ML20214W316

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

ML20214W316
Person / Time
Site:	Braidwood
Issue date:	05/31/1987
From:	Office of Nuclear Reactor Regulation
To:
References
NUREG-1002, NUREG-1002-S03, NUREG-1002-S3, NUDOCS 8706150208
Download: ML20214W316 (35)

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's ,e NOTICE Availability of Reference Materials Cited in NRC Publications Most documents cited in N RC 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 GPO 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.

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NUREG-1002 Supplement No. 3 Safety Evakation Report related to the operation of Braidwood Station, Units 1 and 2 Docket Nos. 50-456 and 50-457 Commonwealth Edison Company i

U.S. Nuclear Regulatory Commission ,

Office of Nuclear Reactor Regulation May 1987 l ,,s~.,

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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. 50-456 and 50-457). The first supplement to NUREG-1002 was issued in September 1986; the second supplement to NUREG-1002 was issued in October 1986. This third supplement to NUREG-1002 reports the status of certain items that remained unresolved at the time e Supplement 2 was published. The facility is located in Reed Township, Will County, Illinois.

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TABLE OF CONTENTS Page ABSTRACT ............................................................... iii 1 INTRODUCTION AND GENERAL DESCRIPTION OF FACILITY .................. 1-1 1.1 Introduction ................................................. 1-1 1.7 Summary of Outstanding Items ................................. 1-1 1.8 Confirmatory Issues .......................................... 1-3 1.9 License Conditions ........................................... 1-4 3 DESIGN CRITERIA FOR STRUCTURES, SYSTEMS, AND COMPONENTS ........... 3-1 3.6 Protection Against Effects Associated with the Postulated Rupture of Piping ...................... .......... 3-1 3.6.1 Plant Design for Protection Against Postulated Piping Failures in Fluid Systems Outside Containment ........................................... 3-1 5 R EACTO R COO LANT SY ST EM . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 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 5.2.4.3 Evaluation of Compliance with 10 CFR 50.55a(g) for Braidwood Unit 1 ........ 5-1 6 ENGINEERED SAFETY FEATURES ........................................ 6-1 6.4 Control Room Habitability .................................... 6-1 6.5 Fission Product Removal and Control System ................... 6-2 6.5.1 Engineered Safety Feature Atmospheric Cleanup System .. 6-2 6.6 Inservice Inspection of Class 2 and 3 Components ............. 6-3 6.6.3 Evaluation of Compliance with 10 CFR 50.55a(g) for Braidwood Units 1 ................................. 6-3 7 INSTRUMENTATION AND CONTROL ....................................... 7-1 7.6 Interlock Systems Important to Safety . . . . . . . . . . . . . . . . . . . . . . . . 7-1 7.6.2 Specific Findings ..................................... 7-1 Braidwood SSER 3 v

TABLE OF CONTENTS (Continued) a le Pag 9 AUXILIARY SYSTEMS ................................................. 9-1 9.5 Other Auxiliary Systems ...................................... 9-1

, 9.5.1 Fire Protection Program ............................... 9-1

9.5.1.2 Fire Protection Program Requirements ......... 9-1 9.5.1.4 General Plant Guidelines ..................... 9-1 9.5.1.5 Fire Protection for Plant-Specific Areas ..... 9-3 9.5.4 Emergency Diesel Engine Fuel Oil Storage and Transfer System ....................................... 9-3 9.5.4.1 Emergency Diesel Engine Auxiliary Support Systems (General) ............................ 9-3 11 RADI0 ACTIVE WASTE MANAGEMENT ...................................... 11-1 11.5 Process and Effluent Radiological Monitoring and Sampling Systems ............................................. 11-1 13 CONDUCT OF OPERATIONS ............................................. 13-1 13.3 Emergency Planning ........................................... 13-1 13.3.1 Atomic Safety and Licensing Board Decision on Emergency Planning Issues ............................ 13-1 APPENDICES 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 I ERRATA TO BRAIDWOOD STATION, UNITS 1 AND 2, SAFETY l

EVALUATION REPORT AND ITS SUPPLEMENTS l

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l Braidwood SSER 3 vi L-

l 1 INTRODUCTION AND GENERAL DESCRIPTION OF FACILITY 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.'50-456 and 50-457). At that time, the staff identified items that had not been resolved with the appli-cant. The first Supplement to NUREG-1002 was issued in September 1986; the second supplement to NUREG-1002 was issued in October 1986. The purpose of this third supplement to the SER is to provide the staff evaluation of the open items that have been resolved to date, to address changes to the SER which re-sulted from the receipt of additional information from the applicant, and to respond to the Atomic Safety and Licensing Board's requirements identified in its May 13, 1987 Partial Initial Decision on Emergency Planning Issues.

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 pro-4 cessing 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 I contains errata to the SER.

Copies of this SER supplement are available for inspection at the NRC Public Document Room, 1717 11 Street, N.W., Washington, D.C., and at the Wilmington Township Public Library, 201 South Kankakee Street, Wilmington, Illinois 60481.

The NRC Project Manager for Braidwood Station, Units 1 and 2, is Ms. Janice A.

Stevens. Ms. Stevens may be contacted by calling (301) 492-4993 or writing:

Janice A. Stevens Division of Reactor Projects III/IV/V and Special Projects U.S. Nuclear Regulatory Commission Washington, D.C. 20555 1.7 Summary of Outstanding Items The current status of the outstanding items listed in the SER follows:

Part A Items Status Section (1) Pump and valve operability Closed in 3.9.3.2**

Supplement 2

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

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

Braidwood SSER 3 1-1

Part A Items (Continued) Status Section (2) Seismic and dynamic qualification of Closed in 3.10*

equipment Supplement 2 (3) Environmental qualification of electrical Closed in 3.11*

and mechanical equipment Supplement 2 (4) Containment pressure boundary components Closed in 6.2.7 Supplement 1 (5) Organizational structure Closed in 13.1, 13.4 Supplement 1 (6) Emergency preparedness plans and facilities Closed in 13.3*

Supplement 1 (7) Procedures generation package (PGP) Closed in 13.5.2 3

Supplement 2 (8) Control room human factors review Partially 18.2*

closed in Supplement 2 (9) Safety parameter display system Opened in 18.3*

Supplement 2 (10) Control room habitability Closed in this 6.4 supplement Part B Items (1) Turbine missile evaluation Closed in 3.5.1.3 Supplement 1

(2) Improved thermal design procedures Closed in 4.4.1 Supplement 1 (3) TMI Action Item II.F.2

