ML20149C664
| ML20149C664 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 12/18/1987 |
| From: | Sands S Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20149C663 | List: |
| References | |
| NUDOCS 8712280150 | |
| Download: ML20149C664 (24) | |
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7590-01 COMMONWEALTH EDIS0N COMPANY BRAIDWOOD STATION, UhlT h0. 2 DOCKET NO. 50-457 NOTICE OF ISSUANCE OF FACILITY OPERATING LICENSE Notice is hereby given that the U.S. Nuclear Regulatory Commission (the Commission or NRC), has issued Facility Operating License No. NPF-75 to Comonwealth Edison Company (the licensee) which authorizes operation of Braidwood Station, Unit No. 2 (the facility) at reactor core power levels not in excess of 3411 rr.egawatts thermal (100 percent rated power). This license is restricted to power levels not in excess of five percent of rated power (170 megawatts thermal).
Braidwood Station, Unit No. 2 is a pressurized water reactor located in Will County, Illinois, about 20 miles south-southwest of Joliet, Illinois, in Reed Township. The license is effective as of the date of issuance.
1 The application for the license complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's regulations. The Commission has made appropriate findings as required by the Act and the Commission's regulations in 10 CFR Chapter 1 which are set forth in the license.
Prior public notice of the overall action involving the proposed issuance of an operating license was published in the Federal Register in December 15,1978(43FR58659).
8712280150 871218 PDR ADOCK 05000457 p
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. The Comission has detemined that the issuance of this license will not result in anf environmental impacts other than those evaluated in the Final Environmental Statement and the Assessment of the Effect of License Duration on Matters Discussed in the Final Environmental Statement for the Braidwood Station, Units 1 and 2 (dated June 1984) since the activity authorized by the license is encompassed by the overall action evaluated in the Final Environmental Statement.
For further details with respect to this action, see (1) Facility Operating License No. NPF-75, with Technical Specifications and the E
Environmental Protection Plan; (2) tha report of the Advisory Comittee on Reactor Safeguards, ' dated February 11, 1985; (3) the Comission's Safety Evaluation Report, dated November 1983, (NUREG-1002), and Supplements 1 through 5; (4) the Final Safety Analysis Report and Amendments thereto:
(5) the Environmental Report and supplements thereto; and (6) the Final Environmental Statement, dated June 1984, (NUREG-1026).
These items are available for inspection at the Comission's Public Document Room located at 1717 11 Street, N.W. Washington, DC 20555 and in the Wilmington Township Public Library, 201 S. Kanakee Street, Wilmington, Illinois 60481. A copy of Facility Operating License NPF-75 may be obtained upon request addressed to the U.S. Nuclear Regulatory Comission, Washington, DC 20555, Attention: Director, Division of Reactor Projects III/IV//.
Copies of the Safety Evaluation Report and Supplements 1 through 5 (NUREG-1002) and r
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. tha Final Environmental Statement (NUREG-1026) may be purchased at current rates from the Natio'nal Technical Information Service, Department of Commerce, 5285 Port Royal Road, Springfield, Virginia 22161, and through the NRC GP0 sales program by writing to the Superintendent of Documents, U.S. Government Printing Office, Post Office Box 37082, Washington, DC 20013-7082.
Dated at Bethesda, Maryland this 18th day of December 1987.
