ML20211F381
| ML20211F381 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 10/31/1986 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-1002, NUREG-1002-S02, NUREG-1002-S2, NUDOCS 8610310144 | |
| Download: ML20211F381 (131) | |
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I NUREG-1002 Supplement No. 2 Safety Evaluation Report related to the operation of Braidwood Station, Units 1 and 2 Docket Nos. 50-456 and 50-457 Commonwealth Edison Company U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation October 1986 l
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8610310144 861031 PDR ADOCK 0500 6
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F NOTICE Availability of Reference Materials Cited in NRC Publications Most documents cited in NRC publications will be available from one of the following sources:
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NUREG-1002 Supplement No. 2 i
Safety Evaluation Report related to the operation of Braidwood Station, Units 1 and 2 Docket Nos. 50-456 and 50-457 Commonwealth Edison Company U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation October 1986 p... ao, y
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.
This second supplement to NUREG-1002 reports the status of certain items that remained unresolved at the time Supplement 1 was published. The facility is located in Reed Township, Will County, Illinois.
I Braidwood SSER 2 iii
TABLE OF CONTENTS Pa!Le ABSTRACT.............................................................
iii 1
INTRODUCTION AND GENERAL DESCRIPTION OF FACILITY.................
1-1 1.1 Introduction................................................
1-1 1.7 Summary of Outstanding Items................................
1-2
- 1. 8 Confirmatory Issues.........................................
1-3 1.9 License Conditions..........................................
1-4 3
DESIGN CRITERIA FOR STRUCTURES, SYSTEMS, AND COMPONENTS..........
3-1 3.8 Design of Seismic Category I Structures.....................
3-1 3.8.3 Other Sei smic Category I Structures..................
3-1 3.9 Mechanical Systems and Components...........................
3-1 3.9.3 ASME Code Class 1, 2, and 3 Components, Component Supports, and Core Support Structures................
3-1 3.9.3.2 Pump and Valve Operability Assurance........
3-1 3.9.6 Inservice Testing of Pumps and Valves................
3-13 3.10 Seismic and Dynamic Qualification of Mechanical and Electrical Equipment........................................
3-14 3.10.1 Seismic and Dynamic Qualification...................
3-14 3.11 Environmental Qualification of Electrical Equipment Important to Safety and Safety-Related Mechanical Equipment...................................................
3-16 4
REACT0R..........................................................
4-1 4.3 Nuclear Design..............................................
4-1 4.3.2 Evaluation Findings..................................
4-1 5
REACTOR COOLANT SYSTEM...........................................
5-1 5.2 Integrity of Reactor Coolant Pressure Boundary..............
5-1 e
Braidwood SSER 1 v
TABLE OF CONTENTS (Continued)
_Page 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 5.2.4.4 Evaluation of Compliance With 10 CFR 50.55a(a)(3) for Braidwood Unit 1....
5-3 6
ENGINEERED SAFETY FEATURES.......................................
6-1 6.4 Control Room Habitability...................................
6-1 6.5 Fission Product Removal and Control System..................
6-1 6.5.1 Engineered Safety Feature Atmospheric Cleanup System...............................................
6-1 6.6 Inservice Inspection of Class 2 and 3 Components............
6-7 6.6.3 Evaluation of Compliance With 10 CFR 50.55a(g) for Braidwood Unit 1.....................................
6-7 7
INSTRUMENTATION AND CONTR0L......................................
.7-1 7.4 Systems Required for Safe Shutdown..........................
7-1 7.4.2 Specific Findings....................................
7-1 7.4.2.2 Remote Shutdown Capability Test.............
7-1 9
AUXILIARY SYSTEMS................................................
9-1 9.5 Other Auxiliary Systems.....................................
9-1
- 9. 5.1 Fire Protection Program..............................
9-1 9.5.1.1 Genera 1.....................................
9-1 9.5.1.3 Administrative Controls.....................
9-1 9.5.1.4 General Plant Guidelines....................
9-1 9.5.1.5 Fire Protection for Specific Plant Areas....
9-12 9.5.1.6 Deviations From BTP CMEB 9.5-1..............
9-13 9.5.1.7 Open Items..................................
9-14 9.5.1.8 Conclusion..................................
9-14 11 RADI0 ACTIVE WASTE MANAGEMENT.....................................
11-1 11.3 Gaseous Waste Management System............................
11-1 11.4 Solid Waste Management Systems.............................
11-1 Braidwood SSER 1 vi l
TABLE OF CONTENTS (Continued)
Page 11.4.1 System Description.................................
11-1 11.4.2 Evaluation and Findings............................
11-2 11.4.2.1 Radiation Doses..........................
11-2 11.4.2.2 Effluents................................
11-4 11.4.2.3 Accidents................................
11-6 1-11.4.2.4 Regulatory Guides........................
11-8 11.4.2.5 Operations...............................
11-9 11.4.2.6 Fire Protection..........................
11-9 11.4.2.7 Conclusions..............................
11-10 11.5 Process and Effluent Radiological Monitoring and Sampling Systems....................................................
11-10 11.5.2 Evaluation and Findings.............................
11-10 13 CONDUCT OF 0PERATIONS............................................
13-1 13.5 Plant Procedures...........................................
13-1 13.5.2 Operating and Maintenance Procedures...............
13-1 14 INITIAL TEST PR0 GRAM.............................................
14-1 15 ACCIDENT ANALYSES................................................
15-1 15.4 Radiological Consequences of Accidents.....................
15-1 15.4.1 Loss-of-Coolant Accident...........................
15-1 15.4.1.2 Post-LOCA Leakage From Engineered Safety Feature System Outside Containment..............................
15-1 15.6 Anticipated Transients Without Scram.......................
15-1 18 HUMAN FACTORS ENGINEERING........................................
18-1 18.2 Main Control Room and Remote Shutdown Pane 1................
18-1 18.3 Safety Parameter Display System............................
18-2 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 EVALUATION REPORT AND ITS SUPPLEMENTS APPENDIX K PRESERVICE INSPECTION RELIEF REQUEST EVALUATION Braidwood SSER 1 vii
TABLE OF CONTENTS (Continued)
LIST OF TABLES Table Page 11.1 Airborne Releases Resulting From Operation of the Braidwood Volume Reduction System..........................................
11-12 11.2 Activity Levels of Salt and Ash in Storage Hopper at Braidwood Station Based on Accident Source Terms...........................
11-12 11.3 Doses Resulting From Rupture of Storage Hopper at Braidwood Station..........................................................
11-13 15.1 Radiological Consequences of Design-Basis Accidents (Revised From SER)........................................................
15-2 15.2 Assumptions Used in.the Calculation of Loss-of-Coolant Accident Doses (Revised From SER)................................
15-3 15.6 Assumptions Used for Estimating the Radiological Consequences Following a Postulated Fuel Handling Accident (Revised From SER).............................................................
15-4 F
f i
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I Braidwood SSER 1 viii I
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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 purpose of this second supplement to the SER is to provide the staff evalua-tion of the open items that have been resolved to date and to address changes to the SER which resulted from the receipt of additional information from the applicant.
The SER stated that the staff had requested that the applicant verify that Braidwood Station meets the pertinent regulatory requirements in 10 CFR 20,50, and 100.
The applicant responded by letter dated October 1,1986, verifying this fact.
Therefore, Confirmatory Issue A(1) is considered closed.
Each of the following sections or appendices is numbered the same as the corres-ponding SER section or appendix that is being updated.
Each section is supple-mentary to and not in lieu of the discussion in the SER unless otherwise noted.
Appendix A continues the chronology of the staff's actions related to the proc-essing of the application for Braidwood Units 1 and 2.
Appendix B lists refer-ences cited in this report.* Appendix F lists principal staff members who con-tributed to this supplement.
Appendix I contains errata to the SER.
Appen-dix K contains a staff report prepared with the technical assistance of the Idaho National Engineering Laboratory evaluating the applicant's preservice inspection relief requests from certain ASME Code Section XI requirements which the applicant has determined to be impractical for systems and compo-nents at Braidwood Unit 1.
Copies of this SER supplement are available for inspection at the NRC Public Document Room, 1717 H Street, N.W., Washington, D.C., and at the Wilmington Township Public Library, 201 South Kankakee Street, Wilmington, 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) 492-7702 or writing:
Janice A. Stevens Division of PWR Licensing-A U.S. Nuclear Regulatory Commission Washington, D.C. 20555
- Availability of all material cited is described on the inside front cover of this report.
Braidwood SSER 2 1-1
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 this 3.9.3.2*
supplement (2) Seismic and dynamic qualification of Closed in this 3.10*
equipment supplement (3) Environmental qualification of electrical Closed in this 3.11*
and mechanical equipment supplement (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 this 13.5.2 supplement (8) Control room human factors review Partially 18.2*
closed in this supplement (9) Safety parameter display system Opened in this 18.3*
supplement (10) Control room habitability Opened in this 6.4 supplement Part B Items Status Section (1) Turbine missile evaluation Closed in 3.5.1.3 Supplement 1 (2) Improved thermal design procedures Closed in 4.4.1 Supplement 1 j
(3) TMI Action Item II.F.2:
Inadequate Core Closed in 4.4.7 Cooling Instrumentation Supplement 1 l
(4) Steam generator flow-induced vibrations Closed in 5.4.2 l
Supplement 1
- This section includes both site-specific-related information and duplicate plant design features.
Braidwood SSER 2 1-2
Part B Items (Continued)
Status Section (5) Conformance of ESF filter system to RG 1.52 Closed in this 6.5.1 supplement (6) Fire protection program Partially 9.5.1 closed in this supplement (7) Volume reduction system Closed in this 11.1, 11.4.2 supplement
- 1. 8 Confirmatory Issues The current status of the confirmatory issues follows:
Part A Items Status Section (1) Applicant compliance with the Commission's Closed in this 1.1, 3.1*
regulations supplement (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 this 5.2.4, 6.6*
supplement (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 i
(8) RPM qualifications Closed in 12.5 Supplement 1 (9) Revision to Physical Security Plan Closed in 13.6 Supplement 1 Part B Items Status Section (1) Inservice testing of pumps and valves Partially 3.9.6 closed in this supplement (2) Steam generator tube surveillance Closed in 5.4.2.2 Supplement 1 Braidwood SSER 2 1-3
Part B Items (Continued)
Status Section (3) Charging pump deadheading Closed in 6.3.2, 7.3.2 Supplement 1 (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 this 7.4.2.2 supplement (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 this 11.4.1 supplement (12) Noble gas monitor Closed in this 11.5.2 supplement (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 this supplement (15) Evaluation of compliance with Closed in this 5.2.4.4 10 CFR 50.55a(a)(3) supplement (16) Steam generator tube failure Opened in 15.4.3 Supplement 1 Braidwood SSER 2 1-4
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1.9 License Conditions The current status of the license conditions follows:
Part A Items Status Section (1) Inservice inspection program Open 5.2.4, 6.6*
(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 Opened in this 11.5.2 Sampling supplement Part B Items Status Section (1) Masonry walls Closed in this 3.8.3 supplement (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 Opened in 9.5.4.1 systems Supplement 1
- This section includes both site-specific-related information and duplicate plant design features.
Braidwood SSER 2 1-5
3 DESIGN CRITERIA FOR STRUCTURES, SYSTEMS, AND COMPONENTS 3.8 Design of Seismic Category I Structures 3.8.3 Other Seismic Category I Structures License Condition B(1) in the SER required that all the questions pertaining to the analysis, design, and erection of masonry walls, including any modifications resulting from the staff's review, be resolved before the beginning of the power operation after the first refueling outage.
Since that time, however, additional information has been obtained from the applicant in letters dated December 5, 1983, and July 16, 1984, which indicates that the walls have been analyzed in compliance with the NRC regulatory require-ments contained in Standard Review Plan (SRP) Section 3.8.4.
Comparison of the maximum calculated stresses to the allowable stresses specified in the SRP in-dicates that the calculated stresses are below allowable stresses.
- Further, the applicant provided a summary of results of tests performed on walls similar to those at the Braidwood plant to estimate the factor of safety against fail-The test results indicate that the average factor of safety is 5.6 for ure.
loads under operating basis earthquake (OBE) load combinations and 3.35 under the safe shutdown earthquake (SSE) load combinations.
By a letter dated August 22, 1986, the applicant has provided results of a sur-vey of masonry walls at Braidwood to detect the presence of structural cracks.
Of 607 walls surveyed, the applicant identified 27 walls that have structural cracks. The applicant's evaluation of these cracks indicates that the identi-fled cracks will not affect the structural integrity of walls as most of the cracks are local and not throughwall.
Six additional walls (of the 607 sur-veyed) were found to have cracks; these walls will be repaired by replacing the blocks that are cracked.
By letter dated October 13, 1986, the applicant stated that these walls had been repaired.
In view of the above, the staff concludes that the design of masonry walls at the Braidwood plant is conservative and complies with the staff's acceptance criteria.
The staff concludes that no additional actions are required regard-ing the masonry wall issue.
Therefore, License Condition B(1) is no longer required.
3.9 Mechanical Systems and Components 3.9.3 ASME Code Class 1, 2, and 3 Components, Component Supports, and Core Support Structures 3.9.3.2 Pump and Valve Operability Assurance To ensure that the applicant has provided an adequate program for qualifying safety-related pumps and valves to operate under normal and accident conditions, the staff performs a two-step review.
The first step is a review of FSAR Sec-tion 3.9.3.2 for the description of the applicant's pump and valve operability Braidwood SSER 2 3-1
assurance program.
This information is compared with SRP Section 3.10.
The information provided in the FSAR, however, is general in nature and not suffi-cient by itself to provide confidence in the adequacy of the applicant's over-all program for pump and valve operability qualification.
To provide this confidence, the Pump and Valve Operability Review Team (PVORT), in addition to reviewing the FSAR, conducts an onsite audit of a small representative sample of safety-related pumps and valves and supporting documentation.
The onsite audit includes a plant inspection of the as-built configuration, a discussion of the normal, accident, and postaccident conditions under which the equipment and systems must operate, and a review of the qualification documenta-tion (status reports, test reports, specifications, etc.).
The PVORT has reviewed the pump and valve operability assurance information con-tained in FSAR Section 3.9.3.2 and conducted two plant site audits to determine the extent to which the qualification of equipment, as installed at Byron Unit 1 (a duplicate of Braidwood Units 1 and 2), meets the current licensing criteria as described in SRP Section 3.10.
Conformance with these criteria provides an acceptable way of meeting the applicable portions of GDC 1, 2, 4, 14, and 30 (of Appendix A to 10 CFR 50), as well as Appendix B to 10 CFR 50.
Since Byron and Braidwood are standardized plants, the staff review of the Braidwood program was restricted to only the safety-related pumps and valves which are site specific for Braidwood.
The applicant's submittal of June 21, 1985, indicated that with the exception of one valve, the active pumps and valves at Braidwood are of the same construction and installation as the corresponding active pumps and valves at Byron.
Furthermore, this submittal also indicated that the composite seismic response spectra used to qualify pumps and valves envelopes both the Byron and Braidwood Stations.
The staff review of this submittal revealed that the identified differences between Byron and Braidwood were not significant.
Therefore, pertinent portions of the Byron SER and its supplements which include details of the PVORT review and the acceptance of the applicant's program for Byron Unit 1 are also valid for Braidwood Station.
The applicant's program for Braidwood meets the previously stated licensing criteria; therefore, the applicant's program for Braidwood is acceptable to the staff and Outstanding Item A(1) is considered closed.
Those details of the PVORT review and the applicant's program for Byron Unit 1 which are applicable to Braidwood Units 1 and 2 follow.
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Background of Previous Audits Two onsite audits were performed for Byron Station, Units 1 and 2:
one during September 9 through 13, 1982, and a reaudit during November 7 through 8, 1983.
i During the audits, a walkdown was conducted to observe the as-built configura-I tion of the selected equipment. Whenever possible, the plant engineers described the features and operating procedures unique to the equipment.
A representative sample of three pumps and seven valves was initially chosen for the September 1982 review.
However, two items were dropped because of time constraints.
The November 1983 reaudit selected one pump and one valve.
l Braidwood SSER 2 3-2 l
During the first PV0RT audit (September 1982), a number of generic and specific concerns were raised which the applicant did not satisfactorily resolve.
This was particularly evident in the three " surprise items," those items for which the applicant only had a few days to prepare.
These generic and specific con-cerns and their current status are evaluated here under " Discussion" and
" Specific Concerns."
The November 1983 reaudit, conducted at the Byron site, was a followup review of the applicant's pump and valve operability program, and was initiated pri-marily to evaluate document retrieval and central file completeness.
This re-view indicated a marked improvement in the areas of generic concerns, especially in the traceability of documentation, its approval stamps / signatures, and its timely retrieval.
The applicant supplied additional supporting information during July and December 1983.
Discussion During the September 1982 audit, the applicant briefly described its Generating Station Maintenance (GSM) History Program which had a reliability-related capa-bility for surfacing troublesome equipment.
The applicant also briefly described its computerized General Surveillance Program (GSP) which initiated in-service surveillance by calendar or usage-time intervals.
These programs were separate from the applicant's spare parts and supplies program.
The execution of these programs will help to satisfactorily address the staff concerns for the opera-tional status of plant equipment.
The November 1983 reaudit along with informa-tion submittals has shown that the applicant is satisfactorily addressing these programs.
These programs, and their effect on the specific equipment audited, are referred to in the " Specific Concerns" section which follows, as applicable.
For some equipment, the concern of environmental effects on operability could not be resolved. This was because the Byron /Braidwood Environmental Qualifi-cation Program was in process and the environmental qualification documents could not be made available to PV0RT at the time of the September 1982 audit.
Another concern was that the written preoperational test procedures for some of the equipment were still in process and others arrived late in the audit.
The applicant, however, demonstrated overall accountability by committing appro-priate personnel to resolve these concerns.
This accountability was verified at the reaudit by the marked improvement shown in the general area of docu-mentation retrieval, and by the applicant's information submittals received subsequent to the audits.
During the plant walkdown, some audited equipment was not completely installed, i.e., several drain pipes were disconnected and temporary pipe supports were in place.
The followup observations of the PVORT reaudit staff relative to pump and valve installations was favorable; the only discropancy was an "N" stamp omissien on an installed pump.
The applicant has demonstrated overall account-ability by committing appropriate personnel to resolve these concerns.
Some equipment was not qualified for operability by testing in the combined fluid dynamic and seismic operability conditions.
In an effort to justify this analytical qualification approach, additional information and analysis were requested by the staff for specific areas of equipment concern.
Also, several of the audited components were qualified by " similarity," by using previously Braidwood SSER 2 3-3
qualified equipment (similar in design, materials, etc.) as a basis for accept-ing the qualification of the equipment installed in the plant. The applicant has submitted additional information and analyses for further staff evaluation.
These evaluations are discussed in the " Specific Concerns" portion below.
For other audited equipment, parts were replaced before qualification testing.
The applicant was asked to submit a comparative analysis to substantiate the replace-ment of parts installed at the plants with similar parts before a qualification test program.
The applicant has satisfactorily addressed these concerns in submittals received after the audit and reaudit.
The concerns expressed above are explained in technical detail in the section that follows.
Specific Concerns Resolution of all equipment-specific items is discussed below.
(1) The Safety Injection Pump--NSSS (Item ISIO1PA) and Its Electric Motor Driver (Model--Lifeline 0-HSDP)
(a) The staff reviewed the shaft deflection analysis performed by Westing-house relative to pump shaft and rotor assembly clearances.
This analysis, which was not part of the initial analytical determinations, was acceptable.
It is determined that the applicant has satisfactor-ily addressed this concern.
(b) A substitute stator used in the pump / motor qualification was justi-fied by comparing the specific materials of the pump stator and the test unit.
These materials were similar.
Additionally, a letter was submitted (Attachment A21, Item 2, CECO July 1983, resnonse to PV0RT and SQRT) which documented the plant-specific applicability of the motor stator test.
It is determined that the applicant has satisfactorily addressed this concern.
(c) The applicant, in response to a PVORT concern, confirmed that in-service periodic testing of this pump is done in conformance with the j
requirements of Subsection IWP-3400 of Section XI of the ASME Boiler and Pressure Vessel Code, 1980 Edition, through the Winter 1980 i
Addenda.
It is determined that the applicant has satisfactorily addressed this concern.
(d) The applicant, in response to a PVORT request, has submitted docu-mentation (Environmental Qualification of Mechanical Equipment--
Byron /Braidwood Units 1 and 2) for the environmental qualification of the safety injection pump.
This equipment was selected as a repre-sentative component to demonstrate the design and qualification of critical soft parts.
This concern has been satisfactorily addressed.
It is determined that the applicant has satisfactorily addressed all of the concerns, and now has the administrative controls in place to assure the opera-bility qualification requirements for this pump.
This item is closed.
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Braidwood SSER 2 3-4 i
(2) Diaphragm Valve--Reactor Coolant Pressurizer--NSS--(Item No. 1RY8028) and Its Air Actuator (Model No. 32101)
(a) During the audit, it was noted that the static deflection seismic /
dynamic functional test plan and report had not yet been approved by Westinghouse for the applicant.
The applicant's July 1983 data sub-mittal indicated that this report had been reviewed (reference the SQRT equipment item as noted in Attachment A20, Item 3, CECO July 1983, response to PVORT and SWRT).
The SQRT staff has reviewed and accepted the functional test plan approved by Westinghouse.
This item is closed.
(b) In response to a PV0RT request for documentation supporting the usage of a substitute valve assembly in the qualification testing, the applicant has forwarded a letter from the vendor certifying the appli-cability of the qualification test report for this valve.
The valve selected represented the worst set of parameters for the particular series of valves being grouped together.
A number of valves were covered by one report:
reference SQRT, Attachment A20, Item 1 for the approved test report (July 1983, CECO submittal).
It is deter-mined that the applicant has satisfactorily addressed this concern.
(c) It was requested that valve closing times (spring-induced) be estab-lished during flow interruption testing of the valve; however, that information was not received.
The applicant stated that the valve is capable of closing within the required 10 seconds, under a 300-psig pressure test, and this pressure is far greater than maximum dynamic flow load which is estimated as 0.18 psig.
Since the dynamic pressure is so small, the staff accepts the static test as a demon-stration of the ability of the valve to close under pressure.
This issue is resolved.
(d) It was noted that although the valve appurtenances (solenoids and limit switches) did undergo exploratory vibration tests, the complete valve body did not.
In response to a staff request, the applicant stated that a static analysis was performed by the valve vendor (ITT Grinnell) to calculate the natural frequency of this valve.
The horizontal natural frequences of this valve were found to be below 33 Hz.
For this reason, the natural frequency analysis reviewed by the audit team was invalidated.
Additional vendor testing has subse-quently determined that the operability of the valve is not impaired for accelerations as high as 10 g.
Westinghouse has since reanalyzed piping systems for this and similar valves in the Byron plant using flexible valve models.
The results indicated that both valve and piping system loads are acceptable from an operability standpoint.
The staff finds operability of this valve adequate, based on the Westinghouse reanalysis and the SQRT reevaluation.
By letter dated September 26, 1984, the applicant indicated the completion of this effort.
This concern is closed.
(e) Although combined fluid dynamic and seismic operability testing was not done on this valve, the applicant has completed static deflection tests.
These tests involved aoplying loads equivalent to those Braidwood SSER 2 3-5
expected from the combined conditions at the center of gravity of the assembly extended structure. While in the deflection position, the valve was cycled to assure freedom of motion and cycling time all at the maximum differential pressure to which the valve is designed.
It was determined that these tests and vendor in-shop mechanical tests along with the analyses performed, have satisfied this concern for valve operability.
(f) In response to a request for an explanation of the valve's environ-mental (aging) program, the applicant has described an overall program developed to ensure that aging of mechanical components will not adversely affect the availability of. safety-related mechanical equip-ment.
It was determined that this response has satisfactorily addressed this concern.
All the concerns about this item were satisfactorily resolved and this item is closed.
(3) Check Valve, Safety Injection System--NSSS (It..n No.1SI89490)
After auditing the documentation for this valve and questioning the environ-mental effects it would experience, it was determined that the applicant has demonstrated and provided assurance that this valve can perform its intended function under normal accident, and postaccident operating conditions. This item is closed.
(4) Gate Valve, Main Steam Isolation--BOP (Item No. 1MS001A) and Its Air Actuator (DWG No. F-4932)
(a)
In response to a concern that the valve accumulators' internal pres-sures could exceed their proof pressure under high-temperature ambient accident conditions, the applicant referred to a design specification pressure of 5000 psi, rather than the normal pressure of 3750 psi.
Since the accumulators were purchased as an ASME Code Section VIII component, the applicant had a vendor certification (ASME "N" stamp) that the accumulators met or exceeded code testing requirements i.e.,
1.5 x design pressure or a minimum test pressure of 7500 psi. The applicant has addressed this concern satisfactorily.
(b) Specification changes were questioned for the external pressures and temperatures (ambient environment) in the turbine safety room, rela-tive to how this turbine room back pressure might affect solenoid l
l pilot valve / actuator operability.
The applicant explained that the initial specification of environmental parameters was too conserva-tive.
A specification amendment dated July 7, 1983, and received for l
staff review on December 29, 1983, indicates that emergency design conditions (maximum duration 1 minute) are:
350 F temperature, 100 psig pressure, and 100% humidity.
At the time of the audit, there l
was a predicted maximum peak pressure of 20 psig.
This peak pressure occurs in less than 1 second after the line break, and is predicated on turbine room doors and ventilation areas being forced open, there-by rapidly venting the area to about ambient pressure in 5 seconds.
The concern about the solenoid pilot valves' operability under the Braidwood SSER 2 3-6
elevated pressure of 20 psig is based on a design limitation of the air actuator.
The applicant has stated that the actuator air supply must be 59 psig above the environmental turbine rcom pressure.
Since actuator supply pressure can be as low as 80 psig, there is only a 1.0 psig margin of safety availaole should a design-basis double-ended break occur.
This small margin is inadequate, especially in light of the questionable pressure rise versus time curve, analytically
~
determined, and predicated on the turbine room doors and ventilation areas blowing open and thereby relieving pressure buildup.
The applicant stated in its submittal of September 26, 1984, that FSAR Section C.3.6 describes the area in question as being a " break exclusion zone." As a result, the scenario postulated regarding ambiert pressure on the actuator is not applicable for the valve operability issue.
This concern is satisfactorily resolved.
(c) In response to a concern about the valve's ability to close against a full flow load, the applicant has referred to vendor testing and anal-ysis.
The vendor provides assurance that the closure time require-ments are met by calculating the force necessary to close the valve against a full flow load and then testing the actuator to assure that adequate closing force was available.
Documentation of this testing is retained by the manufacturer.
The applicant has satisfactorily addressed this concern.
(d) At the audit, the PVORT was unable to verify that the actuator installed at the plant was identical to the qualified unit.
The applicant has confirmed that the actuator qualified by analysis (Model 64324-C) is identical to the model installed.
The applicant has satisfactorily addressed this concern.
(e) It was requested that the applicant provide confirmation that a vibra-tion analysis was performed for the valve, and that it was acceptable.
Additionally, an assessment of the effects of aging and its signifi-cance on valve operability was asked for.
The applicant, in its July 1983 response, confirmed that although this valve was not vibra-tionally tested, it was fully qualified by analysis, and this analysis established that the valve is not sensitive to vibration levels pre-dicted in seismic events. Also, the Byron vibration monitoring pro-gram would detect any unusually high amplitude vibration during pre-operational testing.
With respect to the aging effects on the valve, electrical components of the valve assembly were considered in the qualification program for Class 1E equipment in a harsh environment.
Aging of mechanical components is addressed by the maintenance and surveillance programs which will detect age-related degradation of mechanical components.
This program is an extension of the critical soft parts investigation and tabulation as described for the safety injection pump.
The applicant has satisfactorily addressed these concerns.
(f) The PVORT concern that a gradual loss of accumulator pressure might fail to initiate the solenoid valves' pilot operation was addressed by the applicant in the July 1983 response document.
The applicant explained that the two accumulators are independent, and that two independent failures in a safety grade system is a scenario beyond Braidwood SSER 2 3-7
the required plant design basis.
Additionally, these accumulator pressures are subject to surveillance.
The applicant has satisfac-torily addressed this concern.
It is determined that the applicant has satisfactorily addressed all concerns relative to this component.
Thus, this item is closed.
(5) Safety Relief Valve--Main Steam--BOP (Item No. 1M5013A) (Surprise Component)
(a) In response to a PVORT inquiry about the external / internal allowable valve. leakage rates, the applicant responded by citing this is not a safety concern since it is on the secondary side of the steam supply system.
No external or internal leakage data has been recorded.
(Reference Sargent and Lundy's seismic report EM0003901, Revision 0, June 1, 1976.) Additionally in a telephone conference call on August 7, 1984, the applicant stated that any leakage would lower the relief valve setpoint and would add another conservative factor in its operation.
Also any leakage would be readily noticeable, and corrective action would be taken if operation of the system is affected.
The applicant indicated that past history of operation at Commonwealth Edison Co. (CECO) plants shows that the operability of the safety / relief valve is not affected by leakage.
This item is closed.
(b) The staff requested that a formal acceptance test plan and preopera-tional test plan be written.
A preoperational Ceco test document No. 2.63.10 had indicated that a future preoperational test plan would be written.
The applicant responded in the July 1983 document stating that the Byron Preservice and Inservice Testing Program for valves included the 1MS013A safety / relief valve.
Before startup and 1
at each refueling outage, the valve setpoint will be verified in accordance with IWV-3510 of ASME Code Section XI.
The applicant has adequately responded to the preoperational test inquiry; however, the acceptance test document is actually a seismic document (Phase 3) i with valve operability and qualification determined under vibrational loadings.
l (c)
It was noted that as a consequence of a Phase 1 and 2 seismic investi-gation of this valve's operability, internal damage resulted and a valve redesign was initiated.
The qualification testing of this redesigned valve was identified as Phase 3 and the prototype valve i
tested in Phase 3 was stated to be the same as the production valve l
installed in the Byron plant, except as described in Section 4.2.1 of i
the seismic report EMD003901.
The staff asked to review the Phase 3 results in order to determine the difference between the qualifica-tion test relief valve and the plant-installed equipment.
This information, which was submitted on September 26, 1984, has satisfac-l torily resolved this issue.
(d) The omission of aging tests for this valve was questioned with respect to establishing its qualified life.
The applicant responded that this valve is all metal and does not have any critical nonmetallic parts subject to aging. Therefore, the establishment of a qualified life as a result of aging is not required for this valve.
The appli-cant has satisfactorily resolved this inquiry.
l Braidwood SSER 2 3-8 j
This valve has satisfied all the requirements for operability assurance.
This item is closed.
(6) Pump, Essential Service Water--80P (Item No. 15X01PA) and Its Electric Motor (Model HHS-DPO)
(a)
In response to a staff inquiry about the pump's critical speed and its possible proximity to the pump's operating range, the applicant has stated that the minimal critical speed is 2611 rpm (pump manufac-turer's data). When compared with the 880-rpm normal pump operating speed, it was ascertained that the range between these two conditions is adequate for safe pump operation.
The applicant has satisfactor-ily responded to this inquiry.
(b) A new PVORT form, properly filled out, has been submitted by the applicant in accordance with a staff request.
(c) Test reports / procedures for the initial checkout and operational test-ing of the pump were requested for review.
In the applicant's July 1983 response to open SQRT and PV0RT audit items, Section 9.20 of the preoperational test procedure was submitted for staff review.
This section adequately outlined the initial pump testing.
In addition, it was noted that this pump is included in the Byron Preservice Inspection Testing Program Plan (in accordance with ASME Code require-ments), which includes provisions for monitoring pump vibration, flowrate, discharge pressure, and bearing temperature.
The applicant has satisfied this concern, and has the documentation and controls in place to assure the equipment can be operated safely.
(d) A concern relative to establishing the qualified life and aging of susceptible components was addressed in the applicant's July 1983 submittal.
It was noted that the essential service water pumps are normally operating components located in a mild environment and, therefore, aging would only be due to normal operation.
Also refer-enced was the maintenance and surveillance programs established at Byron.
Preventive maintenance is an important part of this main-tenance and surveillance program.
It is composed of schedule main-tenance procedures where equipment is inspected, monitored, serviced, and replaced at required intervals to ensure no serious equipment malfunctions occur as a consequence of aging for the life of the plant (Reference SQRT Draft SER--Generic Item 6).
The applicant has satisfactorily addressed this concern.
(e) The coupling connecting the pump and motor was an area of concern, and confirmation of a seismic analysis was requested.