Inadequate Core Closed in 4.4.7 Cooling Instrumentation Supplement 1 (4) Steam generator flow-induced vibrations Closed in 5.4.2 Supplement 1 -

(5) Conformance of ESF filter system to RG 1.52 Closed in 6.5.1 Supplement 2 (6) Fire protection program Closed in this 9.5.1 supplement I

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

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Part B Items (Continued) Status Section (7) Volume reduction system Closed in 11.1, 11.4.2 Supplement 2 1.8 Confirmatory Issues The current status of the confirmatory issues follows:

Part A Items (1) Applicant compliance with the Commission's Closed in 1.1, 3.1*

regulations Supplement 2 (2) Site drainage Closed in 2.4.3.3 Supplement 1 (3) Piping vibration test program Closed in 3.9.2.1*

Supplement 1 (4) Preservice Inspection Program Closed in 5.2.4, 6.6*

Supplement 2 (5) Reactor vessel materials Closed in 5.3 Supplement 1 (6) Electrical distribution system voltage Closed in 8.2.4*

verification Supplement 1 (7) Independence of redundant electrical safety Closed in 8.4.4 equipment Supplement 1 (8) RPM qualifications Closed in 12.5 ~

Supplement 1 (9) Revision to Physical Security Plan Closed in 13.6 Supplement 1 Part B Items ,

I (1) Inservice testing of pumps and valves Partially 3.9.6 closed in  !

Supplement 2 (2) Steam generator tube surveillance Closed in 5.4.2.2 Supplement 1  ;

(3) Charging pump deadheading Closed in 6.3.2, 7.3.2 Supplement 1

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

Braidwood SSER 3 1-3 )

Part B Items (Continued) Status Section (4) Minimum containment pressure analysis for Closed in 6.2.1.5 performance capabilities of ECCS Supplement 1 (5) Containment sump screen Closed in 6.2.2 Supplement 1 (6) Containment leakage testing vent and drain Closed in 6.2.6 provisions Supplement 1 (7) _ Confirmatory test for sump design Closed in 6.3.4.1 Supplement 1 (8) IE Bulletin 80-06 Closed in 7.3.2.2 Supplement 1 (9) Remote shutdown capability Closed in 7.4.2.2 Supplement 2 (10) TMI Action Plan Item II.D.1 Partially 3.9.3.3, closed in 5.2.2 Supplement 1 TMI Action Plan Item II.K.3.1 Closed in 7.6.2.7 Supplement 1 TMI Action Plan Item III.D.1.1 Closed in 9.3.5 Supplement 1 (11) SWS process control program Closed in 11.4.1 Supplement 2 (12) Noble gas monitor Closed in 11.5.2 Supplement 2 (13) RCP rotor seizure and shaft break Closed in 15.3.6 Supplement 1 (14) Anticipated transients without scram (ATWS) Partially 15.6 closed in Supplement 2 (15) Evaluation of compliance with Closed in 5.2.4.4 10 CFR 50.55a(a)(3) Supplement 2 (16) Steam generator tube failure Opened in 15.4.3 Supplement 1 1

1.9 License Conditions The current status of the license conditions follows:

Braidwood SSER 3 1-4

Status Section Part A Items (1) Inservice inspection program Closed in this 5.2.4, 6.6*

supplement (2) Natural circulation testing Closed in 5.4.3*

Supplement 1 (3) Response time testing Closed in 7.2.2.5*-

Supplement 1 (4) Steam valve inservice inspection Closed in 10.2*

Supplement 1 (5) Implementation of secondary water chemistry Closed in 10.3.38 monitoring and control program as proposed Supplement 1 by the Byron /Braidwood FSAR (6) TMI Item II.F.1: Iodine / Particulate Closed in this 11.5.2 Sampling supplement Part B Items (1) Masonry walls Closed in 3.8.3 Supplement 2 (2) TMI Item II.B.3 postaccident sampling Closed in 9.3.2 Supplement 1 (3) Fire Protection Program Open 9.5.1 (4) Emergency diesel engine auxiliary support Closed in this 9.5.4.1 systems supplement l

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

Braidwood SSER 3 1-5

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1 3 DESIGN CRITERIA FOR STRUCTURES, SYSTEMS, AND COMP 0NENTS 3.6 Protection Against Effects Associated with the Postulated Rupture of Piping 3.6.1 Plant Design for Protection Against Postulated Piping Failures in Fluid Systems Outside Containment By letter dated May 17, 1985, the licensee informed the staff of revised environmental parameters resulting from a high-energy line break (HELB) in the auxiliary building in either the steam generator blowdown (SD) or auxiliary steam (AS) system. Postulated breaks in these systems have been found to have a greater potential effect on environmental conditions than originally predicted.

If a break cannot be isolated within 30 minutes, equipment qualification temperatures can be exceeded.

The licensee further advised in the May 17, 1985 letter that modifications were in progress to install temperature sensors at strategic locations to detect a temperature rise resulting from the local effects of a HELB. A signal from the sensors would initiate automatic closure of an isolation valve in the affected line, thus terminating the break flow. Design, procurement, installa-tion, and testing of the modifications was to be completed by August 31, 1985.

As an interim measure for continued plant operation, the licensee committed to post a watch at locations where breaks could occur. In such an event, the watch personnel would communicate with the control room for operator action to immediately close either the 50 containment isolation valves or isolate the AS supply line.

The licensee also stated that the modifications had been completed on Byron, Unit 1, and would be completed before fuel loading on Byron, Unit 2, and Braidwood, Units 1 and 2.

The staff evaluation of the adequacy of the modifications to isolate HELBs and prevent auxiliary building temperatures from exceeding equipment qualification temperatures is based on an audit review as prescribed in Standard Review Plan (SRP) Section 3.6.1 (NUREG-0800). It includes a review of the design of the AS and SD systems to ensure conformance with the requirements of GDC 4.

The licensee advised that a temperature of 140 F was the minimum envelope temperature for equipment qualification in the auxiliary building. The modifi-cations for isolating a HELB in the AS and SD systems to prevent environmental temperatures from exceeding 140 F in the auxiliary building consisted of (1) temperature sensors mounted in potentially affected areas (but not t.oo close to the affected equipment), (2) provisions for automatic isolation of the system isolation valves, (3) control room alarms, and (4) appropriate procedures.

The licensee further advised that the teisperature sensors and wiring locations in relationship to HELB locations would not be affected by jet impingement. In addition, only Class 1E instrumentation will be used. The staff finds this acceptable.

