500 lb ) ESW footor, 12,000 lb ) i Turbine Building Turbine cosaponents 125/25 C C C N/A C C C 4 Cranes Unit 1 (* Unit 2) LP Spindles, 294,000 Ib i HP Spindles, 148/25' t j 131,000 lb d HP Cylinder Cover, i j 166,000 8b cL t P cy I Irwse r Cove r, c l 7 172,300 lb a L Other Lighter loads j i l,

._--_.m_.. m _ _ _ _ _ i A i -s 'I as j g IABLE 3.1. (continued) o i CL Guidelines 1 m m .1 2 3 h 1 6 I i m l Weight or Crane Special Crane 1 e-* Capacity Safe load Ope ra to r Lifting Test and Crane j [qttigeent Desienation _ IIe!tyyjoads _fTons) Paths Procedures Trainine Devices Stines Inspection Dqpsien FIS-8 and PIS-9 Circulating Water 30 C C C II/A C C C Pump Mutor, T5,fWN3 Ih Trolley Beam 42 WS Pump, all,300 lb 12 C C C. M/A C C C. l WS 9totor 22,500 Ib k ) C = Applicant action complies with IIUREG-0612 Guideline. i IIC = Applicant action does not comply with 800 REC-0612 Cuideline. I R = Applicant has proposed revisions / modifications designed to comply with leUREG-0612 Cuideline. I = Insufficient Inforestion provided by the Applicant. I 1 I 4 I \\ l 1 l } l i 4 l \\ = V 66 j 3 a 1 u 1 ? l f

4. REFERENCES 1. NUREG-0612, Control of Heavy Loads at Nuclear Power Plants, NRC. 2. V. Stello, Jr. (NRC), Letter to all applicants. Suoject: Request for Additional Information on Control of Haavy Loads Near Spent Fuel, NRC, 17 May 1978. 3. USNRC, Letter to Commonwealth Edison.

Subject:

NRC Request for Additional Information on Control of Heavy Loads Near Spent Fuel, NRC, I 22 December 1980. 4. Commonwealth Edison, Letter to Director of Nuclear Reactor Regulation.

Subject:

Byron Station Units 1 and 2, Braidwood Station Units 1 and 2, 7 April 1982. 5. ANSI B30.2-1976, " Overhead and Gantry Cranes". I 6. ANSI N14.6-1978, " Standard for Lifting Devices for Shipping Containers Weighing 10,000 Pounds (4500 kg) or more for Nuclear Materials". 7. ANSI B30.9-1971, " Slings". 8. CMAA-70, " Specifications for Electric Overhead Traveling Cranes". 9. Commonwealth Edison, Letter to Director of Nuclear Reactor Regulation.

Subject:

Byron Station Units 1 and 2, Braidwood Station Units 1 and 2, Control of Heavy Loads, October 25, 1982. l

10. Commonwealth Edison, Letter to Director of Nuclear Reactor Regulation.

Subject:

Byron Station Units 1 and 2, Braidwood Station Units 1 and 2, Supplemental Information to the Resolution of " Control of Heavy Loads a Nuclear Power Plants," February 10, 1984.

11. Commonwealth Edison, Letter to Director of Nuclear Reactor Regulation.

Subject:

Byron Station Units 1 and 2, Braidwood Station Units 1 and 2, Supplemantal Information to the Resolution of " Control of Heavy Loads a Nuclear Power Plants," April 11, 1984. 4 23099 i Braidwood SSER 1 25 Appendix J

NAC PORu 33 u 5 NUC LE AR Rt GuL ATOR v COvviS5 eon REPOR Y NuvetR tass pa,e ey rioC saa von No .# ea fr NUREG-1002 BIBLIOGRAPHIC DATA SHEET Supplement No.1 2 t ~ -. e-. 3 TITLE AND 5087 87 L 4 4 REciest N T 5 ACCT 55 TON NUM8t H Safety Evaluation Report relate to the operation of , o,,,,,,on,cm ,,,,,o Braidwood Station, Units 1 and SEPTEMBER I,,AR 1986 .0N,. e AUTHOMi5, 7 D AT E REPOR T ISSUED MONin vtAR SEPTEM 1986 9 PROJECT K, WORE UNIT NuM8t R 6 PERFOHMiNG ORG ANII ATiON N Avg AND M AeLtNG ADDHf 55 fiarinde IW~~*r Division of Pressurized Water Reactor icensing-A Office of Nuclear Reactor Regulation iof ~ ~uMata U. S. Nuclear Regulatory Commission / Washington, D. C. 20555 ft $PON50 RANG OMGANil A TION NAVt AND M A'LINU ADDHESS ifacheJe le Codet 12s T YPE OF REPORT Same as 8. above Technical 120 PE H OO COV E RE D isac#ws..e deresJ November 1983 - September 1986 13 5VPPLlut N T AR Y NOTE $ Docket Nos. 50-456 and 50-457 14 A8$7M ACT #100 esords or dessf The Safety Evaluation Report issued in N mber 1 33 provided the results of the NRC staff's review of the Commonwealth Edis. Company' application for licenses-to operate the Braidwood Station, Units 1 and 2. he facilit consists of two pressurized water nuclear reactors located in northeast n Illinois w hin Reed Township, Will County, Illinois. Supplement No. 1 updates the info ' tion contained in he Safety Evaluation and addresses the ACRS Report issued F bruary 11, 1985. \\ 15e nt V WOMD5 ANO DOCuwt NT AN AL Y5is 150 Di $CR IP T OR $ \\ 16 AV AIL ARettiv ST A f t ut NT IF SECURtT V C L AS$1F ICA TION 16 NUMBER OF PAGES UNCLASSIFIED UNLIMITED ,,5ttuR,1,CtAss.,,CA,,0N ,o ,R,C, UNCLASSIFIED s

UNITED STATES - l nast ctans um ~ NUCLEAR REGULATORY COMMISSION . restaaj,= ges caso WASHINGTON, D.C. 20555 wasw. o.c. ' OFFICIAL BUSINESS. PENALTY FOR. PRIVATE USE 8300 120555078877 1 1.N US NRC AC"-]IV 0F TTDC = POLICY s Pg8 vGT HP-PCP f'RtG v n-sci WASHINGTON DC

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