The applicant responded that this analysis was done as part of the qualification document (Reference 6.2 of Mcdonald Engineering Analysis Co. Report No. ME-523, July 2, 1982), which demonstrated the functional capa-bility of the pump.
The applicant has satisfactorily addressed this concern.
The staff finds that the applicant has demonstrated and provided assurance that this pump can perform its intended function under normal, accident, and post-accident operating conditions.
This item is closed.
Braidwood SSER 2 3-9
(7) Pump, Containment Spray--80P (Item No. 1CS01PA) and Its Electric Motor (Model No. VSW-1) (Surprise Component)
(a) In response to a concern relating to an in plant pump replacement that was not properly documented, the applicant has verified that all qualification documentation pertaining to this equipment change has been corrected, approved, and signed off.
The applicant has satis-factorily addressed this concern.
(b) At the audit, the PV0RT was unable to confirm that a program existed which would enable inservice pump testing results to be correlated with preoperational/ shop testing data, the purpose being to monitor the unit for possible performance degeneration.
The applicant, in its July 1983. response to the SQRT and PV0RT audit report, has sub-mitted a section of the preoperational test results for the pump with its corresponding acceptance test criteria.
Additionally, attention was drawn to the Byron preservice/ inservice testing program plan for pumps, which includes this pump.
This program has been developed in accordance with ASME Code Section XI requirements, i.e., it includes provisions for monitoring vibration, flowrate, and discharge pres-sure.
These data, when compared with the preoperational/ acceptance test parameters, will demonstrate the operability of the pump.
The applicant has satisfactorily addressed this concern.
(c) The staff was unable to confirm, at the time of the audit, whether the determination of the 40 year qualified life for this equipment fully considered the environmental and dynamic conditions it would be subject to.
The applicant responded to this concern (July 1983 sub-mittal) by referencing a pump motor environmental / seismic qualifica-tion report by Westinghouse (WCAP-8754, Revision 1, June 6, 1976) and Shop Order 77F14089, and a pump seismic qualification report by Ingersoll-Rand (EAS-TR-7801-IR, Revision 0, January 19, 1978).
In July 1983, the applicant also submitted an environmental qualifica-tion of containment spray pumps by Sargent and Lundy, dated February 25, 1983.
Since there was no previous environmental qualification, this analysis investigated various sources of information relative to nonmetallic components used in the pump, i.e., No. 5 carbon and ethylene propylene terpolymer "0" rings.
The investigation dis-closed that the anticipated radiation corrosion, temperature, humid-ity, and pressure environments would not degrade the 40 year life expectancy of the pump, as long as proper maintenance and inspections are carried out in accordance with vendors' recommended manuals and ASME Code Section XI, Division 1, Article IWP-1000 (10 CFR 50).
The information above and the applicant's maintenance and surveillance programs have satisfactorily addressed this concern.
The applicant has demonstrated and provided assurance that this pump can per-form its intended function under normal, accident, and postaccident operating conditions.
This item is closed.
(8) Butterfly Valve--Essential Service Water--80P (Item No. ISX027A) and Its Actuator (Model No. SMB-00/7.5 H1BC) l i
(a) The applicant submitted a revised PV0RT form; the omissions from the earlier submittal had been filled in, e.g., the correct valve f
Braidwood SSER 2 3-10
mounting method and maximum operating torque required.
The applicant has satisfactorily responded to this data omission.
(b) It was observed that the valve specification (F/L-2884) submitted for review did not specifically cover this 16-inch valve, although it did cover 12, 24, 36, and 48-inch valves.
The applicant's response to this omission indicated that indeed, although not obviously listed, the subject valve was covered under the F/L-2884 specification indirectly.
This valve was procured via purchase order 83068 which references data sheet D5004 Revision 1, and thereby specification F/L-2884.
This purchase order and data sheet were part of the appli-cant's response.
Therefore, the specification will not require revi-sion, and the applicant has satisfactorily responded to this concern.
4 (c) Verification was requested of a proper valve installation relative to the manufactyrer's recommendation for flow direction through the valve.
In response (July 1983), the applicant said the valve is marked with an arrow to indicate the preferred installation direc-tion, but that the installation direction is independent of flow direction, i.e., the valve will close against flow in either direc-1 tion.
The arrow indicates the preferred direction for sealing against flow.
A field check was made which verified that in its installed position, this valve will seal against flow when used to it;olate con-i tainment.
Containment isolation is the primary concern and, therefore, the applicant has satisfactorily complied with this verification request.
(d) The completeness of test documentation such as exploratory vibration and preoperational testing was not verified at the time of the audit.
i The applicant, in response to this PVORT concern (July 1983 response),
has verified that this valve's seismic qualification report had been received, reviewed, and approved.
The Limitorque valve operator had been qualified by test.
This valve is included in the Byron pre-i service and inservice test program (reference page 43 of the Jamesbury Corporation report (JCS82-02, Revision 2, February 14, 1983), sub-mitted on November 4, 1982).
The applicant has satisfactorily addressec this concern.
(e) Concerns relative to the valve torque requirement vs. Limitorque actuator torque output were raised at the audit.
The applicant, in its July 1983 response document, submitted a memorandum of a telephone conversation (June 1983) between Limitorque Corporation and Sargent 1
i and Lundy on this subject.
The valve's operating torque of 1180 ft-lb was compared with the available operator torque of 1250 ft-lb nominal and 1300 ft-lb maximum at 100% voltage requirement.
Sargent and i
Lundy found the actuator adequate for valve operation.
Concern about the small margin of safety from the nominal actuator torque (1.05%),
if the voltage requirement for the Limitorque actuator has an accept-able voltage below 100%, was addressed in the September 26, 1984, submittal.
This submittal indicated that the operator is capable of closing the valve within the specified time at maximum design differ-I ential pressure at the rated voltage of 110%.
The operator will
^
deliver full running torque, not seating torque, without damage, when the voltage drops to 75% of rated voltage.
Braidwood SSER 2 3-11 i
e (f) The qualified life of this valve became a concern during the audit.
This concern was not fully addressed, but was responded to by the applicant in its July 1983 response to SQRT and PVORT audit concerns.
The applicant stated that the valve actuator had been environmentally qualified in the generic Limitorque qualification program, but that an. environmental qualification of the valve itself is not required.
The only nonmetallic parts in the valve.are the valve seat (EPT, i
ethylene propylene terpolymer) and the valve packing (John Crane j
187-I).
I This valve is included in the inservice testing program and the con-tainment isolation valve leak rate testing program; therefore, any degradation of the valve which could affect its ability to isolate the containment will be detected by testing and surveillance.
Addf-I tionally the September 26, 1984, submittal stated that the critical soft parts on this valve (listed below) had been qualified for 40 years at the specified environmental conditions of 320'F (3 minutes) 100 psig, and 2 x 108 rads.
j Valve Seat:
EPT; shaft bearing:
fiberglass epoxy, nylon; shaft i
seal: John Crane 187 (asbestos graphite)
Ordinary maintenance and surveillance will ensure scheduled changeout of parts that may degrade as a result of normal mechanics of wear.
i This valve has satisfied all requirements for operability assurance.
This item is closed.
(9) Relief Valve--RHR Pump Suction--NSSS (Item No. IRH87088)
(a) At the November 1983 plant walkdown, the PVORT noted that the original valve to be reviewed (1RH8708A) had been replaced by the IRH87088 valve which was audited.
Documentation of the acceptability of piping system accelerations and nozzle loads of the "B" valve were submitted and reviewed (reference Westinghouse memorandum MID-PUE-2059 dated i
December 13, 1983).
Also submitted at this time was a document from the residual heat removal (RHR) system design team which indicated that the stresses predicted for this valve in the seismic analysis are not adversely affected by the fact that the valve discharges into a closed system. The applicant has satisfactorily addressed these I
concerns.
I (b) In response to a staff inquiry, a revised table of " active valves" (FSAR Table 3.9-16) which deletes both the IRH8708A and B valves was i
submitted.
This was done because the applicant disclosed that these valves are not required to shut down the plant or mitigate the con-sequences of an accident.
This item is considered closed since the September 26, 1984, submittal confirmed that the valve is not a safety-related item.
I Braidwood SSER 2 3-12
(10) Pump Auxiliary Feedwater--80P (Item No. lAF0-IPB-1) and Its Diesel Driver (Model 16V-14971"V")
(a) At the November 1983 audit, the PVORT noted that the diesel drive for this pump was undergoing qualification tests at Southwest Laboratories under the auspices of the Owners Group.
The final report on the qualification testing was completed as of December 19, 1985.
(b) Although all of the pump documents reviewed bore stamps indicating review and approval by the architect engineer (Sargent and Lundy) it was noted during the plant walkdown that the pump's ASME "N" stamp could not be located.
There were "N" stamps at the nameplates of the heat exchangers associated with this pump, but none were found on the name-plate of the pump itself. Verification of the pump "N" stamp require-ment and stamping was submitted and approved by the NRC previously.
This item is considered closed.
4 This pump has satisfied all the requirements for operability assurance.
This item is closed.
3.9.6 Inservice Testing of Pumps and Valves By letter dated June 20, 1985, the applicant submitted a program for the inservice testing (IST) of pumps and valves.
The applicant also submitted information in a letter dated August 20, 1986, which will allow the review of the Braidwood Unit 1 Pump and Valve Preservice/ Inservice Testing Programs to be done in comparison with the Byron Station Unit 1 programs.
The applicant stated in this letter that the scope of the Braidwood Unit 1 IST program is identical to that of Byron Unit 1 (which is also under detailed review at this time).
The applicant's IST program is required by 10 CFR 50.55a(g) to comply with the ASME Boiler and Pressure Vessel Code,Section XI.
For Braidwood Unit 1, the applicable version of the ASME Code Section XI is the 1980 Edition through the Winter 1981 Addenda.
Pursuant to 10 CFR 50.55a(g)(5), the applicant has requested relief from certain ASME Code testing requirements for specific pumps and valves where the Code requirements are impractical within the limits of design, geometry, and system safety.
The applicant's request for relief includes an explanation and justification for the relief and a proposal for alternative test procedures.
The staff has completed a preliminary review of the Braidwood IST program.
That program includes both baseline preservice testing and periodic inservice testing.
It provides both for functional testing of components in the operat-ing state and for visual inspection to verify proper valve position.
The staff has not yet completed a detailed review of the applicant's submittal (this is Confirmatory Issue B(1)).
However, the preliminary review indicates that it is impractical within the limitations of design, geometry, and system safety for the applicant to meet certain specific requirements of the ASME Code.
Granting of interim relief from those requirements as provided by the regulation will not endanger life, property, or the common defense and security of the public and is in the public interest, giving due consideration to the burden on the applicant that could result if the requirements were imposed.
On the basis of experience at similar plants where no significant adverse health Braidwood SSER 2 3-13
and safety effects were found, the staff concludes that the requirements of 10 CFR 50.55a(g)(6)(i) are satisfied.
Therefore, pursuant to 10 CFR 50.55a(g)(6)(i), the relief that the applicant has requested from certain of the pump and valve testing requirements should be granted on an interim basis until no later than the end of the first refueling outage so that a detailed review of the justifications for each request for relief may be completed.
If the detailed review results in any request for relief being denied, the applicant will be required to comply with the appro-priate Section XI requirements as stated in 10 CFR 50.55a(g).
In addition, if the detailed review identifies any pumps or valves which are not categorized as ASME Code Class 1, 2, or 3 but which perform a safety function, those pumps and valves will be included in the IST program if they are not currently included.
3.10 Seismic and Dynamic Qualification of Mechanical and Electrical Equipment 3.10.1 Seismic and Dynamic Qualification The staff's evaluation of the applicant's program for qualification of safety-related electrical and mechanical equipment for seismic and dynamic loads con-sists of:
(1) determining the acceptability of the procedures used, standards followed, and the completeness of the program in general and (2) auditing the selected equipment items to develop the basis for the staff judgment on the completeness and adequacy of the implementation of the entire seismic and dynamic qualification program. The Seismic Qualification Review Team (SQRT) consisted of engineers from the NRC staff and the Brookhaven National Laboratory (BNL). The SQRT has reviewed the equipment dynamic qualification information contained in the pertinent Final Safety Analysis Report (FSAR) Sections 3.9.2, 3.9.3, and 3.10 and has conducted two site audits to determine the extent to which the qualification of equipment, as installed at Byron Unit 1 (a duplicate of Braidwood Units 1 and 2), meets the current licensing criteria as described in Regulatory Guides (RGs) 1.100 and 1.92, SRP Section 3.10, and Institute of Electrical and Electronics Engineers (IEEE) Standard 344-1975.
Conformance with these criteria is required to satisfy the applicable portions of GDC 1, 2, 4, 14, and 30 (Appendix A to 10 CFR 50), as well as Appendix B to 10 CFR 50 and Appendix A to 10 CFR 100.
Since Byron and Braidwood are standardized plants, the staff review of the Braidwood program was restricted to only those items of the safety-related equipment which are site specific for Braidwood.
The applicant's submittals of June 4 and September 11, 1985, provided detailed information of ten such equipment types (mostly 3/4-inch valves).
The staff's review of these submittals revealed that the identified differences between Byron and Braidwood were not significant.
Therefore, pertinent portions of the Byron SEPs and its supplements which include details of the SQRT review and the acceptance of the applicant's program for Byron Unit 1 are also valid for Braidwood Station.
The applicant's program for Braidwood meets the previously stated licensing criteria as indicated in the applicant's letters dated July 1 and August 16, 1986.
Therefore, the applicant's program for Braidwood is acceptable to the staff. Outstanding Item A(2) is considered closed.
Those details of the SQRT review and the applicant's program for Byron Unit I which are applicable to Braidwood Units 1 and 2 follow.
Braidwood SSER 2 3-14
Discussion on SQRT Review The SQRT conducted an audit at the Byron site from September 13 through September 17, 1982. At the end of the audit, the SQRT concluded that the i
extent of completion of the applicant's qualification program was insufficient for SQRT to draw any conclusions with regard to the acceptability of the seismic qualification of all the safety-related equipment.
The SQRT also informed the applicant that the review team would conduct a second audit when the program was near completion (see the April 4,1983, trip report of the first site audit).
On May 13, 1983, a meeting between the staff and the applicant was held in Bethesda, Maryland, in which the applicant provided a preliminary response to both the generic as well as equipment-specific concerns identified by the SQRT during the above site audit.
As a followup, the applicant provided its formal response in a submittal dated July 7, 1983.
The SQRT reviewed the info'rmation presented by the applicant in the May 13, 1983, meeting and in the July 7 1983, submittal and determined that the informa-ion was still insufficient for the SQRT to reach a conclusion about the adequacy of the applicant's equipment seismic qualification program.
Specifically, a number of equipment items which had not been given favorable review during the above audit were still in the process of being qualified.
The staff, therefore, advised the applicant that a second site audit was needed, i
The second site audit was conducted from November 7 through November 9, 1983.
f The purpose of the audit was twofold:
(1) to review the applicant's proposed resolution to the open items identified during the first site audit and (2) to review the overall completeness of the equipment seismic qualification program.
During this audit, the SQRT reviewed a list of 12 equipment items which were not fully qualified to the SQRT requirements at the time of the first site audit.
This list included seven balance-of plant (80P) items and five nuclear steam supply system (NSSS) items, and consisted of both mechanical and electri-cal equipment. To be assured of the readiness of equipment documents upon request, the SQRT selected two additional equipment items at the site for review.
The second site audit revealed that the applicant's program for the seismic qualification of equipment had been significantly improved since the first audit.
For the 14 items audited, the SQRT found their qualification to be acceptable although some details required clarification.
The only generic concern that remained to be resolved by the applicant was the surveillance and maintenance program for equipment located in a mild environment.
The applicant subsequently submitted a postaudit response addressing the above SQRT concerns.
Further review indicated that both the generic as well as the equipment-specific concerns stated previously had all been satisfactorily resolved by the applicant, with the exception of qualification of the auxiliary feedwater pump and drives.
The SQRT reviewed the qualification plan of this equipment during the second audit and found it to be acceptable.
Braidwood SSER 2 3-15
1 3.11 Environmental Qualification of Electrical Equipment Important to Safety and Safety-Related Mechanical Equipment Equipment which is used to perform a necessary safety function must be demonstrated to be capable of maintaining functional operability under all service conditions postulated to occur during its installed life for the time it is required to operate.
This requirement which is embodied in General Design Criteria (GDC) 1 and 4 of Appendix A to 10 CFR 50, and Sections III, XI, and XVII of Appendix B to 10 CFR 50, is applicable to equipment located inside as well as outside containment. More detailed requirements and guidance relating to the methods and procedures for demonstrating this capability for electrical equipment have been set forth in 10 CFR 50.49, " Environmental Qualification of Electric Equipment Important to Safety for. Nuclear Power Plants"; NUREG-0588, '! Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment," which supplements IEEE Standard 323; and various NRC regulatory guides and industry standards.
NUREG-0588 was issued in December 1979 to promote a more orderly and systematic implementation of equipment qualification programs by industry and to provide guidance to the NRC staff for its use in ongoing licensing reviews.
The positions contained in this report provide guidance on (1) how to establish environmental service conditions, (2) how to select methods which are con -
sidered appropriate for qualifying equipment in different areas of the plant, and (3) other areas such as margin, aging, and documentation.
In February 1980, the NRC requested certain near-term operating license (OL) applicants to review and evaluate the environmental qualification documentation for each item of safety-related electrice equipment and to identify the degree to which their qualification programs comply with the staff positions discussed in NUREG-0588.
IE Bulletin 79-01B, " Environmental Qualification of Class 1E Equipment," issued January 14, 1980, and its supplements dated February 29, September 30, and October 24, 1980, established environmental qualification requirements for operating reactors.
This bulletin and its supplements were provided to OL l
applicants for consideration in their review.
A final rule on environmental qualification of electrical equipment important i
to safety for nuclear power plants became effective on February 22, 1983.
This rule, 10 CFR 50.49, specifies the requirements to be met for demonstrating the environmental qualification of electrical equipment important to safety located in a harsh environment.
In acc:rdance with 10 CFR 50.49, electrical equipment at Braidwood Station may be qualified in accordance with the acceptance l
criteria specified in Category I of NUREG-0588.
I In order te document the degree to which its environmental qualification program complies with the NRC's environmental qualification requirements and criteria, the applicant provided equipment qualification information by letters dated June 21, 1985, and July 1, July 30, August 19, September 25, and October 2, 1986.
The purpose of this supplement is to evaluate the adequacy of the Braidwood Unit 1 environmental qualification program for electrical equipment important to safety as defined in 10 CFR 50.49 and for safety-related mechanical equipment.
Braidwood SSER 2 3-16 1
l
The scope of the staff review was limited to an evaluation of the safety-related mechanical equipment and electrical equipment important to safety at Braidwood Unit 1 that is different from equipment at Byron Unit 1 and which must function in order to mitigate the consequences of a design-basis accident, inside or outside containment, while subjected to the hostile environment associated with these accidents.
By letters dated July 1, July 30, and August 19, 1986, the applicant stated that there are no unique safety-related active mechanical components in harsh environments located at Braidwood Station.
The applicant also stated that, in order to address the requirements of the environmental qualification of Class IE electrical equipment for the Byron and Braidwood Stations, it has instituted an environmental qualification program which establishes the qualification of all Class IE electrical equipment in harsh areas.
By letter dated September 25, 1986, the applicant stated that it has prepared a master list of electrical equipment as required by 10 CFR 50.49(d), and that all equipment which is installed in Braidwood Unit 1, and which has been included on its master list as within the scope of 10 CFR 50.49, is environmentally qualified.
In addition, a maintenance and surveillance pro-gram has been implemented to maintain qualification.
The applicant did not identify any equipment for Braidwood Unit 1 that is different from equipment for Byron Unit 1.
Consequently, on the basis of the information provided by the applicant for staff review of Braidwood Unit 1, the staff has concluded that all equipment within the scope of 10 CFR 50.49 at Braidwood Unit 1 is identical to equipment within the scope of 10 CFR 50.49 at Byron Unit 1 and, therefore, is environmentally qualified.
Conclusions for Braidwood Unit 1 Since the safety related mechanical equipment and the electrical equipment im-portant to safety are identical at Braidwood Unit 1 and Byron Unit 1, the staff review of the environmental qualification program for Byron Unit 1 is applicable to Braidwood Unit 1.
This review includes consideration of the environmental effects of a main steamline break outside containment with superheated steam blowdown.
By letter dated October 15, 1986, the applicant confirmed that the open items which were identified during the review of the Byron program have been resolved.
Therefore, based on the review of the information provided by the applicant for the Braidwood Unit 1 program and the review of the Byron Unit 1 program, the staff concludes that the applicant has demonstrated compli-ance with the requirements for environmental qualification as outlined in 10 CFR 50.49; the relevant parts of GDC 1 and 4; Sections III, XI, and XVII of Appendix B to 10 CFR 50, and with the criteria as specified in NUREG-0588.
Therefore, Outstanding Item A(3) is considered closed.
Those details of the staff review of the environmental qualification program for Byron Unit 1 which are applicable to Braidwood Unit 1 follow.
10 CFR 50.49(b)(3) requires that all installed Regulatory Guide (RG) 1.97, Category 1 and 2, instrumentation located in a harsh environment be included in the equipment qualification program unless adequate justification is provided.
The applicant has stated that additional equipment which may be required to satisfy RG 1.97, or equipment currently installed to satisfy RG 1.97 but for Braidwood SSER 2 3-17
which there is dispute over the need for environmental qualification, will be environmentally qualified and subject to a maintenance and surveillance program in accordance with the schedule accepted by the staff, if so required under 10 CFR 50.49.
The staff finds this acceptable.
The Westinghouse analysis of a main steamline break (MSLB) has for years predicted that saturated steam would be expelled from the break.
- Recently, Westinghouse has determined that under a certain accident scenario (a break in a main steamline (MSL) combined with uncovery of the steam generator tubes),
superheated steam would be expelled from the break and thus result in higher environmental conditions for the area of the plant containing the break.
The resulting temperature and pressure could exceed the temperature and pressure for which safety-related equipment in the area was qualified.
By letters dated July 22 and September 10, 1986, the applicant submitted its evaluation for the Byron and Braidwood Stations of the environmental effects of a main steamline break outside containment with superheated steam blowdown.
This superheat concern was identified in IE Information Notice 84-90.
For certain MSLB accidents, the steam generator tube bundle will be progressively uncovered.
This will result in the release of superheated steam, which will raise the temperature in the safety valve rooms above that previously calculated.
Consequently, the environmental qualification of equipment located in the safety valve rooms needed to be reevaluated.
The mass and energy release data taken from Westinghouse report WCAP-10961 were used as input to the RELAP4/M006 computer code to calculate the temperature profiles in the safety valve rooms.
A thermal lag analysis was then performed to obtain component temperature response.
The applicant postulated a spectrum of 40 cases, covering break sizes from to 4.6 ft, 102% and 70% of full power, and auxiliary feedwater (AFW) 2 0.1 ft2 flow rates of minimum flow rate, 200 gpm, and 300 gpm.
The mass and energy release data were calculated using the Westinghouse computer code LOFTRAN (tabulated as " Category I" in WCAP-10961).
The LOFTRAN code was modified to account for heat transfer to the steam during steam generator tube bundle uncovery.
This modification u described in WCAP-8860 (Supplement 1), which was found acceptable by the staff in the staff's safety evaluation which was transmitted to Westinghouse by letter dated May 27, 1986.
Therefora, ths. staff finds the mass and energy release data used in the subject analysis to be acceptable.
The computer code RELAP4/M006 was used to calculate the compartment temperature profiles. The heat sinks in the valve rooms were conservatively neglected.
The applicant indentified two limiting cases:
0.2 ft2 break at 102% power with 2 break at 102% power and a constant AFW flow of minimum AFW flow and 0.3 ft 300 gpm.
Thermal lag analyses of the internals of the MSIV actuator hydraulic cylinder, the MSIV actuator pneumatic reservoir, and the NAMC0 limit switch were performed in accordance with the guidance in NUREG-0588, Appendix B.
Condensation heat transfer was modeled until the surface temperature of the component reached the saturation temperature corresponding to the pressure in the valve rooms.
Analysis results indicate that condensation heat transfer will last only a very short time and that forced convection heat transfer will occur throughout the remainder of the transient.
Forced convection heat transfer was modeled with flow velocity determined by the blowdown rate.
Braidwood SSER 2 3-18 l
By letter dated October 2, 1986, the applicant provided a report that presents an evaluation based on component surface temperature which demonstrates that the components in question are also qualified using this conservative approach.
On the basis of a review of the methodology and assumptions, computer code input, and analysis results, the staff finds acceptable the applicant's calculated valve room temperature response, as well as component temperature profiles for equipment qualification.
The staff evaluation of the applicant's environmental qualification program for Byron Unit 1 included an onsite examination of equipment, audits of qualifica-tion documentation, and a review of the applicant's submittals for completeness i
and ecceptability of systems and components, qualification methods, and acci-dent environments.
The criteria described in NUREG-0800, Section 3.11, Revi-sion 2; NUREG-0588, Category I; and 10 CFR 50.49 form the bases for the staff evaluation of the adequacy of the applicant's qualification program.
The staff audited the applicant's qualification documentation and installed electrical equipment at Byron Station on June 21-23, 1983.
The audit consisted of a review of eleven files containing information regarding the equipment qualification.
The staff's findings during the audit are discussed in detail in " Environmental Qualification Audit," below.
Completeness of Equipment Important to Safety" 10 CFR 50.49 identifies three categories of electrical equipment which are required to be qualified in accordance with the provisions of the rule:
(1) safety-related electrical equipment i.e., equipment relied upon to remain functional during design-basis events (2) non-safety-related electrical equipment whose failure under the postulated environmental conditions could prevent satisfactory accomplishment of the safety functions by the safety-related equipment (3) Regulatory Guide 1.97, Revision 2, Category 1 and 2, postaccident monitoring equipment The applicant has provided information addressing compliance with this requirement of 10 CFR 50.49.
The systems identified by the applicant for the environmental qualification program as being required to function to mitigate the consequences of design-basis accidents (DBAs) and with components located in a harsh environment were compared to Table 3.2-1 of the FSAR, " Safety Category and Quality Group Classifications for Structures and Components."
Omission of systems from the harsh environment program were adequately justified by the applicant (such as all equipment located in a mild environment).
FSAR Table 3.11-4 lists the systems identified and their Class 1E function.
To address conformance with 10 CFR 50.49(b)(2) concerning non safety-related equipment whose failure under postulated accident conditions could prevent the satisfactory accomplishment of safety functions. the applicant referred to staff reviews of the responses to IE Information Notice 79-22. " Qualification Braidwood SSER 2 3-19
of Control Systems."
In addition, the staff has reviewed and evaluated the applicant's conformance with Regulatory Guide 1.75, " Physical Independence of Electric Systems," and found it acceptable.
On the basis of this, the staff concludes that the applicant's conformance to 10 CFR 50.49(b)(2) is acceptable.
Qualification Methods (1) Electrical Equipment in a Harsh Environment Detailed procedures for qualifying safety-related electrical equipment in a harsh environment are defined in NUREG-0588.
The criteria in this staff report are also applicable to other equipment important to safety defined in 10 CFR 50.49. Type testing of equipment in a sequence consisting of preaging (thermal, radiation, and mechanical), seismic and dynamic loading, and exposure to LOCA!
HELB (loss-of-coolant accident /high-energy-line break) conditions (where applicable) is the principal method of qualification.
(2) Safety-Related Mechanical Equipment in a Harsh Environment Although there are no detailed requirements for mechanical equipment, GDC 1
" Quality Standards and Records," and GDC 4, " Environmental and Missile Design Bases," and Appendix B to 10 CFR 50, " Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants" (Section III, " Design Control," and Section XVII, " Quality Assurance Records"), contain the following requirements related to equipment qualification.
Components shall be designed to be compatible with the postulated environ-mental conditions, including those associated with loss-of-coolant accidents.
Measures shall be established for the selection and review for suitability of application of materials, parts, and equipment that are essential to safety-related functions.
Design control measures shall be established for verifying the adequacy of design.
Equipment qualification records shall be maintained and shall include the results of tests and materials analyses.
The results of the safety-related mechanical equipment qualification program have been submitted to the staff for review.
In addition, qualification documentation for three items of safety-related mechanical equipment has been submitted by the applicant and has been reviewed by the staff.
The staff review has verified that the requirements for environmental qualification of safety-related mechanical equipment have been adequately addressed.
Service Conditions NUREG-0588 defines the methods to be utilized for determining the environ =cntal
(
conditions associated with loss-of-coolant accidents or high-energy-line breaks, I
inside or outside containment.
The review and evaluation of the adequacy of l
Braidwood SSER 2 3-20
these environmental conditions are described below.
The staff has reviewed the qualification documentation to ensure that the qualification conditions envelope the conditions established by the applicant.
(1) Temperature, Pressure, and Humidity Conditions Inside Containment i
The applicant provided the LOCA/MSLB profiles used for equipment qualification.
The peak values resulting from these profiles are as follows:
Maximum Maximum Condition temperature, *F pressure, psig Humidity, %
LOCA 265'F 50 psig 100%
MSLB 330 F 38 psig 100%
The staff has reviewed these profiles and finds them acceptable for use in equipment qualification; i.e., there is reasonable assurance that the actual pressures and temperatures will not exceed these profiles anywhere within the specified environmental zone (except in the break zone).
(2) Temperature, Pressure, and Humidity Conditions Outside Containment The applicant has provided the temperature, pressure, and humidity conditions associated with high energy-line breaks outside containment.
The criteria used to define the size and location of breaks are described in FSAR Section 3.6 and in the response to Question 010.40.
The auxiliary building and steam tunnel outside containment are subject to a harsh environment following a high-energy-line break.
i (3) Submergence The maximum submergence levels have been established by the applicant for various plant areas.
Inside containment, an elevation of 382 feet 2 inches, or approximately 5 feet above the containment floor, is postulated as a result of transferring the volume of the refueling water storage tank to the containment.
Equipment which is required post-LOCA and subject to submergence is or will be qualified for this condition.
The outside containment flooding analysis is discussed in Section 3.6 of the FSAR and in the response to Question 010.47.
It has been reviewed and evaluated in Section 3.6.1 of the SER.
(4) Chemical Spray Chemical spray may be utilized during an accident for containment heat removal.
The applicant has included this parameter in the evaluations of equipment located inside containment.
(5) Aging The aging Program requirements for Byron electrical equipment are defined in Section 4, Category 1, of NUREG-0588.
The degrading influences of temperature radiation, vibration, and electrical and mechanical stresses should be consid-ered and included in the aging program.
Any justifications for excluding Braidwood SSER 2 3-21
preaging of equipment in type testing should be established on the basis of equipment design and application, or on state-of-the-art aging techniques.
A qualified life is to be established for each equipment item.
In addition to the above, a maintenance / surveillance program should be implemented to identify and prevent significant age-related degradation of electrical and mechanical equipment.
The applicant has committed to follow the recommendations in RG 1.33, Revision 2, " Quality Assurance Program Requirements (Operation),"
which endorses American National Standard ANS 3.2/ ANSI N18.1976,
" Administrative Controls and Quality Assurance for the Operational Phase of Nuclear Power Plants," in the FSAR. This standard defines the scope and content of a maintenance / surveillance program for safety-related equipment.
Provisions for preventing or detecting age-related degradation in safety grade equipment are specified and include (a) utilizing experience with similar equipment, (b) revising and updating the program as experience is gained with the equipment during the life of the plant, (c) reviewing and evaluating mal-functioning equipment and obtaining adequate replacement components, and (d) establishing surveillance tests and inspections based on reliability analyses, frequency and type of service, or age of the items as appropriate.
The applicant has stated that a maintenance / surveillance program is in effect at Byron.
(6) Radiation (Inside and Outside Containment)
The applicant has provided values for the radiation levels postulated to exist following a LOCA.
The application and methodology employed to determine these values were presented to the applicant in NUREG-0588 and NUREG-0737, "Clarifi-cation of TMI Action Plan Requirements." The staff review determined that the values to which equipment was qualified enveloped the requirements identified by the applicant.
The values specified for use in equipment qualification in the containment is an integrated dose of 2 x 108 rads.
In the auxiliary building, doses of up to rads were used in areas with recirculating fluid lines.
These values are 107 acceptable for use in the qualification equipment.
Environmental Qualification Audit The staff and EG&G Idaho personnel conducted an audit of the Byron plant quali-l fication files and installed equipment on June 21-23, 1983.
The following observations and conclusions and their subsequent resolutions were made as a result of the audit:
Not all essential equipment potentially exposed to flooding has been identified.
For example, it was observed during the plant walkdown that the junction box for a valve motor operator required to operate post-LOCA was located below the flood level in containment but had not been reviewed for submergence qua'ification.