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Detailc of the staff's evaluation of HELBs in the auxiliary steam system and ,

the steam generator blowdown system are discussed below. l By letter dated May 20, 1985, the staff accepted the interim measures until August 31, 1985 (when the permanent modifications were to be completed). In addition it was requested that the licensee submit the details of the modi-fications for staff review. l By letter dated August 2,1985, the licensee submitted a technical description of the permanent modifications at Bpon 1 that would aid in isolating the HELB for the AS or SD systems in the auxiliary building. They also advised that similar modifications, dependent on break location, would be made on Byron Unit 2 and Braidwood, Units 1 and 2, before their fuel loading dates.  ;

By letter dated August 23, 1985, the licensee advised NRC of installation problems resulting from equipment interference and delays in delivery of temperature sensors. A completion date for the modifications was forecast to the Office of Inspection and Enforcement (IE) to be no later than September 30, l 1985. By letter dated August 27, 1985, the staff accepted the schedule, '

provided that personnel would be kept posted at locations where breaks could occur.

By letter dated December 11, 1985, the licensee responded to a staff request for additional information and also advised of additional modifications that were included to' accommodate postulated single active failures. A rescheduled completion date for Byron, Unit 1, was set for February 28, 1986.

By letter dated April 29, 1986, the licensee provided the following additional information with regard to the proposed modifications.

Auxiliary Steam (AS) System Auxiliary boiler steam or main turbine extraction steam is supplied to the AS system. The AS system provides low pressure steam to various auxiliary build-ing loads that include the boric acid and radwaste systems. A tie line allows for the interconnection of Units 1 and 2.

The licensee has installed redundant temperature sensors in the vicinity of postulated break locations in the auxiliary building where break effects may not be confiaed to non-safety areas. At a setpoint of 140 F, the sensors will initiate (1) an alarm in the control room and (2) automatic isolation of the AS system from the auxiliary building by, closure of the parallel, pressure-regulating valves AS013 and AS167.

To ensure that steam flow is isolated in the event of failure of either of these valves, the licensee stated in a letter dated August 2, 1985 that pro-cedures were being developed to require local manual closure of valve AS012 on receipt of a control room alarm. This valve is positioned upstream and *n series with valves AS013 and AS167. By letter dated December 11, 1985, in response to a staff inquiry, the licensee advised that sufficient time for manual operator action may not be available in the event of a single failure of either of valves AS013 or AS167, and stated that further modifications would be made. By letter dated April 29, 1986, the licensee committed to install a redundant automatic isolation valve (AS286) downstream, in series with valves Braidwood SSER 3 3-2

AS013 and AS167. The location of the isolation valves ensures that upon a HELB, auxiliary steam from both the affected unit and unaffected unit (via the-crosstie) is isolated.

The staff finds that the licensee has provided adequate, redundant isolation valves to ensure isolation of the AS system in the event of a HELB. This will ensure that the temperature in the auxiliary building will not exceed equipment qualification temperatures.

Steam Generator Blowdown (SD) System The SD system contributes to maintaining proper water chemistry in the steam generators. Each unit's steam generators blow down to a condenser located in the auxiliary building. A crosstie between Units 1 and 2 allows for the blowdown of all steam generators to one blowdown condenser. This may be desired during maintenance of a blowdown condenser.

The licensee has installed redundant temperature sensors in the vicinity of postulated break locations in the auxiliary building where the break may affect safety-related equipment. At a setpoint of 140*F these sensors will initiate (1) an alarm in the control room and (2) automatic isolation of the SD system of the unit with the activated sensors. In response to staff questions by letters dated December 11, 1985 and April 29, 1986, the licensee committed to install redundant automatic isolation valves in each steam generator blowdown line. This will ensure that in the event of a HELB, concurrent with a single active failure, the SD system wili be isolated and prevent auxiliary building temperatures frcm exceeding equipment qualification temperatures.

Automatic isolation of the SD system may not occur, however, if the crosstie between units is being used. For example, in the event of a HELB in one unit, temperature sensors will only initiate isolation of the affected unit's SD system; the unaffected unit's SD system could feed the break via the crosstie.

In response to a staff inquiry regarding this, the licensee committed by letter dated April 29, 1986 to revise HELB procedures involving the SD system. Now, on control room annunciation of a 140*F condition, an operator of the affected unit will verify the crosstie valve positions. If the valves are open, operator action will be required at the main control board to close the automatic isolation valves of the unaffected unit's steam generator blowdown lines.

t The staff finds that the licensee has provided acequate redundant isolation valves and procedures to ensure isolation of the SD system in the event of a' HELB, thus preventing auxiliary building temperatures in excess of equipment qualification temperatures. ,

From the staff's evaluation of the licensee's plant modifications and proce-dures, it concludes that the designs of the AS and SD systems for protection against postulated piping failures in the auxiliary building are acceptable, and meet the requirements of General Design Criterion (GDC) 4 with respect to accommodating the effects of postulated pipe ruptures.

V Braidwood SSER 3 3-3

5 REACTOR COOLANT SYSTEM 5.2 Integrity of Reactor Coolant Pressure Boundary 5.2.4 Reactor Coolant Pressure Boundary Inservice Inspection and Testing 5.2.4.3 Evaluation of Compliance with 10 CFR 50.55a(g) for Braidwood Unit 1 By letter dated April 3, 1987, the licensee committed to submit the Braidwood Unit 1 inservice inspection (ISI) program by October 17, 1987. The initial 120-month inspection interval will comply with tha American Society _of Mechani-cal Engineers (ASME) Boiler and Pressure Vessel Code (the Code),Section XI, 1983 Edition, including the summer 1983 addenda. . This ISI program will be similar to the ISI program established for Byron Unit 1. Therefore, License Condition A(1) is no longer required.

Braidwood SSER 3 5-1

6 ENGINEERED SAFETY FEATURES 6.4 Control Room Habitability By letter dated June 3, 1986, the licensee provided an analysis to demonstrate that the chlorine monitors for the control room intake are not required. ~The analysis included a re-survey of the offsite sources of chlorine within 5 miles of the station, a calculation of the dispersion of the toxic gas and control room infiltration, and a probability evaluation of the rupture of a chlorine tank car on the Norfolk and Western Railroad.

The licensee's analysis stated that the probability of a chlorine tank car ,

rupture within a 5-mile radius of the plant is significantly less than 10 s.

per year. The licensee's analysis presumed that no toxic gas monitors existed and that the station operator would not don the self-contained breathing apparatus until the chlorine odor was detected (3.5 parts per million (ppm)).

