ML20052D000
| ML20052D000 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 03/31/1982 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-0876, NUREG-0876-S01, NUREG-876, NUREG-876-S1, NUDOCS 8205060186 | |
| Download: ML20052D000 (30) | |
Text
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I NUREG-0876 Supplement No.1 Safety Evaluation Report related to the operation of Byron Station, Units 1 and 2 t
Docket Nos. STN 50-454 and STN 50-455 l
Commonwealth Edison Company I
1 U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation March 1982 se "%,
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NOTICE Availability of Reference Materials Cited in NRC Publications Most documents cited in N RC publications will be available from one of the following sources:
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- 1. The NRC Public Document Room,1717 H Street, N.W.
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$2.75 GPO Printed wpy price.
NUREG-0876 Supplement No.1
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Safety Evaluation Report related to the operation of Byron Station, Units 1 and 2 Docket Nos. STN 50-454 and STN 50-455 a
Commonwealth Edison Company U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation March 1982 y ** %,
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TABLE OF CONTENTS 1
INTRODUCTION AND GENERAL DESCRIPTION OF FACILITY...............
1-1 1.1 Introduction..............................................
1-1
- 1. 7 S umma ry o f O u ts ta nd i ng I s s u e s.............................
1-2 1
1.8 C o n f i rmato ry I s s u e s.......................................
1-2 i
.i 1.9 License Conditions........................................
1-2 i
3 DESIGN CRITERIA FOR STRUCTURES, SYSTEMS, AND COMPONENTS........
3-1 1
i 3.2 Classification of Structures, Components, and Systems.....
3-1 j
3.2.1 Seismic Classification.............................
3-l' 3.2.2 System Quality Group C1 2, ification................
3-2 i
3.5 Missile Protection........................................
3-4 i
3.5.3 Barrier Design Procedures..................
3-4 1
3.9 Mechanical Systems and Components.........................
3-4 3.9.2 Dynamic Testing and Analysis of Systems Components, and Equipment........................
3-4 3.9.3 ASME Code Class 1, 2, and 3 Components, Component Supports, and Core Support Structures..............
3-5 4
REACTOR.......................................................
4-1 4.4 Thermal-Hydraulic Design..................................
4-1 5
REACTOR COOLANT SYSTEM...........
5-1 5.2 Integrity of Reactor Coolant System.
5-1 5.2.1 Compliance with 10 CFR Part 50.55a.................
5-1 6
ENGINEERED SAFETY FEATURES.....................................
6-1 6.3 Emergency Core Cooling System.............................
6-1 t
J 6.3.5 Emergency Core Cooling System Performance l
Evaluation.......................................
6-1 9
AUXILIARY SYSTEMS..............................................
9-1 9.5 Other Auxiliary Systems...................................
9-1 i
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9.5.6 Emergency Diesel Engine Starting Systems...........
9 Byron SSER #1 i
TABLE OF CONTENTS (Continued)
Page 12 RADIATION PROTECTION.................
12-1 12.3 Radiation Design Features.
12-1 12.3.1 Facility Design Features....
12-1 13 CONDUCT OF OPERATIONS..
13-1 13.1 Organizational Structure of Applicant.............
13-1 7
13.1.2 Operating Organization.
13-1 13.6 Industrial Security................................
13-1 19 REVIEW BY THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS.
19-1 APPENDICES Page APPENDIX A Chronology of NRC Staff Radiological Safety Review.
A-1 APPENDIX F NRC Staff Contributors and Consultants.
APPENDIX G Report of the Advisory Committee on Reactor F-1 Safeguards..
C-1 APPENDIX H Errata to Byron Station Safety Evaluation Report.
H-1 LIST OF TABLES 4.1 Available Thermal Margin.
4-1 12.1 Component Decontamination Experience..
12-2 W
I Byron SSER #1 ii
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1 IkfRODUCTION AND GENERAL DESCRIPTION OF FACILTIY 1.1 Introduction
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The Nuclear Regulatory Commission's Safety Evaluation Report (NUREG-0876) in the matts of Commonwealth Edison Company's application to operate the Byron l
Station Units 1 and 2, was issued in February 1982.