The applicant must therefore, conduct a plant walkdown to ieentify all equipment and interfaces (junction boxes, splices, etc.) whict are below the postulated flood levels and either relocate these items or demonstrate qualification for submergence.
The applicant has subsequently informed the staff that all essential equipment have been relocated above flood level.
Braidwood SSER 2 3-22
In several of the equipment reviews, it was determined that insufficient attention had been given to the acceptance criteria for qualification tests and their applicability to plant specific requirements (see discus-sions of Marathon terminal blocks and Conax penetrations below).
The acceptance criteria for all equipment in the program should, therefore, be reviewed.
The applicant has subsequently reviewed the acceptance criteria for all equipment and has provided the status of all files in its revised submittal.
A number of discrepancies existed between the qualification summary sheets supplied with the environmental qualification submittal and the information in the plant files, most of which were satisfactorily addressed during the audit.
The summary sheets 'were furnished to the staff on June 17, 1982, and did not reflect the most recent qualification data.
The applicant has since provided ttye revised and updated submittal.
In the review of individual items of equipment during the audit, several ques-tions could not be satisfactorily resolved.
These questions, which are listed below, were addressed by the applicant before Byron Unit 1 was licensed.
Anchor / Darling main steam isolation valve--The audit team was unable to resolve questions concerning the correct values for postulated pressure and temperature during a DBA, the requir'ed operability time, and time period after an accident during which failure may not occur.
In addition, the failure modes and effects analysis and identification of valve accessories should be clarified.
Since then, the applicant has submitted the revised file for the equipment.
The staff has reviewed the file and finds it acceptable.
Marathon 1600 Series terminal blocks--The staff reviewed this item for instrumentation applications.
The acceptance criteria specified included only the ability to withstand an applied voltage and current and to not exceed a specified level of leakage current during exposure to LOCA conditions.
Insulation resistance values were specified in the design specification but were not measured during LOCA exposure.
In addition, the leakage current tests results indicated that insulation resistance although not directly measured, was probably less than the value required for instrumentation circuits.
The applicant stated that the test results can apply to control circuits only and has replaced terminal blocks with splices in this application as a result of the review Conax electrical penetrations--The qualification file did not contain results of insulation resistance measurements during exposure to LOCA conditions, as required by IEEE 317-1976.
The checklist in the file did not address this omission and accepted the existing incomplete test data as sufficient.
The applicant contacted the vendor during the audit and deter-mined that these data were available and demonstrated the acceptability of the instrumentation penetrations for this application.
The applicant has received the information from Conax and incorporated it into the equipment qualification (EQ) file.
The applicant also committed to providing information on surveillance to be used to monitor the condition of the penetrations during the life of the plant.
The applicant has siace provided the surveillance information to the staff.
The staff finds the information acceptable.
Braidwood SSER 2 3-23
Rosemount 11538 transmitter--The applicant should confirm that the transmitter will be replaced at proper intervals.
Since then, the applicant has confirmed the replacement interval, thus resolving the staff's concern.
Reliance fan motor for RCFC--The comparison of postulated chemical spray conditions vs. tested conditions should be furnished to the staff for review.
Since then, the applicant has provided the analysis to demonstrate that the test condition exceeded the postulated condition.
The staff finds the applicant's response acceptable.
Okonite and Dekorad cables--The applicant committed to provide information on surveillance techniques to be utilized for cables inside containment.
The applicant has stated that Byron cables are not susceptible to any significant age-related degradation.
The existing maintenance and surveil-lance program will identify any age-related problem. The staff finds the applicant's response acceptable.
Conclusions for Byron Unit 1 The staff has reviewed and evaluated the Byron Station program for the environ-mental qualification of electrical and mechanical equipment.
This review has been performed to ensure that the systems selected for qualification, the envi-ronmental conditions resulting from design-basis accidents, and the methods used for qualification are in compliance with applicable regulations and stan-dards. The Bryon Unit 1 license was conditioned to require the applicant to environmentally qualify all electrical equipment within the scope of 10 CFR 50.49 by November 30, 1985.
By letter dated October 15, 1986, the applicant confirmed that this license condition had been met.
Braidwood SSER 2 3-24
4 REACTOR 4.3 Nuclear Design 4.3.2 Evaluation Findings In a letter dated October 24, 1983, the applicant submitted information describ-ing the nuclear analysis methods it used in support of control rod worth measure-ments using the rod swap technique for its Zion, Byron, and Braidwood reactors.
On March 12, 1981 (letter from S. A.-Varga, NRC, to J. S. Abel, CECO), the staff approved use of the rod swap technique for control rod worth measurements
)
for Zion Station, Units 1 and 2, provided the predictions were done by
~
The rod swap technique has been used for four reload cycles on the Zion Units. The applicant is now requesting approval to its nuclear analysis methods to do the predictions for rod swap starting with the Zion i
Unit 1 Cycle 8 reload.
To support its request, the applicant has performed a rod swap benchmark study j
in which rod swap analyses were performed for the four cycles of Zion data.
The computer codes are as described in the applicants' topical report,
" Benchmark of PWR Nuclear Design Methods," NFSR-0016, which has recently been approved by the NRC.
The calculational methodology used to generate the rod swap parameters is identical to that used by Westinghouse as described in i
Section 3.2 of WCAP-9863-/, " Rod Bank Worth Measurements Utilizing Bank Exchange."
The staff has reviewed the summary of the Edison rod swap benchmark study, in the form of the percent ditferences between measured or inferred rod worth and the Edison predictions. The staff has compared these results with the results-obtained using Westinghouse predictions.
In general, the differences between i
measured or inferred and predicted rod worths were smaller for the CECO data for than for the Westinghouse data. The CECO differences measurement prediction x 100 prediction tended to be both positive and negative; the Westinghouse differences are almost totally negative.
Of the 64 cases (4 cycles x 8 banks x 2 predictors),
there were no differences greater than the design criteria of 115%.
Only two differences were larger than 10%.
Staff review showed that over the four cycles analyzed, the average difference between the measured and predicted total rod worth values was -2.3% for CECO and -5.38% for Westinghouse.
The i
average difference for individual control rod banks was -1.0% for CECO and
-5.09% for Westinghouse.
l Because the rod swap calculations and measurement technique are so intricate, previous NRC approval for rod swap use has required a boron dilution versus rod swap comparison.
The applicant has not performed a boron dilution versus rod Braidwood SSER 2 4-1
swap comparison to validate its calculational ability.
However, the benchmart study that the applicant did perform is quite extensive and the results show the applicant's ability to perform the rod swap analysis with results comparable to better than those previously approved.
On this basis,the staff approves the applicant's use of the nuclear analysis methods in support of control rod worth measurements using the rod swap technique. Use of the rod swap technique is still subject to the other conditions of the March 12, 1981, approval, namely:
(1) All banks (control and shutdown) will be measured.
(2) Procedures as outlined in WCAP-9863-A will be followed.
(3) Design Criteria, Safety Criteria,and Remedial Action as stated in a March 12, 1981, letter to the applicant and in two letters from the applicant (February 4 and March 5, 1981) will be followed.
(4) A report comparing measured and predicted rod worths will be submitted to the NRC within 45 days of completion of the rod worth tests for the first use of rod swap on a reload at each unit.
Braidwood SSER 2 4-2
5 REACTOR COOLANT SYSTEM 5.2 Integrity of Reactor Coolant Pressure Boundary 5.2.4 Reactor Coolant Pressure Boundary Inservice Inspection and Testing This section was prepared with the technical assistance of U.S. Department of Energy (DOE) contractors from the Idaho National Engineering Laboratory (INEL).
5.2.4.3 Evaluation of Compliance With 10 CFR 50.55a(g) for Braidwood Unit 1 This evaluation supplements conclusions in this section of the Braidwood Safety Evaluation Report (SER) (NUREG-1002) which addressed the definition of examina-tion requirements and the evaluation of compliance with 10 CFR 50.55a(g).
The application for the Byron and Braidwood Stations was submitted and accepted for review under the Commission's standardization policy statement using the dupli-cate plant option. Therefore, the staff considered the review of the preservice inspection (PSI) program to be a confirmatory issue based on the staff review of the Byron Unit 1 PSI program, which was determined to be acceptable, and contingent upon the applicant:
(1) demonstrating that either (a) the Byron Unit 1 and Braidwood Unit 1 PSI programs are essentially the same or (b) the Braidwood Unit 1 PSI program is different but meets the requirements of 10 CFR 50.55a(g)(3)
(2) submitting all relief requests with a supporting technical justification (3) submitting conclusions regarding the ability to examine the cast stainless steel pipe elbows Since the SER was issued, the staff has completed reviewing the following information:
(1) the FSAR through Amendment 47 dated April 1986 (2) the Interim Report on Ultrasonic Examination of Welds in Cast Stainless Steel Components at Byron and Braidwood Stations," submitted June 25, 1986 (3) the results of the staff meeting at the Braidwood plant site June 26, 1986,-
to discuss the preservice inspections of the primary coolant system's stat-ically cast stainless steel fittings (4).the applicant's August 8, 1986, submittal noting the differences between the Braidwood Unit 1 and Byron Unit 1 PSI Program Plans (5) the summary report, " Ultrasonic Examination of Cast Stainless Component Welds at Byron Unit 2 and Braidwood Unit 1," submitted September 2, 1986 Braidwood SSER 2 5-1
(6) requests for relief from the ASME Code Section XI requirements that the applicant has determined to be impractical for Braidwood Unit 1, submitted July 31 and September 2, 1986 (7) clarifications and revisions to the relief requests, along with a new re-quest for relief, received in the applicant's submittal dated September 18, 1986 Considering the construction permit issuance date of December 31, 1975, 10 CFR 50.55a(g)(3) requires that the PSI Program be developed and implemented using at least the edition and addenda of Section XI of the ASME Boiler and Pressure Vessel Code applied to the construction of the particular components.
The applicant has prepared a program based on the requirements of the ASME Code Section XI, 1977 Edition with Addenda through Summer 1978.
The use of later-referenced Code editions is acceptable as specified by 10 CFR 50.55a(g)(3).
In the August 8, 1986, subuittal, the applicant verified that the scope and re-quirements of the Braidwood Unit 1 PSI Program, as well as the augmented inspections, are identical to those of Byron Unit 1.
The Braidwood Unit 1 Code Class 1 components exempted from preservice examination per Section XI, paragraph IWB-1220, correspond directly to the Byron Unit 1 exempted components.
It was also reported that the automated ultrasonic examination of the Braidwood Unit 1 reactor pressure vessel was performed to the requirements of NRC Regulatory Guide (RG) 1.150.
On June 26, 1986, the staff met with the applicant at the Braidwood plant site for a specific demonstration to determine the effectiveness of the applicant's ultrasonic examinations on the statically cast stainless steel elbows using qualified procedures on the Byron /Braidwood calibration blocks, a 0.5-inch-deep mechanical fatigue crack in a specimen obtained from the Westinghouse Owners Group, machined notches in a pipe-to-elbow weld obtained from the cancelled Marble Hill site, and actual plant welds.
The applicant's examinations, using dual 1.0-inch-diameter, 1.0-MHz, alpha series flat-faced transducers mounted on contoured removable wedges that produce an approximately 40 to 45* refracted longitudinal wave focused slightly beyond the weld ID (interior diameter) sur-face, were completed from the statically cast fitting side of the welds.
The wrought side (pipe side) of the welds had been previously examined during PSI using conventional shear wave techniques.
On the basis of discussions and demonstrations during the meeting at the Braidwood plant site, the staff reached the following conclusions regarding the preservice ultrasonic examination of the cast stainless steel fitting welds:
(1) The examinatinn procedures meet the methodology requirements of Section XI of the ASME Code.
(2) The ultrasound penetrated the region of the weld subject to examination and produced reflections from inherent geometrical conditions in the pipe that could be interpreted.
(3) The detection of significant construction-type defects, if present, was pos-sible with the ultrasonic signal-to-noise ratios observed, Braidwood SSER 2 5-2
On the basis of the above, the staff has determined that the fitting and piping welds in the primary coolant system at the Braidwood Unit 1 plant have suffi-ciently good acoustical properties to permit a valid ultrasonic examination with state-of-the-art instrumentation.
Therefore, the staff considers the issue of the preservice ultrasonic examination of welds in the primary coolant system to be resolved.
Requests for relief from the ASME Code Section XI requirements which the appli-cant has determined to be impractical for systems and components within the reactor coolant pressure boundary at Braidwood Unit 1 were contained in submit-tals dated July 31 and September 2, 1986.
Clarifications and revisions to these relief requests, along with a new request for relief, were received in the Sep-tember 18, 1986, submittal from the applicant.
The July 31, 1986, submittal contained a comparison and cross-reference between the Braidwood Unit 1 and the Byron Unit 1 relief requests and detailed any differences between the individual items.
Of the 16 relief requests submitted, 4 are plant specific for Braidwood Unit 1, 11 are common to both plants, and 1 was deleted by the applicant.
All of these relief requests were supported by information pursuant to 10 CFR 50.55a(a)(3).
Therefore, the staff evaluated the ASME Code-required examinations which the applicant stated to be impractical and determined that the applicant has demonstrated that either (1) the proposed alternatives would provide an acceptable level of quality and safety or (2) compliance with the requirements would result in hardships or unusual difficulties without a compensating in-crease in the level of quality and safety.
On the basis of review of the appli-cant's submittals and the granting of relief from these preservice examination requirements, the staff concludes that the PSI program for Braidwood Station, Unit 1, is acceptable and in compliance with 10 CFR 50.55a(g)(3).
The detailed evaluation supporting this conclusion is provided in Appendix K to this report.
Therefore, Confirmatory Issue A(4) is considered closed.
The applicant has not submitted the initial inservice inspection program.
This program will be evaluated, as a condition to the license, based on 10 CFR 50.55a(g)(4) which requires that the initial 120-month inspection interval shall comply with the requirements in the latest edition and addenda of the ASME Code incorporated by reference in paragraph 50.55a(b) on the date 12 months before the date of issuance of the operating license.
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.
This is License Condition A(1).
5.2.4.4 Evaluation of Compliance With 10 CFR 50.55a(a)(3) for Braidwood Unit 1 Supplement 1 to the Braidwood SER (SSER 1) contained an evaluation of the appli-cant's request to waive weld repair of two flaws that exceed the preservice ex-amination acceptance standard of Section XI of the ASME Code.
One unacceptable indication was found in the upper shell-to-transition cone circumferential weld of the loop 1 steam generator.
Another unacceptable indication was identified in the upper-middle-shell-to-lower-middle-shell circumferential weld of the pressurizer.
The safety evaluation concluded that an acceptable level of quality and safety without weld repair was demonstrated subject to certain specified examination and test requirements.
Braidwood SSER 2 5-3
m By letter dated October 8, 1986, the applicant committed to examine the areas containing the flaws in accordance with the inspection interval requirements of IWB-2420 of Section XI of the ASME Code.
In addition, the applicant committed to perform the loop 1 steam generator secondary side hydrotests and leak tests at temperatures greater than 165 F and 150 F, respectively, and perform the pressurizer primary side hydrotests and leak tests at temperatures greater than 120 F.
i j
On the basis of the fracture mechanics evaluations and commitment to the examina-tion and test requirements discussed in SSER 1, the applicant has demonstrated compliance with the criteria in 10 CFR 50.55a(a)(3).
Accordingly, the pressur-izer and loop 1 steam generator may be placed into service without weld repair 1
of the two flaws that exceed the preservice acceptance standard of Section XI of the ASME Code.
Therefore, Confirmatory Issue B(15) is considered closed.
1 l
l l
Braidwood SSER 2 5-4
1 6 ENGINEERED SAFETY FEATURES 6.4 Control Room Habitability As a result of the staff's re-review of the Byron control room heating, venti-lation, and air conditioning (HVAC) system (VC), it was concluded that the postaccident radiological dose criteria of General Design Criterion (GDC) 19 of Appendix A to 10 CFR 50 was not met.
Consequently, the applicant committed to upgrading the VC recirculation system and making other improvements at both Byron and Braidwood.
The applicant submitted the calculation of the revised radiological doses in a letter dated April 1, 1986.
In a June 3, 1986, letter'to the staff, the applicant submitted an analysis which demonstrated that the chlorine detectors presently installed at Braidwood are not required.
The staff is presently reviewing this analysis.
This review will not be completed before startup.
However, because the detectors are currently installed, removal of the detectors is an issue that does not need to be resolved before operation at Braidwood.
In an August 26, 1986, letter, the applicant submitted its interim operating auxiliary building ventilation (VA) system at Braidwood.
This plan called for operation of the reactor while the nonaccessible area filter ventilation system and the fuel handling filter ventilation system are inoperable.
With these systems inoperable, the doses to the control room operator would increase.
The applicant provided a dose evaluation which showed that based upon an unfiltered inleakage rate of 25 cfm into the control room, 30-day continuous leakage at a rate of 1 gpm from the emergency core cooling system (ECCS) pumps and associated equipment, and a 50 gpm leak rate for 30 minutes starting 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the accident, GDC 19 is met if the reactor is limited to 20% of rated power operation with the VA system inoperable.
The staff has reviewed the control room design and has determined that the control room meets GDC 19 when the VA system is operable with (1) the unfiltered inleakage limited to 25 cfm, (2) ECCS pump leakage at 1 gpm, and (3) 50 gpm leakage for 30 minutes. If the VA system is inoperable, the design meets GDC 19 if the reactor power is limited to 20% of rated power.
Therefore this issue is resolved except for the question of removal of the chlorine derectors (Outstand-ing Item A (10)).
6.5 Fission Product Removal and Control System 6.5.1 Engineered Safety Feature (ESF) Atmospheric Cleanup System The SER indicated that the applicant had not included either moisture separators or heaters in the nonaccessible area exhaust filter system or the fuel handling building exhaust filter system.
In the SER-CP (NUREG-75/023), the staff took the position that relative humidity control to 70% was required for the incoming air to the engineered safety feature (ESF) filter systems.
The applicant submitted two studies.
One study Braidwood SSER 2 6-1
provided an analysis showing that moisture separators are not required in the nonaccessible area exhaust filter system; the other study provided an analysis of the relative humidity anticipated in the inlet air to the fuel handling building exhaust filter system and to the nonaccessible area exhaust filter system.
Because the staff had not completed its review of the relative humidity analy-sis at the time the SER was issued, the staff had credited the above filter systems with a removal efficiency of 90% for elemental forms of radioiodine and removal efficiencies for organic forms of methyl iodine of 50% and 70% for the nonaccessible area and fuel handling building exhaust filter systems, respec-tively.
At that time, the staff indicated in the SER that the adsorber effi-ciency for organic radioiodines may be increased for the fuel handling building filter system and the nonaccessible area exhaust filter system upon completion of the staff's review.
After the SER was issued, the staff took the position on a licensing action in-solving another plant that no air filtration unit could be credited as an ESF grade system unless the system included moisture separators.
Since the filter systems did not meet the specifications in Regulatory Guide (RG) 1.52, the charcoal adsorbers could fail to remove the amount of radioiodine assumed in the accident evaluations.
Such a failure could result in doses exceeding the criteria of 10 CFR 100.
The applicant was asked to respond to the staff's concerns about the exclusion of the moisture separators and the applicant did so in its October 4, 1984, letter.
The staff has evaluated the applicant's analysis on the relative humidity ex-pected in the inlet air to the nonaccessible area and the fuel handling building exhaust filter systems.
On the basis of this evaluation, the staff concludes that the relative humidity expected in the inlet air to the fuel handling building exhaust filter system will be greater than 70%; that to the nonaccessible area exhaust filter system would be less than 70%.
The staff concluded that entrained moisture would not be a problem if a fuel handling accident occurred in the fuel handling building; therefore, moisture separators are not requirec' for the filtration unit associated with that building.
For the nonaccessible area exhaust filter system, the staff concluded, on the basis of the review of the applicant's analysis, that there was adequate dilution of the entrained water resulting from a 50 gpm pump seal failure in one of the nonaccessible area cubicles for 30 minutes, so that moisture separators are not required for the nonaccessible area exhaust filter system either.
With these conclusions, the staff has determined that the appropriate removal efficiencies for radiciodine for the nonaccessible area exhaust filter system are 95% for both elemental and organic forms of radioiodine and that the allow-able methyl iodide penetration for the laboratory test of the charcoal is 1%.
For the fuel handling building filter exhaust system, the staff determined that the appropriate removal efficiencies for the elemental and organic forms of radioiodine would be 90% and 30%, respectively, and the allowable penetration for the methyl iodide test would be 10%.
Since the SER was issued, the applicant has amended the Byron /Braidwood FSAR on several occasions.
Some of these changes have made it necessary for the staff to review conclusions presented earlier in the SER to ensure that those conclu-sions have not been negated.
Some of the changes that the applicant has l
Braidwood SSER 2 6-2 l
l l
4
proposed involve the exceptions to RG 1.52.
The applicant's initial conform-ance to this guide was covered in Section 6.5.1 of the SER.
Some of the changes that the applicant has proposed with respect to these exceptions are as follows:
The auxiliary building and fuel handling building exhaust filter housings would not be leak tested to ANSI N509-1976 requirements because (1) the housings are at negative pressure with respect to their surroundings, (2) the housings are located in the auxiliary building general area which is a low airborne radiation area, and (3) any inleakage from the general area will not adversely affect releases.
Filter mounting frame leak tests will be performed in accordance with ANSI N510-1980.
The control room emergency makeup air system filter housings would not be tested in accordance with ANSI N509 because the housings are at negative pressure and are located within the control room boundary. Therefore, any inleakage would be from the control room environment into the housings.
No alternative testing was proposed for any of the filter housings; however, filter mounting frame leak tests would be performed in accordance with ANSI N510-1980.
The remaining ESF system's ductwork would not be leak tested in accordance with ANSI N509-1976 because the ductwork is at negative pressure with re-spect to its surroundings and any inleakage would be filtered before release.
No alternative testing for the ductwork in lieu of testing in accordance with ANSI N509 was proposed.
Previously, the staff had accepted that the flow rate through the ESF fil-ter trains would not be recorded because the fans are fixed-speed fans and the applicant has a curve of flow rate versus pressure drop.
Since Ap across all high-energy particulate air (HEPA) filters was to be recorded and because the applicant was to have a Technical Specification requirement that would verify flow rate as a function of pressure drop, this was acceptable.
In Amendment 42 to the Byron /Braidwood FSAR, the applicant indicated that now only Ap across the HEPA filters, upstream of the charcoal adsorbers, would be recorded.
The applicant has stated that airflow from the nonaccessible area exhaust fil-ters and the control room emergency makeup air filters will be continuously sensed and controlled to maintain constant airflow.
Therefore, flow rates through the filter trains will not be affected by variations in pressure drops across the filters within the train.
High and low fan differential pressure alarms are provided on the main control panel to alert the operator to high or low airflow conditions, and airflow indicators are provided on local control panels in accessible areas within the control rooms so that actual flow rates can be obtained.
The applicant did not address the recording of flow rates through the fuel handling building filtration unit.
The applicant has indicated that the Ap alarms on the upstream HEPA filters will have setpoints which will indicate a deviation of i 10% from the rated flow.
The staff has reviewed changes made to the FSAR since the SER was published, including the new and revised exceptions taken to RG 1.52, and the applicant's October 18, 1984; July 2, 1986; and August 21, 1986, letters.
With respect to these exceptions, the following comments apply.
Braidwood SSER 2 6-3
4 Although it is commonly assumed that all leakage will be inleakage in a nega-tive pressure system, outleakage can occur under some conditions even when the system is operating at its design negative pressure and particularly when the system is down.
Therefore, it is important that the filter housings be leak-tight. Any inleakage would be drawn from the particular room or cubicle in which the filtration unit is located.
i The applicant has committed to performing a mounting-frame pressure leak test.
A mounting-frame pressure leak test verifies that there are no leaks through the HEPA filter and adsorber mounting frames or through the seal between the mounting frames and the housing.
The test also verifies that there exists no bypassing of the mounting frames through electrical conduits, drains, compressed air connections, and other inadvertent leak paths.
Typical sources of leaks are weld cracks and incomplete welds.
The staff finds the performance t
of the mounting-frame pressure leak test acceptable.
The applicant has indicated that the ductwork will not be leak tested in accor-dance with ANSI N509-1976 and has not proposed any alternative testing for the ductwork, not even the less rigorous testing of ANSI /ASME N509-1980. The staff initially accepted the applicant's exception to leak testing the ductwork because all radioactivity would leak into the ductwork and would be filtered.
However, further consideration of this exception has raised other questions.
l Obviously, the additional inleakage to the ducts will result in an increase in the quantity of radioactivity released off site in the event of an accident.
However, more importantly, a greater problem exists with respect to the poten-tial degradation of the charcoal in the ESF filter trains from the unknown transmission of fumes from painting, fires, or chemical releases via the leaky ductwork.
Such a transmission could be chrcnic and plant personnel may never laboratory-test the charcoal until the scheduled refueling outage.
In the intervening months, the plant may have been operating with charcoal incapable of performing at the efficiency assumed in the staff's SER.
The staff is also concerned that the relative humidity seen by the charcoal adsorbers may be altered because of the variation in the flow rate brought about by this duct-work inleakage from the various sources to the filtration units and that the analyses presented by the applicant justifying the exclusion of electrical heaters may be invalidated.
In a meeting with the applicant, the staff was told that because of the duct routing, the direction of airflow from clean to dirty, and system operation, leakage into the accessible or nonaccessible area exhaust ductwork in the general access areas would be clean, and in relation to the overall flow rates, should be of a relatively small quantity.
In addition, inleakage into the non-accessible area ductwork, along with the exhaust from other nonaccessible cubicles, would be monitored by radiation monitors located in branch ducts at j
the inlet to the filter plenums.
This exhaust air would then be directed to charcoal adsorbers if levels exceeded monitor setpoints.
In addition, plant vent stack radiation monitors would measure and record particulate, noble gas, and iodine concentrations, and alarm conditions when they exceeded monitor setpoints.
With respect to degradation of the carbon adsorber from unknown transmission of fumes, the applicant stated that the charcoal adsorbers would not be in the airflow path unless airborne radioactive material was present in the Braidwood SSER 2 6-4
nonaccessible area exhaust in excess of monitor setpoints.
The applicant indicated that leakage into the ductwork would not result in any significant increase in transmission of chemical fumes to the adsorbers.
Because of (1) the general layout of the building, (2) the principal air movement pattern in the building (i.e., direction of airflow from general areas to cubicles),
(3) the open hatches and stairwells throughout the building, and (4) the existence of natural thermal ducts through the general area, if a release did occur, the fumes would end up on the adsorbers even if there were no inleakage.
Therefore, it has to be assumed that a release in any area, other than inside an accessible cubicle, could end up on the charcoal adsorbers if the release is not removed by other means (i.e., plateout or other absorption mechanisms).
The applicant also stated that since gross leakage will be detected through the duct inspection and testing and balancing, the remaining leakage for the type of duct construction at Braidwood will result in very insignificant leakage compared with the total system flow rate.
The applicant also indicated that the correlation of audible leaks with actual measurements of leakage has led to the conclusion that by eli'minating all audible leaks, the total leakage will be less than 1% of the system capacity.
ANSI N509-1980 requires the sealing of all audible leaks.
The applicant has stated that all ESF system ductwork will be visually inspected and audibly checked for leaks and that all audible leaks will be sealed.
Although the Technical Specifications will require that the adsorber material be tested following painting, fire, or chemical release, the staff is concerned that the transmittal of such fumes may proceed undetected through such pathways as inleakage through the ductwork.
In addition, it is likely that such fumes may proceed to the adsorbers through damper leakage.
Therefore, the applicant will post signs throughout the auxiliary building and control room envelope stating that, before any painting, the control room operator shall be contacted to determine whether the adsorbers are operating.
The signs will also state that, af ter any fire or chemical release event, the control room operator or individual cognizant of the Technical Specification surveillance requirements be contacted and a determination be made whether the integrity of the adsorber material could have been compromised.
The applicant perfarmed an analysis to determine whether inleakage into the ductwork resulted in a change in the relative humidity of the air in the various nonaccessible area cubicles and, thus, negated a previous analysis provided by the applicant which justified the exclusion of electrical heaters In this analysis, the applicant assumed 10% inleakaga into the exhaust ducts in the general area, adjusted the flow rates assumed in the moisture content calculations accordingly, and assumed that the inleakage air was at 100% RH (relative humidity).
The applicant calculated that the relative humidity would be increased from 51% to approximately 59% and that leakage rates up to 30% of the airflow rate could be tolerated and still, the exhaust air relative humidity would be below 70%.
On the basis of this analysis, the staff concluded that the inleakage associated with the ductwork would not result in the negation of the applicant's previous submittal justifying the exclusion of electrical heaters.
The staff finds acceptable the applicant's method for ensuring that the flow rate in the nonaccessible area exhaust filters systems and the control room emergency makeup air filters is maintained within i 10% of its design flow, Braidwood SSER 2 6-5
l l
provided the op alarm setpoints on the upstream HEPA filters are established to indicate a deviation in flow of 10%.
l The applicant informed the staff (July 2 and August 21, 1986, letters) that silicone sealants are utilized in ESF filter systems contrary to Regulatory Position C.S.c of RG 1.52.
In an August 26, 1986, letter, the applicant addressed the acceptability of operating Unit 1 with portions of the auxiliary building ventilation system (VA) (which includes the nonaccessible area and fuel handling building filter exhaust systems) inoperable because of construc-tion going on at Unit 2.
The staff has reviewed these submittals.
On the basis of this review, the staff has concluded that:
(1) The exceptions to RG 1.52 are acceptable.
(2) Utilization of silicone sealant is acceptable in this case because its use in the control room ventilation system (VC) involves ductwork all within the control room envelope. Were degradation of the sealant to occur, there would be only minimal consequences.
The c.cmmitment by the applicant to monitor the control room ventilation flow on a daily basis and to calibrate this monitor on an 18-month basis provides additional assurance that the system integrity can be maintained.
For the VA system, leakage into or out of the nonaccessible and fuel handling building ventilation system as a result of sealant degradation occurs before reaching the filtration unit.
The filtration unit is in the last portion of the exhaust ductwork from the auxiliary building and the booster fans for the filter units are within the filter plenum.
Consequently, the leakage would be processed before release.
The nonconformance of the ESF filter systems to RG 1.52 has been reviewed and found acceptable.
There-fore, Outstanding Item B(5) is considered closed.
(3) The interim operating plan for the VA system outlined in the August 16, 1986, letter, is acceptable.
However, the manner in which the ECCS equip -
ment leak rate will be determined must be provided to the staff along with a program for reverifying this leakage rate on a quarterly basis.
This must be provided before exceeding 20% of rated power.
The staff estimates that operation without the VA system is limited to 29% of rated power based upon a LOCA doses with ECCS pump room leakage at 1 gpm and a 50 gpm leak for 30 minutes at the end of the first day following the accident.
In an August 21, 1986, letter, the applicant committed to testing all original or replacement charcoal to the requirements of Table 5-1 of ANSI N509-1980 except that the laboratory test for methyl iodine penetration which is to be conducted at 30 C and 95% relative humidity will have an acceptance criterion of <1% penetration rather than 3%.
This is acceptable to the staff.
On September 9,1986, the applicant proposed an interim operation plan for the control room ventilation system (VC).
This plan was intended for use during i
the startup of Braidwood Unit 1.
The highlights of this plan cover operation during fuel loading and reactor system testing before initial criticality; initial criticality and during Unit 1 operation up to 5% of rated reactor power; and above 5% of rated reactor power.
i Braidwood SSER 2 6-6
I During fuel loading and reactor system testing before initial reactor criticality, one train of the VC emergency makeup filter system will be available as will an associated chiller system and control room air handling unit.
After initial criticality and up to 5% of rated reactor power, the applicant has proposed that both trains of the VC system will be operab1? and that the control room envelope will be maintained at a slightly positive pressure to allow Unit 2 cable pulling to be completed.
Above 5% reactor power level, the applicant has proposed to adequately seal the control room envelope to maintain the upper cable spreading room at a positive pressure of 0.02 inch WG and all other portions of the envelope at a positive pressure of 0.125 inch.
The staff has reviewed the applicant's proposal and finds it acceptable for operation of Braidwood Unit 1.
However, upon operation of Braidwood Unit 2, the upper cable spreading room will be required to be maintained at a positive pressure 0.125 inch WG.
6.6 Inservice Inspection of Class 2 and 3 Components This section was prepared with the technical assistance of U.S. Department of Energy (DOE) contractors from the Idaho National Engineering Laboratory.