FOR THE NUCLEAR REGULATORY COMMISSION h b' C
Stepher'P. Sands, Project Manager Project Directorate III-2 Division of Reactor Projects - III, IV, V and Special Projects i
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Docket No.50-45E AMENDMENT TO INDEMNITY AGREEMENT NO. B.-102 AMENDMENT NO. 5 December 18, 1987 Effective
, Indemnity Agreement No. B-102, between Commonwealth Edison Company and the Nuclear Regulatory Commission, dated October 8, 1985, as amended, is hereby further amended as follows:
Item 3 of the Attachment to the indemnity agreement is deleted in its entirety and the following substituted therefore:
Iten 3 - License number or numbers SNM - 1938 (From 12:01 a.m., October 8, 1985, to 12 midnight, October 16, 1986, inclusive)
SNM - 1945 (From IP:01, July 27, 1987, to 12 midnight, December 17, 1987, inclusive)
NPF - 59 (From 12:01 a.m., October 17, 1986, to 12 midnight, May 20, 1987, inclusive)
NPF - 70 (From 12:01 a.m., May 21, 1987, to 12 midnight, July 1, 1987, inclusive)
NPF - 72 (From 12:01 a.m., July 2, 1987)
NPF - 75 (From 12:01 a.m., December 18,)1987 FOR THE U. S. NUCLEAR REGULATORY COMMISSION A$0, M mm Cecil 0. Thomas, Jr., Chief Policy Development and Technical Support Branch Program Management, Policy Development and Analysis Staff Db$k${$$ff, PD$
Accepted By COMMONWEALTH EDISUN COMPANY P'
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NUCLEAR REGULATORY COMMISS!ON o
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Docket No. 5,0-45F AMENDMENT TO INDEMNITY AGREEMENT N0. B.-102 AMENDMENT NO. 5 Effective December 18,1,9S7ndemnity Agreement No. B-102, between Comonwealth Edison Company and the Nuclaar Reaulatory Comission, dated October 8,1985, as amended, is hereby further amended as follows:
Item 3 of the Attachment to the indemnity agreement is deleted in its entirety and the following substituted therefore:
Item 3 - License number or numbers SNM - 1938 (From 12:01 a.m., October 8, 1985, to 12 midnight, October 16, 1986, inclusive)
SNM - 1945 (From 12:01, July 27, 1987, to 12 midnight, December 17, 1987, inclusive)
NPF - 59 (From 12:01 a.m., October 17, 1986, to 12 midnight, May 20, 1987, inclusive)
NPF - 70 (From 12:01 a.r.t., May 21, 1987, to 12 midnight, July 1, 1987, inclusive) 1 NPF - 72 (From 12:01 a.m., July 2, 1987) i NPF - 75 (From 12:01 a.m., December 18,)1987 FOR THE U. S. NUCLEAR REGULATORY COMMISSION 0 4 D. C h a n.
Cecil 0. Thomas, Jr., Chief Policy Development and Technical Support Branch Program Management, Policy Development and Analysis Staff Accepted By COMMONWEALTH EDISON COMPANY
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i u,l NUREG-1002 s
Supplement No. 5 a
! Safety Evaluation Report i
relatec' to the 03eration of 2
g Braic'wooc Sta':lon, Units 1 and 2 8
Docket Nos. STN 50-456 and STN 50-457 i
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O 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 was issued in October 1986; the third supplement was issued in May 1987; the fourth supplement was issued in July 1987.
This fifth supplement to NUREG-1002 is in support of the low power license for Unit 2 and provides the status of certain items that remained unresolved at the time Supplement 4 was published.
The facility is located in Reed Township, Will County, Illinois.
i Braidwood SSER 5 iii r.
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TABLE OF CONTENTS 4
Pagee 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......................................................
1-5 5
REACTOR COOLANT SYSTEM............................................
5-1 1
5.2 Integrity of Reactor Coolant Pressure Boundary...............
5-1 5.2.4 Reactor Coolant Pressure Boundary Inservice Inspection and Testing...........................................
5-1 5.2.4.5 Evaluation of Compliance With 10 CFR 50.55'a(g) for Braidwood Unit 2.........................
5-1 6
ENGINEERED SAFETY FEATURES........................................
6-1 6.4 Control Room Habitability..................................
6-1 6.6 Inservice Inspection of Class 2 and 3 Components.............
6-2 6.6.4 Evaluation of Compliance With 10 CFR 50.55a Braidwood Unit 2..........................(.g) for 6-2 9
AUXILIARY SYSTEMS.................................................