At that time, the staff identified items that were not yet resolved with the applicant.
These items were categorized as:
1.
Outstanding items which needed resolution prior to the issuance of an operating license.
Items for which the staff had completed its review and had determined j
2.
positions for which there appeared to be no significant disagreement between the applicant and the staff.
Further information was needed, however, to confirm these positions.
3.
Items for which the staff had taken positions and would require implementa-tion and/or documentation after the issuance of the operating license.
These would be conditions to the operating license.
At its 263rd meeting on March 4, 1982, the Advisory Committee on Reactor Safe-guards completed its review of the application.
The Committee in its March 9, 1982 letter to Chairman Palladino of the NRC concluded that if due considera-tion is given to the items mentioned in its letter, and subject to the satis-factory completion of construction, staffing, and preoperational testing, there is reasonable assurance that the Byron Station Units 1 and 2 can each be operated at power levels up to 3425 thermal megawatts without undue risk to the health and safety of the public.
The purpose of this supplement to the Safety Evaluation Report is to provide the staff evaluation of the open items that have been resolved, to address changes to its Safety Evaluation which resulted from the receipt of additional information from the applicant, and to address those recommendations that are contained in the Advisory Committee on Reactor Safeguards letters of March 9, 1982.
That letter is included as Appendix G to this supplement to the Safety Evaluation Report.
The staff's response to the recommendation in the Committee's letter is given in Section 19 of this supplement.
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Copies of this SER supplement are available for inspection at the NRC Public Document Room, 1717 H Street, NW., Washington, D.C., and at the Rockford Public Library, Rockford, Illinois.
Single copies may be purchased from the sources indicated en the inside front cover.
The NRC Project Manager assigned to the Operating License application for Byron Station is Stephen H. Chesnut.
Mr. Chesnut may be contacted by calling (301) 492-7150 or writing:
i Byron SSER #1 1-1
Stephen H. Chesnut Division of Licensing U.S. Nuclear Regulatory Commission Washington, D.C.
20555 1.7 Summary of Outstanding Items The partial or complete resolution of some of the outstanding items identified in the SER is described in appropriate sections of this supplement.
One open item, the moisture content of the diesel air starting system 1as been resolved and is discussed in Section 9.5.6 of this supplement.
The our. standing items remaining in the staff operating license review are listed below.
Additional information requested by the staf f must be supplied by the applicant before the balance of the outstanding items can be resolved.
The resolution of these items will be discussed in a future supplement to the SSER.
The staff will complete its review of these items before the operating license is issued.
(1) Additional information to confirm pipeline foundation design.
(2) Turbine missile evaluation (3) High and moderate energy pipe break analysis outside containment (4) Pump and valve operability assurance program (5) Baseplate flexibility and anchor bolt loading (6) Seismic and dynamic qualification of equipment (7) Environmental qualification of electrical equipment (8)
Improved thermal design procedures (9)
Inadequate core cooling instrument (TMI-action item II.F.2)
(10) Steam generator flow induced vibrations (11) Reactor pressure vessel forces and moments analysis (12) Equipment and floor drainage system for internal flood protection (13) Fire protection program (14) Volume reduction system (15) Emergency preparedness plans (16) Control room human factors review
- 1. 8 Confirmatory Issues The following is an update of each of those confirmatory issues in Section 1.8 of the Safety Evaluation Report which have been completed.
Item 2 Category I manhole protection from tornado missiles (Section 3.5.3)
Item 5 Snubber inspection and testing program (Section 3.9.2.1)
Item 6 Seismic reassessment of components and supports (Section 3.9.2.2)
Item 33 Revis:'n to ohysical security plan (Section 13.6) 1.9 License Conditions The following is an update of the status of the license conditions in Section 1.9 of the Safety Evaluation Report.
Item 10 Personnel on shift with previous commercial PWR experience during startup phase.
(See Section 13.1.2.1.)