The licensee calculated that the concentration existing in the control room 2 minutes after detection would be 13.3 ppe (toxicity is~15.0 ppm). The li-censee stated that this allows adequate time for the operators to don self-contained breathing apparatus.

The NRC staff has independently reviewed the probability of a chlorine tank car rupture within the 5-mile radius and calculated the concentration in the-control room assuming no toxic gas monitors and detection by odor at 3.5 ppe.

The staff's analysis of the probability of a chlorine tank car rupture within 5 miles of the Braidwood facility was 1.3 X 10 8 per year, which was a factor of approximately 30 greater than the probability calculated by the licensee.

The reason for this disparity was attributed to the fact that the licensee calculated the probability of an accident based on the frequency of chlorine tank car shipments per year and release per car mile (any type). The calcula-tion should have been based on the probability of a release per chlorine car mile.

The staff's analysis also showed that if the licensee depended on the operators to detect the chlorine at a concentration of 3.5 ppe and then to initiate action to protect themselves, an insufficient period of time existed for the operators to don self-contained breathing apparatus before toxicity limits were exceeded. Consequently, with the absence of chlorine monitors, the staff believes that if a release occurs within 5 miles of the Braidwood facility, the control room should be placed in the isolation mode before chlorine odor is detected by the control room' operators. This can~be accomplished by early notification of a chlorine tank car accident.

By letter dated December 23, 1986, the licensee indicated that representatives-from Will County, Illinois, agreed to provide notification to the Braidwood Station in the event of a chlorine accident. The staff believes that this notification process is an acceptable means of alerting the Braidwood Station so that the control room can be isolated. In addition, the station wil.1 be required to include.(1) with their control room technical specifications a '

surveillance requirement to demonstrate on an 18-month basis that the control Braidwood SSER 3 6-1 w.g r .

room envelope can be isolated, and (2) a procedure to demonstrate, on an 18-month basis, that control room envelope integrity is' maintained (i.e., infiltration into the control room envelope in the isolation mode does not negate the toxic gas analysis and thus, the capability to protect the operators). The first demonstration that the control room envelope integrity is maintained must be completed before the fuel loading date for Braidwood Unit 2.

Although the probability of the occurrence of a chlorine release is small, compensatory actions are in place that would be used to mitigate the conse-quences of such an event. On the basis of the small probability of a chlorine event, the licensee's commitment to maintain prompt notification communications with the county, the requirement to isolate the control room when notified of a toxic gas incident, and the requirement to demonstrate control room integrity on a routine basis, the staff has determined that the removal of chlorine monitors from the Braidwood Station control room intakes is consistent with the guidelines of SRP Section 6.4 and is, therefore, acceptable. By letters dated

April 2, 1987 and May 4, 1987, the licensee committed to demonstrate, en an 18-month basis, that the control room envelope can be isolated and that its integrity is maintained. Therefore, Outstanding Item A(10) is considered closed.

6.5 Fission Product Removal and Control System 6.5.1 Engineered Safety Feature Atmospheric Cleanup System l

By letter dated November 17, 1986, the licensee indicated that laboratory analysis of the charcoal to be used in their safety (emergency safety feature) and non-safety grade air filtration units, when tested at 30 C and at 95%

relative humidity, demonstrated an average penetration of greater than 1% in nine out of ten batches. The average penetration was 1.5%, which was higher than the maximum of 1% to which the licensee had committed in a letter to the NRC staff dated October 21, 1986. The staff has evaluated this data and has determined that this adsorber material is acceptable. The material has an efficiency greater than that assumed in the safety analysis and, although its efficiency does not reflect the safety factor of seven presented in Regulatory Guide (RG) 1.52,, it does reflect a safety factor that is greater than 3.

Although the staff considers the use of this charc ol acceptable, the licensee may need to consider purchasing future batches of cnerceC that are impregnated with more than just potassium iodide (KI), such as charcoal impregnated by TEDA (triethylenediamine). Such charcoal exhibits greater adsorption capabil-ities than charcoal impregnated with KI alone.

The licensee also indicated in the letter dated November 17, 1986 that the carbon provided to fill the last six carbon trays and twenty test canisters is not identical to the type of carbon that is in the other trays and, in addi-tion, the American Society for Testing and Materials (ASTM) Standard D-4069 test for carbon properties showed that the carbon is slightly out of specifi-cation in that the particle size for the material retained by the #16 sieve is 39% rather than the required 40%. This charcoal, which is impregnated with TEDA and KI, when laboratory tested at 95% relative humidity and 30*C showed a penetration of 0.22% for methyl radiciodine. This TEDA- and KI-impregnated charcoal is approved by the staff for use at the Braidwood facility provided that the TEDA- and KI-impregnated charcoal is not mixed with the charcoal impregnated with KI alone and provided that the test canisters use TEDA- and Braidwood SSER 3 6-2

KI-impregnated charcoal if the cells use TEDA- and KI-impregnated charcoal.

The test canisters must contain charcoal that represents the type in the ventilation system filter units. The staff does not consider the small difference in particle size distribution as significant in this application.

Therefore, use of this charcoal is approved with the conditions noted above.

6.6 Inservice Inspection of Class 2 and 3 Components 6.6.3 Evaluation of Compliance with 10 CFR 50.55a(g) for Braidwood Unit 1 By letter dated April 3, 1987, the licensee committed to submit the Braidwood Unit 1 ISI program by October 17, 1987. The initial 120-month inspection interval will comply with the ASME Code,Section XI, 1983 Edition, including the summer 1983 addenda. This ISI program will be similar to the ISI program established for Byron Unit 1. Therefore, Licensee Condition A(1) is no longer required.

Braidwood SSER 3 6-3

7 INSTRUMENTATION AND CONTROL 7.6 Interlock Systems Important to Safety 7.6.2 Specific Findings By letter dated August 2, 1985, the licensee described the design details of a system being installed in Byron Units 1 and 2 and Braidwood Units 1 and 2 to detect and automatically isolate high-energy line breaks (HELBs) in the steam generator blowdown (SD) and auxiliary steam (AS) systems. The intent of the isolation system is to ensure that an HELB in either the SD or AS system will not result in a significant temperature increase in the auxiliary building above the qualification temperatures of safety-related equipment located within the building. Further design information was provided in a letter from the licensee dated April 29, 1986 covering a modification to the isolation system.

The following documents the NRC staff's review and evaluation of the licensee's HELB isolation system based on SRP Section 7.6.