9-1 9.5 Other Auxiliary Systems......................................
9-1 l
9.5.1 Fire Protection Program...............................
9-1 APPENDICES APPENDIX A CONTINUATION OF CHRON0 LOGY OF NRC STAFF RADIOLOGICAL SAFETY REVIEW 0F BRAIDWOOD STATION, UNITS 1 AND 2 APPENDIX F NRC STAFF CONTRIBUTORS APPENDIX K PRESERVICE INSPECTION RELIEF REQUEST EVALUATION q
Braidwood SSER 5 v
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1 INTRODUCTION AND GENERAL DESCRIPTION OF FACILITY 1.1 Introduction i
In November 1933, the Nuclear Regulatcry 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 Braidwcod 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 applicant.
The first supplement to NUREG-1002 was issued in September 1986; the second supplement to NUREG-1002 was issued in October 1986; the third supplement to NUREG-1002 was issued in May 1987; the fourth supplement was issued in July 1987.
This fifth supplement to the SER provides the staff evaluation of the open items that have been resolved to date and addresses changes to the SER that resulted from the receipt of additional information from Commonwealth i
Edison Company (licensee); in addition, this supplement supports the issuance l
of the low power license for Unit 2.
Each section or appendix that follows is numbered the same as the correspond-ing 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-cessing of the application for Braidwood Units 1 and 2.
Appendix F lists principal staff members who contributed to this supplement.
Appendix K provides the staff evaluation of the licensee's request for relief from performing the Code-required volumetric examination on two welds.
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 l
Township Public Library, 201 South Kankakee Street, Wilmington, Illinois 60481.
The NRC Project Manager for Braidwood Station, Units 1 and 2, is Mr. Stephen P.
Sands.
Mr. Sands may be contacted by calling (301) 492-8298 or writing:
Stephen P. Sands Office of Nuclear Reactor Regulation Project Directorate III-2 U.S. Nuclear Regulatory Commission Washington, D.C. 20555
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1.7 Summary of Outstanding Items The current status of the outstanding items listed in the SER follows:
Braidwood SSER 5 1-1 a
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l Part A Items Status Section (1) Pump and valve operability Closed in 3.9.3.2*
Supplement 2 (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 i
(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*
I Supplement 1 (7) Procedures generation package (PGP)
Closed in 13.5.2 Supplement 2 (8) Control room human factors review Closed in 18.2*
Supplement 4' (9) Safetyparamejerdisplaysystem Closed in 18.3*
Supplement 4 (10) Control room habitability Closed in 6.4 Supplement 3 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 9.5.1 Supplement 3
- This section includes both site-specific-related information and duplicate-plant design features.
Braidwood SSER 5 1-2 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 l' (6) Electrical distribution system voltage Closed in 8.2.4*
verification Supplement 1 (7) Independence of redundant electrical safety Closed in 8.4.4 i
equipment Supplement 1 (8) RPM qualifications Closed in 12.5 Supplement 1 (9) Revision to Physical Security Plan Closed in 13.6 Supplement 1 (10) Control room human factors review Opened in 18.2*
Supplement 4 (11) Safety parameter display system Opened in 18.3*
Supplement 4 Part B Items (1) Inservice testing of pumps and valves Partially 3.9.6 closed in Supplement 2
- This section includes both site-specific-related information and duplicate-plant design features.
Braidwood SSER 5 1-3
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Part B Items (Continued)
Status Section (2) Steam generator tube surveillance Closed in 5.4.2.2 Supplement 1 (3) Chargidgpumpdeadheading Closed in 6.3.2, 7.3.2 Supplement 1 (4) Minimum containment pressure analysis for Closed in 6.2.1.5 performance c6pabilities 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 l
Supplement 1 Braidwood SSER 5 1-4
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1.9 License Conditions The current status of the license conditions follows:
Part A Items Status Section (1) Inserviceinspectionprogram Closed in 5.2.4, 6.6*
Supplement 3 (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 10.3.3*
monitoring and control program as proposed Supplement 1 by the Byron /Braidwood FSAR (6) TMI Item II.F.1:
Iodine / Particulate Closed in 11.5.2 Sampling Supplement 3 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 9.5.4.1 systems Supplement 3
- This section includes both site-specific-related information and duplicate-plant design features.