6.6.3 Evaluation of Compliance With 13 CFR 50.55a(g) for Braidwood Unit 1 This evaluation supplements conclusions in this section of the SER which ad-dressed the definition of examination requirements and the evaluation of com-pliance with 10 CFR 50.55a(g).
The application for the Byron and Braidwood Stations was submitted and accepted for review under the Commission's standard-i ization policy statement using the duplicate plant option.
Therefore, the staff considered the review of the preservice inspection (PSI) program to be a con-firmatory issue based on the staff review of the Byron Unit 1 PSI program, which was determined to be acceptable, and contingent upon the applicant (1) demonstrating that either (a) the Byron Unit 1 and Braidwood Unit 1 PSI programs are essentially the same or (b) the Braidwood Unit 1 PSI program is different but meets the requirements of 10 CFR 50.55a(g)(3)
(2) submitting all relief requests with a supporting technical justification (3) committing to perform an augmented examination of the Code Class 2 piping welds in the residual heat removal (RHR), emergency core cooling (ECC),
and containment heat removal (CHR) systems for Braidwood Unit 1.
Since the SER was issued, the staff has completed its review of the following information:
(1) the FSAR through Amendment 47 dated April 1986 (2) the applicant's August 8, 1986, submittal noting the differences between the Braidwood Unit 1 and Byron Unit 1 PSI Program Plans (3) requests for relief from the ASME Code Section XI requirements that the applicant has determined to be impractical for Braidwood Unit 1, submitted July 31 and September 2, 1986 Braidwood SSER 2 6-7
(4) the applicant's September 18, 1986, submittal containing clarifications and revisions to the relief requests, a new relief request, and a com-mitment for an augmented examination of Class 2 welds Considering the construction permit issuance date of December 31, 1975, the regu-lation [10 CFR 50.55a(g)(3)] requires that the PSI Program be developed and imple-mented using at least the edition and addenda of Section XI of the ASME Boiler and Pressure Vessel Code applied to the construction of the particular components.
The applicant has prepared a program based on the requirements of the ASME Code Section XI, 1977 Edition with Addenda through Summer 1978.
The use of later-referenced Code editions is acceptable as specified by 10 CFR 50.55a(g)(3).
In the August 8,1986, submittal, the applicant verified that the scope and re-quirements of the Braidwood Unit 1 PSI Program, as well as the augmented inspec-tions, ere identical to those of Byron Unit 1.
The Braidwood Unit 1 Code Class 2 and Class 3 components exempted from preservice examination per Section XI, paragraph IWC-1220 and Table IWD-2500-1, correspond directly to the Byron Unit 1 exempted components.
As a result of the staf f's concern about the exemption of Class 2 lines in the RHR, ECC, and CHR systems based on the exemption criteria of IWC-1220, the applicant (in the September 18, 1986, submittal) committed to examine a random sampling of 7.5% of the large-bore (greater-than-4-inch) piping circumferential welds in the safety injection (SI), chemical and volume control (CV), and con-tainment spray (CS) systems for Braidwood Unit 1.
The applicant stated that the RHR system will not be included in the augmented examinations as the RHR system does not include any piping greater than 4-inch diameter.
The applicant also committed to perform these baseline examinations during the first refuel-ing outage and to examine the welds over the 10 year inspection interval as described in the ISI Program and track them for the life of the plant.
On the basis of this commitment, the staff considers this issue resolved.
Requests for relief from the ASME Code Section XI requirements which the appli-cant has determined to be impractical for Class 2 and Class 3 components at Braidwood Unit 1 were contained in a submittal dated July 31, 1986.
This sub-mittal also contained a comparison and cross-reference between the Braidwood Unit 1 and the Byron Unit I relief requests and detailed any differences be-tween the individual items.
The applicant received clarifications and revi-sions to these relief requests, along with a new relief request, in the sub-mittal dated September 18, 1986.
Of the 16 relief requests submitted, 4 are plant specific for Braidwood Unit 1, 11 are common to both plants, and 1 was deleted by the applicant.
All of these relief requests were supported by in-formation pursuant to 10 CFR 50.55a(g)(3); therefore, the staff evaluated the ASME Code-required examinations that the applicant determined to be impractical and determined that the applicant has demonstrated that either (1) the proposed alternatives would provide an acceptable level of quality and safety or (2) com-pliance with the requirements would result in hardships or unusual difficulties without a compensating increase in the level of quality and safety.
On the basis of review of the applicant's submittals and the granting of relief from these preservice examination requirements, the staff concludes that the PSI program for Braidwood Station, Unit 1, is acceptable and in compliance with 10 CFR 50.55a(g)(3).
The detailed evaluation supporting this conclusion is 1
l l
Braidwood SSER 2 6-8
i l
i l-provided in Appendix ~K to this report.
Therefore, Confirmatory Issue A(4) is considered closed.
The applicant has not submitted the initial inservice inspection program.
This program will be evaluated, as a condition to the license, based on 10 CFR 50.55a(g)(4) which requires that the. initial 120-month inspection interval shall comply with the requirements in the latest edition and addenda of the Code incorporated by reference in 10 CFR 50.55a(b) on the date 12 months before the date of issuance of the operating license.
This program will be evaluated i
after the applicable ASME Code edition and addenda can be determined and before the first refueling outage when inservice inspection commences. This is License Condition A(1).
2 1
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1 4
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I Braidwood SSER 2 6-9 1
7 INSTRUMENTATION AND CONTROL
- 7. 4 Systems Required for Safe Shutdown 7.4.2 Specific Findings 7.4.2.2 Remote Shutdown Capability Test In the SER, the staff required that the applicant verify the remote shutdown capability of the plant.
The applicant stated in a letter dated January 26, 1982, that the plant startup test program included a one-time demonstration of the ability to maintain the plant in a safe-shutdown condition from outside the control room following a plant trip from above 10% of rated reactor power.
This demonstration test was satisfactorily completed during the 30% power level test sequence for Byron Station, Unit 1 [see Inspection Report No. 50-454/
85024(DRS)], dated July 10, 1985.
Since the design of the remote shutdown system is identical for all four Byron /Braidwood units, the test conducted on Byron Unit 1 is sufficient to verify the adequacy of the Braidwood design.
Therefore, Confirmatory Issue B(9) is considered closed.
By letter dated June 23, 1986, the applicant proposed to eliminate this test from the startup test program for Braidwood Units 1 and 2.
The staff does not find this acceptable, as is discussed in Section 14 of this supplement, since this test is necessary to verify proper operation of the remote shutdown capa-bility of each unit.
NRC Region III staff will assure that this test is com-pleted for each of the Braidwood units.
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Braidwood SSER 2 7-1
i J
9 AUXILIARY SYSTEMS 9.5 Other Auxiliary Systems 9.5.1 Fire Protection Program J
l 9.5.1.1 General i
In the SER,' the staff stated that it had not yet audited the fire protection 2
program at Braidwood because construction had not progressed to the level j
where a visit would be meaningful.
Between August 18 and 22, 1986, the staff assisted Region III personnel in conducting a fire protection program inspec-tion at Braidwood Station, Unit 1.
During the inspection, the staff raised a number of questions and concerns about certain aspects of the fire protection program.
t In addition, in Amendment 7 to the Fire Protection Report (FPR) and by letters dated May 2 and 27, June 2, July 30, and August 4 and 14, 1986,'the applicant provided additional information of a plant-specific nature concerning the Braidwood Station, Unit 1, fire protection program.
The staff's evaluation of this information is contained in the following paragraphs.
1
- 9. 5.1. 3 Administrative Controls l
Fire Brigade and Fire Brigade Trainina In Amendment 7 to the FPR, the applicant deleted a previous commitment to have i
fire brigade members take physical examinations which show them capable of un-i restricted activity.
The staff was concerned that in the absence of such an j
examination, physically unqualified individuals could-become brigade members.
By letter dated August 4, 1986, the applicant responded to this concern by re-storing the commitment to subject fire brigade members to more strenuous exami-c i
nations.
This conforms with Section C.1.a.(5) of Branch Technical Position j
(BTP) CMES 9.5-1 and is, therefore, acceptable.
During the inspection, the staff evaluated the training of fire brigade members with regard to handling charged and flowing hose streams.
There was a concern that nozzle pressures comparable to those expected within the plant were not used in training.
However, the applicant supplied lesson plans which demon-l strated that fire brigade training, including evolutions and drills, emphasizes expected nozzle pressures and plant conditions.
The staff finds this' acceptable.
I l
9.5.1.4 General Plant Guidelines Building Desian In the SER, the staff evaluated deviations from Section C.S.a. of BTP CMEB 9.5-1 to the extent that it requires redundant shutdown-related systems i
to be separated / protected by fire-rated barriers and/or automatic fire protec-l tion systems.
In the auxiliary building, redundant divisions are separated i
j Braidwood SSER 2 9-1
between elevations by floor / ceiling assemblies which feature open stairways and hatchways.
The applicant committed to provide open stairways and certain hatch-ways with automatic sprinkler protection.
Contrary to the above, the staff observed that sprinklers had not beer. installed to protect the stairway and hatchway in Zone 11.2-0 on elevation 346 feet of the auxiliary building.
In addition, sprinklers were not installed in accordance with National Fire Pro-tection Association (NFPA) Standard 13, for the open hatchway between eleva-tions 401 feet and 426 feet in the auxiliary building.
Also, draft curtains were not installed in conjunction with the sprinkler protection per Para-graph A-4-4.8.2 of the above standard.
However, by letter dated September 22, 1986, the applicant committed to install sprinklers at the hatchway opening before exceeding 5% of rated power.
Additionally, draft curtains will be added to the hatchsay no later than 6 months after fuel load.
As an alterna-tive to draft curtains, the applicant proposed to cover the hatchway with steel plates that are caulked into place, before exceeding 5% of rated power.
The applicant also committed to institute hourly fire watches when the steel plates are removed.
The staff finds that the applicant's commitments to pro-tect the hatchway provide reasonable assurance that the effects of a fire would not propagate thorough the hatchway, and are acceptable.
This issue is, therefore, closed.
In the SER, the staff approved the use of unrated steel hatch covers in lieu of fire-rated construction in several floor / ceiling assemblies, on the basis that the steel cover would form a continum c.credmbustible barrier against the vertical propagation of smoke and hot g g However, during the insper-tion, the staf f observed that steel hatch cove Nn floors on elevations wHch are not considered " water tight," contain gr, ano
%s which under fire e.on-ditions would allow products of combustior o rise intu' N vertically an oin-ing fir areas.
By letter dated Septemb' 22, 1986, the ap "
ant comm,c.ted tar. caulk to seal gaps and openings in the steel atch covers with fire r e
before exceeding 5% of rated power.
dismodificationadequatelyp jects against the vertical spread to smok and hot gases and is, therefor, 3 eptable.
On this basis, the staff consider this issue closed.
f During the inspection, the s % f observed a number of unproy. sed steel beams that were framed into fire-'ated barriers, such CS a,b. ions 467 feet and the penetration area on..'evation 414 feet of the au' iiary building.
The ap-plicant reaffirmed that all such steel will be protected with "fireproofing" g-to an equivalent rating of the barrier.
This work was not complete at the time of the inspection, but will be done before fuel load.
On this basis, the staff considers this issue closed.
The staff requested design details concerning the steel "fireproofing" to con-firm that the in plant assemblies were consistent with the rating of fire walls and floor / ceiling assemblies.
The original design of steel protection was based on Underwriter's Laboratories (UL) tested configuration D717.
However, this assembly was subsequently de-listed because the cementitious mixture used con-tained asbestos.
The applicant is now using another mixture which has been tested to UL designs N723 and N724.
On the basis of the tests performed and the configuration of the fireproofing that has already been installed (and was observed during the inspection), the staff concludes that the steel protection conforms with Section C.S.a. of BTP CMEB 9.5-1 and is, therefore, acceptable.
Braidwood SSER 2 9-2
During the inspection, the staff expressed concern that the doors to the eleva-tor shaf t and the undersize access doors in the wall separating Zones 11.6-1 from 11.6-0 in the auxiliary building were not fire rated.
However, the appli-cant produced documentation that these doors are 1 -hour (Class B) and 3-hour (Class A) fire rated, respectively.
This conforms with Section C.S.a. of the staff's fire protection guidelines and is, therefore, acceptable.
The staff observed that when the hollow metal fire doors had been installed in masonry and concrete walls, caulk was used to seal gaps between the fire door frame and the wall.
There was concern that in the event of a fire of signifi-cant magnitude the caulking would fail.
However, the applicant supplied design details during the audit which confirm that hollow metal door jambs are filled with grout.
The staff was also concerned that gaps between fire doors and the floors exceeded the limits of NFPA Standard 80.
However, the applicant provided documentation confirming that all gaps are reviewed during inspection and are within acceptable limits.
The staff finds the above items acceptable.
Alternative Shutdown Capability The applicant's method of achieving safe shutdown during a fire relies upon cer-tain manual actions, such as operating valves.
In some circumstances, a plant operator would need to reenter the fire area.
The staff was concerned that an operator would be required to enter a hazardous environment to effect manual actions necessary for shutdown.
By letter dated August 4, 1986, the applicant responded to this concern by affirming that in no instance is it necessary for an operator to reenter an area that has experienced a fire, before at least I hour has elapsed.
Because the fire loading in these locations is limited and because the existing level of fire protection will assure early fire detection and suppression, the staff finds this response acceptable.
During the inspection, the staff expressed a related concern that shutdown pro-cedures did not take into account the possible loss of habitability in multiple
" zones" within a single fire area should the products of combustion spread through non-fire-rated walls and floor / ceiling assemblies.
The applicant re-sponded that there is only one case in which an operator would have to travel through a fire area to reach a location where manual actions are required.
For a fire on elevation 383 feet of the auxiliary building, manual action is I
required on elevation 364 feet of the same fire area.
However, the operator could reach this location via fire-rated, enclosed stairwells.
Because this one area is the only location where shutdown procedures require an operator to traverse a fire area and the operator has a protected path of travel, this issue is considered closed.
During the inspection, the staff expressed
- concern that for a fire in the con-trol room, power to electronic card readers and door locks could be interrupted, preventing operators from reaching locations necessary for achieving safe shut-down.
The applicant could not resolve this concern during the audit.
By letter dated September 16, 1986, the applicant provided information regarding the opera-bility of electronic card readers in the event of a fire in the control room.
Additionally, the applicant has committed to modify the operating procedure, before exceeding 5% of rated power, to instruct the operators to take keys with them when control room evacuation is required.
A fire in the control room would, therefore, not prevent operators from reaching locations necessary for achieving safe shutdown.
This issue is considered closed.
Braidwood SSER 2 9-3
Safe-Shutdown Capability In Amendment 7 to the FPR, the applicant identified seseral additional devia-tions from Section C.S.b. of the staff's fire protection guidelines pertaining to the separation / protection of redundant shutdown-related systems.
At elevation 364 feet of the auxiliary building, redundant charging pumps and cubicle coolers are separated by approximately 65 feet.
The area is protected by automatic fire detectors and manual fire-fighting equipment.
The staff was concerned that because one division is not completely separated from the other by a fire barrier and the area is not protected by an automatic fire suppression system, a fire in this location could result in the loss of redundant shutdown systems.
However, one division of charging pump cables will be protected by a 3-hour fire-rated cable wrap until the cables enter the charging pump cubicle, Zone 11.3D-1.
This cubicle is separated from the remaining portion of the fire area by masonry walls, and all pipe and cable penetrations there are sealed with a fire-rated material.
The doorway is protected by a steel door.
The only un-protected opening into this cubicle is a penetration for a heating, ventilation, and air conditioning (HVAC) system duct.
It is the staff's judgment that because of the construction of the charging pump cubicle, the existence of fire detectors throughout the fire area, the fire-rated cable wrap, and the distance between the cubicle and the redundant unprotected cables (located about 65 feet away),
a fire is not likely to damage both the cables in Zone 11.3-0 and the shutdown systems in Zone 11.3D-1.
Therefore, this condition represents an acceptable deviation from Section C.5.b. of BTP CMEB 9.5-1.
1 Within containment, the eight reactor coolant system hot-leg temperature instru-mentation cables have a minimum, horizontal separation of approximately 15 feet.
Intervening combustibles are present in the form of cable trays.
The staff was concerned that because of the limited spatial separation, a fire could result in damaging all hot-leg temperature instrumentation cables.
By letter dated August 4, 1986, the applicant responded to this concern by committing to pro-tect one division of cables with a 1-hour fire barrier in such a manner as to achieve the minimum separation distance required by Section C.5.b(2)(b) of BTP CMEB 9.5-1.
Combustible cable insulation will remain in the intervening space between the instrumentation cables.
Because the combustibles are widely dispersed and sources of ignition are limited, the staff does not expect a fire of significant magnitude or duration to occur.
Smoke and hot gases from a postulated fire would be dissipated and cooled through the large open areas of containment.
It is the staff's judgment that, under these conditions, a fire would, at most, cause damage to systems from one shutdown division, but would not be able to propagate horizontally and damage the redundant division before being self-extinguished or suppressed by the plant fire brigade.
Therefore, the presence of intervening combustible materials within containment is an acceptable deviation from Section C.S.b(2) of BTP CMEB 9.5-1.
During the inspection, the staff observed that redundant RHR systesis (including pumps, motors, unit coolers, and cables) located on elevation 346 feet of the auxiliary building, were not separated / protected per the requirements of Sec-tion C.S.b(2) of the staff's fire protection guidelines.
The applicant stated in Amendment 7 that the "RHR system" could be repaired within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
(Act-ually, the 72-hour time limit is the prescribed time to achieve and maintain
-Braidwood SSER 2 9-4
cold shutdown, and includes estimated times to complete any repairs.) Allow-ance for repairs is consistent with the guidelines of Section C.S.b(1) of BTP CHEB 9.5-1.
However, the applicant's repair procedures only cover repair of cables; the cable repair procedure has been reviewed and found acceptable.
The applicant apparently has no procedures for repairing other potentially vulner-able components of the RHR system.
Because of this and because the RHR system is vulnerable to fire damage, the plant may not be able to achieve and maintain 1
cold shutdown within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> as stipulated in Section C.S.b(1).
By letter dated September 22, 1986, the applicant committed to upgrade the wall between the redundant RHR pumps to one having a 1-1/2-hour fire rating.
Before exceed-ing 5% of rated power, the applicant will notify the staff if the upgrade of the wall is not completed.
On the basis of the limited combustible content, the staff concludes that the 1-1/2-hour rated wall is acceptable protection for the redundant RHR pumps. Therefore, this issue is considered closed.
The staff stated in the SER that the fire protection program, including safe-shutdown and alternative shutdown capability, for Braidwood Station, Unit 1, is the same as that for Byron Station, Unit 1 (NUREG-0876), except, as noted.
However, in Amendment 7 to the Fire Protection Report (FPR) and by letters dated May 2 and 27, June 2, July 30, and August 4 and 14, 1986, the applicant provided additional information of a plant-specific nature, concerning the Braidwood Station, Unit 1.
The applicant's safe-shutdown analysis in Amend-ment 7 to the FPR, states that systems needed for hot standby include redun-j dant trains and that one of the redundant trains would be free of fire damage I
or an alternative shutdown capability would be available, such as:
(1) the charging system, using the refueling water storage tank as its water source; (2) the auxiliary feedwater system, including the condensate storage tank, the steam generator safety valves, and the steam generator atmospheric relief valves; (3) the emergency diesel generators and essential switchgear; (4) the essential service water system, including tower fans; (5) instrumentation, including pressurizer pressure and level, reactor coolant temperature, and steam generator pressure and level indications; and (6) various support com-ponents, including essential ventilation components.
Furthermore, for cold shutdown, the applicant's safe-shutdown analysis states that at least one train of the residual heat removal system (RHRS) and component cooling water system (CCWS) would be available.
The RHRS and CCWS would be utilized for long-term decay heat removal and would provide the capability to achieve cold shutdown within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
The staff stated in the SER that the applicant performed a cable separation study as part of the safe shutdown ana-lysis to ensure that at least one train of the above equipment and essential instrumentation would be available either from the control room or from an alternate location in the event of a fire in an area which might affect these components.
In Amendment 7 to the FPR, the applicant stated that the cable separation study performed for the Byron Station is valid for Braidwood Station, Unit 1, with the exception of different cable routing.
For the Braidwood plant, routing points in the cable tray system are identified at prescribed intervals.
Each routing point is assigned the number of the fire zone in which it is located.
A computerized cable tray data base of all cables in the Braidwood plant was utilized to correlate fire zones and cable routing.
For each fire zone, a listing of safe shutdown cables was generated.
Conduit routing was manually added to the fire zone list.
Braidwood SSER 2 9-5
__ -~
The applicant has stated that adequate fire protection is provided for fire zones which contain redundant equipment or cabling.
The applicant also con-sidered associated circuits by verifying that fire-induced failures in cabling for equipment not required to achieve safe shutdown, would not adversely impact safe shutdown after a fire occurs.
The applicant has stated that adequate coor-dinated circuit protection exists to ensure the availability of power supplies necessary for safe shutdown after a fire occurs.
The applicant has stated that a systems review shows that spurious valve operations identified for Braidwood Station are the same as those identified for Byron Station, except that Braid-wood does not have valves 05X162-A, B, C, and D.
In Amendment 7 to the FPR, the applicant requested three deviations from the criteria of SRP Section 9.5-1.
These deviations, identified in Amendment 7 as c.1, c.6, and c.7, were reviewed and accepted by the staff during the Byron Sta-tion review. The applicant has stated that these deviations are the same for Braidwood Station.
On this basis, the staff finds the deviations acceptable.
During the inspection, the staf f walked through the 1 BWOA-PRI-5, Revision 1, procedures for safely shutting down the plant in the event the control room becomes inaccessible.
The applicant's procedures specify the use of electrical and pneumatic jumpers to block spurious safety injection signals which may be caused by a fire in the auxiliary electrical equipment room (AEER), which re-quires evacuation of the control room, to maintain hot standby.
These actions are identified in Appendices C and E of the PRI-5 procedures and wouM be im-plemented in the AEER shortly after reactor trip.
These actions are in contra-diction to the requirements of Appendix R, Sections III.G.1 and III.L, to the extent that, while repairs are permissible to achieve and maintain cold shutdown, they are not permissible to maintain hot standby.
Therefore, the current pro-cedures call fbr actions which are inadvisable in the event of a fire in the AEER.
By letters dated September 25 and 30, 1986, the applicant revised the PRI-5 procedures (which will prevent operators from using the jumpers to attain hot standby from the AEER) for the safety injection system.
The revised pro-cedures will call for performing the functions manually at the SI pumps, as stated in Attachment K of the PRI-5 procedures.
The staff finds this acceptable, and, therefore, this item is closed.
In Amendment 7 to the FPR, the applicant committed to complete the installation of all communication and emergency lighting systems to facilitate safe shutdown before fuel load.
The staff finds this commitment acceptable.
The applicant has also stated that the associated circuit analysis for the soleniod-operated reactor vessel head valves will be performed and submitted to the staff by the time the plant reaches 5% of rated power.
The applicant's commitment is accept-able, but the matter will remain an unresolved item pending completion of the staff's review.
During the walkthrough of the PRI-5 procedures, the staff observed that the currently written procedures do not clearly identify to the operators the instruments and controls at the fire hazards panel that are electrically iso-lated from the control room when the transfer switches are placed in the remote positions to prevent spurious actuations.
The staff's concern is over the clarity of the procedures and the adequacy of the operators' training in understanding the implications of their actions on safe shutdown.
This remains an unresolved item.
Braidwood SSER 2 9-6
Control of Combustibles Duetag the inspection, the staff observed that styrofoam plastic was being used as damming material during the installation of penetration seals.
There was some concern that this material would be left in place and would constitute a fire hazard. The applicant responded that this combustible material would be removed during the QA/QC (quality assurance / quality control) inspection of the 4
seals as part of a' standard procedure.
On the basis that all plastic material will be removed, the staff finds the applicant's rssponse acceptable.
The staff observed piping identified as " oxygen" and " extremely flammable gas" routed in the turbine and auxiliary building.
There was concern that this piped gas system may not have been designed in accordance with the staff's fire protection guidelines.
By letter dated September 22, 1986, the applicant pro-vided additional information regarding this piped gas system.
This system, which provides propane gas to burners in the rad-chem labs in the auxiliary building, is only temporary.
The applicant has committed to abandon this gas line and system in place by fuel load.
The staff finds the applicant's response acceptable, and, therefore, in compliance with the guidelines of Section C.S.d.
of BTP CMEB 9.5-1.
This issue is considered closed.
The staff observed the protection for the hydrogen seal oil unit, which includes fire detectors and an automatic fire-suppression system.
The staff was ini-tially concerned that the protection did not. include a dike around the unit to confine potential oil leakage during a fire.
However, based on the limited oil inventory, the presence of drains near the unit, the fire suppression for the seal oil unit, as well as the ceiling-level sprinkler system, the staff concludes that the lack of a dike is acceptable.
Electric Cable Construction, Cable Trays, and Cable Penetrations In Amendment 7 to the FPR, the applicant described the nature of fire seals inside conduits.
The seals will either be installed at the fire barrier or at both ends of the conduit.
The staff's concern with conduit penetrations of fire walls is that they can act as an avenue for flame and hot gases to pass from one fire area to another; the air gap between the conduits and the fire walls will be sealed at the point of penetration.
This satisfies the guide-j lines of Section C.51 to BTP CMEB 9.5-1.
The applicant's method of protecting the inside of conduits provides reasonable assurance thut products of combustion will not be conveyed from one side of the barrier to the other through the con-duit.
This represents an acceptable deviation from Section L.5.a of BTP CMEB 9.5-1.
During the inspection, the staff noted that spare conduit sleeves with screw-type metal caps were not going to be sealed.
This represents a deviation from Section C.S.a. of BTP CMEB 9.5-1.
The staff will require that spare conduit sleeves be sealed with a fire-resistant material in accordance with the above-referenced guidelines.
By letter dated September 22, 1986, the applicant com-mitted to seal spare conduits with fire-resistant material to form a 3-hour fire-rated barrier.
The 3-hour rating for this application is supported by a fire test report.
The applicant com.nitted to seal the spare conduits consist-ent with its penetration seal program.
Any seals not completed before fuel load will be monitored with hourly fire watches.
This issue is considered closed.
Braidwood SSER 2 9-7
During the review of Amendment 7 to the FPR, the staff expressed concern over the apparent lack of specificity concerning the nature of the fire barrier penetration seals.
The staff was concerned that if a non-fire-rated seal mate-rial was used, fire might propagate through the seal with resultant loss of safe-shutdown capability.
By letter dated August 4,1986, the applicant re-sponded to this concern by affirming that where walls and floor / ceiling assem-blies are relied on to prevent fire spread, all openings are protected. Where seals are installed to achieve this end, the seal material possesses a 3-hour fire rating.
On this basis, and because of the fact that the fire rating is determined on the basis of a standard fire test, the staff concludes that the penetration seals conform with Section C.S.a(3) of BTP CMEB 9.5-1.
J During the inspection, the staff requested that the applicant provide verifica-tion that the " gypsum"/"thermafiber" penetration seal was fire rated.
The applicant provided a copy of a test report (dated January 20, 1986) detailing the results of a fire test and hose stream test on the seal assembly utilized in the plant. On the basis of the test report, the staff concludes that the above-referenced penetration seal is acceptable.
Ventilation In Amendment 7 to the FPR, the applicant requested approval for a deviation from Section C.S.b of BTP CMEB 9.5-1 to the extent that auxiliary building HVAC systems are not separated / protected to ensure that one division remains free of fire damage.
The staff was concerned that if these systems were lost, the shutdown-related equipment in the auxiliary building would not function as needed.
The applicant responded by letter dated August 4, 1986, stating that shutdown-related systems in the auxiliary building do not require the above-referenced HVAC system to achieve and maintain safe-shutdown conditions.
On this basis, no deviation exists.
This issue is considered closed.
During the site audit, the staff expressed concern that the fire damper OVA-495Y on elevation 426 feet exceeded the maximum size for a single-section, fire-rated damper.
However, the applicant supplied information during the audit which confirmed that this damper is fire rated. This issue is considered closed.
Fire Detection In its review of Amendment 7 to the FPR, the staff observed that fire detection systems had not been provided in all locations where they were provided at the Byron Station.
The staff was concerned that if a fire occurred in safety-related areas or in locations which represented a significant fire exposure to safety-related systems, the fire would burn undetected until major damage resulted.
By letter dated August 4, 1986, the applicant responded to this concern.
In certain locations (such as diesel fuel oil storage rooms and the auxiliary r
feedwater pump diesel room), two different types of detectors were installed at Byron Station.
This level of detection exceeded the requirements of National Fire Protection Association (NFPA) Standards and the staff's fire protection guidelines.
The applicant evaluated the hazards in these locations and con-cluded that redundant detectors were not necessary.
The staff agrees that a single thermal detection system provides reasonable assurance of early fire detection.
The system conforms with Section C.6.1 of BTP CMEB 9.5-1 and is, therefore, acceptable.
Braidwood SSER 2 9-8
I In the remaining locations, no safety-related systems are present.
The areas are protected by an automatic sprinkler system as delineated in the FPR.
If a fire should occur, it would either be discovered by plant operators and/or the sprinkler system water flow alarm would annunciate in the control room.
The fire brigade would then be dispatched to put out the fire using manual fire-fighting equipment.
Because no safety-related systems are in these areas, safe plant shutdown could be achieved and maintained using systems which are physi-cally and electrically independent of these locations.
On this basis, the absence of fire detectors in the areas delineated in the applicant's August 4, 1986, letter is an acceptable deviation from Section C.6.a of BTP CMEB 9.5-1.
The staff noted that certain plant locations featured large floor-to-ceiling heights.
There was concern that smoke from a fire would tend to stratify at a level below that of ceiling-mounted detectors.
The applicant responded to this concern by affirming that the potential for smoke stratification had been con-sidered in the design of the smoke-detection systems.
The systems were designed in accordance with NFPA Standard 72E, 1985 Edition.
The spacing between detec-tors was reduced for high air velocity and high ceiling areas.
In addition, the smoke detectors were staggered (i.e., detectors were mounted alternately at ceiling level and at a distance below the ceiling) in high ceiling areas.
On this basis, the staff concludes that the applicant has adequately addressed smoke stratification in the design of the smoke-detection systems.
This issue is considered closed.
During the site audit, the staff observed smoke detectors installed within the cable spreading rooms.
There was concern that the installation did not conform to the requirements of NFPA 72E.
However, the applicant provided an evaluation of the fire detector layout for these areas, which demonstrated that the detec-tors, as designed, provide reasonable assurance that a fire will be promptly detected.
This conforms with the staff's fire protection guidulines and is, therefore, acceptable.
During the inspection, the staff expressed concern that an alarm condition for the heat detection and fire-suppression systems which protect charcoal filter units were not all annunciated in the control room. The applicant demonstrated that such alarms are annunciated in the control room.
This issue is considered j
closed.
The staff requested information to confirm that the fire-alarm system was sup-plied by a reliable power supply and would not be rendered inoperable if the primary source of power were lost.
The applicant responded by affirming that, except for the fire-detection / fire protection circuits for the carbon dioxide systems, all other fire-alarm system-related circuits are fed off the 1E bus.
If offsite power is lost, the emergency diesel generators would supply power to these circuits.
The fire-detection / fire protection circuits for the carbon dioxide systems are fed off the engineered safety features (ESF) station batteries, the adequacy of which has previously been evaluated by the staff and documented in the SER.
This issue is considered closed.
Fire-Protection System Water Supply By letter dated June 2, 1986, the applicant identified certain deviations from NFPA Standard 20 pertaining to the installation of fire pumps. Paragraph 6-3.1.1 Braidwood SSER 2 9-9
of this standard stipulates that fire pump feeder conductors inside buildings shall be enclosed by 2 inches of concrete or equivalent 1-hour fire resistance.
At Braidwood, conductors are in steel conduit.
The staff expressed concern that a single fire could render both fire pumps inoperable.
However, by letter dated August 4, 1986, the applicant responded that the diesel-driven fire pump is completely independent from the motor-driven pump.
All circuitry and com-ponents needed to start the diesel pump are located within the same area, which is separated from the motor-driven pump by 3-hour fire-rated barriers. On this bases, this deviation from NFPA 20 has no safety significance and is, therefore, acceptable.
During the inspection, the staff observed automatic sprinkler systems and manual hose station standpipes supplied from the same header.
BTP CMEB 9.5-1 provides guidance regarding water sprinklers and hose standpipe systems.
Header arrange-ments should be such that no single failure can impair both primary and backup fire protection systems.