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Byron SSER #1 1-2
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3 DESIGN CRITERIA FOR STRUCTURES, SYSTEMS, AND COMPONENTS 3.2 Classification of Structures, Components, and Systems Sections 3.2.1, " Seismic Classification" and 3.2.2, " System Quality Group Classification" were included in the SER.
However, several paragraphs were inadvertently printed out of order.
Those sections are reprinted correctly below.
3.2.1 Seismic Classification GDC 2, " Design Bases for Protection Against Natural Phenomena," of 10 CFR 50, Appendix A, in part, requires that nuclear power plant structures, systems, and components important to safety be designed to withstand the effects of earth-quakes without loss of capability to perform their safety function.
These plant features are those necessary to ensure (1) the integrity of the reactor coolant pressure boundary, (2) the capability to shut down the reactor and maintain it in a safe shutdown condition, or (3) the capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to 10 CFR 100 guidel ne exposures.
The earthquake for which these plant features are designed is defined as the safe shutdown earth-quake (SSE) in 10 CFR 100, Appendix A.
The SSE is based on an evaluation of the maximum earthquake potential and is that earthquake which produces the maximum vibratory ground motion for which structures, systems, and components important to safety are desioned to remain functional.
Those plant features that are designed to remain functional if an SSE occurs are designated seismic Category I in Regulatory Guide 1.29.
Regulatory Guide 1.29, " Seismic Design Classification," is the principal document used in staff review for identifying those plant features important to safety that, as a minimum, should be designed to seismic Category I requirements.
The July 1981 edition of the SRP includes Section 3.2.1, " Seismic Classification."
The Byron Station Units 1 and 2 were reviewed in aceordance with SRP 3.2.1.
The results of this review are con-tained in thic sER.
The structures, systems, and components important to safety of the Byron Sta-tion Units 1 and 2 that are required to be designed to withstand the effects of an SSE and remain functional have been identified in an acceptable manner in Table 3.2-1 of the FSAR.
Table 3.2-1, in part, identifies the major components in fluid systems, mechanical systems, and associated structures designated as j
seismic Category I.
In addition, piping and instrumentation diagrams in the FSAR identify the interconnecting piping and valves and the boundary limits of each system classified as seismic Category I.
The staff has reviewed FSAR Table 3.2-1 and the fluid system piping and instrumentation diagrams, and concludes that the structures, systems, and components important to safety of Byron Units 1 and 2 have been properly classified as seismic Category I items in conformance with Regulatory Guide 1.29, Revision 3, except for the following item.
Byron SSER #1 3-1 l
Regulatory Position C.1 of the guide identifies structures including their foundations and supports, that perform a safety function as seismic CateCory I.
The seismic Category I supply and return lines of the essential service water system at Byron Units 1 and 2 are routed through the basemat of the turbine building, a nonceismic Category I structure.
The acceptance of the routing of these supply and return lines through the basemat of the turbine building is discussed in Section 3.8.3 of this SER.
In its review of Section 3.9 of the FSAR, the staff confirmed that acceptable design interfaces exist between seismic Category I and nonseismic portions of piping systems.
All other structures, systems, and components that may-be required for operation of the facility are not required to be designed to seismic Category I requirements, including those portions of Category I systems such as vent lines, fill lines, drain lines, and test lines on the downstream side of isolation valves and portions of these systems that are not required to perform a safety function.
The staff concludes that the structures, systems, and components important to safety of Byron Station Units 1 and 2 are properly classified as seismic Category 1 items in accordance with Regulatory Guide 1.29 and constitute an acceptable basis for satisfying, in part, the requirements of GDC 2, and are, therefore, acceptable.
3.2.2 System Quality Group Classification GDC 1, " Quality Standards and Records," of 10 CFR 50, Appendix A requires that nuclear power plant systems and components important to safety be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety function to be performed.
These fluid system pressure-retaining components are part of the reactor coolant pressure boundary and other fluid systems important to safety, where reliance is placed on these systems:
(1) to prevent or mitigate the consequences of accidents and mal-functions originating within the reactor coolant pressure boundary, (2) to permit shutdown of the reactor and maintain it in a safe shutdown condition, and (3) to retain radioactive material.