Braidwood SSER 5 1-5 O*
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5 REACTOR COOLANT SYSTEM 5.2 Integrity of Reactor Coolant pressure Boundary 5.2.4 Reactor Coolant Pressure Boundary Inservice Inspection and Testing 5.2.4.5 Evaluation of Compliance With 10 CFR 50.55a(g) for Braidwood Unit 2 This evaluation supplements conclusions in Section 5.2.4.3 of Supplement No. 2 of the Braidwood Safety Evaluation Report (SER), NUREG-1002, dated October 1986.
In Supplement No. 2 of the SER, the staff evaluated the preservice inspection program for Braidwood Unit 1 and concluded that the preservice inspection program is acceptable and in compliance with 10 CFR 50.55a(g).
By letter dated September 23, 1987, the applicant submitted the preservice inspection program for Braidwood Unit 2 and stated that the Unit 2 preservice inspection requirements are the same as those used for Unit 1.
Except for Relief Request 2NR-8, the relief requests for Unit 2 are identical to those for Unit 1, which were evaluated in Appendix K of Supplement No. 2.
Relief Request 2NR-8 applies only to Unit 2 and is evaluated in Appendix K.
All technical issues related to the Braidwood Unit 2 preservice inspection program have been addressed and the staff, therefore, concludes that it is acceptable.
By letter dated April 28, 1987, the licensee committed to submit the Braidwood Unit 2 inservice inspection (ISI) program within 12 months from the date of issuance of the first facility operating license for Braidwood Unit 2.
This program will be evaluated on the basis of 10 CFR 50.55a(g)(4) which requires that the initial 120-month in.cpection interval shall comply with requirements in the latest edition and adoenda of the ASME Code incorporated by reference in paragraph 50.55a(b).
This program will be evaluated after the applicable ASME Code edition and addenda can be determined and before the first refueling outage when inservice inspection commences.
Braidwood SSER 5 5-1
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6 ENGINEERED SAFETY FEATURES 6.4 Control Room Habitability In a letter dated October 30, 1987, the licensee proposed an interim operating plan for the control room ventilation (VC) system.
Section 6.5.1 of Supplement No. 2 (SSER 2) of the Braidwood Safety Evaluation Report (SER), NUREG-1002, dated October 1986, provided the staff evaluation and acceptance of the plan that was intended for use during the startup of Braidwood Unit 1.
This acceptance in-cluded the provision, during fuel loading and reactor system testing before initial criticality, that one train of the VC emergency makeup filter system be available, including an associated chiller system and control room air handling unit.
In a letter dated March 26, 1987, from S. C. Hunsader to H. R. Denton, the licensee provided a plan to utilize, on a temporary basis in the early summer cf 1987, the service building chilled water system in lieu of the control room chiller.
Section 6.4 of SSER 4, dated July 1987, provided the NRC evaluation and acceptance of this plan.
On the basis of its review of the licensee's pro-posal, the staff concluded that the proposed cross-tie met General Design Cri-terion (GDC) 4 of Appendix A to 10 CFR 50 as described in the Standa'rd Review
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Plan Section 9.4.1 (NUREG-0800) and was, therefore, an acceptable means for providing adequate temporary cooling to the control room envelope.
The licensee has proposed the same plan to again utilize the service building chiller system in an effort to complete the retubing work of the VC system 1
1 chillers, before the initial criticality of Braidwood Unit 2.
A temporary cross-tie of the service building and control room chilled water systems is again being considered.
The proposed configuration is the same as the one presented in SSER 4.