Automatic sprinkler systems and manual hose station standpipes should have independent connections to supply headers. The applicant could not demonstrate that the condition of automatic sprinkler systems and manual hose station standpipes supplied from the common headers is an acceptable deviation from Section E.3.a of BTP CMEB 9.5-1.
By letter dated September 22, 1986, the applicant identified that for all plant locations where a single break or failure in piping would cause a loss of the water supply to both sprink-1er systems and standpipe outlets, compensatory measures, including the use of adjacent unaffected hose stations, would be implemented.
The applicant requested approval of the deviation from Section C.6.c of BTP CMEB 9.5-1 to the extent that it requires independent water supplies to the sprinkler and standpipe systems in these areas.
To compensate for a pipe break that might render both systems inoperable, the applicant will provide temporary interior hose connec-tions from adjacent locations using fire brigade hose carts, or will supply water via hose connections to outside hydrants.
The applicant would also im-plement actions prescribed in the fire protection report by posting a fire watch where sprinkler systems are affected.
Inside containment, plant Technical Specifications require that if the fire-suppression system water supply becomes inoperable, a backup fire-suppression system water supply must be established within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
If a backup water supply cannot be established, the applicant would have to bring the plant to a shutdown condition as stipulated in the plant Technical Specifications.
The staff considers this acceptable and, therefore, the issue is closed.
In the June 2,1986 letter, the applicant also identified a deviation from NFPA Standard 24, to the extent that it requires thrust blocks or other suitable means of restraint at each change of direction of the underground fire-3 protection system water-supply piping.
The applicant was not able to verify that thrust restraint had been provided.
The staff was concerned that the absence of thrust restraint could result in loss of pipe integrity with corres-ponding loss of capability to supply water to standpipes, sprinkler systems, and yard hydrants.
The applicant, by letter dated August 4, 1986, responded to this concern by confirming that the underground fire-main system at Braidwood is welded steel, which by formal interpretation (FI) 81-4 of NFPA 24 does not require thrust restraint except at the flanged hydrant connections. The appli-cant has also developed a procedure to monitor leakage at the hydrant connec-tions.
In the event that a leak develops, the affected portions of the system will be repaired.
If any hydrant was rendered inoperable, the applicant would Braidwood SSER 2 9-10
be able to provide compensating protection from adjacent hydrants until the affected hydrant is returned to service.
On this basis, the staff concludes that the apparent lack of thrust restraint at the flanged hydrant connections is an acceptable deviation from NFPA Standard 24.
Water Suppression and Standpipe Systems During the inspection, the staff observed manual deluge fire-suppression systems installed in various charcoal filter units.
There was concern that sufficient drainage of the charcoal filter units did not exist in the event of deluge sys-tem actuation.
The applicant should confirm that adequate drainage, sized to remove expected fire-fighting water flow, exists where fixed water-suppression systems are installed in charcoal filter units.
In response, the applicant has committed to implement safeguards, including pre-fire planning, to ensure that the charcoal filter unit housings do not become overfilled during deluge system operation.
The procedure. change will be made before exceeding 5% of rated power.
This response was accepted in Inspection Report 50-456/86026 (DRSS),
dated August 14, 1986.
Therefore, this issue is considered closed.
During the inspection, the staff observed that sprinkler heads were obstructed in areas including, but not limited to elevation 364 feet of the auxiliary building and elevations 364 feet and 426 feet of the turbine building.
The staff requires that sprinkler heads be repositioned /added to provide effective protection from an exposure fire.
The applic' ant has committed to relocate obstructed sprinkler heads, to clear the obstruction in accordance with NFPA Standard 13 criteria or provide technical justification for leaving the obstruc-tions.
The technical justification for leaving obstructed sprinklers in place will include assurance that adjacent sprinklers provide sufficient overlapping spray coverage to the obstructed area.
The applicant has committed to relocate required sprinklers within safety-related buildings before exceeding 5% of rated power.
Turbine building sprinkler obstructions will be evaluated, and sprinklers will be relocated, as required, before the end of the first refuel-ing outage.
During the inspection, the staff observed that fixed standpipe outlets were not provided to permit the fire brigade to reach all tunnel areas with no more than 100 feet of fire hose, as prescribed in Section C.6.c of BTP CMEB 9.5-1.
By letter dated September 22, 1986, the applicant committed to provide additional hose at a hose station to provide coverage to one of the valve rooms within the tunnel, before exceeding 5% of rated power.
The other valve room within the tunnel can be served by a hose from an outside hose house.
The applicant has verified that adequate pressures and flow are available to accommodate the additional hose.
The applicant has committed to revise its pre-fire plans to account for this additional hose before exceeding 5% of rated power.
Halon Suppression Systems In Amendment 7 to the FPR the applicant described the mechanism used to manually inithte the halon fire suppression system.
The staff was concerned that this design was not UL listed and represented an unjustified deviation from NFPA Standard 12A.
The applicant responded by letter dated August 4, 1986 that the actuating switches are UL listed and that no deviation exists.
The staff con-sidered this response satisfactory.
This issue is considered closed.
Braidwood SSER 2 9-11
Carbon Dioxide Suppression Systems The staff requested the results of the acceptance tests on the carbon dioxide fire-suppression systems. Not all of the tests had been performed at the time of the audit.
For those systems where testing was complete, the test results confirmed that the design concentrations and " soak times" were achieved and maintained.
9.5.1.5 Fire Protection for Specific Plant Areas Control Room In Amendment 4 to the FPR, the staff 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, the staff concluded that the smoke-detection system in the control room was acceptable.
During the inspection, the staff observed that smoke detectors were not installed in the main control console, although they were installed in the return air vents and ceiling.
Pending installation of smoke detectors in the main control console, the ade-quacy of the smoke-detection system in the control room is considered open.
Other Areas The staff observed previously unidentified chemical storage areas on elevation 401 feet of the turbine building.
There was concern that these areas were not provided with sufficient protection to prevent a significant fire from deselop-ing.
Pending receipt and evaluation of a fire hazards analysis for these areas, the staf f considers the adequacy of the level of fire protection in these loca-tions an open item.
During the review of Amendment 7 to the FPR, the staff expressed concern that the applicant may not have identified all significant deviations from NFPA codes and standards.
In response to this concern, the applicant, by letter dated June 2, 1986, identified a number of deviations from the following NFPA codes:
NFPA 13 Automatic Sprinkler Systems NFPA 14 Standpipe Systems NFPA 15 Water Spray Systems NFPA 20 Fire Pumps NFPA 24 Private Fire Service Mains NFPA 72E Fire Detectors A number of the sigqificant deviations have previously been evaluated and found acceptable in the SER and in this supplement.
The remaining deviations concern fire protection system hardware requirements, design parameters, or installation practices which are not applicable in a nuclear power plant, which are super-seded by other staff guidelines, and/or which have no safety significance.
The staff, therefore, concludes that the NFPA Code deviations identified in the June 2, 1986, letter are acceptable.
By letter dated September 30, 1986, the applicant submitted its evaluation of NFPA compliance with the remaining NFPA codes referenced in staff guidelines, identifying and justifying deviations in safety-related areas.
Based on this Braidwood SSER 2 9-12
letter, the staff identified the following features of the fire protection program as being incomplete (the current status is listed in parentheses):
4 (1) the installation of penetration seals in fire walls and fire-rated floor /
ceiling assemblies (to be completed before exceeding 5% of rated power)
(2) the installation of fireproofing for steel structural elements (to be completed before exceeding 5% of rated power)
(3) the installation of fire hose nozzles, hard rubber hose, and hose-house fire-fighting equipment (completed)
(4) the installation of fire detectors (installation completed; fully opera-tional before exceeding 5% of rated power)
(5) the installation of automatic sprinklers and draft stops (sprinkler in-stallation to be completed before exceeding 5% of rated power; draft stop installation to be completed 6 months after fuel load) 3 (6) the installation of fire doors (installation completed; fully operational before initial criticality)
(7) the installation of fire-rated cable wraps (to be completed before exceed-ing 5% of rated power) i (8) the removal of covered construction shacks which obstruct sprinkler head water discharge (completed)
NRC Region III personnel will confirm that the remaining unresolved issues are completed by the applicant.
Compensatory measures, such as hourly fire watch patrols, continuous fire watches and operational hose stations for manual back-up suppression capability will be in effect, as required by the plant Technical Specifications, until the above installation and requisite testing activities-are completed.
The staff finds this approach acceptable.
9.5.1.6 Deviations From BTP CMEB 9.5-1 In the SER, the staff approved the following deviations from the guidelines contained in BTP CMEB 9.5-1:
(1) protection of structural steel as described in Section 9.5.1.4*
(2) continuity of floor / ceiling assemblies as described in Section 9.5.1.4*
(3) acceptance criteria for fire barrier penetrations as described in Section 9.5.1.4*
(4) unlisted fire doors as described in Section 9.5.1.4*
(5) design of the fire pumps controller as described in Section 9.5.1.4*
(6) seismic design of the standpipe system as described in Section 9.5.1.5*
- This supplement.
Braidwood SSER 2 9-13
(7) absence of pressure reducers for the standpipe system as described in Section 9.5.1.5*
(8) fire protection for containment as described in Section 9.5.1.5*
(9) fire protection for the control room complex as described in Section 9.5.1.5*
(10) fixed fire suppression systems in the cable spreading room as described in Section 9.5.1.5*
(11) the separation of cooling water lines for the diesel generators in Room 18 as described in Section 9.5.1.5*
(12) deviations from BTP CMEB 9.5-1 in other plant areas as described in Sec-tion 9.5.1.5*
(13) implementation of fire protection modifications as described in Sec-tion 9.5.1.5*
On the basis of the evaluations contained in this supplement, the staff concludes that the following additional deviations are acceptable:
(14) internal conduit seals as described in Section 9.5.1.4*
(15) lack of fire detectors in certain fire areas as described in Section 9.5.1.4*
(16) NFPA Code deviations identified in Section 9.5.1.5*
9.5.1.7 Open Items Four issues require resolution:
(1) lack of smoke detectors in the main control console (2) identification / justification of NFPA Code deviations (3) associated circuit analysis of reactor vessel head vent valves regarding spurious actuation (4) revision of procedures (PRI-5) to clearly identify those instruments and controls that are electrically isolated from the control room when transfer switches are placed in remote positions 9.5.1.8 Conclusion On the basis of its review, the staff concludes that the fire protection program, with approved deviations but with the exception of the open items identified in Section 9.5.1.7, meets BTP CMEB 9.5-1, satisfies GDC 3 of Appendix A to 10 CFR 50, and is, therefore, acceptable.
This is Outstanding Item B(6).
By letter dated August 29, 1986, the applicant indicated that the approved fire-protection program is included by reference in the FSAR as required by Generic i
i RThis supplement.
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Letter 86-10.
Also included in the August 29 letter, was the proposed technical specification changes (which are currently under review) required for removal of the fire protection technical specifications in accordance with Generic Let-ter 86-10.
The staff will condition the operating license to require that the 3
applicant implement and maintain in effect'all provisions of the approved fire-
~
protection program.
This is License Cond'ition B(3).
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1 Braidwood SSER 2 9-15 i
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11 RADIOACTIVE WASTE MANAGEMENT 11.3 Gaseous Waste Management System Since the SER was issued, the applicant has amended the Final Safety Analysis Report (FSAR) on several occasions.
Some of these revisions have made it neces-sary for the staff to review previous conclusions presented in the SER to ensure that those conclusions a;'s still valid.
Some of the changes that the applicant has made involve exceptions to the regulatory positions of Regulatory Guide (RG) 1.140, " Design, Testing and Maintenance Criteria for Normal Ventilation Exhaust System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants." The staff has reviewed these changes through FSAR Amendment 47 and finds them acceptable.
11.4 Solid Waste Management Systems 11.4.1 System Description The SER stated that (1) the staff had not reviewed the polymer binder solidifi-cation system for conformance with Standard Review Plan (SRP) Section 11.4 and RG 1.143 because the applicant had not provided all pertinent information on the system design for staff review and (2) the applicant had not provided the process control programs (PCPs) for Braidwood Station, Units 1 and 2, for staff review.
Subsequently, the applicant was asked to provide (1) additional design informa-tion on the p~olymer binder solidification system and the Braidwood PCPs for the cement solidification system and (2) the Braidwood PCP for the polymer binder solidification system.
In response, the applicant provided the re-quested information in letters dated October 11 and 16, 1984.
The staff has reviewed the polymer binder solidification portion of the Braid-wood solid waste management system and the Braidwood PCPs for the cement and i
polymer binder solidification systems.
The cement solidification portion of the Braidwood solid waste management system was reviewed in the SER and was found acceptable.
The scope of review of the polymer binder solidification system and the PCPs included piping and instrument diagrams, descriptive information, the appli-cant's proposed design criteria and design bases, and the applicant's analysis of those criteria and bases.
The capability of the proposed system to process the types and volumes of vastes expected during normal operation and anticipated operational occurrences in accordance with General Design Criterion (GDC) 60, provisions for the handling of wastes relative to the requirements of 10 CFR 20 and 71 and of applicable U.S. Department of Transportation regulations, and the applicant's quality group classification and seismic design relative to RG 1.143 have also been reviewed.
The applicant's proposed methods of ensuring complete solidification have been reviewed, and the processing and design fea-tures meet Branch Technical Position (BTP) ETSB 11-3 and 10 CFR 61.
Braidwood SSER 2 11-1
The staff concludes that the design of the system meets the requirements of 10 CFR 20.106, 10 CFR 50.34a, and GDC 60, 63, and 64, and that the Braidwood PCPs for cement and polymer binder solidification systems meet the requirements in 10 CFR 61 and 71 and BTP ETSB 11-3, Revision 2.
The basis for acceptance has been conformance of the applicant's designs, design criteria, and design bases for the solid radwaste system to the regulations and guides referenced above, as well as to staff technical positions and industry standards.
On the basis of the foregoing evaluation, the staff concludes that the proposed solid radwaste system is acceptable.
Therefore, Confirmatory Issue B(11) is consid-ered closed.
The SER stated in Section 11.4 that no determination could be made as to the acceptability of the volume reduction (VR) system until the staff had com-pleted the review of the Aerojet Energy Conversion Company (AECC) topical re-port (AECC-2-P), which had been referenced by the applicant, and had received and reviewed additional information from the applicant covering site-specific application of the AECC system to Braidwood Station.
The staff has reviewed the AECC-2-P topical report, and it has been approved in a letter dated Novem-ber 21, 1984, from Cecil 0. Thomas (NRC) to Dr. Ramon Garcia (AECC).
In usually presented in Sections 11.1, 11.2, 11.3, 11.5, and 12.2 of the staff's SER. The information that would have been placed in these sections has been incorporated into this section of the SER for the purpose of maintaining con-tinuity.
Because the AECC-2-P system has been approved generically by the NRC, this supplement will discuss only its plant-specific application ta Braidwood Station. The Braidwood VR system has been reviewed to determine whether it conforms to 10 CFR 20.101, 20.103, 20.105, 20.106, and 20.305; GDC 60, 61, 63, and 64 of Appendix A to 10 CFR 50; 10 CFR 50.36a; Appendix I to 10 CFR 50; and 10 CFR 61 and 100.
In addition, the VR system's conformance to RG 1.140, 1.143, and 8.8 has also been determined.
11.4.2 Evaluation and Findings 11.4.2.1 Radiation Doses The staff has reviewed the Braidwood VR system and has evaluated the capability of the VR system to maintain radiation doses to individuals in restricted areas to the levels of 10 CFR 20.101.
The staff has reviewed the shielding provided for the various components of the VR system and has determined that the exposure to plant personnel will meet the requirements of this section.
This part of the Braidwood VR system review was covered under Sections 12.3 and 12.4 of the SER.
10 CFR 20.103 establishes limitations on the concentration that individuals in restricted areas may inhale and/or absorb through their skin.
The staff evaluated the VR system to determine whether suitable detection methods exist to measure airborne concentrations of radioactive materials and that access to radiation areas are suitably controlled with caution signs, labels, etc., in conformance with 10 CFR 20.103.
The staff determined that the applicant has incorporated such controls in the VR system areas at Braidwood Station.
This part of the Braidwood VR system review was addressed in a general nature in Section 12.3 of the SER.
Braidwood SSER 2 11-2
The permissible levels of radiation in unrestricted areas are addressed in 10 CFR 20.105.
These levels are based on hourly, 7-day, annual, and 40 CFR 190 dose limitations.
Conformance is determined by calculating the effluents expected to be released from the plant including all potential sources such as the VR system. The staff has performed this evaluation, and it was originally addressed in Section 11.1 of the SER and in Appendix D of the Braidwood Final Environmental Statement (FES) (NUREG-1026).
The staff has subsequently revised its calculations on the estimated releases occurring from the operation of the VR system; however, this has not altered the staff's conclusions presented in the above referenced sections of the SER and FES.
The radiation levels at Braidwood Station still meet the criteria of 10 CFR 20.105.
Additional details on the calculation of effluents released from the VR system during normal operation are contained in Section 11.4.2.2 of this supplement.
Radioactivity levels in effluents to unrestricted areas are limited by 10 CFR 20.106.
Because effluent concentrations are a function of all plant effluent streams, compliance must be determined on a plant specific basis.
The staff has determined that plant effluents from Braidwood Station meet the criteria of 10 CFR 20.106 with the operation of the VR system.
This conclusion had been previously expressed in the FES and Section 11.1 of the SER.
The pre-viously mentioned determination in the calculated effluents from the VR system did not alter this conclusion.
10 CFR 20.305 precludes applicants from treat-ing or disposing of licensed material by incineration except for (1) 0.05 micro-curie (pCi) or less per gram of medium of H3 or CH used for liquid scintilla-tion counting, (2) 0.005 pCi per gram or less of animal tissue averaged over the weight of the entire animal, or (3) approval pursuant to 10 CFR 20.106(b) and 20.302.
If the applicant has made a reasonable effort to minimize the radioactivity contained in effluents to unrestricted areas, and exposure to these effluents would not result in exposure of an individual to concentrations exceeding the limits specified in Appendix B, Table II, of 10 CFR 20, then 10 CFR 20.106(b) is met.
The applicant complies with 10 CFR 20.302 if it ap-plies for approval of a proposed disposal procedure and includes in the appli-cation a description of the radioactive material involved including quantity, kinds, activity levels, and the proposed manner and conditions of disposal.
The procedures to be observed to minimize the risk of unexpected or hazardous exposures must also be included.
The staff has evaluated the applicant's proposed method for incineration and has finds it conforms to 10 CFR 20.305 and all its referenced parts.
GDC 60 requires a nuclear power plant design to include means to control suit-ably the release of radioactive materials in gaseous and liquid effluents and to handle radioactive solid wastes produced during normal operation, including anticipated operational occurrences.
GDC 60 also requires that sufficient holdup capacity be provided for the retention of gaseous and liquid effluents containing radioactive materials.
The acceptability of the method used to control the release of radioactive gaseous effluents was addressed in the staff's SER for the AECC-2-P system.
For Braidwood, GDC 60 must be addressed with respect to the manner in which the solid radioactive wastes, which are produced during operation of the VR system, are handled, and with respect to the treatment of the overflow from the VR sys-tem condenser.
The latter is collected in a floor drain sump and then treated I
i Braidwood SSER 2 11-3
by Braidwood's liquid radwaste treatment system; the salt generated by the fluidized bed dryer and the ash generated by the incinerator of the VR system will be solidified utilizing a polymer binder system.
The staff has determined that the treatment of the condenser sump overflow resulting from operation of the VR system is in accordance with GDC 60.
The handling of solid wastes in accordance with GDC 60 has been addressed and found acceptable in the SER.
GDC 61 requires that radioactive waste and other systems that may contain radio-activity be designed to ensure adequate safety under normal operating conditions and postulated accident conditions.
The systems are to be designed (1) with the capability to permit periodic inspection and testing of components important to safety, (2) with suitable shielding for radiation protection, and (3) with ap-propriate containment, confinement, and filter systems.
The VR system components are housed in a series of individually shielded cubicles so that periodic in-spection and testing can be conducted.
Mechanical devices such as pumps and blowers have been isolated so that periodic inspection and maintenance can be conducted with a minimum ~ radiation exposure to operating personnel.
AECC pro-vided a radioisotope inventory of each component of the VR system to the appli-cant so that appropriate shielding could be designed.
Mechanical components, such as pumps and blowers, that require periodic maintenance are located in cubicles separate from nonmechanical components of the VR system, such as the fluidized bed dryer.
This allows the mechanical components to be decontami-nated and maintained in a low radiation field without need for decontaminating the entire system.
The staff has determined that the applicant's design meets GDC 61.
This was reviewed as a part of Sections 12.1 and 12.3 of the SER.
GDC 63 requires that appropriate systems be provided for radioactive waste systems and in associated handling areas in order that conditions that may result in excessive radiation levels may be detected and that appropriate safety actions may be initiated.
The VR system's instrumentation, its capa-bility to monitor various parameters in components and process lines of the system, and its capability to sense abnormal occurrences and to activate alarms on such an occurrence were discussed in the staff's review of Topical Report AECC-2-P.
Area radiation monitoring that must be addressed on a plant-specific basis was reviewed and discussed for Braidwood in Section 12.3 of the SER.
The area radiation monitoring system was found to conform with GDC 63.
11.4.2.2 Effluents GDC 64 requires that means be provided for.nonitoring effluent discharge paths for radioactivity released from normal operations, including anticipated opera-tional occurrences, and from postulated accidents.
At Braidwood the offgas from the VR system will be routed to the plant ventilation exhaust system, and the plant monitoring system will be used to monitor the raleases. With respect to GDC 64, the staff finds this approach acceptable.
The staff reviewed the acceptability of this monitoring scheme in Section 11.5 of the SER.
Appendix I to 10 CFR 50 establishes numerical guides for design objectives and limiting conditions for operation to meet the 10 CFR 50.36a criterion of as low as is reasonably achievable for radioactive material in nuclear power plant ef-fluents.
The effluents from the VR system must be considered in conjunction Braidwood SSER 2 11-4 l
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with the effluents from the remaining portions of the plant and the effect from the total effluents must be within the guidelines of Appendix I.
The operation of the Braidwood VR system will result in additional liquid and gaseous releases from the nuclear power plant above those that would occur if the VR system were not used.
However, these releases are small and, in the case of liquids, insignificant.
Airborne effluents would result directly from the operation of the dry waste processor and the fluidized bed dryer.
These re-leases would be discharged from the offgas filter assembly.
Liquid effluents would not occur directly from the VR system, but would occur indirectly.
Over-flow from the condenser goes to the plant's floor drain system.
Various pumps such as the scrubber /preconcentrator recirculation pump and the waste recircu-lation pumps all have drains associated with them that would take any leakage to the station's floor drain system.
In addition, the decontamination solutions are sent back to the evaporator feed tank for processing.
All of these wastes are treated by the station's liquid radwaste treatment system, with some being discharged off site after treatment.
For those wastes that are treated by evaporators, the bottoms will again be treated by the fluidized bed dryer and the process will begin anew.
Airborne Effluents The processing of evaporator bottoms in the fluidized bed dryer and dry acti-vated waste and contaminated oil in the dry waste processor will result in the discharge of airborne effluents from the offgas filter assembly.
The activity in the feed to the system will be dispersed among (1) the material removed at the gas / solids separator (2) the scrub solution collected in the bottom of the scrubber /preconcentrator (3) the condensate collected in the condenser sump (4) the exhaust air from the condenser The exhaust air from the condenser will be split into two streams.
Approxi-mately two-thirds of this exhaust will be discharged through a high-efficiency particulate air (HEPA) filter, followed by the equivalent of 6 inches of char-coal adsorber material (KI and TEDA impregnated), and another HEPA filter.
The other one-third of the exhaust air is recycled back to the fluidized bed dryer.
The distribution of activity with the various streams is a function of the particular radionuclide and removal capability or DF associated with the various VR system exhaust gas treatment components.
In its review of AECC-2-P, the staff credited the system with DFs of 100 for radioiodine and particulates.
Table 11.1 presents the airborne effluents that were calculated for release from Braidwood Station.
These releases are on a per reactor basis and were based on processing 5800 ft3 per reactor of combustible waste, 500 gallons per reactor of contaminated oil, and 9.63 x 105 gallons per reactor of evaporator bottoms. On the basis of the demographics of the Braidwood site and the releases indicated in Table 11.1, the thyroid dose was calculated to be 1.1 mrem /yr to an infant from all sources.
Half of this dose was calculated to occur as a re-sult of releases from the VR system.
The release of Cs-134 from the VR system was also calculated and found to contribute to the bone of an individual a dose of approximately 1.0 mrem /yr.
With respect to effluents originating from the VR system, this was the largest organ dose other than that to the thyroid.
Braidwood SSER 2 11-5
Liquid Effluents All of the liquid condensate generated by the operation of the VR system would be collected in the floor drain system, treated by the liquid radwaste system, and then reprocessed in the VR system with a portion of that liquid processed by the floor drain system discharged as a liquid effluent.
The staff calculated that the contribution of the conder. sate from the VR system to the liquid effluents at Braidwood would be negligible.
Solid Waste 10 CFR 61 establishes performance objectives for land disposal of waste, i.e.,
technical requirements for the waste form that waste generators must meet for the land disposal of waste and classification of waste.
The staff calculated the expected waste volumes and activity that would be anticipated from Braid-wood Station as a result of using the VR system.
The staff determined that the classification of waste would most likely be determined by the Cs-137 content of the fluidized bed dryer salt.
On the basis of the anticipated concentra-tion levels, the staff expects most of the VR waste product from Braidwood Station to be a Class A waste.
The applicant has indicated that the VR waste will be solidified using polymer binder. The staff has evaluated the use of the binder as a solidification agent for this product and for its conformance to 10 CFR 61.
The proposed manner of solidification has been found to be acceptable and in conformance with 10 CFR 61.
11.4.2.3 Accidents 10 CFR 100 establishes criteria that guide the NRC in its evaluation of the suitability of proposed sites for stationary power and testing reactors sub-ject to 10 CFR 50.
Through its use of the regulatory guides and standard re-view plans, the staff has established acceptable guidelines in terms of poten-tial offsite doses resulting from various potential accidents at a nuclear power plant. The magnitude of these doses varies with accidents, with the more probable accidents having the lower acceptable dose guidelines.
For rad-waste systems such as a VR system, the staff has determined that the allowable consequences of an accident are those that would result in a discharge of liquids off site with a concentration less than that stated in 10 CFR 20 and a discharge of gases off site with doses of no more than 0.5 rem to the whole body, 3.0 rems to the skin, bone, or thyroid, or 1.5 rnns to any other organ.
The Braidwood VR system was reviewed for potential accidents and the sources of radioactivity for those accidents.
Potential sources of radioactivity that were identified for this evaluation were those tanks or process equipment that could contain radioactivity, and the solid radwaste storage tank that will store the salts from the fluidized bed dryer and the ash from the dry waste processor.
The consequences of these accidents are a function of the feed stream activity to the VR system.
i The potential accidents that were analyzed in the evaluation of the Braidwood VR system were the rupture of the waste tanks containing the largest inven-tory, in terms of curies, of liquid and solid wastes. The staff's review of the VR system determined that the rupture of the storage hopper would result in the largest quantity of airborne activity and the largest dose.
The storage i
i Braidwood SSER 2 11-6
hopper collects the salts generated by the operation of the fluidized bed dryer and the ash generated by the operation of the dry waste processor.
For liquid releases, the release that would result in the highest dose con-sequences would occur because of the failure of piping associated with the scrubber /preconcentrator.
Rupture of Storage Hopper In the evaluation of potential airborne releases resulting from the operation of the VR system at Braidwood Station, it was determined that the rupture of the storage hopper would result in the largest quantity of airborne activity of any incident and also the largest dose.
For this analysis, it was assumed that the associated radioactivity of the feed streams to the VR system was based on 1% failed fuel in the reactor and that these streams had been pro-cessed by evaporators.
For this accident analysis, two sources of VR system product had to be con-sidered.
The first involved dry salts generated by the operation of the fluidized bed dryer, and the second involved dry ash generated by the opera-tion of the dry waste processor.
The feed to the fluidized bed dryer is the contents of the waste concentrate tanks.
The source of this waste is the evaporator bottoms.
The concentration of radioactivity in the evaporator bottoms is a function of the source of the bottoms.
For Braidwood Station, two cases were analyzed.
For the first, it i
was assumed that all the activity came from the boron recycle evaporators.
When the evaporator bottoms were calcined, it was assumed that the volume of evaporator bottoms was reduced by a factor of 11 and the increase in activity of radioiodine and particulates in the dry salt is by a factor of 11 over that in the initial feed stream.
The second case assumed that all of tiie activity in the evaporator bottoms originated from processing performed in the radwaste evaporators.
For this case, it was assumed that the volume of evaporator bottoms was reduced by a factor of 5 and the activity of radioiodine and particulates in the dry salt was increased by a factor of 5 over the initial feed stream activity.
The second source of activity to the storage hopper is that originating from the operation of the dry waste processor.
In this analysis, all of the material in the storage hopper was assumed to have originated from the incin-eration of dry activated waste.
For the dry activated waste, the volume of waste was assumed to be reduced by a factor of 200 and the activity in the ash was increased by a factor of 200.
Table 11.2 presents the activity levels of the salt and ash for the two cases and the two sources described above.
It was assumed that the rupture of the storage hopper would result in its entire contents being released off site.
On the basis of an offsite X/Q value of 5.6 x 10 4 sec/m2 and a breathing rate of 3.47 x 10 4 m3/sec, the maximum dose to an individual resulting from rupture of the storage hopper is presented in Table 11.3.
The resultant doses are acceptable.
Braidwood SSER 2 11-7 4
Accident Involving Discharge of a Liquid Two potential accidents involving the rupture of a vessel or piping in the VR system containing liquid radioactivity were considered which could result in an event more severe than that which has been previously analyzed in the SER.
These were rupture of the pipe from the scrubber /preconcentrator vessel to the scrubber recirculation pump and rupture of the line from the scrubber recircula-tion pump.
In either event, the concentration of the activity in the release would be a factor of 2.8 greater than the activity concentration in the feed stream.
The worst case was the rupture of the discharge piping from the scrubber /preconcentrator.
The maximum volume that could be released is approx-imately 216 gal.
In such an event, the liquid will be collected by the floor drains and routed back to the liquid radwaste system to be reprocessed.
11.4.2.4 Regulatory Guides The staff has reviewed tile VR system with respect to conformance to RGs 1.140, 1.143, and 8.8.
With respect to RG 1.140, the applicant stated that the Braid-wood VR system conformed to the regulatory positions associated with the guide as discussed in AECC's response to NRC Questions 63 and 80 in Topical Report AECC-2-P, Amendment 2.
The staff has reviewed these responses and has deter-mined that the applicant must test the HEPA filters and charcoal adsorbers with the frequency noted in RG 1.140.
In addition, the applicant must add the AP instrumentation to charcoal adsorbers' in accordance with RG 1.140.
The applicant has committed to implement the required testing and instrumentation.
Therefore, the Braidwood system meets the requirements of this guide.
The staff expressed a concern that the inlet air filter to the fluid bed dryer (0VR02M) did not include a AP indicator and a AP alarm.
The applicant indicated that the normal mode of operation would have no flow through this filter and that the filter is only used during purge conditions during shutdown operations.
The staff will not require a AP indicator and a AP alarm provided this filter unit is not used as a source of makeup air to the fluid bed dryer for the pur-pose of maintaining normal operation of the fluid bed dryer.
Filter OVR08F has a AP indicator.
In lieu of equipping this filter with a AP alarm, the applicant has committed to logging this parameter twice per shift.
The staff finds this acceptable.
With respect to RG 1.143, the applicant has indicated that the VR system con-forms to the guide's regulatory positions except for Regulatory Position 1.2, those exceptions taken by AECC in Topical Report AECC-2-P(NP) and its amendments, and those exceptions in response to NRC Questions 55 and 79.
The staff reviewed these exceptions and concluded that these exceptions were acceptable.
The staff has reviewed the Braidwood VR system with respect to its conformance to RG 8.8.
The general radiation protection design features of the Braidwood VR system are consistent with the guidelines of RG 8.8, "Information Relevant to Ensuriag That Occupational Radiation Exposures at Nuclear Power Stations i
Will Be As Low As Is Reasonably Achievable" (Revision 3).
These features are intended to ensure that occupational radiation exposure to personnel involved with the operation, maintenance, and inspection of the radwaste system is main-tained as low as is reasonably achievable (ALARA).
This Braidwood VR system conforms with RG 8.8.
Braidwood SSER 2 11-8
11.4.2.5 Operations The operation of the AECC-2-P system was addressed in the staff's review of the topical report submitted by AECC.