Steam Generator Blowdown System Each steam generator blowdown line is provided with redundant, series, air-operated isolation valves located in the isolation valve room. The valves, which are held in the open position, close when the plant instrument air supply is removed either by loss of the air source or by de-energization of the solenoid valves. Redundant electrical trains of the HELB isolation function die provided that de-energize the solenoid valves (one solenoid per valve per train).

The HELB isolation signals for the SD system are developed from two temperature switches per train in each of two areas within the auxiliary building. Closure of any of the four temperature switches within a train on a high temperature (approximately 134 F) energizes a corresponding relay that, in turn, de-energizes the four solenoid valves (one per blowdown line) in the train. Annunciation is provided in the control room on closure of any temperature switch. Electrical power for the isolation signals is provided from redundant safety-related sources. Appropriate surveillance and limiting conditions of operation for the system will be included in the technical specifications.

Auxiliary Steam System The auxiliary steam system line is provided with two parallel-flow, air-operated isolation valves in series with a redundant air-operated isolation valve located in the turbine building. These valves are held open and close when the plant instrument air supply is removed either by loss of the air source or by de-energization of the solenoid valves. Redundant electrical trains of the HELB isolation function are provided that de-energize the solenoid valves (one solenoid for the parallel valves for one train; another solenoid for the series valve for the other train).

i Braidwood SSER 3 7-1

. ._ _ . . - - - - ~

The HELB isolation signals for the AS system are developed from one temperature switch per train in each of six areas within the auxiliary building. Opening

!. .of any of the six temperature switches within a train on.a high temperature 1

(approximately 134*F) de-energizes a corresponding-relay that, in turn,.de-E energizes the solenoid valve in the train. Annunciation is provided -in' the.

control on actuation of any temperature switch. Electrical power for the isolation signals is provided from redundant safety related sources. Appropri-ate surveillance and limiting conditions of operation for the isolation system

, are included in the technical specifications.

SRP Section 7.6 states that "The objectives of the review are to confirm that design considerations such as redundancy,-independence, single failures,-

i . qualification, bypasses,- status' indication, and testing are consistent with the design bases of these systems and commensurate with the importance of the 4 safety functions to be performed." Further, the SRP states that specific cri-teria of the Institute of Electrical and Electronics Engineers (IEEE) Standard 279 are to be used as guidance for systems reviewed under SRP Section 7.6.

The staff's review indicates that the HELB isolation system design satisfies 1

most of the considerations of the criteria in IEEE Std. 279 because it is a

, redundant, safety-related system comprised of all Class 1E components that are qualified for the appropriate environment. It is automatically ir,itiated when i setpoints for the area temperatures are exceeded and there are no control /

j protection system interactions. The system will be tested as~ required by the

! technical specifications. There are no operating or manual bypasses in the-

j system and no multiple setpoints. Once initiated,-the system is isolated until system operation is restored by the operator through the use of reset switches.

' Although there is no manual initiation capability for the system, manual isolation of the SD and AS systems is possible using other solenoids for some of the isolation valves or other isolation valves in the lines. Although valve position indication is not provided in the control room for the-isolation j valves, flow indication and the annunciation from the temperature switches is 4 available to indicate that the isolation function has' occurred. On the basis i that the system design satisfies most of the considerations of IEEE Std. 279 as i discussed above, the staff finds it acceptable.

Although the staff finds the design acceptable, two deviations from IEEE Std. ,

279 criteria are apparent.- First, one annunciator is provided for.each of the I i systems (AS and SD)-isolated. These annunciators are common to the system- '

l. related temperature swi.tches and:no separate indication at the sensor. level is i provided.' Second, the isolation function is generated effectively by one-out-

of-eight (SD system) or-one-out-of-twelve (AS system) logic. No provision has

! been provided to bypass a failed temperature switch or.to prevent system isolation during testing. In response to these two issues, the licensee.

indicated that the information available to the operator on system status is i sufficient and that the AS and SD systems would remain in an-isolated condition until a failed temperature switch is repaired. The staff finds this acceptable.

)

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Braidwood SSER 3 7-2

9 AUXILIARY SYSTEMS 9.5 Other Auxiliary Systems 9.5.1 Fire Protection Program In Braidwood SSER 2, the staff identified three unresolved issues, namely, an incomplete-high/ low pressure interface analysis of the reactor head vent valves and excess letdown lines, the' lack of smoke detectors in the main control console, and a need for revising procedures (PRI-5) to clearly identify those instruments and controls that are electrically isolated from the control room when using transfer switches placed in remote positions. The following discussion addresses the resolution of those three issues. Therefore, Out-standing Item B(6) is considered closed.

9.5.1.2 Fire Protection Program Requirements By letter dated August 29, 1986, the licensee submitted a proposed revision to FSAR Section 9.5-1. This revision will (1) incorporate the fire protection program that has been approved by the staff, including the fire hazards analysis and commitments that form the basis for the fire protection program, into the FSAR by reference to specific previous submittals, and (2) include a commitment to establish limiting conditions for operation, action statements, and surveil-lance requirements for the fire protection program within the Braidwood plant operating procedures. The licensee stated that these procedures will provide a level of protection equivalent to that provided by the fire protection sections of the Westinghouse Standard Technical Specifications (NUREG-0452, Rev. 4).

Furthermore, in the August 29, 1986 letter, the licensee committed to incor-porate this revision to Section 9.5-1 into the Braidwood FSAR by means of a future amendment.

9.5.1.4 General Plant Guidelines Safe Shutdown Capability By letters dated September 29, October 23, November 3, and December 12 and 15, 1986, the licensee submitted additional information regarding the high/ low pressure interface analysis of the reactor head vent valves and excess letdown lines. In the November 3, 1986 submittal, the licensee proposed to remove power from the reactor head vent valves in the event of a fire (which is stated in the PRI-5-Rev. 54 procedures). The staff finds this consistent with the guidelines of Branch Technical Position (BTP) CMEB 9.5-1, Section C.S.b and, therefore, finds this action acceptable.

By letter dated October 23, 1986, the licensee provided the results of an analysis concerning spurious operation of the valves that form the high/ low pressure interface for the excess letdown lines. The licensee has stated that the valves are arranged in such a way that at least four spurious valve fail-ures would have to occur simultaneously to create an uncontrolled coolant loss to the reactor coolant drain tank (RCDT) in the event of a fire. Because of Braidwood SSER 3 9-1 j

I the number of valves involved, their simultaneous spurious operation resulting from fire is not deemed credible. Therefore, the design is consistent with the guidelines of BTP CMEB 9.5-1, Section C.S.b. On this basis, the staff finds the licensee's analysis for the excess letdown lines acceptable. This item is, therefore, considered closed.