The evaluated finding of safety significance will remain unchanged since Braidwood Unit 1 will be in cold shutdown (Mode 5) and Unit 2 is expected to be in a condition no more than its pre-critical testing phase.
There are no changes necessary to the Braidwood Technical Specifications since Technical Specification 3/4.7.6 currently includes a note that renders it not applicable before initial criticality on Cycle 1.
The licensee requests that i
this note remain in place for Braidwood Unit 2 in order that Unit 2 pre-critical testing can proceed during the anticipated Unit 1 outage and concurrent VC t
chiller retubing work.
Additionally, Braidwood Unit 1 will be maintained in cold shutdown during the retubing effort and no positive reactivity changes or core alterations will be permitted.
On the basis of its review of the licensee's proposal, the staff concludes that the proposed cross-tie of the service building and control room chilled water systems is acceptable. The retubing of the VC chillers must be completed and satisfactorily tested before entering Mode 5 for Unit 1.
The licensee will maintain compliance with the appropriate action statement requirements of Technical Specification 3.7.6.
Braidwood SSER 5 6-1
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In SSER 3, dated May 1987, 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, Braidwood Station was required to include with its control room technical specifications:
(1) a surveillance requirement.to demonstrate, on an 18-month basis, that the control room envelope can be isolated and (2) a procedure to demonstrate, on an 18-month basis, that contr91 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 was to be completed before the fuel loading date for Braidwood Unit 2.
However, the licensee has proposed that this demonstration be deferred until the surveillance outage for Unit 1, scheduled for January 1988.
Otherwise, Unit 1 would need to shut down as per Technical Specification 3/4.7.6.
On the basis of its review of the licensee's proposal, the staff concludes that the proposed deferral is acceptable because of the low probability of the occur-rence of a chlorine release and the fact that compensatory measures are in place that would be used to mitigate the consequences of such an event.
However, the control room envelope integrity demonstration must be completed before Unit 1 l
can enter into Mode 4.
The licensee made this commitment by letter dated December 11, 1987 from S. C. Hunsader to T. E. Murley.
6.6 Inservice Inspection of Class 2 and 3 Components 6.6.4 Evaluation of Compliance With 10 CFR 50.55a(g) for Braidwood Unit 2 This evaluation supplements conclusions in Section 6.6.3 of SSER 2.
In SSER 2, the staff evaluated the preservice inspection program (PSI) for Braidwood Unit 1 and concluded that the PSI program is acceptable and in compliance with 10 CFR 50.55a(g).
By letter dated September 23, 1987, the applicant submitted the PSI l
program for Braidwood Unit 2 and stated that the Unit 2 PSI requirements are 1
the same as those used for Unit 1.
The relief requests for Class 2 components of Braidwood Unit 2 are identical to those for Braidwood Unit 1; these were evaluated in Appendix K of SSER 2.
The staff therefore concludes that the PSI program for Class 2 and 3 components for Braidwood Unit 2 is acceptable.
By letter dated April 28, 1987, the licensee committed to submit the Braidwood Unit 2 inservice inspection (ISI) program within 12 months from the date of is-suance of the first facility operating license for Braidwood Unit 2.
This pro-gram will be evaluated on the basis of 10 CFR 50.55a(g)(4) which requires that the initial 120-month inspection interval shall comply with requirements in the latest edition and addenda of the ASME Code incorporated by reference in para-graph 50.55a(b).
This program will be evaluated after the applicable ASME Code edition and addenda can be determined and before the first refueling outage when ISI commences.
Braidwood SSER 5 6-2
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9 AUXILIARY SYSTEMS 9.5 Other Auxiliary Systems 9.5.1 Fire Protection Program By letter dated June 3, 1987, the licensee submitted Amendment No. 10 to the Fire Protection Report for Byron /Braidwood Units 1 and 2.
The staff review relates to those changes that are specific to Braidwood Unit 2.
Amendment 10 provides a new Section 2.4 and Appendix AS.8 relating to the Braid-wood Unit 2 safe shutdown analysis and Appendix R deviations, respectively.