The staff indicated there that site-specific application of this system would require the applicant to address hew it will limit the feed to the dry waste processor to less than 1% by weight of halogen-ated plastics and to 0.3% by content of sulfur in the contaminated oil.
The applicant is planning on segregating those materials that are acceptable for incineration from those that are not by (1) sorting large items of dry active waste (DAW) at the point of discard and (2) manually sorting bags of DAW at a special table equipped with a hood and a HEPA filter.
The HEPA filter would discharge to the room containing the sorting table.
In addition, to reduce the amount of sorting required and associated dose to workers, the applicant has committed to limit as much as possible its stock items to those with ac-ceptable levels of halogenated plastics and sulfur.
In its review, the staff has approved this method of segregating the waste and the applicant's approach, provided that station workers assigned to manually sort the DAW receive specific training on proper monitoring and ALARA procedures.
The staff has expressed a concern to the applicant about the number of indivi-duals who would operate the VR system during the year and the means for ensur-ing that the system would be maintained with so many people operating it.
The applicant estimated that an equipment attendant, who will operate the VR system, would operate the equipment once every 4 weeks; the staff estimates this figure to be once every 6 weeks.
In addition, because the VR system is only expected to be operated half of the time, equipment attendants may only operate the sys-tem on the average M 1 week out of 12.
The staff believes that this is too infrequent, and as a con:,qence, most of the equipment attendants will never develop a thorough working knowledge of the system.
The applicant has re-sponded to the staff's concern by committing to augment the VR system staff during the first 6 months of operation.
Following this initial period of opera-tion, the applicant will review system performance and the staff in terms of complexity of operation, problems experienced, ALARA considerations, etc.
On the basis of this review, the applicant will reassess its long-term staffing requirements to ensure safe operation of the VR system.
The staff concurs with the applicant's approach and will follow closely the operation of the VR system.
The staff's review of the AECC-2-P topical report also noted that the system design was based on decontaminating with hot water and not organic decontami-nating solutionr.
The Braidwood VR system is restricted to decontamination by hot water.
11.4.2.6 Fire Protection The system design and operation of the Braidwood VR system was reviewed with re-spect to its capability to conform to BTP CMEB 9.5-1.
The operation of the VR system does not result in any nonconformance with this BTP as noted in Sec-tion 9.5.1 of the SER and this supplement.
Braidwood SSER 2 11-9 t
11.4.2.7 Conclusions The staff has determined that (1) The Braidwood VR system can safely and adequately process liquid radio-active wastes, contaminated oil, and compactible trash.
(2) The design, construction, and quality group classification of the Braid-wood VR system are in accordance with RG 1.143 as noted in Section 11.4.2.4.
(3) The design, construction, and quality group classification of the offgas filter system are in accordance with RG 1.140 as noted in Section 11.4.2.4.
(4) The Braidwood VR system can operate and meet the in plant ALARA criteria of RG 8.8.
(5) The Braidwood VR system can operate without jeopardizing the operation of the remainder of the plant or the safety of the general public.
(6) The Braidwood VR system can operate within the fire protection criteria of BTP CMEB 9.5-1 (7) The Braidwood VR system conforms with the following: 10 CFR 20.101, 20.103, 20.105, 10 CFR 20.106, and 20.305; 10 CFR 50, Appendix A, GDC 61 and 64; 10 CFR 50, Appendix I; and 10 CFR 100.
(8) Incineration of polyvinyl chloride in the feed to the AECC-2-P system should be limited to 1% by weight, and the sulfur content in the oil should be limited to 0.3% by weight.
(9) Only hot water solutions may be used to decontaminate VR system components.
(10) The VR system results in a product that is capable of conforming to 10 CFR 61.
On the basis of the above findings, the volume reduction system is approved.
Therefore, Open Item B(7) is considered closed.
11.5 Process and Effluent Radiological Monitoring and Sampling Systems 11.5.2 Evaluation and Findings In the SER, the staff indicated that the applicant had not provided the calibra-tion techniques nor the energy dependence of response of the noble gas monitor as required by TMI Action Plan Item II.F.1, Attachment 1.
The applicant has provided this information in its October 17, 1984, letter, and it has bean found acceptable.
Therefore, Confirmatory Issue B(12) is considered closed.
The inspection report enclosed in the July 10, 1984, letter from C. J. Paperiello to C. Reed identified concerns by the NRC inspectors on the ability of installed equipment at Byron which is the same as at Braidwood to adequately meet the re-quirements of NUREG-0737, Item II.F.1, Attachment 2, regarding the sampling and analysis of iodine and particulate effluents.
In particular, recent research Braidwood SSER 2 11-10
into the deposition of airborne radiciodine on metal surfaces indicates that the Braidwood design may not provide a representative sample.
By letter dated August 17, 1984, the applicant committed to resolve this mat-ter before startup following the first refueling outage.
The staff found this schedule acceptable because the noble gas monitor may be used to project the magnitude of radioiodine releases if an accident occurs during the first cycle.
Therefore, the Braidwood license is being conditioned to require that, before startup following the first refueling outage, the applicant demonstrate that the operating iodine / particulate sampling system will perform its intended function.
This is License Condition A(6).
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i Braidwood SSER 2 11-11
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Table 11.1 Airborne releases resulting from operation of the Braidwood volume reduction system (Ci/yr/ unit)
Fluid bed Dry waste Radionuclide dryer operation processor operation Total H-3 2.23(2)*
2.23(2)
Mn-54 3.36(-4) 1.22(-3) 1.56(-3)
Fe-55 1.84(-4) 1.84(-4)
Fe-59 8.17(-4) 8.17(-4)
Co-58 1.49(-4) 1.26(-2) 1.27(-2)
Co-60 2.37(-3) 1.66(-2) 1.91(-2)
Sr-89 3.04(-4) 3.04(-4)
Sr-90 1.16(-5) 1.16(-5)
I-131 7.57(-2) 7.57(-2)
Cs-134 5.02(-2) 2.44(-3) 5.26(-2)
Cs-136 7.63(-3) 7.63(-3)
Cs-137 3.72(-2) 6.50(-3) 4.37(-2)
- Exponential notation: 2.23(2) = 2.23 x 102, Table 11.2 Activity levels of salt and ash in storage hopper at Braidwood Station based on acci-dent source terms (Ci)
Boron Radwaste Dry active Radionuclide recycle evaporator evaporator waste H-3 2.78(1)*
1.03(1)
Mn-54 9.22(-4) 1.93(-3) 3.37(-1)
Fe-55 5.24(-3) 1.05(-2)
Fe-59 1.70(-3) 4.76(-3)
Co-58 3.49(-2) 8.61(-2) 3.48 Co-60 5.73(-3) 1.36(-2) 4.58 Sr-89 5.24(-3) 1.47(-2)
Sr-90 2.74(-4) 5.55(-4)
I-131 9.49(-1) 3.74 Cs-134 9.97 1.36 6.73(-1)
Cs-136 9.73(-1) 2.72(-1)
Cs-137 5.73 9.97(-1) 1.79
- Exponential notation:
2.47(1) = 2.47 x 101 f
i Braidwood SSER 2 11-12 l
(
J Table 11.3 Doses resulting from rupture of storage hopper at Braidwood Station (rem)
Whole Maximum Source body Skin organ Boron recycle evaporator bottoms 2.78(-3)*
3.68(-3) 2.97(-1)(a)
Radwaste evaporator bottoms 6.25(-4) 8.40(-4) 1.07(b)
Dry active waste 2.16(<3) 2.84(-3) 7.50(-1)(c)
(a) liver.
(b) thyroid.
(c) lung.
- Exponential notation:
2.78(-3) = 2.78 x 10 8 t
I i
1 1
1 2
I Braidwood SSER 2 11-13
.~.,- _,---_ _ __
13 CONDUCT OF OPERATIONS 13.5 Plant Procedures 13.5.2 Operating and Maintenance Procedures After the accident at Three Mile Island (TMI), the staff developed the TMI Action Plan (NUREG-0660 and NUREG-0737), which required licensees of operating reactors to reanalyze transients and accidents and upgrade emergency operating procedures (EOPs) (Item I.C.1).
The plan also required the NRC staff to develop a long-term plan that integrated and expanded efforts in the writing, reviewing, and monitoring of plant procedures (Item I.C.9).
NUREG-0899, " Guidelines for Preparation of Emergency Operating Procedures," represents the staff's long-term program for upgrading E0Ps, and describes the use of a Procedures Genera-tion Package (PGP) to prepare E0Ps.
Submittal of the PGP was made a requirement by Supplement 1 to NUREG-0737, " Requirements for Emergency Response Capability" (Generic Letter 82-33).
Generic Letter 82-33 requires each licensee to submit to the NRC a PGP, which includes:
(1) plant specific technical guidelines (2) a writer's guide (3) a description of the programs to be used for the validation and verification of E0Ps (4) a description of the training program for the upgraded E0Ps This report describes the staff's review of the applicant's response to Sec-tion 7 of Generic Letter 82-33 related to the development and implementation of E0Ps.
Criteria for the review of a PGP are in Standard Review Plan (SRP) Section 13.5.2.
Further guidance is contained in NUREG-0899, the reference document for the E0P upgrade portion of Generic Letter 82-33.
The staff determined that the PGP is acceptable for full power operation.
The applicant has committed to address a list of items in the longer term, as dis-cussed in the following sections, consistent with generic resolution of Westing-house Owners Group (WOG) Emergency Response Guidelines (ERGS) long-term issues.
Discussion and Evaluation By letter dated March 26, 1985, the applicant submitted its PGP for Byron /
Braidwood.
By letter dated September 16, 1986, the applicant provided addi-tional information.
The PGP contained an introduction and the following sections:
Plant-Specific Technical Guidelines Abnormal and Emergency Operating Procedures Writer's Guide BRAIDWOOD SSER 2 13-1
E0P Verification Program E0P Validation Program E0P Training Program (1) Plant-Specific Technical Guidelines (P-STGs)
All licensees and applicants are required to submit P-STGs.
These guidelines may be based on generic technical guidelines (prepared by the owner's group) or on a plant-specific reanalysis of transients and accidents as described in TMI Action Plan Item I.C.l.
In either case, the P-STG should be based on the identification of plant systems and frnctions, and should be supported by an analysis of operator tasks to identify operator information and control needs.
If generic technical guidelines are referenced, additional task specification may be needed, depending on the level of task information provided by the generic technical guidelines and the nature of deviations from the guidelines.
Examples of deviations that should be reviewed and documented are as follows:
any modification to the mitigative strategy of the generic technical guidelines (e.g., for a Westinghouse plant, initial depressurizing of the reactor coolant system (RCS) following a steam generator tube rupture without first having conducted a limited cooldown in accordance with the guidelines to establish a margin to saturation) differences in equipment operating criteria (e.g., reactor coolant pump (RCP) trip criteria, safety injection (SI) termination criteria) differences in equipment operating characteristics (i.e., between the plant-specific equipment and that assumed in the generic analyses, such as SI that can be throttled vs. only on/off) identification of methods and equipment used to address the technical areas of the generic guidelines that are specified as " plant-specific" plant-specific setpoints or action levels that are calculated or deter-mined in the manner other than specified in the generic technical guidelines NOTE:
Plant-specific setpoints (e.g., setpoints associated with automatic initiation of the emergency core cooling system) called for by the generic guidelines need not be included in the P-STG submittal.
actions that are taken in addition to those specified in the generic guidelines and that affect the mitigative strategy differences that affect the equipment's ability to adequately provide the necessary mitigative function use of different instruments or control parameters than those specified in l
the generic technical guidelines or determining instrumentation and control characteristics in a manner different than, or with a different basis than, j
that specified in the generic technical guidelines BRAIDWOOD SSER 2 13-2
identification of items not covered by the NRC-approved generic technical guidelines (e.g., plant-specific conditions, equipment, operations, or bracketed [ ] information from the generic technical guidelines that re-late to systems, functions or methods)
The purpose of the review of the technical guidelines submittal is to determine that the following general objectives are adequately addressed:
(1) The E0Ps will be based on acceptable, validated technical guidelines de-rived from approved analyses of transients and accidents as described in NUREG-0660, Items I.C.1 and I.C.9 (as clarified by NUREG-0737 and Supple-ment 1 to NUREG-0737).
The P-STG along with the generic guidelines (if referenced) and supporting documentation provide E0P writers with all the technical information necessary for preparing E0Ps which direct operators' actions to mitigate the consequences of transients and accidents without needing to first diagnose an event to maintain the plant in a safe condi-tion (function orientation).
(2) The PGP describes an adequate method to identify information and control needs to be used as a basis for identifying control room instrumentation and controls necessary to perform the tasks specified in the technical guidelines.
By letter dated March 26, 1985, the applicant.provided information on plant emergency response guidelines and referenced the Westinghouse Emergency Re-sponse Guidelines (ERGS), Revision 1 (high pressure version) which were authorized for implementation by the staff in a letter dated December 27, 1984, supersede those based on the Westinghouse ERGS, Revision 0.
The applicant identified the following source documents for use in generating E0Ps for Byron /
Braidwood (B/B):
8/B E0P Writers Guide WOG ERG (Revision 1) and background documents B/B electrical drawings B/B piping and instrumentation drawings B/B FSAR licensing commitments relating to E0Ps Westinghouse Bulletins and Memos (as appropriate) vendor technical manuals (as appropriate) 8/B plant descriptions B/B instrumentation description B/B administrative procedures B/B system operating procedures B/B general operating procedures B/B abnormal operating procedures existing B/B emergency operating procedures 8/B precautions, limitations, and setpoints document In the September 16, 1986, letter, the applicant provided (1) clarification of the PGP identified instrumentation and controls needed to support use of the B/B E0Ps, (2) a sample plant E0P with accompanying deviation, training, and validation documentation for staff audit, and (3) a written commitment to ex-pand the step deviation documentation to include justification for the bracketed, Braidwood SSER 2 13-3 l
l
I plant-specific steps by 6 months following the issuance of a full power license for either Braidwood Unit 1 or Byron Unit 2, whichever comes first.
From its review, the staff concludes that the B/B PGPs reflect essentially the same objective and mitigative strategies as the referenced WOG ERGS which were approved for implementation.
1 The staff has noted concerns regarding treatment of steam generator tube rup-ture (SGTR) events, partially attributable to the lack of status tree identifi-cation of " radioactivity control" as a critical safety function, and regarding a questionable interface with the emergency action levels (EALs) of plant emer-gency plans in the referenced ERGS.
However, these concerns are attenuated by provisions identified in the staff's review of the material submitted by the applicant.
Attachment A, Table 2, of the B/B PGP identifies the availability of steamline radiation monitors which would assist in the diagnoses of SGTR events.
The sample E0P which the staff audited has an added note that indicates an in-terface with the site emergency plan.
From its review of the B/B PGP and its audit of the sample E0P and its support-ing documentation, the staff finds that the B/B P-STGs provide appropriate tech-nical variance to account for plant-specific differences from the reference plant design, and to include other emergency operational considerations, e.g.,
interaction with EALs, not covered by the generic guidelines.
On the basis of its review (discussed above), the staff finds that the plant-specific emergency procedures guidelines described in the B/B PGP are techni-cally acceptable for implementation.
In support of this finding, the applicant has committed to expand the technical deviations documentation to include brack-eted items from the WOG ERGS.
Concerns related to the treatment of SGTRs, pro-vision of a radiation control safety function, and interaction with EALs are attenuated by B/B P-STG feature, and will be pursued further consistent with generic resolution of WOG ERG long-term issues.
The B/B P-STGs, expanded as committed, contain plant-specific technical infor-mation which is necessary as a reference to supplement the EGPs, as a basis for the plant task analysis program, and as documentation of status in amending the P-STGs in the future.
In consideration of the above, the staff concludes that the technical content of the B/B PGPs is consistent with the requirements o' SRP Section 13.5.2 and the B/B PGPs are, therefore, acceptable for implementation. Therefore, Out-standing Item A(7) is considered closed.
(2) Writer's Guide Applicants are required to submit a writer's guide that details the specific methods to be used in preparing E0Ps which are based on the P-STGs.
NUREG-0899 provides objectives and intent for the writer's guide.
Because of the variety of available technical writing style guides and other references pertaining to the presentation of information, the specific information found in the writer's guide is expected to vary considerably among plants.
For this reason, the staff did not perform a generic review of the human factors aspects of the Westinghouse Owners Group Writer's Guide.
Each applicant has to submit a plant-specific writer's guide for staff review.
Braidwood SSER 2 13-4
The purpose of the evaluation is to determine if acceptable methods are described for accomplishing the following general objectives:
The writer's guide provides sufficient information for developing E0Ps from the P-STG, which are usable, accurate, complete, readable, convenient to use, and acceptable to control room personnel.
The writer's guide supports upgrading of the procedures and long-term consistency within and between procedures.
By letter dated March 26, 1985, the applicant submitted its PGP.
This package included the Byron /Braidwood E0P Writer's Guide.
By letters dated September 16 and 29, 1986, the applicant prcvided additional information.
On the basis of its review, the staff concludes that the Writer's Guide submit-ted by the applicant provides acceptable information for developing E0Ps which are usable, accurate, complete, readable, convenient to use, and acceptable to control room personnel, meets the requirements of SRP Section 13.5.2, and there-fore is acceptable.
(3) Program for Validation / Verification (V/V)
The purpose of evaluating the applicant's V/V program is to determine whether the applicant has provided evidence that the upgraded E0Ps are technically correct, are written to accurately reflect the plant-specific Writer's Guide, are usable, correspond to control room / plant equipment, are compatible with the minimum number, qualifications, training, and experience of the operating staff, and provide a high level of assurance that the procedures will work as a component of the accident mitigation system.
By letter dated March 26, 1985, the appli: ant described its V/V program.
The program implemented by the applicant is systematic and comprehensive,
- however, certain details and criteria that were used in the program were omitted from the PGP.
These details would be necessary to replicate the V/V program properly in future years.
The applicant has committed in its letter of September 16, 1986, to add these details to the PGP by January 31, 1987.
The staff finds this commitment acceptable because of the low probability of needing to consult the Write 's Guide regarding this issue before January 31, 1987.
The next major revision of E0Ps will not be completed until 6 months after a full power license has been issued.
On this basis, the staff concludes that the applicant has developed a V/V pro-i gram that provides adequate assurs'ce that E0Ps are technically correct and usable, follow the Writer's Guide, correspond to the control room / plant hard-4 ware, and are compatible with the minimum number, qualifications, training, and experience of the operating staff, meets the requirements of SRP Section 13.5.2, and is therefore acceptable.
(4) Program for Operator Training and E0Ps The purpose of evaluating the applicant's training program on E0Ps is to ensure that operators understand the philosophy, mitigation strategies, and technical bases of the E0Ps; that they have a working knowledge of the technical content BRAIDWOOD SSER 2 13-5
of the E0Ps, and are capable of successfully executing the E0Ps under emergency conditions.
By letter dated March 26, 1985, the applicant described a program of classroom and simulator training directly aimed at enabling operators to understand the structure, basis, and limitations of the E0Ps and to provide a working know-ledge of the technical content of the E0Ps by practicing E0Ps under simulated emergency conditions. The staff finds that the program is acceptable, but has not been fully documented in the PGP.
By letter dated September 16, 1986, the applicant committed to revise the PGP to provide further detail regarding the training program by January 31, 1987.
The staff finds this commitment accept-able because of the low probability of needing the revised guidance before that date. The staff concludes that implementation of the described training pro-gram should result in the operator understanding the philosophy behind the approach to the E0Ps, understanding the mitigative strategy and technical basis 1
of the E0Ps, having a working knowledge of the technical content of the E0Ps, and having the capability to execute the E0Ps under operational conditions.
The training program meets the requirements of SRP Section 13.5.2, and is acceptable.
Conclusions (1) Plant-Specific Technical Guidelines (P-STGs)
The staff concludes that because the applicant's E0Ps are based on the Westing-house Emergency Response Guidelines that have been approved for implementation, and because they retain the basic mitigation strategies of the Westinghouse ERGS, they contain adequate technical basis.
The applicant has committed to add plant-specific information as discussed above.
The B/B PGP is consistent with the applicable technical requirements of SRP Section 13.5.2 and is accept-able.
Therefore, Outstanding Item A(7) is considered closed.
(2) Writer's Guide The applicant has committed to an acceptable Writer's Guide that provides infor-mation for developing E0Ps from the P-STGs, which are usable, accurate, complete, readable, convenient to use, and acceptable to control room personnel.
The Writer's Guide will be revised to comply with staff recommendations by Janu-l ary 31, 1987.
This commitment is acceptable because of the low probability of I
needing the revised guidance before that date.
(3) V/V Program i
The applicant has conducted activities which meet the objectives of a V/V pro-gram.
The applicant has committed to fully document its V/V process in a revi-sion to the PGP by January 31, 1987.
This commitment is acceptable because of i
the low probability of needing the revised guidance before that date.
l l
(4) E0P Training Program The E0P training progam described in the applicant's letter of March 22, 1986 (with the commitment regarding the E0P training program), meets the requirements l
of SRP Section 13.5.2 and is, therefore, acceptable.
l BRAIDWOOD SSER 2 13-6
By letter dated September 16, 1986, the applicant has committed to provide further information regarding the E0P training program in the PGP by January 31, 1987.
This commitment is acceptable because of the low probability of needing the revised guidance before that date.
1 BRAIDWOOD SSER 2 13-7
14 INITIAL TEST PROGRAM By letter dated June 23, 1986, the applicant proposed to revise one test and eliminate five others from the Startup Test Program for Braidwood Station, Units 1 and 2.
The requests to eliminate the tests for shutdown from outside the control room and for loss of offsite power are not acceptable; the other four proposals are acceptable.
The staff evaluation for all six proposals follows.
Shutdown From Outside The Control Room The applicant has proposed to modify the test summary to indicate that it will be performed on Byron Station, Unit 1, only.
The applicant justifies this change on the bais of preoperational tests of the remote-shutdown systems, which require the plant to be maintained in the hot-standby condition for at least 30 minutes, and which provide all necessary design information regarding remote-shutdown capability.
Regulatory Guide (RG) 1.68.2, Section C, states, in part, that the test program should verify that "the nuclear power plant can be safely shut down from outside the control room."
The applicant has provided no alternative test that will demonstrate this capability.
In addition, RG 1.68.2 states, in part, that
" licensees... conduct a test program to demonstrate remote shutdown capability for each unit of their plants."
This demonstration is necessary to verify proper operation of the remote-shutdown capability on each unit.
Experience at other facilities has demonstrated that preoperational and subsystem level tests do not achieve these objectives.
This change is, therefore, not acceptable.
Loss of Offsite Power The applicant has proposed to modify the test summary to indicate that it will be performed on Byron Unit 1 only.
The applicant justifies this on the basis of the test performed at Byron Unit 1 and the preoperational test program at Byron Unit 2 and Braidwood Units 1 and 2 which verifies that onsite power systems are functional.
RG 1.68, Appendix A, Section 5.J.-j, states that the test program " demonstrate that the dynamic response of the plant is in accordance with design for a condi-
{
tion of loss of turbine generator coincident with loss of all sources of offsite power (i.e., station blackout)." Although the test performed at Byron Unit 1 demonstrated that the plant design is adequate, each unit must be tested to ver-l ify that the hardware performance at the system level is as expected.
Experience at other facilities has demonstrated that preoperational and subsystem level tests are not adequate to demonstrate that the dynamic response of each unit is i
in accordance with design.
This change is, therefore, not acceptable.
Rod Orop Measurements This test summary requires measurement of rod drop times at cold no-flow, cold full-flow, hot no-flow, and hot full-flow conditions following core loading.
The applicant proposed to modify the test summary to indicate that it applies to Braidwood SSER 2 14-1
Byron Unit 1 only, and to add a new summary for Byron Unit 2 as well as Braidwood Units 1 and 2 which requires measurement of rod drop times at hot, full-flow conditions only.
The applicant's justification for this change is that the West-inghouse acceptance criteria apply to the bot full-flow condition only, and that from previous experience the applicant has found rod drop times at other test conditions fall under the hot full-flow values. The staff finds this justifi-cation bounds the other test conditions, these additional tests are not neces-sary, and therefore the proposed modification is acceptable.
Pseudo Rod Ejection The applicant has proposed to modify the test summary to indicate that it applies to Byron Unit 1 only. Verification of core design parameters for Byron Unit 2 and Braidwood Units 1 and 2 will be achieved through control rod worth measure-ments, boron worth measurements, and flux mapping at zero power.
The purpose of this test is to verify calculational models and accident analysis assumptions. These design features have been verified on Byron Unit 1.
RG 1.68, Appendix A, Item 5.e, specifically allows the test to be deleted for facilities using calculational models and designs identical to prototype facilities.
This change is acceptable.
Flux Asymmetry Evaluation The applicant proposed to delete this test from the Startup Test Program for Byron Unit 2 and Braidwood Units 1 and 2.
The performance of this test on Byron Unit 1 has confirmed that the core thermal and nuclear parameters are in accor-dance with predictions.
The staff finds this change acceptable.
Turbine Trip From 25% Power The applicant proposed to delete this test from the Startup Test Program for Bryon Unit 2 and Braidwood Units 1 and 2.
This test is not required since a 100% power full-load rejection test will be performed in accordance with RG 1.68, Appendix A, Item 5.n.n.
This change is, therefore, acceptable.
Requests for Schedular Relief By letter dated September 19, 1986, the applicant requested schedular relief for certain preoperational tests described in FSAR Section 14.2.
RG 1.68, Para-graph C.2, statt,, in part:
" Tests designated in the FSAR as preoperational tests should bt 6.npleted and the results of such tests evaluated and approved by the applican' prior to issuance of the operating license." However, 10 CFR 50.57(b) states:
"Each operating license will include appropriate pro-visions with respect to any uncompleted items of construction and such limita-tions or conditions as are required to assure that 6peration during the period of the completion of such items will not endanger public health and safety."
The applicant's request distinguishes between tests of the following systems that will be operational when required by the Technical Specifications and those systems that will not be operational when required.
Braidwood SSER 2 14-2
(1) Systems That Will Be Available When Required by Technical Specifications AF-10, auxiliary feed PS-10, primary sample These systems will be available before system criticality.
They will not be considered to be operational until testing has been completed.
Because the fission product inventory is low before initial criticality, the consequences of an accident before initial criticality are minimal.
Tharefore, the staff finds it acceptable to defer completion of these tests beyond fuel load on the condi-tion that all testing, retesting, data review, and approval are completed before initial criticality.
(2) Systems That Will Not Be Available When Required by Technical Specifications WQ-10, Control Room Chill Water The control room chill water subsystem is required for operability of the con-trol room ventilation system.
The control room chillers have been tested to the extent possible without the VC available as a heat load.
Because all testing, retesting, data review, and approval of the control room chillers need to be completed for operability of the control room ventilation system, the staff finds it acceptable to defer the schedule for testing the chillers until such time as the control ventilation system is available.
RY-10, Pressurizer The pressurizer pressure control system was tested during hot functional testing.
Post-test data review determined that the spray valve full-open setpoint and the control deviation alarm, which tripped at the same controller output required adjustment and retesting.
The applicant has proposed to perform the calibration, retesting, data review, and approval during the initial plant heatup.
In re-sponse to the staff's request, the applicant provided the pertinent hot functional test data in a letter dated October 3, 1986.
The staff has reviewed these data and determined that the setpo;nt deviation from nominal is approximately 1.5%,
which does not endanger the public health and safety.
In addition, during mode 3 of the initial plant startup, there are no fission products in the core and, therefore, no decay heat sources.
An overpressure condition might result in lifting a pressurizer power operated relief valve and a consequent loss of primary coolant.
However, with no decay heat sources in the core, this is not a safety concern.
Therefore, the staff finds it acceptable to operate in mode 3 during the initial plant heatup with the pressurizer pres-sure control functions out of tolerance.
EF-10, ESF Actuation The ESF (engineered safety features) actuation system will be operable before fuel load except for the outputs to the auxiliary building ventilation system.
The applicant has requested deferral of the operability requirement for the ventilation system.
Because there is no effect on the public health and safety if the ventilation system actuation system is inoperable during the time that the ventilation is inoperable, the staff finds this deferral of testing acceptable.
Braidwood SSER 2 14-3
15 ACCIDENT ANALYSES 15.4 Radiological Consequences of Accidents Section 6.5.1 of this supplement revises several filter efficiencies.
As a result, Section 15.4.1.2 and Tables 15.1, 15.2, and 15.6 of the SER have been revised.
15.4.1 Loss-of-Coolant Accident 15.4.1.2 Post-LOCA Leakage From Engineered Safety Feature System Outside Containment In FSAR Table 15.6-Sa, the applicant has identified a value of 3760 cc/hr as the leak rate from emergency core cooling system (ECCS) equipment following an accident.
Following the guidance of the Standard Review Plan, the staff evaluated the potential radiological consequences from this release pathway assuming a routine leakage rate of twice the applicant's value (7520 cc/hr).
The resultant radiological consequences to the thyroid were estimated to be 0.8 rem at the exclusion area boundary and 0.5 rem at the low population zone (LPZ). Whole-body doses, as indicated in Table 15.1, are estimated to be small.
15.6 Anticipated Transients Without Scram The Westinghouse generic guidelines (Revision 0) approved by the staff in Generic Letter 83-22, dated June 3, 1983, adequately address the subject of anticipated transients without scram (ATWS).
As discussed in Section 13.5.2, the staff concludes that the applicant's implementation of the latest update of these Westinghouse generic guidelines (Revision 1) is acceptable for licensing and full power operation.
The ATWS rule,10 CFR 50.62, " Requirements for Reduction in Risk From Antici-pated Transients Without Scram (ATWS) Events for Light-Water-Cooled Nuclear Power Plants" requires that each Westinghouse pressurized-water reactor must have equipment from sensor output to final actuation device, which is diverse from the reactor trip system, to automatically initiate the auxiliary feedwater system and initiate a turbine '-ip under conditions indicative of an ATWS.
This equipment must be designed to perform its function in a reliable manner and be independent (from sensor output to the final actuation device) from the existing reactor trip system.
In accordance with 10 CFR 50.62(d), the applicant has submitted a schedule for implementing the requirements of the ATWS rule.
In a letter dated October 10, 1985, the applicant stated that the ATWS imple-mentation design for Braidwood Unit 1 is scheduled to comply with the require-ments of the ATWS rule before startup following the first refueling outage.
Implementation for Braidwood Unit 2 will occur before a license authorizing operation above 5% of rated power for that unit is issued.
The staff has not completed its review of the Braidwood design for compliance with the ATWS rule; however, staff review and approval are not a requirement for plant licensing.
The staff has reviewed the Westinghouse Topical Report Braidwood SSER 2 15-1
4 i
WCAP-10858, "AMSAC Generic Design Package," and has concluded that the generic designs presented in WCAP-10858 adequately meet the requirements of 10 CFR 50.62 and follow the staff's review guidelines.