By letter dated December 12, 1986, the licensee provided additional information concerning clarifications in the PRI-5 procedures. The licensee has stated that those instruments and controls available at the fire hazards panel (FHP 1PL10J) that may be electrically isolated from the control room panels are clearly labeled. Further, the instruments and controls are powered from separate, isolated power supplies. On this basis, the staff finds that the licensee's proposed changes to the PRI-5 procedures are consistent with the guidelines of BTP CMEB 9.5-1, Section C.S.b, and, therefore, are acceptable.

This item is considered closed.

By letter dated February 17, 1987, the licensee provided additional information concerning a design change in the power feed to the emergency battery pack lighting units. The problem with the units not turning on in the event of a loss of offsite power is correctable by switching the power feed from the ac (alternating current) essential lighting to non-essential lighting. For Braidwood Unit 1, the battery packs will be installed before reaching initial criticality, and the associated rewiring will be completed before exceeding 5%

power. For Braidwood Unit 2, installation of the battery packs and completion of the power feed design change will be completed before fuel loading.

The staff finds that with the proposed power feed design change, the emergency battery pack lighting system design is consistent with the guidelines of Sec-tion C.5.g.(2) of BTP CMEB 9.5-1, and is, therefore, acceptable.

In SSER 2, the staff stated that the applicant committed (by letter dated September 22, 1986) to install sprinklers at the hatchway openings on eleva-tions 346, 401, and 426 feet before exceeding 5% of rated power. Additionally, draft curtains were to be added to the hatchways no later than 6 months after fuel loading. However, by letters dated February 17 and March 16, 1987, the licensee notified the staff that 6-inch-thick, removable concrete plugs will be installed at the above hatchway elevations and at elevation 383 feet. The concrete plugs for the hatchway at elevation 426 feet will be completed before 5% rated power for Unit 1. However, the concrete plugs at hatchway elevations 364, 383, and 401 feet will be installed before fuel loading at Braidwood Unit 2 because the hatchways are located on the Unit 2 side of the auxiliary building.

The proposed installation of concrete plugs at the hatchways will provide fire protection equivalent to the sprinklers and draft curtains. Further, the licensee will institute compensatory measures (such as a fire watch) for the hatchways on elevations 364, 383, and 401 feet until the concrete plugs are installed. On the basis of the staff's review and the licensee's commitments, the staff has found the proposed resolution acceptable; therefore, this item is considered closed.

Braidwood SSER 3 9-2

l l

1 9.5.1.5 Fire Protection for Plant-Specific Areas l

Control Room l

Section C.7.b of BTP CMEB 9.5-1 states that smoke detectors should be provided in the control room cabinets and consoles. In Amendment 4 to the Fire Protec-tion Report (FPR), the licensee stated that fire detectors were installed in the main control console, in the vents of the cabinets and at the ceiling of the control room. On this basis, tne staff concluded that the smoke detection system for the control room was acceptable. During the fire protection inspec-tion between August 18 and 22, 1986, the staff observed that smoke detectors were not installed in the main control room console. This issue remained open in SSER 2.

By letter dated September 29, 1986, the applicant described the capability of the smoke detection system to detect fires in the main control room console.

The console panels are provided with return air ducting from the control room ventilation system that terminates at the top of each panel. Each panel return has at least one duct smoke detector to detect smoke within the panels.

This provides discrete annunciation of smoke alarms by panel. The duct smoke detectors are installed in accordance with manufacturer's recommendations with respect to air flow velocity and detector location. There are no physical barriers separating the interiors of each panel that would restrict smoke movement to the return air ducting. The control room ventilation system is powered by an emergency safety feature (ESF) bus to ensure that the ventila-tion system remains operable and that return air from the panels passes the j detectors. -

The applicant's method of detecting fires within the main control room consoles provides reasonable assurance of early fire detection. On this basis, the installation of detectors for the main control room consoles as delineated in the applicant's September 29, 1986 letter is an acceptable deviation from Section C.7.b of BTP CMEB 9.5-1.

9.5.4 Emergency Diesel Engine Fuel Oil Storage and Transfer System 9.5.4.1 Emergency Diesel Engine Auxiliary Support Systems (General)

The license for Unit 1 contained a condition that required satisfactory comple-tion of dynamic qualification of the instrumentation and controls on the diesel generator control panel. The concern was discussed in Supplement 1.

My letter dated November 18, 1986, the licensee provided additional information in response to the staff's August 23, 1985 letter.

The basic dynamic qualification of the control panels for Braidwood Units 1 and 2 was established on the basis of similarity with the panel including its in-strumentation and other devices at the LaSalle plant. The LaSalle control panel was qualified by actual testing. The staff performed a review of the additional information provided in the November 18, 1986 letter and concluded the following:

Braidwood SSER 3 9-3

(

(1) During vibration testing of the panel, acceleration responses were re-corded at seven locations where safety-related devices are mounted on the panel. This is satisfactory since a direct response measurement at the device location was obtained vs the excitation at the base of the panel f only.

(2) On the basis of the acceleration plots on the LaSalle panel, it is evident that driving frequencies during generator operation are in the 10- to 1000-Hz range, which is significantly beyond seismic input range. There-fore, the margin available at the zero period acceleration for the safe shutdown earthquake is adequate to account for the combined effects of the earthquake and operational vibration.

(3) Similarity between the LaSalle components and Braidwood Units 1 and 2 components was indicated in a component by component listing. That is satisfactory.

(4) By providing the details of methods and results of calculations per-formed, the licensee has established fatigue endurance of the Braidwood Units 1 and 2 control panels. The fatigue damage potential for the Braid-wood Units 1 and 2 panels subjected to a synthetic time history was eval-uated by comparing them to the fatigue damage induced by measured time histories for the LaSalle panels (qualified by fatigue testing).

On the basis of the above, the staff concludes that the investigation conducted )

by the licensee on the dynamic qualifications of-its diesel generator control i panels, including the instrumentation and control devices, is satisfactory, and there is reasonable assurance that the subject equipment should perform its safety function adequately. Therefore, License Condition B(4) is no longer required.