The licensee also provided three attachments to the submittal.
Attachment A is an itemized summary and explanation of all changes included in the amendment.
Appendix B compares the Braidwood Unit 2 safe shutdown analysis to the Braid-wood Unit 1 and Byron Unit 2 analyses.
Appendix C compares the Braidwood Unit 2 Appendix R deviations to the Byron Unit 2 deviations.
The staff has reviewed this submittal with respect to the fire protection pro-gram against the corresponding analysis that was previously reviewed as docu-mented in the Safety Evaluation Report and its supplements for Byron' Units 1 and 2 and Braidwood Unit 1.
The Braidwood Unit 2 submittal does not indicate any significant change from the methodology previously reviewed and accepted at the other three nuclear reactor facilities.
Further, the Braidwood Unit 2 alternate shutdown approach is identical to that of the other three nuclear reactor facilities.
On the basis of this review, the staff concludes that the same level of fire protection is being provided at Braidwood Unit 2 for the post-fire safe shut-down and alternate shutdown capability as was previously approved for Byron Units 1 and 2 and Braidwood Unit 1.
The staff concludes that the previous safety evaluation documented in the Braidwood SER remains valid.
Braidwood SSER 5 9-1 l
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V APPENDIX A CONTINUATION OF CHRONOLOGY OF NRC STAFF RADIOLOGICAL SAFETY REVIEW 0F BRAIDWOOD STATION, UNITS 1 AND 2 j
June 20, 1987 Representatives from NRC, Commonwealth Edison, and Busi-ness and Professional People for Public Interest meet in Bethesda, Maryland to provide information to assist in determination of significant hazards consideration regard-ing transfer of ownership.
(Summary issued on July 13, 1987.)
June 26, 1987 Letter to licensee advising that appropriate offsite emer-
)
gency response plan be revised to reflect provisions of Federal Emergency Management Agency (FEMA) Guidance Memorandum.
July 7, 1987 Letter from licensee concerning additional welds where Code Case N-340 is used at facility.
July 8, 1987 Letter from licensee transmitting Final Safety Analysis Report (FSAR) changes.
July 9, 1987 Letter to licensee concerning Generic Letter 87-12, loss i
of residual heat removal (RHR) while reactor coolant system (RCS) partially filled.
July 9, 1987 Letter from licensee transmitting additional information to justify FSAR changes.
July 10, 1987 Letter to licensee concerning Generic Letter 87-13, in-tegrity of requalification exams at nonpower reactors.
July 14, 1987 Letter to licensee transmitting Supplement 4 to the Safety Evaluation Report (SSER 4) (NUREG-1002) regarding operation of facility.
July 14, 1987 Letter from licensee concerning list of 11 concern areas on emergency procedures.
July 15, 1987 Letter from licensee concerning changes to be made to plant FSAR.
July 16, 1987 Letter to licensee concerning proposed additions and clarifications regarding conformance with criteria of Reg-ulatory Guides 1.52 and 1.140.
I l
Braidwood SSER 5 1
Appendix A
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July 22, 1987 Letter from licensee transmitting Technical Specifications regarding surveillance testing of emergency diesel generators.
July 22, 1987 Letter from licensee concerning information regarding five separate issues, including DOE /NRC Forms 741 and 742 and 4
review of safeguards classification information on forms.
July 23, 1987 Letter to licensee concerning amendment to Technical Speci-fication 4.8.1.1.2, diesel generator.
July 30, 1987 Letter from licensee concerning applications for amendments.
July 30, 1987 Letter from licensee submitting "Final Summary Report of Human Factors Engineering Review for Byron and Braidwood Stations SPOS."
August 4, 1987 Letter to licensee concerning Generic Letter 87-14 regard-ing operator licensing exams.
August 5, 1987 Letter from licensee transmitting Amendment 48 to FSAR.