This is Confirmatory Issue B(14).
t Table 15.1 Radiological consequences of design-basis accidents (revised from SER)
Exclusion area low population i
Postulated accident Thyroid Whole body Thyroid Whole body Loss of coolant:
Containment leakage l
0-2 hr 141 4
0-8 hr 31 0.9 8-24 hr 12 0.2 24-96 hr 11 0.1 96-720 hr 13 0.1 Total containment leakage 141 4
67 1.3
)
ECCS component leakage 1
<0.01 1
<0.01 Total 142 4
68
- 1. 3 Steamline break outside secondary containment:
j Long-term operation case j
(DEI-131 at 1 pCi/gm) 10
<1.0
- 2. 6
<1.0 Short-term operation case (DEI-131 at 60 pCi/gm) 13
<1.0 2.5
< 1. 0 Control rod ejection:
Containment leakage pathway 41
< 1. 0 47
< 1. 0 Secondary system release pathway 36
<1.0 0.4
<1.0 Fuel-handling accident in fuel-handling area 40 0.5 4.2
<0.1 I
Small line break 4.7
<0.1 0.5
<0.1 i
Steam generator tube rupture:
Long-term operation case j
(DEI-131 at 1 pCi/gm) 9
<0.1 1.5
<0.1 Short-term operation case j
(DEI-131 at 60 pCi/gm) 50
<0.1 6.7
<0.1 i
l Braidwood SSER 2 15-2 1
i
- - ~
i Table 15.2 Assumptions used in the calculation of loss of-coolant accident doses (revised from SER) i Parameter and unit of measure Quantity Containment leakage Power level, MWt 3,565 Operating time, yr 3
Fraction of core inventory available for containment leakage, %
Iodine 25 Noble gases 100 Initial iodine composition in containment, %
Elemental 91 Organic 4
Particulate 5
Containment leak rate, %/ day 0-24 hr 0.1 After 24 hr 0.05 Containment volume, ft3 Sprayed volume 2.35 x 108 Unsprayed volume 4.1 x 105 Containment mixing rate from cooling fan operation, cfm 180,000 Containment spray system Maximum elemental iodine decontamination factor 100 Spray removal coefficients, hr 1 Elemental iodine 10 Particulate iodine 0.45 Organic iodine 0
Relative concentration values, sec/m3 0-2 hr at the exclusion area boundary 5.6 x 10 4 0-8 hr at the low population zone (LPZ) boundary 5.9 x 10 5 8-24 hr at the LPZ boundary 4.4 x 10 5 24-96 hr at the LPZ boundary 2.3 x 10 5 96-720 hr at the LPZ boundary 9.4 x 10 8 ECCS leakage outside containment Power, MWt 3,565 Sump volume, gal 484,000 Flash fraction 0.1 Leak rate, gph (twice the maximum operational leakage defined in FSAR Table 15.6-15a) 2.1 Leak duration, hr 720 Delay time, hr 0.50 Filter efficiency for iodine, %
Elemental and particulate 95 Organic iodine 95 Braidwood SSER 2 15-3
~
\\ \\
Table 15.6 Assumptions used for estimating the radiological consequences following a postulated fuel handling accident (revised from SER) x Parameter and unit of measure Quantity Power level, MWt-3,565 Number of' fuel rods damaged 314 Total number of fuel rods in core 60,602 Radial peaking factor of damaged ro'd 1.65 Shutdown time, hour 100 Inventory released from damaged rods (iodines and noble gases), %
10 Pool decontamination factors Iodin 9 100 Noble gases 1
Iodine fractions released from pool, %
Elemental 75 Organic 25 Iodine removal efficiencies for ABGTS (spent fuel pool area), %
Elemental 90 Organic 30 3
x/Q values, sec/m 0-2 hr at 485 m (exclusion area boundary) 5.6 x 10 4 0-8 hr at 1,810 m (low population zone) 5.9 x 10 5 l
P l
Braidwood SSER 2 15-4 l-
.m__,
i
)
18 HUMAN FACTORS ENGINEERING 18.2 Main Control Room and Remote Shutdown Panel The SER referenced the applicant's Preliminary Design Assessment (PDA) dated November 12, 1981; the staff's onsite Control Room Design Review / Audit (CRDR/A) at Byron Station of November 1981; and the CRDR/A report transmitted to the applicant January 11, 1982.
The applicant developed resolutions to human engin-eering discrepancies (HEDs) applicable to both Byron and Braidwood and proposed schedules for implementation of control room improvements.
These were trans-mitted to the staff by letter dated May 9, 1983.
All reports were reviewed by the staff and several issues were clarified with the applicant during a meeting in Bethesda, Maryland, on October 26, 1983.
A subtittal by the applicant docu-menting the clarifications was transmitted by letter dated January 6, 1984.
All but one HED were resolved.
The one unresolved iteshake&%c mivrai. ion of the range and volume controls p%e struWe-range nuclear instrument from the nuclear instrument cabinet IPM07J to the main control board IPM05J where they are needed during startup.
The ap-plicant had proposed to relocate the controls from panel 2PM07J to 2PM05J at Byron Station before its preoperational test, and if the test results indicated no technical problems, such as electrical noise interference, the change would be made permanent.
The applicant, in its letter of October 2,1986, stated that the test was successful and the controls have been relocated at Braidwood Unit 1 and will be relocated at Braidwood Unit 2 before its fuel load.
In addition, by letter dated September 30, 1983, the applicant committed to con-duct a PDA on Braidwood site-specific items which included site-specific panels, and environmental systems.
The findings, proposed corrective actions, and imple-mentation schedules would be reported to the staff for review and approval at least 120 days before issuance of the operating license.
In its letter dated June 4, 1986, the applicant provided a list of HEDs and pro-posed implementation schedules for corrective actions applicable to the Braidwood control room.
It was unclear from the list which HEDs applied to the Byron /
Braidwood identical panels and instrumentation, and which were applicable to the Braidwood site-specific panels and instrumentation.
Since panel differences have been described as minimal, the staff finds it acceptable that the site-specific reviews should be included in the detailed control room design review (DCRDR) and reported in the DCRDR Summary Report due for submittal December 1, 1986.
In the letter of June 4,1986, the applicant further committed to take environ-mental measurements before plant criticality.
Since the staff would prefer that these measurements be taken under more realistic conditions than during construc-tion, this schedule is satisfactory.
The staff requires that the applicant sub-mit for staff review and approval, the report of the environmental survey includ-ing a description of the survey findings along with any proposed corrective actions and implementation schedules as part of the DCRDR Summary Report.
N Braidwood SSER 2 18-1
The staff is satisfied with all other proposed control room improvements and the schedule for implementation. The staff concludes that, with these improve-ments, the potential for operator error leading to serious consequences as a result of human factors considerations in the control room will be sufficiently low to permit safe operation of Braidwood Unit 1.
This completes the prelicensing staff evaluation of the Braidwood control room and the preliminary design assessment portion of TMI Action Plan Item I.D.1.
The plant must still be subjected to a DCRDR.
Requirements for the DCRDR are identified in Supplement 1 to NUREG-0737, " Requirements for Emergency Response Capability," dated December 17, 1982.
In addition, the DCRDR for Braidwood must address all PDA issues which the staff agreed could be postponed until that re-view. This shall include the environmental survey. The applicant, in its let-ter of September 24, 1984, has committed to submitting a Summary Report of the DCRDR by December 1, 1986.
In its letter of October 10, 1986, the applicant committed to including the report of the environmental survey in the DCRDR Summary Report. This is Outstanding Item A(8).
18.3 Safety Parameter Display System All holders of operating licenses issued by the Nuclear Regulatory Commission (licensees) and applicants for an operating license (OL) must provide a safety The parameter display system (SPDS) in the control rooms of their plants.
Commission-approved requirements for the SPDS are defined in Supplement 1 to NUREG-0737.
The purpose of the SPDS is to provide a concise display of critical plant var-iables to control room operators to aid them in rapidly and reliably determin-ing the safety status of the plant.
NUREG-0737, Supplement 1, requires licen-sees and applicants to prepare a written safety analysis describing the basis on which the selected parameters are sufficient to assess the safety status of each identified function for a wide range of events, including symptoms of severe accidents.
Licensees and applicants shall also prepare an implementa-tion plan for the SPDS which contains schedules for design, development, in-stallation, and full operation of the SPDS as well as a design Verification /
Validation (V/V) Plan.
The staff conducted an audit of the Byron SPDS on July 24-26, 1985, and sent the results of its audit to the applicant in a letter dated October 30, 1985.
The audit verified that the design of the Byron SPDS, which is identical to the Braidwood SPDS, should meet the requirements of Supplement 1 to NUREG-0737.
However, the audit team noted there were several problems. The corporate V/V project for the Byron SPDS had not been fully developed at the time of the audit.
The applicant should submit the corporate V/V report so that the staff can complete its evaluation of this issue.
The audit team also noted three human engineering problems:
(1) there was no clear way of determing whether the wide-range or narrow-range iconic display was on the screen; (2) the red alarm bars at the end of each iconic spoke were dif ficult to detect; and (3) the wide-range steam generator level spoke did not cover the full range at plant operation.
The applicant attempted to correct these problems.
Inspection Report No. 50-454/86021(DRP), dated July 15, 1986, stated How-that the applicant had satisfactorily corrected the last two discrepancies.
ever, the applicant's attempt to distinguish between the wide-and narrow-range Braidwood SSER 2 18-2
4 displays by coloring solid the center of the wide-range display was found unacceptable.
It is the staff's position that identification of the wide-range and narroe-range displays should be clear, concise, and unambiguous in order to satisfy the.'equirements of Supplement 1 to NUREG-0737.
The applicant, it. its letter of October 10, 1986, has committed to submitting the Verification / Validation Report and correcting the identification of the wide-range and narrow-range displays before startup following the first refuel-ing outage of Braidwood Unit 1.
This is Outstanding Item A(9).
i 1
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Braidwood SSER 2 18-3 1
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APPENDIX A CONTINUATION OF CHRONOLOGY OF NRC STAFF RADIOLOGICAL SAFETY REVIEW 0F BRAIDWOOD STATION, UNITS 1 AND 2 August 5, 1986 Letter from applicant concerning essential services water system.
August 8, 1986 Letter from applicant concerning National Fire Protection Association Code deviations.
August 8, 1986 Letter from applicant concerning preservice inspection.
August 13, 1986 Letter from applicant concerning comments on Proof and Review Technical Specifications.
4 August 15, 1986 Letter from applicant concerning Braidwood Licensed Operator Hot Participation Experience Risum6s.
August 15, 1986 Letter from applicant concerning Proof and Review Technical Specification changes.
August 15, 1986 Letter from applicant concerning description of Appendix R conformance as provided by Appendix A5.7 of the Byron /
Braidwood fire protection report.
August 19, 1986 Letter from applicant concerning Office of Inspection and Enforcement (IE) Information Notices 86-02 and 86-03,
" Compliance With the EQ Rule," and "Information on Seismic and Dynamic Qualification," respectively.
August 19, 1986 Letter from applicant concerning integrated leakage rate test results.
August 20, 1986 Letter from applicant concerning Preservice/ Inservice Inspection Pump and Valve Program.
Comparison is made to Byron Unit 1 program.
August 21, 1986 Letter from applicant transmitting supplemental information for exceptions to FSAR Appendix A Regulatory Guides 1.52 and 1.140.
l August 21, 1986 Letter from applicant to NRC Region III transmitting a response to Inspection Report Nos. 50-456/86-021 and 50-457/86-019.
August 22, 1986 Letter from applicant concerning masonry walls.
August 22, 1986 Letter to applicant concerning final Draft Technical Specifications for Braidwood Station, Units 1 and 2.
Braidwood SSER 2 1
Appendix A
- - _ - _,. = - _ - _
August 22, 1986 Letter from applicant concerning emergency plan exercise.
August 22, 1986 Letter from applicant transmitting supplemental response to supolemental safety evaluation report concerning emer-gency planning for Braidwood.
August 22, 1986 Letter from applicant transmitting response to FSAR Question 010.65.
August 26, 1986 Letter from applicant transmitting the Interim Operation Plan for the auxiliary building ventilation system.
August 29, 1986 Letter from applicant concerning fuel protection for Byron and Braidwood Stations, Units 1 and 2, and transmitting application for Amendment to Facility Operating License, NPP-37, Appendix A, Technical Specifications Byron Unit 1.
1 September 2, 1986 Letter from applicant transmitting a report on the inspec-tion of cast stainless steel component welds with relief i
requests.
September 3, 1986 Letter from applicant transmitting revised pages of Braidwood and Dresden Station annexes.
September 3, 1986 Letter from applicant transmitting revised pages of Braidwood and Dresden Station annexes.
September 5, 1986 Letter from applicant concerning Braidwood Station, Unit 1, transmitting request pursuant to 10 CFR 50.57(c).
September 8, 1986 Letter from applicant concerning request pursuant to 10 CFR 50.57(c).
September 9, 1986 Letter from applicant concerning control room ventilation system Interim Operation Plan.
September 10, 1986 Letter from applicant transmitting responses to NRC open items from the Braidwood fire protection inspection in August 1986.
September 10, 1986 Letter from applicant transmitting supplemental information on IE Information Notice 84-90, " Environmental Effects of Main Steam Line Break Outside Containment."
September 10, 1986 Letter from applicant transmitting final Draft Technical Specification changes.
September 11, 1986 Letter from applicant concerning chloride content for Category I concrete structures at Braidwood Station.
September 15, 1986 Letter from applicant concerning final Draft Technical Specification on seismic instrumentation system.
l l
Braidwood SSER 2 2
Appendix A t
-..--.-n
September 16, 1986 Letter from applicant concerning pump and valve operability assurance.
September 16, 1986 Letter from applicant revising the expected fuel load date for Braidwood Station, Unit 1.
September 16, 1986 Letter from applicant concerning proposed Technical Specification for high energy line break (HELB) isolation sensors.
September 16, 1986 Letter from applicant concerning installation of seismically qualified 7300 Series cards at Braidwood Station, Unit 1.
September 16, 1986 Letter from applicant responding to NRC open items from the Braidwood fire protection inspection in August 1986 for Braidwood Station, Unit 1.
September 16, 1986 Letter from applicant concerning interim guidance on Emergency Planning Standard 10 CFR 50.47(b)(12) for Byron Station, Unit 2, and Braidwood Station, Units 1 and 2.
September 16, 1986 Letter from applicant concerning Emergency Operating Procedures.
September 16, 1986 Letter from applicant concerning emergency plan exercise.
September 16, 1986 Letter from applicant concerning preservice inspection (PSI) program revisions and clarifications for Braidwood Station, Unit 1.
September 19, 1986 Letter from applicant concerning deferral of limited aspects of the preoperational test program for Braidwood Station, Unit 1.
September 19, 1986 Letter from applicant concerning Braidwood Licensed Operator Hot Participation Experience Rssumss.
September 22, 1986 Letter from applicant concerning fire protection, August 1986 Audit Resolution of Issues, Braidwood Station; Unit 1.
September 22, 1986 Letter from applicant concerning final Draft Technical Specification changes.
September 25, 1386 Letter from applicant concerning Braidwood Station, Unit 1, environmental qualification in compliance with 10 CFR 50.49.
September 25, 1986 Letter from applicant concerning Technical Specifications.
September 25, 1986 Letter from applicant concerning fire protection, August 1986 audit, Resolution of Issues, Braidwood Unit 1.
Braidwood SSER 2 3
Appendix A 4
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September 29, 1986 Letter from applicant concerning Technical Specifications, seismic instrumentation.
September 29, 1986 Two letters from applicant concerning fire protection, August 1986 audit, Resolution of Issues, Braidwood Unit 1.
September 29, 1986 Letter from applicant concerning, Section 6.0, final Draft Technical Specification changes.
September 30, 1986 Letter from applicant concerning final NFPA deviation list.
September 30, 1986 Letter from applicant concerning fire protection, August 1986 audit, Resolution of Issues, Braidwood Unit 1.
September 30, 1986 Letter from applicant concerning Technical Specification changes.
October 1, 1986 Letter from applicant concerning satisfaction of licensing requirements, Braidwood Unit 1.
October J, 1986 Letter from applicant concerning Technical Specifications.
October 8, 1986 Letter from applicant concerning inservice inspection examination commitment on loop 1 steam generator and the pressurizer, Braidwood Station, Unit 1.
October 9, 1986 Letter from applicant concerning interim Technical Specifications to 5% power, test deferrals, Braidwood Unit 1.
October 9, 1986 Letter from applicant concerning an exemption from the requirements of 10 CFR 50, Appendix A, General Design Criteria 13 and 17.
October 10, 1986 Letter from applicant concerning preliminary design assessment of Braidwood Unit 1 control room.
October 10, 1986 Letter from applicant concerning safety parameter display system.
October 13, 1986 Letter from applicant concerning masonry walls.
October 13, 1986 Letter from applicant concerning deferral of limited aspects of the preoperational test program, Braidwood Unit 1.
Braidwood SSER 2 4
Appendix A
APPENDIX B BIBLIOGRAPHY Ingersoll-Rand, EAS-TR-7801-IR, Revision 0, " Pump Seismic Qualification Report,"
January 19, 1978.
Jamesbury Corporation, JCS82-02, Revision 2, February 14, 1983.
Limitorque, Generic Report B0058, January 11, 1980.
McDonnel Engineering Co., MES23, July 2, 1982.
i Sargent and Lundy Engineers, EM0003901, Revision 0, Seismic Report, June 1, 1976.
U.S. Nuclear Regulatory Commission, Generic Letter 86-10, " Implementation of Fire Protection Requirements," April 24, 1986.
--, NUREG-75/023, " Safety Evaluation Report on the Byron Station, Units 1 and 2, and the Braidwood Station, Units 1 and 2," April 1975.
--, NUREG-0588, " Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment," November 1979.
--, NUREG-0660, "NRC Action Plan Developed As a Result of the TMI-2 Accident,"
Vol. 1, May 1980.
--, NUREG-0737, " Clarification of TMI Action Plan Requirements," November 1980.
--, NUREG-0737, Supplement 1, " Clarification of TMI Action Plan Requirements:
Requirements for Emergency Response Capability," January 1983.
--, NUREG-0800, " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants," LWR Edition, July 1981.
--, NUREG-0876, " Safety Evaluation Report Related to the Operation of Byron Station, Units 1 and 2," February 1982; Supplement 1, March 1982; Supplement 2, January 1983; Supplement 3, November 1983; Supplement 4, May 1984; Supplement 5, October 1984; Supplement 6, February 1985.
I
--, NUREG-0899, Guidelines for Preparation of Emergency Operating Procedures,"
August 1982.
--, NUREG-1026, " Final Environmental Statement Related to the Operation of Braidwood Station, Units 1 and 2," June 1984.
J l
Braidwood SSER 2 1
Appendix B 1
i
--, Office of Inspection and Enforcement, Bulletin 79-01B, " Environmental Qualification of Class 1E Equipment," January 14, 1980, and supplements dated February 29, September 30, and October 24, 1980.
1
--, Information Notice 79-22, " Qualification of Control Systems," September 14, i
1979.
--, Information Notice 84-90, " Main Steam Line Break Effect on Environmental j
Qualification of Equipment," December 7, 1984.
--, Inspection Report 50-454/85024 (DRS).
--, Inspection Report 50-456/86026 (DRSS), August 14, 1986.
1
--, Inspection Report 50-454/86021 (DRP), July 15, 1986.
i Westinghouse Electric Corp., WCAP-8754, Revision 1, June 1976.
--, WCAP-8860, Supplement 1, R. Land, " Mass and Energy Releases Following Steam Line Rupture."
--, WCAP-9863-A (Proprietary), " Rod Bank Worth Measurements Utilizing Bank j
Exchange," May 1982.
i
--, WCAP-10858, "AMSAC Generic Design Package."
--, WCAP-10961.
l INDUSTRY CODE AND STANDARDS American National Standards Institute 4
ANSI N509-1976 ANSI N510-1980 I
ANSI /ASME N509-1980 American Nuclear Society ANS 3.2/ ANSI N18.1976, " Administrative Controls and Quality Assurance for the Operational Phase of Nuclear Power Plants."
i American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section XI, " Rules for Inservice Inspection of Nuclear Power Plant Components."
i l
Division 1, Article IWP-1000 Section XI, 1980 Edition, Winter 1981 Addenda Subsection IWP-3400, 1980 Edition through Winter 1980 Addenda Braidwood SSER 2 2
Appendix 8 1
I 2
Institute of Electrical and Electronics Engineers Standard 317-1976, " Electrical Penetration Assemblies in Containment Structures for Nuclear Power Generating Stations."
l Standard 323-1974, " Standard for Qualifying Class IE Equipment for Nuclear Power Generating Stations," February 28, 1974.
Standard 344-1975, " Recommended Practices for Seismic Qualification of Class 1E Equipment for Nuclear Power Generating Stations," January 31, 1975.
National Fire Protection Association Code 12A Code 13, " Standard for Installation of Sprinkler Systems," 1976.
Code 14, " Standpipe and Hose System for Sizing, Spacing, and Pipe Support j
Requirements," 1974.
s i
Code 15, " Standard for Water Spray Fixed System," 1973.
1 Code 20, " Standard for the Installation of Centrifugal Fire Pumps," 1973.
Code 24 4
Code 72E, 1985 Edition Code 80 l
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i Braidwood SSER 2 3
Appendix 8
1 APPENDIX F NRC STAFF CONTRIBUTORS AND CONSULTANTS NRC STAFF Name Title Review branch
- Francis M. Akstulewicz, Jr.
Nuclear Engineer Integrated Safety Assess-ment Directorate (PWR-B)
Goutam Bagchi Section Leader Engineering (PWR-A)
Frederick H. Burrows Reactor Engineer Electrical / Instrumentation and Control Systems (PWR-A)
Margaret S. Chatterton Nuclear Engineer Reactor Systems (PWR-A)
Nilesh C. Chokshi Reliability and Risk Reliability and Risk l
Analyst Assessment (PWR-A)
Kenneth C. Dempsey Nuclear Engineer Engineering (PWR-A)
Richard J. Eckenrode Human Factors Electrical / Instrumentation l
Engineer and Control Systems (PWR-A)
Barry J. Elliot Materials Engineer Engineering (PWR-A)
Ronald N. Gardner Section Chief Region III Jchn J. Hayes, Jr.
Nuclear Engineer Plant Systems (PWR-A)
Donald E. Hickman Training Assessment Facility Operations (PWR-A)
Specialist (SRO)
George Johnson Materials Engineer Engineering (PWR-A)
Dennis J. Kubicki Fire Protection Electrical / Instrumentation Engineer and Control Systems (PWR-B)
George W. Lapinsky, Jr.
Engineering Facility Operations (PWR-A)
Psychologist James J. Lazevnick Electrical Engineer Electrical / Instrumentation and Control Systems (PWR-A)
Samson S. Lee Materials Engineer Engineering (PWR-A)
- Reflects reorganization since SER was issued.
Braidwood SSER 2 1
Appendix F
Name Title Review branch
- Chang-Yang Li Mechanical Engineer Plant Systems (PWR-A)
Jerry L. Mauck Electrical Engineer Electrical / Instrumentation and Control Systems (PWR-A)
Leonard N. Olshan Byron Project Manager Project Directorate #3 (PWR-A)
Frank R. Orr Reactor Systems Facility Operations (PWR-A)
Engineer Robert L. Perch Project Manager, Facility Operations (PWR-A)
Technical Specifications Madelyn M. Rushbrook Licensing Assistant Project Directorate #5 (PWR-A)
Amarjit Singh Mechanical Engineer Plant Systems (PWR-A)
Edmund J. Sullivan, Jr.
Section Leader Engineering (PWR-A)
Thomas M. Tongue Senior Resident Region III Inspector Harold Walker Mechanical Engineer Electrical / Instrumentation and Control Systems (PWR-A)
Seymour H. Weiss Section Leader Electrical / Instrumentation and Control Systems (PWR-A)
CONSULTANTS Name Organization D. Arnold Franklin Research Center B. Brown Idaho National Engineering Laboratory R. Gruel Battelle Dacific Northwest Laboratory A. Sutthoff Battelle Pacific Northwest Laboratory 1
- Reflects reorganization since SER was issued.
Braidwood SSER 2 2
Appendix F
APPENDIX I ERRATA TO BRAIDWOOD STATION, UNITS 1 AND 2, SAFETY EVALUATION REPORT AND ITS SUPPLEMENTS SER Change Page vii Change "3.11 Environmental Qualification of Safety-Related Electrical Equipment" to "3.11 Environmental Qualification of Electrical Equipment Important to Safety and Safety-Related Mechanical Equipment" l
Braidwood SSER 2 1
Appendix I
l r
i j
APPENDIX K COMMONWEALTH EDISON COMPANY l
BRAIDWOOO GENERATING STATION - UNIT 1 i
DOCKET NUMBER 50-456 SAFETY EVALUATION REPORT SUPPLEMENT PRESERVICE INSPECTION RELIEF REQUEST EVALUATION I.
INTRODUCTION j-j This section was prepared with the technical assistance of 00E contractors from the Idaho National Engineering Laboratory.
For nuclear power facilities whose construction permit was issued on or after July 1, 1974, 10 CFR 50.55a(g)(3) specifies that components shall meet the preservice examination requirements set forth in editions and addenda of Section XI of the ASME Boiler and Pressure Vessel Code applied to the construction of the particular component. The provisions of 10 CFR 50.55a(g)(3) also state that
]
components (i.ncluding supports) may meet the requirements set forth in subsequent editions and addenda of this Code which are incorporated by reference in 10 CFR 50.55a(b) subject to the limitations and modifications listed therein.
Requests for relief from the ASME Code Section XI requirements which the Applicant has determined to be impractical for systems and components at Braidwood Unit 1 were contained in the submittals dated July 31, 1986 and September 2, 1986. Clarifications and revisions to these relief requests, along with a new request for relief, were received in the September 18, 1986 submittal from the i
Applicant. The July 31, 1986 submittal contained a comparison and l
cross-reference between the Braidwood Unit I and the Byron Unit 1
}
relief requests and detailed any differences between the individual i
items. Of the sixteen relief requests submitted, four are plant j
specific for Braidwood Unit 1, eleven are common to both plants, and i
one was deleted by the Applicant.
These relief requests were all supported by information pursuant to 10 CFR 50.55a(a)(3).
Therefore, the staff evaluation consisted of reviewing the j
submittals to the requirements of the applicable Code edition and addenda and determining if relief from the Code requirements was justified.
II.
TECHNICAL REVIEW CONSIDERATIONS i
A. The construction permit for Braidwood Generating Station, Unit 1 I
was issued on December 31, 1975.
In accordance with 10 CFR 50.55a(g)(3), components (including supports) which are j
classified as ASME Code Class 1 and 2 have been designed and i
provided with access to enable the performance of required preservice examinations.
I i
Braidwood SSER 2 1
Appendix K 1
i
B. Verification of as-built structural integrity of the primary pressure boundary is not dependent on the Section XI preservice examination. The applicable construction codes to which the primary pressure boundary was fabricated contain examination and testing requirements which by themselves provide the necessary assurance that the pressure boundary components are capable of performing safely under all operating conditions reviewed in the FSAR and described in the plant design specification. As a part of these examinations, all of the primary pressure boundary full penetration welds were volumetrically examined (radiographed) and the system was subjected to hydrostatic pressure tests.
C. The intent of a preservice examination is to establish a reference or baseline prior to the initial operatic, of the facility. The results of subsequent inservice exan,irations can then be compared with the original condition to determine whether changes have occurred.
If the inservice inspection results show no change from the original condition, no action is *equired.
In the case where baseline data are not available, all flaws must be treated as new flaws and evaluated accordingly.
Section XI of the ASME Code contains acceptance standards which may be used as the basis for evaluating the acceptability of such flaws.
D. Other benefits of the preservice examination include providing redundant er alternative volumetric examination of the primary pressure boundary using a test method different from that employed !aring the component fabrication. Successful performance of the preservice examination also demonstrates that the welds so examined are capableaof subsequent inservice examination using a similar test method.
In the case of Braidwood Generating Station Unit 1, a large portion of the preservice examination required by the ASME Code was performed. Failure to perform a 100% preservice examination of the welds identified below will not affect the assurance of the initial structural integrity.
E. In some instances where the required preservice examinations were not performed to the full extent specified by the applicable ASME Code, the staff may require that these examinations or supplemental examinations be conducted as a part of the inservice inspection program. Requiring supplemental examinations to be performed at this time would result in hardships-or unusual difficulties without a compensating increase in the level of quality or safety. The performance of supplemental examinations, such as surface examinations, in areas where volumetric examination is difficult will be more meaningful after a period of operation. Acceptable preoperational integrity has already been established by similar ASME Code Section III fabrication examinations.
In cases where parts of the required examination areas cannot be effectively examined because of a combination of component design or current examination technique limitations, the development of Braidwood SSER 2 2
Appendix K
new or improved examination techniques will continue to be eval uated. As improvements in these areas are achieved, the staff will require that these new techniques be made a part of the inservice examination requirements for the components or welds which received a limited preservice examination.
Several of the preservice inspection relief requests involve examination of less than the required volume of a specific weld.
The inservice inspection (ISI) program is based on the examination of a representative sample of welds to detect generic service-induced degradation.
In the event that the welds identified in the PSI relief requests are required to be examined again, the possibility of augmented inservice inspection will be evaluated during review cf the Applicant's initial 10 year ISI program. An augmented program may include increasing the extent and/or frequency of examination of accessible welds.
III. EVALUATION OF RELIEF REQUESTS The Applicant requested relief from specific preservice inspection requirements in submittals dated July 31, 1986 and September 2, 1986. Clarifications and revisions to these relief requests, along with a new request for relief, were received in Applicant's September 18, 1986 submittal. The July 31, 1986 submittal contained a comparisen and cross-reference between the Braidwood Unit 1 and the Byron Unit i relief requests and detailed any differences between the individual items. Of the sixteen relief requests submitted, four are plant specific for Braidwood Unit 1, eleven are common to both plants, and one was deleted by the Applicant. All of these relief requests were supported by information pursuant to 10 CFR 50.55a(a)(3). Based on the ir. formation submitted by the Applicant and the staff's review of the design, geometry, and materials of construction of the components, certain preservice inspection requirements of the ASME Boiler and Pressure Vessel Code,Section XI have been determined to be impractical to perform. The Applicant has demonstrated that either (i) the proposed alternative would provide an acceptable level of quality and safety or (ii) compliance with the specified requirements of this section would result in hardships or unusual difficulties without a compensating increase in the level of quality and safety. Therefore, pursuant to 10 CFR 50.55a(a)(3), conclusions that these preservice requirements are impractical are justified as follows. Unless otherwise stated, references to the Code refer to the ASME Code,Section XI, 1977 Edition including Addenda through Summer 1978.
A. Relief Recuest No. INR-1 (Rev. 2), Examination Category C-F, Items C5.31 and C5.32, Class 2 Pressure Retaining Branen Connection Welds in the Main Steam, Safety injection, and Residual Heat Removal Systems Code Requirement:
Section XI, Table IWC-2500-1, Examination Category C-F, Items C5.31 and C5.32 require a 100!. surface Braidwood SSER 2 3
Appendix K i
examination on Class 2 pipe branch connections as defined by Figures IWC-2520-9 and IWC-2520-7, respectively.
Code Relief Request: Relief is requested from performing 100% of the Code-required surface examination on the following 24 branch connection welds:
Line Number Weld Numbers 1MS07AA-28" MS-04-25, 26, 27, 28, and 29 1MS07AB-28" 1MS-06-43, 44, 45, 46, and 47 1MS07AC-2B" 1MS-08-25, 26, 27, 28, and 29 t
1MS07AD-28" 1MS-02-37, 38, 39, 40, and 41 ISIO6BA-24" 1SI-24-23BA and 23ABA ISIO6BB-24" 15I-24-23BB and 23ABB Reason for Request: The above listed welds are inaccessible for a 100% surface examination due to the location of saddle plates over the pressure retaining welds. The Applicant has committed to a surface examination (liquid penetrant) and visual examination (leak test) on the saddle plate fillet welds in lieu of the required surface examinations for the cressure retaining welds listed above.
Staff Evaluation: This relief request is acceptable based on the following considerations:
1.
The branch pipe circumferential welds listed above have received radiographic volumetric examinations in accordance with the ASME Code Section III, Class 2, requirements during fabrication.
2.
The as-built component geometry makes the required Section XI examination impractical. Removal of the weldea reinforcement collars to make the area accessible for a preservice surface examination would result in hardship or unusual difficulties without a compensating increase in the level of quality and safety since the radiography performed during construction on the branch oipe circumferential welds verifies the preservice structural integrity. Based on the above, the staff has determined that performing a surface and visual examination of the saddle plate fillet weld is an acceptable alternative to the Code-required surface examination.
B. Relief Request No. 1NR-2 (Rev. 2), Examination Category B-J, Item B9.11, Class 1 Cast Stainless Steel Elbow-to-Cast Stainless Steel Pumo or Valve Welds (fitting-to-fitting)
Code Requirement:
Section XI, Table IWB-2500-1, Examination Category B-J, Item B9.11 requires a 100% surface and volumetric examination on Class 1 pressure retaining welds in piping 4 inch Braidwood SSER 2 4
Appendix K
and greater nominal pipe size as defined by Figure IWB-2500-8.
In addition, Appendix III, Supplement 7, requires that ultrasonic examination sensitivity be established using I.O. notches with a depth of 10*. wall thickness.
Code Relief Request: Relief is requested from the examination sensitivity being established to resolve 10% I.D. notches.
Relief is also requested from performing 100% of the Code-required volumetric examination on the following 8 elbow-to-pump or valve welds:
Line Number Weld Number Line Number Weld Number 1RC02AA-31 1RC-01-17 1RC01AA-29 1RC 4 1RCO2AB-31 1RC-02-31 1RC01AB-29 1RC-02-18 1RCO2AC-31 1RC-03-17 1RC01AC-29 1RC a 1RC02AD-31 1RC-04-18 1RC01AD-29 1RC 5 Reason for Reouest: The above listed welds join cast stainless elbows to either Cast pumps or cast valves (fitting-to-fitting).
Therefore, because of the large contoured wedge search units and the weld geometry, these welds experience axial and circumferential scanning limitations.
In addition, the optimized ultrasonic technique used for the statically cast stainless steel welds will only reliably detect flaws 25% or greater through the wall. This sensitivity is less than the 10% required by the Code.