Regulatory Guide 1.26, " Quality Group Classification and Standards for Water, Steam, and Radioactive Waste Contain-ing Components of Nuclear Power Plants," is the principal document used in the staff's review for identifying on a functional basis the components of those systems important to safety that are Quality Groups B, C, and D.
10 CFR 50.55a identifies those ASME Section III, Class 1 components that are part of the reactor coolant pressure boundary (RCPB).
Conformance of these RCPR components with 10 CFR 50.55a is discussed in Section 5.2.1.1 of the SER.
These RCPB components are designated in Regulatory Guide 1.26 as Quality Group A.
Certain other RCPB components that meet the exclusion requirements of footnote 2 of the rule are classified Quality Group B in accordance with Regulatory Guide 1.26.
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The July 1981 edition of the SRP for the Review of Safety Analysis Reports for Nuclear Power Plants, includes Section 3.2.2, " System Quality Group Classifica-tion." The Byron Units 1 and 2 were reviewed in accordance with SRP 3.2.2.
The results of this review are contained in this SER.
The systems and components important to safety of Byron have been identified in an acceptable manner in c ble 3.2-1 of the FSAR.
Table 3.2-1, in part, identi-a fies the major components in fluid systems such as, pressure vessels, heat Byron SSER #1 3-2
l exchangers, storage tanks, pumps, piping, and valves and mechanical systems, such as cranes, refueling platforms, and other miscellaneous handling equipment.
In addition, the piping and instrumentation diagrams in the FSAR identify the Quality Group classification boundaries of the interconnecting piping and valves.
The staff has reviewed Table 3.2-1 and the fluid system piping and instrumentation diagrams and concludes that pressure-retaining components have been properly classified as Quality Group A, B, C, or D com-ponents in conformation with Regulatory Guide 1.26, Revision 3, except for the following items.
Regulatory Position C.1.a(2) of the guide identifies systems or portions of systems important to safety that are designed for postaccident containment heat removal as constructed to Quality Group B standards.
The reactor containment fan coolers for Byron Units 1 and 2 are constructed to Quality Group C standards.
To be acceptable, the staff requires the applico t to upgrade the reactor con-tainment fan ccolers to Quality Group B standards to the extent practical.
The cooling coils which are an integral part of each anit are the only pressure-retaining portion of the fan coolers that are affected by this upgrading to Quality Group B standards.
Therefore, in order to provide a level of quality equivalent to Quality Group B standards (ASME Section III, Class 2), the applicant will perform a volumetric examination on the welding neck flange of each cooling coil.
Other portions of the cooling coils such as brazed joints and fillet welds are acceptable, as the examination requirements are the same for Quality Group B and C components.
In addition to this upgrading of the containment fan coolers, the influent and ef fluent lines of the essential ser-vice water system (ESWS) extending from the outermost containment isolation valve up to the inlet and outlet flanges of each cooling coil of the fan coolers will be upgraded from Quality Group C to Quality Group B standards.
As required by Quality Group B standards (ASME Section III, Class 2), volumetric examination will be performed on the circumferential weld joints in the piping of that portion of the ESWS identified above.
In reclassifying this portion the ESWS to Quality Group B standards, the applicant has provided a closed system which meets the requirements of GDC 57.
The staff finds these actions taken by the applicant to upgrade the containment fan coolers and that portion of the ESWS which is a closed system within containment to the equivalent of Quality Group B standards to be acceptable.
The containment fan coolers and the portion of the ESWS within containment will be inspected as Quality Group B components as part of the plant Inservice Inspection Program Section XI require-ments during plant operation.
The codes and standards used in the construction of Quality Group A, B, C, or D components are identified in Table 3.2-2 of the FSAR.
The staff finds this summary list of codes and standards used in the construction of components to be acceptable.
Quality Group A components of the RCPB have been constructed
- in accordance with the ASME Boiler and Pressure Vessel Code,Section III, Division 1, Class 1.