August 6, 1987 Letter to licensee transmitting Safety Issues Management System (SIMS) printouts for review for each pla,nt.
August 7, 1987 Letter from licensee concerning proposed revisions to test in startup test program involving control rod drop measurements.
August 12, 1987 Letter from licensee transmitting additional proposed exception to FSAR, Appendix A.
August 24, 1987 Letter from licensee transmitting page 6-3 of final sum-mary report of human factors review not included in initial submittal of facilities emergency reponse report.
August 31, 1987 Letter from licensee transmitting SER (NUREG-1002, Supple-ment 4) Section 18 response, addressing detailed control room design review (DCRDR) and safety parameter display system (SPDS) items.
August 31, 1987 Letter from licensee transmitting Revisions 6 and 7 to inservice testing pump and valve programs, respectively, for Byron Station and Revisions 3 and 3a to inservice testing pump and valve programs.
September 1, 1987 Letter from licensee transmitting final report "Evalua-tion of Plant Variables for Compliance," per Regulatory Guide 1.97 and Supplement 1 to NUREG-0737.
September 8, 1987 Letter from licensee clarifying the author's withdrawal of applications for amendments to Licenses NPF-66 and NPF-72.
Braidwood SSER 5 2
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September 8, 1987 Letter from licensee forwarding August 3, 1987 safety issues management system (SIMS) update.
September 9, 1987 Letter to licensee transmitting cost analyses for operat-ing license (0L) application reviews.
September 9 1987 Letter to licensee requesting that replacement parts en diesel generator auxiliary equipment be classified Safety Category 1, but Quality Group G rather than Quality Group C acceptable.
September 9, 1987 Letter to licensee transmitting request for additional information regarding Generic Letter 83-28, Items 4.2.3 and 4.2.4 concerning Salem anticipated transient without scram (ATWS).
September 10, 1987 Letter from licensee requesting approval for deviation to schedule for submittal of startup test as delineated in Appendix B of Revision 2 to Regulatory Guide 1.68.
September 18, 1987 Letter from licensee requesting one-time exemption to utility February 19, 1986 licensed operator requalification program topical report.
September 18, 1987 Letter from licensee transmitting additional iriformation, requested per telephone conversations, on Item 1 of utility December 1, 1986 response to request for additional infor-mation about changes made in Amendment 47 to FSAR.
September 23, 1987 Letter from licensee transmitting Revision 0 to "Braidwood Unit 2 Preservice Inspection Program."
September 23, 1987 Letter from licensee transmitting updated response for Items 2.1 and 4.5.2 of Generic Letter 83-28.
September 25, 1987 Letter from licensee transmitting information assessing safe operation of pressurized-water reactors (PWRs) when reactor coolant systems (RCS) water level is below top of reactor vessel, per Generic Letter 87-12.
September 30, 1987 Letter from licensee transmitting revised emergency plan annexes for generating station emergency procedure manual.
October 6, 1987 Letter from licensee transmitting Policy 1 to MAELU Certificate M-115 and NELIA Certificate N-115.
October 7, 1987 Letter from licensee informing staff that fuel will be loaded on December 11, 1987 as scheduled.
Braidwood SSER 5 3
Appendix A A-
CN U
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October 13, 198) 1.etter from licensee informing staff that J. Gallo and P. P. Steptoe will present oral arguments in proceeding on November 21, 1987 regarding harassment issues presented on appeal and issues concerning plan construction assessment program and overinspection results compiled by Pittsburgh Testing Laboratory.
October 19, 1987 Letter to licensee transmitting safety evaluation regarding SPDS.
October 26, 1987 Letter to licensee informing it of upcoming NRC site visit on November 9-13, 1987 to examine MESAC system.
October 26, 1987 Letter to licensee advising it to schedule submittal for startup tests, per Appendix B of Regulatory Guide 1.68.
{
November 25, 1987 Letter from licensee informing staff that portions of construction activities regarding plant fire protection program in safety-related areas not expected to be com-pleted by start of fuel loading.