Staff Evaluation:
This relief request is acceptable for PSI based on the following considerations:
1.
The subject welds received both volumetric examination by radicgraphy and surface examinations during fabrication in accordance with ASME Code Section III requirements.
2.
The staff has determined that the Applicant has developed, within the state-of-the-art, ultrasonic equipment and procedures for an effective ultrasonic examination of the cast stainless steel welds.
3.
The staff will continue to evaluate the development of new or improved procedures and will require that these enhanced p*ocedures be made a part of the inservice examination requi*ements.
Based on the above, the staff concludes that the Section III fabrication examinations, supplemented by the Section XI surface examination and the state-of-the-art Section XI volumetric examination, provides an acceptable level of preservice structural integrity 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.
l Braidwood SSER 2 5
Appendix K
C. Relief Request No. INR-3 (Rev. 2), Examination Category B-M-2, item B12.40, Class 1 Valve Bodies in the Reactor Coolant, pressurizer, Safety Injection, and Residual Heat Removal Systems Code Requirement: Section XI, Table IWB-2500-1, Examination Category B-M-2, Item B12.40 requires a visual (VT-1) examination of the valve body internal surfaces on valves exceeding 4 inch nominal pipe size. Examinations are limited to one valve within each group of valves that are of the same constructional design, e.g., globe, gate or check valve, manufacturing method, and that are performing similar functions in the system, e.g., containment isolation and system overpressure protection.
Code Relief Request: Relief is requested from disassembly of an operable valve for the sole purpose of performing a preservice visual examination (VT-1).
Reason for Request: The requirement to disassemble an operable valve for the sole purpose of performing a visual examination (VT-1) of the internal pressure retaining boundary is impractical and not commensurate to the increased safety achieved by this inspection. Class 1 valves are installed in their respective systems and many have completed functional testing. To disassemble these valves would provide a very small potential for increasing plant safety margins with a very disproportionate impact on expenditures of plant manpower and resources.
The Applicant states that the manufacturer's test data will be used in lieu of a preservice visual examination (VT-1). This includes documentation of examinations performed curing fabrication and installation of the subject valves. The examinations perforced may include volumetric, surface, and visual examinations, as required by ASME Section II, " Material Specifications for Ferrous and Nonferrous Materials."
The Applicant also states that the integrity of the pressure retaining boundary of both carbon steel and stainless steel valve bodies has been excellent. Class 1 valve bodies cannot historically be linked to breaching of the pressure retaining boundary in plant systems. Class 1 valves are subjected to numerous types of nondestructive testing and a rigorous quality assurance program during all stages of fabrication, storage, and installation. These valves have been found acceptable by the manufacturer, the ASME Authorized Nuclear Inspector, and Commonwealth Edison's Quality Assurance.
Staff Evaluation: The staff concludes that disassembly of these val.ves at this time solely to perform the required Section XI preservice visual examination of the internal surface is impractical. The staff has determined that the nondestructive examinations and functional tests performed to date significantly exceed the requirements of the Section XI visual examination and, therefore, these examinations and tests are an acceptable alternative to the Code requirement; relief is granted as reauested.
Braidwcod SSER 2 6
Appendix K
- 0. Relief Request No. INR-4 (Rev. 2), Examination Category B-0, Items B3.120 and B3.140, Inside Radius Sections on Pressurizer and Steam Generator Vessel Nozzles Code Requirement: Section XI, Table IWB-2500-1, Examination Category B-D Items B3.120 and B3.140 require a 100% volumetric examination on pressurizer and steam generator nozzle inside radius sections as defined by Figure IWB-2500-7.
Code Relief Reouest:
Relief is requested from performing the ultrasonic examiration of the Code-required volume of the following nozzle inner radii (14 items total):
Component Number Weld Numbers 1RC01BA Primary Nozzles (2) 1RC01BB Primary Nozzles (2) 1RC01BC Primary Nozzles (2) 1RC0180 Primary Nozzles (2) 1RYO15 RY-1, 2, 3, 4A, 48, and 4C Reason for Reouest: These nozzles all contain inherent geometric constraints and clad inner surfaces which limit the ability to perform meaningful volumetric examinations.
In an attempt to develop a technique to locate flaws in the nozzle inner radii area, a mock-up was used with little success. All pressure retaining components were hydrostatically tested to the requirements of ASME Code Section III.
Staff Evaluation: This relief request is acceptable for PSI based on the following considerations:
1.
I All pressure retaining components were hydrostatically tested to the requirements of ASME Code Section III prior to plant startuo.
2.
The staff review of the design configuration of the nozzle inner radius has concluded that the Code-required volumetric examination is impractical. The staff has determined that performing the ASME Section III hydrostatic test is an acceptable alternative.
3.
The staff will continue to evaluata the development of new or improved procedures and will require that these enhanced procedures be made part of the ISI examination requirements.
Braidwood SSER 2 7
Appendix K
i E. Relief Recuest No. INR-5 (Rev. 2), Examination Category B-F, Item B5.30, Steam Generator Nozzle-to-Safe End Welds Code Requirement: Section XI, Table IWB-2500-1, Examination Category B-F, Item B5.30 requires a 100% surface and volumetric examination on steam generator nozzle-to-safe end welds as defined by Figure IWB-2500-8.
In addition, Appendix III, Supplement 7, requires that ultrasonic examination sensitivity be established using I.D. notches with a depth of 10% wall thickness.
Code Relief Request: Relief is requested from the examination sensitivity being established to resolve 10% I.D. notches.
Relief is also requested from performing 100% of the Coce-required volumetric examination on weld IRC-02-23. The following welds are appliuable for this relief request:
Line Number Weld Number Line Number Weld Number 1RC01AA-29" 1RC 8 1RC02AA-31" 1RC 9 1RC01AB-29" 1RC-02-19 1RCO2AB-31" 1RC-02-23 1RC01AC-29" 1RC 8 1RC02AC-31" 1RC 9 1RC01AD-29" 1RC 9 1RC02AD-31" 1RC-04-10 Reason for Request: The above listed welds join cast austenitic stainless steel (SA-351-CF8A) to ca'st carbon steel (SA-261 GR-WWC) and are clad with austenitic stainless steel. The optimized ultrasonic technique used for the cast stainless steel side of these welds will only reliably detect flaws 25% or greater through the wall. This sensitivity is less than the 10%
required by the Code.
In addition, Weld IRC-02-23 had limited circumferential scans near the weld toe area. This limitation is due to the search units inability to maintain coupling while scanning over the weld / base metal transition.
Staff Evaluation: This relief request is acceptable for PSI based on *.he following considerations:
1.
The subject welds received both volumetric examination by radiography and surface examinations during fabrication in accordance with ASME Code Section III requirements.
2.
The staff has determined that the Applicant has developed, within the state-of-the-art, ultrasonic equipment and procedures for an effective ultrasonic examination of the cast stainless steel welds.
3.
The Code-required ultrasonic examination was completed from the carbon steel side of the weld and a "best effort" ultrasonic examination using state-of-the-art techniques was completed from the cast stainless side.
4.
The staff will continue to evaluate the development of new or improved procedures and will require that these Braidwood SSER 2 8
Appendix K
enhanced procedures be made a part of the inservice examination requirements.
Based on the above, the staff concludes that the Section III fabrication examinations, supplemented by the Section XI surface examination and the state-of-the-art Section XI volumetric examination, provide an acceptable level of preservice structural integrity 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.
F. Relief Request No. INR-6 (Rev. 2), Examination Category B-J, Item B9.11, Class 1 Wrouant Stainless Steel Pipe-to-Cast Stainless Steel Elbow, Pumo, or Valve Welds Code Requirement:
Section XI, Table IWB-2500-1, Examination Category B-J, Item B9.11 reauires a 100% surface and volumetric examination on Class 1 pressure retaining welds in piping 4 inch and greater nominal pipe size as defined by Figure IWB-2530-8.
In addition, Appendix III, Supplement 7, requires that ultrasonic examination sensitivity be established using I.D. notches with a depth of 10% wall thickness.
Code Relief Reauest: Relief is requested from the examination sensitivity being established to resolve 10% I.D. notches.
Relief is also requested from performing 100% of the Code-required volumetric examination on the following welds listed with an asterisk:
Pice-to-Elbow helds Line Number Weld Numbers 1RC02AA-31" 1RC-01-10, 11,'12, 16 1RC02AB-31" 1RC-02-24, 25, 26, 30 1RC02AC-31" 1RC-03-10, 11*, 12, 16 1RC02AD-31" 1RC-04-11, 12, 13*, 17 1RC03AA-27.5" 1RC-01-30 1RC03AB-27.5" 1RC-02-12 1RC03AC-27.5" 1RC-03-31 1RC03AD-27.5" 1RC-04-31 Pipe-to-Valve Welds Line Number Weld Numoer 1RC01AA-29" 1RC 3 1RC01AB-29" 1RC-02-17*
1RC01AC-29" 1RC 3*
1RC01AD-29" 1RC 4*
1RC03AA-27.5" 1RC-01-22, 23*
1RC03AB-27.5" 1RC 5*, 6*
1RC03AC-27.5" 1RC-03-22, 23 1RC03AD-27.5" 1RC-04-23, 24*
Braidwood SSER 2 9
Appendix K
Pipe-to-Pump Welds Line Number Weld Number 1RC03AA-27.5" 1RC-01-18*
1RC03AB-27.5" 1RC 1*
1RC03AC-27.5" 1RC-03-18*
1RC03AD-27.5" 1RC-04-19*
Reason for Request: The above listed welds join cast stainless steel components to wrought stainless steel pipe. The optimized ultrasonic technique used for the statically cast stainless steel welds will only reliably detect flaws 25% or greater through the wall. This sensitivity is less than the 10% required by the Code.
In addition, because of the large contoured wedge search units and the weld geometry, those welds listed with an asterisk, experienced axial and circumferential scanning limitations.
Staff Evaluation: This relief request is acceptable for PSI based on the following considerations:
1.
The subject welds received both volumetric examination by radiography and surface examinations during fabrication in accordance with ASME Code Section III requirements.
2.
The staff has determined that the Applicant has developed, within the state-of-the-art, ultrasonic equipment and procedures for an effective ultrasonic examination of the cast stainless steel welds.
3.
The Code-required ultrasonic examination was completed from the wrought stainless pipe side of the weld and a "best effort" ultrasonic examination using state-of-the-art techniques was completed from the cast stainless side.
4.
The staff will continue to evaluate the development of new or improved procedures and will require that these enhanced procedures be made a part of the inservice examination requirements.
Based on the above, the staff concludes that the Section III fabrication examinations, supplemented by the Section XI surface examination and tne state-of-the-art Section XI volumetric examination, provide an acceptable level of preservice structural-integrity 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 2 10 Appendix K
G. Relief Reauest No. INR-7 (Rev. 2), Examination Category B-J, Item B9.11, Reactor Pressure Vessel Safe End-to-Cast Stainless Steel Elbow Welds Code Reauirement:
Section XI, Table IWB-2500-1, Examination Category B-J, Item B9.11 requires a 100% surface and volumetric examination on Class 1 pressure retaining welds in piping 4 inch and greater nominal pipe size as defined by Figure IWB-2500-8.
In addition, Appendix III, Supplement 7, requires that ultrasonic examination sensitivity be established using I.O. notches with a depth of 10% wall thickness.
Code Relief Request:
Re'ief is requested from the examination sensitivity being established to resolve 10% I.D. notches.
Relief is also requested from performing 100% of the Code-required volumetric examination on the following welds:
Elbow-to-Safe End Welds Line Number Weld Numbers 1RC03AA-27.5" 1RC-01-31 1RC03AB-27.5" 1RC-02-13 1RC03AC-27.5" 1RC-03-32 1RC03AD-27.5" 1RC-04-32
Reason for Request
The above listeo welds all join cast austenitic stainless steel elbows to reactor vessel nozzle safe ends. The optimized ultrasonic technique used for the statically cast stainless steel elbow welds will only reliably detect " laws 25% or greater through the wall.
This sensitivity is less than the 10% required by the Code.
In addition, because of the large contoured wedge search units and the weld geometry, these welds experienced axial and circumferential scanning limitations.
Staff Evaluation: This relief request is acceptable for PSI based on the following considerations:
1.
The subject welds received both volumetric examination by radiography and surface examinations during fabrication in accordance with ASME Code Section III requirements.
2.
The staff has determined that the Applicant has developed, within the state-of-the-art, ultrasonic equipment and procedures for an effective ultrasonic examination of the cast stainless steel welds.
3.
The Code-required ultrasonic examination was completed from the non-cast (safe end) side of the weld and a "best effort" ultrasonic examination using state-of-the-art techniques was completed from the cast Stainless side.
4.
The staff will continue to evaluate the development of new or improved procedures and will require that these Braidwood SSER 2 11 Appendix K
enhanced procedures be made a part of the inservice examination requirements.
Based on the above, the staff concludes that the Section III fabrication examinations, supplemented by the Section XI surface examination and the limited state-of-the-art Section XI volumetric examination, provide an acceptable level of preservice structural integrity 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.
H. Relief-Request No. INR-8 This request for relief has been deleted by the Applicant.
I. Relief Request No. INR-9 (Rev. 3), Examination Category B-A, Pressure Retaining Welds in the Reactor Vessel Code Requirement: Section XI, Table IWB-2500-1 Examination Category B-A, Items Bl.11, 81.21, Bl.30, and Bl.40 require a 100%
volumetric examination of the subject reactor pressure vessel welds as defined by Figures IWB-2500-1, 3, 4, and 5 respectively.
Code Relief Request: Relief is requested from performing preservice volumetric examination of the inaccessible portions of the following four reactor pressure vessel welds:
Weld Numbers RV-001 RV-002 RVCH-001 RVCH-002 Reason for Reouest: Configuration, permanent attachments, and/or structural interferences prohibit 100% ultrasonic examination coverage of the required volume.
1.
Lower Disk-to-Dutchman weld RV-001 has 58 instrument tubes which physically obstruct the search unit and/or search unit positioning device.
2.
Lower Shell Course-to-Dutchman weld RV-002 has six core support guide lugs welded tc the interior surface of the reactor vessel approximately 3.50 inches above the weld.
These lugs restricted the automated inspection tool from inspecting the required volume in the areas of the lugs.
All of the weld and heat affected zone received 100%
coverage from at least one direction, however tne required base metal was not fully inspected in the area of the core support guide lugs.
Braidwond SSER 2 12 Appendix K
t 3.
Closure Head Flange-to-Dutchman Forging weld RVCH-001 has a flange which physically obstructs the ultrasonic transducer from performing the required scan.
Part of the three larger lifting lugs also fall in the required scan area.
4.
Dutchman Forging-to-Closure Head Dome weld RVCH-002 has 6 lifting lugs which physically obstruct the ultrasonic transducer from performing the required scan.
Staff Evaluation: The staff has reviewed the above information, including the figures submitted with Relief Request INR9 which show the areas receiving the required examination and the areas which could not be examined, and concluded that this relief request is acceptable based on the following considerations:
1.
A significant portion of the above listed welds received the preservice volumetric examination in accordance with the ASME Code Section XI.
Completion of the remaining portions of the required examination is impractical and would result in undue hardship without a compensating increase in safety.
2.
All of the reactor pressure vessel welds passed volumetric examinations during fabrication in accordance with ASME Code Section III.
3.
All of these welds will be subjected to a system pressure test in accordance with Section XI requirements.
Therefore, the limited Section XI ultrasonic examination, the radiography performed during fabrication and the hydrostatic test provide an acceptable level of preservice structural integrity.
J. Relief Request No. INR-10 (Rev. 01, Examination Category C-A, Item C1.10, Pressure Retaining We:ds in the Chemical and Volume Control, Horizontal Letdown Heat Exchanger Code Recuirement:
Section XI, Table IWC-2500-1, Examination Category C-A, Item C1.10 requires a 100% volumetric examination on the circumferential pressure retaining shell welds in Class 2 pressure vessels as defined by Figure IWC-2520-1.
Code Relief Recuest:
Relief is requested from performing the Code-required ultrasonic circumferential scan for reflectors transverse to the weld seam on weld number HLHXC-01.
Reason for Request
The circumferential scan could not be performed due to flange bolting extending over the weld crown.
An ASME Code Section XI ultrasonic examination for reflectors parallel to the weld seam and an alternative surface examination will be performed.
Braidwood SSER c 13 Appendix K
Staff Evaluation: This relief request is acceptable for PSI based on the following considerations:
1.
The subject weld received radiographic examination and a hydrostatic test during fabrication in accordance with ASME Code Section III requirements.
2.
The staff has reviewed the design configuration of the flange, the wall thickness of the shell, and the condition of the weld crown and has determined that disassembly of the bolting solely for the purpose of PSI examinati;n would result in hardships or unusual difficulties without a 5
compensating increase in the level of quality and safety.
The staff has also determined that the radiography, surface examination, and limited ultrasonic examination established an acceptable level of preservice structural integrity.
3.
However, in the event the bolted connection is disassembled for repair or maintenance during service, the staff will require that the preservice examination be performed.
K. Relief Request No. INR-11 (Rev. 1), Examination Category C-A, Item C1.10, Chemical and Volume Control - Excess Letdown Heat Exchanger Shell Circumferential Weld Code Recuirement: Section XI, Table IWC-2500-1, Examination Category C-A, Item C1.10 requires a 100% volumetric examination on pressure retaining circumferential shell welds on Class 2 pressure vessels as defined by Figure IWC-2520-1.
Code Relief Request: Relief is requested from performing 100% of the Code-cequired preservice volumetric examination on the inaccessible portion of the Excess Letdown Heat Exchanger Weld Number ELHXC-03.
Reason for Request: Ultrasonic examination of weld ELHXC-03 was limited for approximately 70% of the weld length due to four branch connections welded to the vessel. A liquid penetrant test will be performed as an alternative test.
Staff Evaluation: This relief request is acceptable for NSI '
based on the following considerations:
1.
An alternative surface examination was performed in addition to the limited ultrasonic examination.
2.
The ASME Section III radiographic and hydrostatic test, along with the limited Section XI ultrasonic examination and alternative surface examination, demonstrate an acceptable level of preservice structural integrity.
Braidwood SSER 2 14 Appendix K
L. Relief Recuest No. INR-12 (Rev. 2), Examination Category C-8, Item C2.20, Pressure Retaining Class 2 Nozzle Welds in the Steam Generator and Residual Heat Exchanger Code Reouirement:
Section XI, Table IWC-2500-1, Examination Category C-B, Item C2.20 requires a 100% surface and volumetric examination on Class 2 nozzles in vessels over 1/2 inch nominal wall thickness as defined by Figure IWC-2520-4.
This figure requires volumetric examination of the nozzle-to-vessel weld and, for pipe aizqs over 12 inches, an examination of the nozzle inner radii.
Code Relief Request:
Relief is requested from performing the Code-required surface and volumetric examination on the following 10 steam generator and residual heat exchanger nozzle welds:
Comoonent No.
Nozzle No.
Restricted Exam 1RC01BA SGN-02, 03 Inner radii 1RC01BB SGN-02, 03 Inner radii 1RC01BC SGN-02, 03 Inner radii 1RC01BD SGN-02,.03 Inner radii 1RH02AB RHXN-01, 02 Inner radii and nozzle to vessel weld Reason for Reauest: The nozzles listed above contain inherent geometric constraints which limit the ability to perform meanirigful ultrasonic examinations. The main steam nozzles (SGN-03's) have an internal multiple venturi-type flow restrictor.
This design does not have a nozzle inner radii as described in Figure IWC-2520-4.
This nozzle has seven individual inner radii, corresponding to each venturi, none of which could be examined by ultrasonic examination. The main feedwater nozzles (SGN-02's) also have an internal multiple venturi-type flow restrictor but have a thermal sleeve in addition. This design could not be examined due to the geometry of the nozzles' internal design.
The Applicant reports, however, that the increased safety margin afforded by these nozzles makes them a desirable part of plant design.
The Residual Heat Removal Heat Exchanger is aporoximately 7/8 inch nominal wall thickness with nozzles of 14 inch diameter and approximately 3/8 inch nominal wall thickness.
The configuration is best characterized as a fillet welded nozzle using an internal reinforcement pad and, thereby, is not analogous to a full penetration butt welded nozzle as shown in Figure IWC-2520-4.
In addition, the inner radius of the reinforcement pad would be representative of the nozzle inner radius required for inspection. The inherent geometric constraints of the nozzle design prevent the performance of the required ultrasonic examinations of tne nozzle-to-shell weld and the nozzle inner radius.
Braidwood SSER 2 15 Appendix K
Ultrasonic examination of the above listed nozzle inner radii is l
not practicable and the inner radii are not accessible to direct contact for surface examination or even remote visual examination. However, these nozzles have been examined at the point of attachment to the vessel by radiography per ASME l
Section III, and by ultrasonic examination per ASME Section XI.
l In addition, a system hydrostatic test, at 125% of the design pressure, has been performed in accordance with ASME Section III.
Staff Evaluation: This relief request is acceptable for PSI based on the following considerations:
1.
The subject weld area received radiographic examination and a hydrostatic test during fabrication in accordance with ASME Code Section III requirements. An ultrasonic examination has been performed on the nozzle-to-vessel welds per ASME Code Section XI requirements.
2.
The staff review of the design configuration of the nozzle inner radius has concluded that the
)
Code-required volumetric examination is impractical.
i The staff has determined that the ASME Section III examinations demonstrate,an acceptable level of preservice structural integrity.
M. Relief Request No. 1NR-13 (Rev. 2), Examination Category B-J, Item B9.11, Circumferential Pressure Retaining Welds in Class 1 Piping Code Requirement:
Section XI, Table IWB-2500-1, Examination Category B-J, Item B9.11 requires a 100% volumetric and surface examination on circumferential pressure retaining welds in Class 1 piping of 4 inch and greater nominal pipe size as defined by Figure IWB-2500-8.
Code Relief Request: Relief is requested from performing the Code-required volumetric and/or surface examination on the following four welds:
Line Number Weld Number Rel Reg'd Interference IRC21AD-8" 1RC-15-13 vol/sur Permanent Restraint 1RY018-6" 1RC-16-1 vol Nozzle - Reducer 1SIA48-8" 1SI-02-35 vol Reducer - Valve 15I040-8" 1S1-11-22 vol Reducer - Valve Reason for Recuest: Weld No. IRC-15-13 is encased in a permanent whip restraint making a preservice ultrasonic or surface examination impractical. The geometry of welds 1RC-16-1, 1S1-02-35, and 1SI-11-22 prohibit meaningful ultrasonic examination. The Applicant also reports that in performing a visual examination (leak test) on IRC-15-13 and surface examinations on IRC-16-1, 1S1-02-35, and 151-11-22, an acceptable level of structural integrity for system operation is provided.
Austenitic Type 304 stainless steel possesses a high degree of Braidwood SSER 2 16 Appendix K
b toughness. Crack propagation through this material is slow and leak testing is a viable method of identifying flaws prior to weld failure in this type of material.
In addition, acceptable surface and volumetric (radiography) examinations were performed on these welds during construction.
Staff Evaluation: This relief request is acceptable for PSI based on the following considerations:
1.
The subject welds received radiographic and surface examinations during fabrication in accordance with ASME Code Section III requirements.
2.
The staff notes that complete examinations which met the requirements af ASME Code Section XI were performed on similar ucids using the same inspection techniques, equipment, and procedures as these uninspected welds. Since these welds will see the same operating and environmental conditions as the inspected welds, a reasonable assurance of the structural integrity of the welds for which relief is requested has be9n attained.
3.
All of these welds uill be subjected to a system pressure test in accordance with Section XI requirements.
The staff therefore concludes that the Section III fabrication examinations and the Section XI surface examination provide an acceptable level of preservice structural integrity 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.
N. Relief Recuest Nos. INR-14 (Rev. 3), Examination Category C-F, Pressure Retainino Welds in Class 2 Piping Code Requirement:
Section XI, Table IWC-2500-1, Examination Category C-F, Items C5.21 and C5.22 require a 100% surface and volumetric examination on pressure retaining welds in Class 2 piping over 1/2 inch nominal wall thickness as defined by Figure IWC-2520-7.
Code Relief Request: Relief is requested from performing 100% of tne Code-required preservice surface and/or volumetric examinations on the inaccessible portions of the following welds:
Weld Number Rel Reo'd Weld Number Rel Reo'd 1MS-02-34 vol 1MS-04-43 vol 1MS-05-7 vol/sur 1MS-06-40 vol 1MS-08-43 vol 1FW-02-08 vol/sur 1FW-03-11 vol/sur ISI-04-12 vol 15I-04-58 vol ISI-14BA-33B vol ISI-24BA-34 vol 1SI-24BB-28 vol/sur 1SI-24BB-30B vol ISI-24BB-30C vol 1SI-24BB-30A vol 151-26-1 vol 15I-26-61 vol Braidwood SSER 2 17 Appendix K
Reason for Recuest: Physical limitations due to geometric configuration of the welds (i.e. reducer-to-valve) or complete inaccessibility of the weld prevents the Code-required examination. The Applicant reports that the subject welds received the Code-required Section III examinations during fabrication.
Staff Evaluation: This relief request is acceptable for PSI based on the following considerations:
1.
The staff notes that complete examinations which met the requirements of ASME Code Section XI were performed on welds of similar configuration using the same inspection techniques, equipment, and procedures as the partially inspected or uninspected welds. Since the partially inspected or uninspected welds will see the same operating and environmental conditions as the inspected welds, a reasonable assurance of the structural integrity of the welds for which relief is requested has been attained.
2.
All of the subject welds received the ASME Code Section III examinations during fabrication.
Based on the above review, the staff has concluded that the Section III fabrication examinations, supplemented by the Section XI surface examination and the limited Section XI volumetric examination as apolicable, provide an acceptable level of preservice structural integrity 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.
O. Relief Request No. 1NR-15 (Rev. 3), Examination Category C-F, Items C5.11 and C5.12, Pressure Retaining Welds in Class 2 Pioing Code Requirement: Section XI, Table IWC-2500-1, Examination Category C-F, Items C5.11 and C5.12 require a 100% surface examination on pressure retaining welds in Class 2 piping with 1/2 inch or less nominal wall thickness as defined by Figure IWC-2520-7.
Code Relief Request: Relief is requested from performing 100% of the Code-required preservice surface examination on the following welds:
Line Number Weld Number 1RH02AA-8" 1RH-05-128 1RH02AA-8" 1RH-05-13 1RH02AA-8" 1RH-05-13A 1FWO87CB-6" 1FW-06-16 Braidwood SSER 2 18 Appendix K
Reason for Recuest: The Applicant reports that these welds are inaccessible for the Code-required preservice surface examination as the subject welds are located inside a missile wall. The Applicant also reports that the subject welds received the Code-required Section III examinations during fabrication.
Staff Evaluation: Paragraph IWB-2200 of ASME Code Section XI allows shop and field examinations to serve ia lieu of the on-site PSI examination provided that such examinations are conducted under conditions and with equipment and techniques equivalent to those which are expected to be employed for subsequent inservice examinations, and that the shop and field examination records are or can be documented and identified in a form consistent with those required in paragraph IWA-6000. The staff therefore concludes that if these conditions have been met, as reported by the Applicant, relief is not required.
P. Relief Recuest No. 1NR-16 (Rev. 0), Examination Category B-J, Item B9.11, Circumferential Pressure Retaining Welds in Class 1 Piping Code Requirement: Section XI, Table IWB-2500-1, Examination Category B-J, Item 89.11 requires a 100% volumetric and surface examination on circumferential pressure retaining welds in Class 1 piping of 4 inch and greater nominal pipe size as defined by Figure IWB-2500-8.
Code Relief Recuest:
Relief is requested from performing the Code-required volumetric examination on inaccessible portions on the following nineteen welds:
% of Weld Length Line No.
Weld No.
Not Examined 1RY01C-4" 1RC-16-2 16%
1RY01AA-4" 1RC-17-22 16%
1RYO3AA-6" 1RC-32-1 16%
1RY03BA-6" 1RC-32-6 16%
1RYO3AB-6" 1RC-32-7 16%
1RYO3BB-6" 1RC-32-12 16%
1RYO3AC-6" 1RC-32-13 16%
1RH01AB-12" 1RH-01-20 16%
1RC35AA-6" 15I-02-37 8%
1RC04AA-12" 1SI-02-42 8%
ISIO5DB-6" 1SI-06-10 16%
ISIO9BC-10" 151-09-4 41%
ISIOSDC-6" 1SI-10-11 8%
1SIO50C-6" 1SI-10-18A 8%
ISICSDC-6" 15I-10-22 16%
ISIO9BD-10" 1SI-13-1 8%
Braidwood SSER 2 19 Appendix K
Reason for Request: The above listed welds have interfering conditions on each side of the weld (i.e. elbow-to-flange welds, elbow-to-nozzle, elbow-to-tee, etc.) which can cause: poor coupling of the transducer, limited movement of the transducer, redirecting of the sound beam and, in some cases, complete restriction of a particular scan. These conditions sufficiently limit the axial scans so as to leave the above listed percentage of the weld length uninspected.
The Applicant reports that all of the subject welds received the Code-required surface examination for PSI in addition to the limited volumetric examination, and that these welds received the ASME Code Section III. radiographic examination during fabrication.
Staff Evaluation: This relief request is acceptable for PSI based on the following considerations:
1.
The staff notes that a significant percentage of the Code-required volumetric examination was completed and that all of the subject welds received radiographic examination during fabrication in accordance with ASME Code Section III requirements.
2.
Complete examinations which met the requirements 'of ASME Code Section XI were performed on similar welds using the same inspection techniques, equipment, and procedures as these partially inspected welds. Since these welds will see the same operating and environmental conditions as the inspected welds, a reasonable assurance of the structural integrity of the welds for which relief is requested has been attained.
3.
All of these welds will be subjected to a system pressure test in accordance with Section XI requirements.
The staff therefore concludes that the Section III radiographic examination, the Section XI surface examination, and the limited Section XI volumetric examination provide an acceptable level of preservice structural integrity 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.
IV.
CONCLUSIONS Based on the foregoing, pursuant to 10 CFR 50.55a(a)(3), the staff has determined that certain Section XI required preservice examinations are impractical. The Applicant has demonstrated that either (i) the proposed alternatives would provide an acceptable level of quality and safety or (ii) compliance with the requirements would result in hardships or unusual difficulties without a compensating increase in the level of quality and safety.
Braidwood SSER 2 20 Appendix K
The staff technical evaluation has not identified any practical method by which the existing Braidwood Generating Station Unit I can meet all the specific preservice inspection requirements of Section XI of the ASME Code. Requiring compliance with all the exact Section XI required Inspections would delay the startup of the plant in order to redesign a significant number of plant systems, obtain sufficient replacement components, install the new components, and repeat the preservice examination of these components.
Examples of components that would require redesign to meet the specific preservice examination provisions are:
the reactor pressure vessel, the letdown heat exchanger, the steam generators, and a number of the piping and component support systems.
Even after the redesigr efforts, complete compliance with the preservice examination requirements probably could not be achieved. However, the as-built structural integrity of the existing primary pressure boundary has already been established by the construction code fabrication examinations.
Based on the staff review and evaluation, it is concluded that the public interest is not served by imposing certain provisions of Section XI of the ASME Code that have been determined to be impractical. Pursuant to 10 CFR 50.55a(a)(3), relief is allowed from these requirements which are impractical to implement.
1 Braidwood SSER 2 21 Appendix K
FORM FM u 5 NvCLE Aa REGuLATon y COYY'55 ION t R E PCa t NyV8E R (Ass,,aed Or ItDC ### VO' %O
'8 8"r3 NUREG-1 C'12 BIBLIOGRAPHIC DATA SHEET Supplement No. 2 2
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TITLE AND Sustif LE
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Safety Evaluatio Report related to the operation of Braidwood Station, Units 1 and 2
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j'E^a OCT0fR 1986 AU T HQR e $1 F DATE POR T ISSUE D
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Division of PWR-Licensing A 5 OfficeofNuclearReactorRegu\\ation U. S. Nuclear Regulatory Commisbion s
Washington, D. C. 20555
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$UPPLE VE N T AR Y NOTE S Docket Nos. STN 50-546 and STN 50-547 A95T A AC 7 #100 words or wasJ Supplement No. 2 to the Safety Evaluation Repo.related to operation of the Braidwood Station, Units 1 and 2 addresses the items rel t g to the issuance of fcel loading and precriticality testing for Unit No. I of he aidwood Station.
The report relates to the application filed Commdswealth Edison Company for licenses to operate the Braidwood Station, Units 1 an 2 loca in Reed Township, Will County, Illinois.
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