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Braidwood SSER 3 9-4

l 11 RADI0 ACTIVE WASTE MANAGEMENT 11.5 Process and Effluent Radiological Monitoring and Sampling Systems The Unit 1 license contained a condition that the licensee demonstrate that the iodine / particulate sampling system will perform its intended function. This issue was discussed in Supplement 2.

In a letter dated December 29, 1986, the licensee submitted a description of the modifications that were made at Braidwood Station, Units 1 and 2, to demon-strate the capability to sample for gaseous radioiodine in conformance with THI Action Plan Item II.F.1. This information was submitted to provide the NRC staff sufficient detail to evaluate whether the concerns identified in Supplement 2 had been alleviated.

The licensee replaced its 1/4-inch diameter high-range sample line with a 3/4-inch-diameter heat-traced line that samples under both low and mid/high-range operating conditions. An auxiliary pump skid was added with automatic isokinetic flow control through the use of a flow splitter manifold. This auxiliary pump is automatically started by the radiation monitor's coprocessor and stopped by the high flow sample pump. The flow splitter manifold directs the entire sample to the high flow path during low range radioactivity opera-tion when the hi path during mid/gh flow sample pump is operating and to the low flow samplehigh rang auxiliary pump are running.

The staff believes that the use of this high flow sample line (3/4-inch) elimi-nates the concern expressed in the Region III inspection report enclosed in the July 10, 1984 letter from C. J. Paperiello to C. Reed (Inspection Report 50-454/84-33). However, the NRC has contracted Pacific Northwest Labora-tories (PNL) to study the problems associated with obtaining representative samples from these sampling lines. Although the staff finds the present modi-fications to the Braidwood facility acceptable, additional modifications may be required based on the recommendations of the PNL study. For the present, the licensee's modifications are deemed acceptable. Therefore, License Condi-tion A(6) is no longer required.

Braidwood SSER 3 11-1 l

l

l 13 CONDUCT OF OPERATIONS 13.3 Emergency Planning 13.3.1 Atomic Safety and Licensing Board Decision on Emergency Planning Issues The Atomic Safety and Licensing Board hearings on the emergency preparedness contention were conducted on October 29, 1985 and March 11-12, 1986. The hear-ing record was closed on March 12, 1986, and the Board's Partial Initial Deci-sion on Emergency Planning Issues was issued on May 13, 1987. In this decision the Board stated:

The Board concludes that with respect to all matters in controversy, the offsite emergency response plan for the Braidwood Station com-plies with the applicable provisions of 10 CFR @ 50.47 and 10 CFR Part 50, Appendix E, and provides reasonable assurance that adequate protective measures can and will be taken in the event of a radiolog-ical emergency; provided, that Applicant shall include in the next annual revision of its booklet, " Emergency Information - Braidwood" a ,

discussion of (a) the physical characteristics of a radioactive plume; (b) the significance of wind speed and direction in the movement of the plume; (c) the relationship between weather condi-tions and the selection of optimum evacuation routes, the latter topic to be covered in the section of the booklet dealing with evacuation.

In a May 18, 1987 letter to the NRC, the licensee committed to comply with the Board's instructions within approximately 12 weeks, which is the next scheduled date for the revision and distribution of the licensee's public information booklet.

On the basis of this commitment, the staff finds that the licensee has satis-factorily responded to the Board's requirements with respect to this issue.

Braidwood SSER 3 13-1

-- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _a

APPENDIX A CONTINUATION OF CHRON0 LOGY OF NRC STAFF RADIOLOGICAL SAFETY REVIEW OF BRAIDWOOD STATION, UNITS 1 AND 2 October 16, 1986 Notice of Hearing-Change issued by the ASLB noting that the evidentiary hearing reconvened in Courtroom #1743 at the Federal Building, 219 South

Dearborn Street,

Chicago, Ill.

50604. It would continue from day to day during the week.

October 17, 1980 Letter from licensee concerning Final Draft Technical Specification Seismic Instrumentation System.

October 27, 1986 Letter to licensee transmitting 20 copies of NUREG-1002, Supplement No. 2 (Braidwood Station SSER 2).

October 23, 1986 Letter from licensee concerning results of analysis of spurious operation of valves.

October 29, 1986 Letter from licensee concerning environmental qualification of electrical penetrations.

October 30, 1986 Letter from licensee concerning Final Safety Analysis Report (FSAR) update.

November 6, 1986 Order denying licensee's petition for review of an Appeal Board decision (ALAB-817, 22 NRC 470).

November 6, 1986 Letter from licensee concerning emergency diesel generators.

i November 7, 1986 Letter from licensee concerning emergency core cooling system operability.

November 7, 1986 Order concerning disclosure of certain investigatory mate-rials related to Order dated July 31, 1986.

November 7, 1986 Protective Order concerning disclosure of certain inves-tigatory materials related to order dated July 31, 1986.

l November 17, 1986 Letter from licensee concerning atmospheric cleanup filter

units, Regulatory Guides 1.52 and 1.140.

i l November 19, 1986 Letter from Philip P. Steptoe, Isham, Lincoln & Beale (IL&B), to NRC acknowledging Commission's Protective Order of November 7, 1986.

November 20, 1986 Transcript of hearing held on November 20, 1986 in Chicago, Illinois.

Braidwood SSER 3 1 Appendix A

November 21, 1986 Letter from IL&B concerning Affidavit of Nondisclosure related to Protective Order entered on December 6, 1985.

December 1, 1986 Letter from licencee concerning Initial Test Program.

December 3, 1986 Letter from IL&B concerning Protective Order dated November 7, 1986.

December 9, 1986 Transcript of December 9,1986 telephone conference with licensee, NRC, and intervenor.

December 11, 1986 Letter from IL&B concerning Protective Order dated November 7, 1986.

December 11, 1986 Letter from licensee concerning Initial Test Program.

December 12, 1986 Letter from licensee concerning Resolution of Fire Pro-tection Program Unresolved Items.

December 17, 1986 Order admitting licensee's Exhibit 188 and closing record.

December 19, 1986 Letter from licensee concerning Revision 14 of the Security Plan.

December 19, 1986 Letter from Business & Professional People for the Public Interest concerning Exhibit 98.

December 23, 1986 Letter from licensee concerning habitability of control room following postulated accidents involving shipments of chlorine.

December 29, 1986 Letter from licensee concerning plant effluent sampling.

January 5, 1987 Letter from IL&B concerning licensee's motion for extension of page limit for brief.

January 16, 1987 Letter to licensee concerning control room habitability--

utilization of charcoal absorber material.

January 22, 1987 Letter from licensee concerning Endorsement No. 12 to facility form policy NF-294.