November 25, 1987 Letter from licensee requesting schedular relief for com-pletion of review and evaluation of five preoperational tests and completion and review of retests beyo,nd fuel loading.
November 25, 1987 Letter to licensee forwarding Amendments 12 (Byron Unit 1),
12 (Byron Unit 2), and 2 (Braidwood Unit 1) to Licenses NPF-37, NPF-66, and NPF-72, respectively.
Amendments re-vise Technical Specifications to allow one-time extention to 32 months for interval for performing 18 month instru-ment surveillance.
Braidwood SSER 5 4
Appendix A
O O
APPENDIX F s
NRC STAFF CONTRIBUTORS Name Title Review Branch
- John W. Craig Branch Chief Plant Systems Branch, DEST Richard J. Eckenrode Human Factors Engineer Human Factors Assessment Branch, DLPQ George Johnson Materials Engineer Materials Engineering Branch, DEST Dennis J. Kubicki Fire Protection Engineer Plant Systems Branch, DEST Linda L. Luther Licensing Assistant Project Directorate III-2 Rayleona F. Sanders Technical Editor Policy & Publications Management, DPS Jared S. Wermiel
[
Section Leader Plant Systems Branch, DEST
- Reflects reorganization since SER was issued.
Braidwood SSER 5 1
Appendix F
O b
APPENDIX K COMMONWEALTH EDIS0N COMPANY BRAIDWOOD GENERATING STATION - UNIT 2 DOCKET NUMBER 50-457 SAFETY EVALUATION REPORT SUPPLEMENT PRESERVICE INSPECTION RELIEF REQUEST EVALUATION Q.
RELIEF REQUEST NO. 2NR-8 (REV. 0), EXAMINATION CATEGORY B-J, ITEM NO. 89.31, PRESSURE RETAINING BRANCH CONNECTION WELDS IN CLASS 1 PIPING CODE REQUIREMENTS:
Examination Category B-J, Item B9.31 requires a sur-face and volumetric examination of the areas described in Figures IWB-2500-9 thru IWB-2500-11 for pipe branch connections greater than 2 in, nominal pipe size.
This examination includes essentially 100% of the weld length.
Code Relief Request:
Relief is requested from performing the Code required volumetric examination on the following welds:
Line Number Weld Number 2RC04AB-12" 2RC-11-05 2RC04AA-12" 25I-02-45
Reason for Request
The applicant reports that the above listed welds are 316 stainless steel
\\
weldolets.
Due to the weld geometry and the metallurgical properties of the material, ultrasonic examination of these welds is not practical.
Staff Evaluation:
This relief request is accep' le for PSI based on the following:
1.
The subject welds received radiographic u s e T -- avaminations during fabrication in accordance with.u ; Code n:x ements.
2.
The welds are subjected to a system pres rc R:,
r.
ccordance with Section XI requirements.
The staff therefore concludes that fabrication ext... e<
and Section XI surface examinatior, provide assurance of the preser /.ce structural integrity of the branch connection welds and that compliance with the specific requirements of Section XI would result in hardship or unusual difficulties without a compensating increase in the level of quality and safety.
Braidwood SSER 5 1
Appendix K I
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November 1983 -
December 1987 12 $UPPL4Wl%T AR Y %QTf $
Docket Nos. STN 50-456 and STN 50-457 13 AS$IR ACT 12M ocrfs pr 4ssf 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 14ense to operate Braidwood Station, Units 1 and 2 (Docket Hos. 50-456 and 50-457).
The first supplement to NUREG-1002 was issued in September 1986; the second supplement was issued in October 1986; the third supplement was issued in !!ay 1987; the fourth supplement was issued in July 1987.
This fifth supplement to NUREG-1002 is in support of the low-power license for Unit 2 and provides the status of certain items that remained unresolved at the time Supplement 4 was published. The facility is located in Reed Township, Will County,
- Illinois, e
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