Components in fluid systems important to safety that are classified Quality Group B have been constructed
- in accordance with ASME Boiler and Pressure
- Constructed, as used herein, is an all-inclusive term comprising materials certification, design, fabrication, examination, testing, inspection, and certification required in the manufacture and installation of components.
Byron SSER #1 3-3 1
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Vessel Code,Section III, Division 1, Class 1.
Components in fluid systems j
important to safety that are classified Quality Group B have been constructed in accordance with the ASME Boiler and Pressure Vessel Code,Section III, Division 1, Class 2.
Components in fluid systems that are classified Quality Group D have been constructed to the following codes as appropriate:
ASME Boiler and Pressure Vessel Code,Section VIII, Division 1 and 2; ANSI B31.1.0 Power Piping; and storage tank codes such as API-620, API-650, AWA-D100, or ANSI B96.1.
The staff concludes that constructior, of the components in fluid systems impor-tant to safety in conformance with the ASME Code, the Commission's regulations, and the guidance provided in Regulatory Guide 1.26 provides assurance that component quality is commensurate with the importance of the safety function of these systems and constitutes an acceptable basis for satisfying the require-ments of GDC 1 and is, therefore, acceptable.
3.5 Missile Protection 1
3.5.3 Barrier Design Procedures In the SER, the staff reported that the applicant should ensure that Category 1 manhole covers should be adequately protected against the impact of tornado missiles.
In a letter dated March 15, 1982, the applicant indicated that the i
Category 1 manhole covers had been modified and designed in accordance with SRP 3.5.3 and utilized 100 KSI ductile steel.
The staff finds those design changes acceptable and considers this item to be closed.
3.9 Mechanical Systems and Components l
3.9.2 Dynamic Testing and Analysis of Systems, Components, and Equipment 3.9.2.1 Snubber Examination Program
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In Section 3.9.2.1 of the Byron Safety Evaluation Report, the staff stated that the applicant's detailed program on preservice examination and testing of snubbers would be submitted in March 1982.
In a letter from T. R. Tramm to H.
R. Denton dated March 19, 1982, the applicant provided this information.
In this letter, the applicant has committed to a snubber preservice and exami-nation program in accordance with our pc ition as stated in FSAR Question 110.63.
The staff finds the applicant's proposec' program to be acceptable.
3.9.2.2 Seismic Qualification Testing of Safety-Related Mechanical Equipment In Section 3.9.2.2 of the Byron Safety Evaluation Report, the staff discussed the seismic reevaluation of piping and components required for safe shutdown.
This reevaluation was reviewed by the staff and found to be acceptable except for identification of those components and supports which were designed at or close to the allowable stress or strain limits.
The applicant has now provided this information and the staff has completed its review of the seismic reassess-ment of the Byron safety related piping and mechanical equipment.
The back-ground of this issue, together with a discussion of our review, is presented below.
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Byron SSER #1 3-4
The design of the Byron structures and components required for safe shutdown was reassessed by the applicant using Regulatory Guide 1.60 input at the CP stage the staff approved a safe-shutdown earthquake design of 0.2 g using deconvolution.
Since then, staff concluded that the deconvolution procedures are not acceptable and that Regulatory Guide 1.60, " Design Response Spectra,"
should be applied at the foundation level.
The Marble Hill design forces and spectra were used as the reassessment basis for Byron since the Marble Hill design is based on Regulatory Guide 1.60 spectra for a safe-shutdown earthquake of 0.2 g.
The reassessment covered piping runs, BOP and NSSS supplied mechanical equipment, components and supports required for safe shutdown.
Qualification procedures for all Nuclear Steam Supply System (NSSS) and Balance of Plant (B0P) supplied mechanical equipment in these safe shutdown systems were reviewed.
A comparison was made between the qualification levels of these components and the Marble Hill seismic levels at the location of the component in the plant.
In a large majority of cases, Byron equipment qualification reports demonstrated that the equipment was qualified to withstand greater accelerations than those calculated for Marble Hill.