January 23, 1987 Letter to licensee concerning Generic Letter 83-28, Item 4.5.2 (reactor trip system reliability--online testing).

February 3, 1987 Order granting page limit increase and extensions of time.

February 6, 1987 Letter to licensee concerning high-energy line breaks in the auxiliary steam and steam generator blowdown systems for Byron /Braidwood stations.

February 11, 1987 Letter to licensee concerning reconstitution of a fuel assembly, request for additional information.

Braidwced CSEP. 3 2 Appendix A

l l'

February 11, 1987 Letter from IL&B concerning abstracts for the quality con-trol (QC) inspectors.

February 13, 1987 Letter from IL&B concerning abstracts for October 23, 1986.

February 17, 1987 Letter from licensee concerning Appendix R, Emergency Battery Pack Lighting.

February 18, 1987 Filing of reply brief of licensee, Commonwealth Edison Company.

February 20, 1987 Letter from IL&B concerning abstracts for August 25 and 27 and fo: November 6, 12, 13, 20, 21, and 26, 1986.

! February 25, 1987 Letter from IL&B concerning abstracts for November 18 and

19, 1986.

1 February 25, 1987 Letter from licensee concerning revisions to FSAR.

4 February 26, 1987 Letter to licensee concerning 50% load reduction startup test.

March 4, 1987 Letter to licensee concerning removal of control room

, chlorine monitors.

1 March 4, 1987 Letter from licensee concerning Application for Amendment to Facility Operating Licenses NPF-37 and NPF-66, Appen-l dix A, Technical Specifications.

1 March 10, 1987 Letter from IL&B concerning licensee's exhibits.

March 11, 1987 Letter to licensee concerning TMI Action Item II.K.3.31.

I  ;

March 11, 1987 Letter from IL&B concerning exhibits list.

March 16, 1987 Letter from licensee concerning fire protection program.

March 17, 1987 Letter from licensee concerning emergency diesel generator l

missile effects.

i March 17, 1987 Letter from licensee concerning diesel engine generator rocker arm failure, j March 20, 1987 Letter from licensee concerning Nuclear Site Property Damage Insurance.

March 23, 1987 Letter from licensee concerning application for amendment regarding reconstitution of a fuel assembly.

March 24, 1987 Letter from licensee concerning application for amendment regarding source range flux instrumentation.

March 25, 1987 Letter from licensee concerning proposed exceptions to FSAR Appendix A, Regulatory Guides 1.52 and 1.140.

Braidwood SSER 3 3 Appendix A J

k 4

~

4 March'26, 1987 Letter from licensee concerning control room ventilation interim operation plant.

March 26, 1987 Letter to licensee concerning performance testing of

]

relief and safety valves.

March 27, 1987 Letter from licensee concerning interim operation of heating, ventilating, and air conditioning (HVAC) systems.

March 30, 1987 Order from Atomic Safety and Licensing Board concerning requesting welding procedure documents.

April 2, 1987 Letter from licensee concerning minor corrections to I

Technical Specifications.

April 2, 1987 Letter from licensee concerning removal of control room

. chlorine monitors.

April 2, 1987 Letter from licensee concerning diesel engine generator rocker arm failure.

l April 3, 1987 Letter from licensee concerning inservice inspection (ISI) program.

May 18, 1987 Letter from licensee concerning Atomic Safety and Licensing j Board's decision on emergency planning.

I 1

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I 4

I 4

i l

4 Braidwood SSER 3 4 Appendix A l

i l'

i APPENDIX B BIBLIOGRAPHY l

U.S. Nuclear Regulatory Commission, NUREG-0452, " Westinghouse Standard Technical Specifictions," Rev. 4, November 1981.

i -- , NUREG-0800, " Standard Review of Safety Analysis Reports for Nuclear Power

, Plants," LWR Edition, July 1981.

i

-- , Office of Inspection and Enforcement, Inspection Report "

50-454/84-33, j July 10, 1984.

l INDUSTRY CODE AND STANDARDS American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel j Code,Section XI, 1983 Edition.

l American Society for Testing and Materials (ASTM), D-4069, " Standard Specifica-

- tion for Impregnated Activated Carbon used To Remove' Gaseous Radio-iodines from Gas Streams," 1981.

Institute of Electrical and Electronic Engineers (IEEE), 279, " Criteria for j Protection Systems for Nuclear Power Generating Stations," 1971.

4 l

i I

I i

Braidwood SSER 3 1 Appendix B l

.. .. . _ . . ~ _ . __ . . . . . _ _. _ _ .

l 1

APPENDIX F l I NRC STAFF CONTRIBUTORS AND CONSULTANTS NRC STAFF Name Title' Review Branch

• Goutam Bagchi Branch Chief Structural and Geosciences, DEST

, Frederick H. Burrows Reactor Engineer Instrumentation and Control systems, DEST John J. Hayes, Jr. Nuclear Engineer Project Directorate II-1 DRP-I/II

! Marthe E. Harwell Acting Branch Chief ' Policy & Publications ,

l Management, DPS I George Johnson Materials Engineer Materials Engineering, DEST Calvin W. Moon Senior Reactor Engineer Technical Specifications, i 00EA

AmarjitSingh Mechanical Engineer Inspection, Licensing i'

and Research Integration, PMAS Catherine S. Vogan Licensing Assistant ProjectDirectorate 1 1-1, DRP-1/2 i

I i

1 i

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• Reflects reorganization since SER was issued.

i i

t Braidwood SSER 3 1. Appendix F l

[

APPENDIX I ERRATA TO BRAIDWOOD STATION, UNITS 1 AND 2, SAFETY EVALUATION REPORT AND ITS SUPPLEMENTS Page Line Change Supplement 2 6-8 22 Delete sentence beginning "The applicant stated that the RHR system...".

9-4 14 Change "a fire-rated Material" to "non-combustible l material".

9-4 24 Change "a minimum horizontal separation of approximately i

15 feet" to "a minimum horizontal distance of 52 feet".

9-5 29 Delete " including tower fans".

l 9-6 10 Change "05X162-A, B, C and D" to "0SX162-A, B, C, and D".

i i

l Braidwood SSER 3 1 Appendix I

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The first supplement to NUREG 0 2 was issued in September 1986; the second supplement to NUREG-1002 was ssu d in October 1986. This third supplement to NUREG-1002 reports the s tus certain items that remain-d unresolved

at the time Supplement 2 w publi ed. The facility is located in l Reed Township, Will Coun , Illinoi l

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