In those cases where the Marble Hill level exceeded the levels in the qualification reports, stresses, deflections, and margins were calculated for the Marble Hill accelerations by scaling the calculated seismic stresses and deflections by the required Marble Hill-to-Byron accelera-tion ratio.
A comparison of the resultant stresses with the ASME Code allowable stresses indicated that no components were overstressed or had inadequate margins.
The following equipment was identified to have resultant stresses above 75% but below the ASME Code Allowable stresses.
A considerable margin-to-failure is stili available when a component is at its Code Allowable Stress, as discussed in NUREG/CR-2137, " Realistic Seismic Design Margins of Pumps, Valves and Piping."
1.
Auxiliary Building Supply Fan 2.
Cooling Tower Piping 3.
Cooling Tower U-bolt and Anchor Bolts 4.
Diesel Generator Exhaust Silencer 5.
Diesel Generator Lube Oil Circulation Pump 6.
HVAC Local Control Panels The procedures and analyses used to evaluate these items assure that the revised floor response spectra were conservatively assumed and that stresses and deflections resulting from a response spectrum analysis using the revised spectra are acceptable.
Based on this review and the review of the applicant's submittal conducted earlier and discussed in SER 3.9.2.2, we conclude that no piping runs, or BOP or NSSS supplied mechanical components and supports in systems required for safe shutdown were overstressed or had inadequate safety margins for the reanalyzed spectra.
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 The staf f has reviewed the applicant's pump and valve operability assurance program as discussed in Section 3.9.3.2 of the FSAR and compared this informa-tion with Section 3.9.3 of the Standard Review Plan.
Based on the review, the Byron SSER #1 3-5 l
applicant has provided information to define how active pumps and valves are generally qualified with respect to operability.
It should be noted that for those components where qualification and/or operability assurance is by analysis alone, some question remains as to the confidence level assured by this methodology.
The necessity Sr additional component testing is being considered and cannot be established without an inspection at the plant site.
Therefore, for the staff to determine the adequacy of the implementation of the applicant's pump and valve operability assurance program, an onsite audit of the equipment and supporting documenta-tion is required.
The onsite audit will include a plant inspection to observe the as-built con-figuration and installation of the equipment.
Also during the audit, the staff will review qualifying documentation, e.g., test reports and analysis, which are described in the applicant's program.
Thus, the staff's overall review includes an FSAR review and an onsite audit of the equipment.
Both phases of the staff review must be determined acceptable to ar-ive at a favorable conclusion on the applicant's overall pump and valve operability assurance program.
Before the audit is conducted, 85 to 90 percent completion should be attained j
for both the equipment documentation and the onsite installation of the equip-ment.
The applicant has indicated that they can support a June 1982 audit for Unit 1, and the staff will perform the audit at about that time.
The com-ponents to be included in the audit will be selected by the staff based on information the applicant has been requested to provide.
Once the applicant has indicated that work is substantially complete on Unit 2, the staff will conduct an onsite audit for that unit shortly thereaf ter.
The results of the onsite audits will be reported in a supplement to the SER.
Byron SSER #1 3-6
4 REACTOR 4.4 Thermal-Hydraulic Design In Section 4.4.2 of the SER, the staff noted that the applicant had margin available to offset the DNBR reduction due to rod bowing.
The staff also stated that if the applicant intended to use this margin, a description and the amount of margin must be included in the technical specifications.
Subse-quently, the applicant supplied an amended response.
The acceptability of the proposed method used to offset the DNBR reduction due to rod bowing is discussed balow.
The applicant presented the following relationship for determining the amount of thermal margin available to offset the DNBR reduction based on the difference between the safety analysis DNBR limit and the design DNBR limit.
Amount of Margin = (Safety Analysis DNBR Limit)-(Design DNBR Limit)
(Safety Analysis DNBR Limit)
Since the DNBR limits are different for typical and thimble cells, the amount of available margin is dependent on the cell type.
Using the relationship given above, the applicant has calculated 10.7% margin for a typical cell and 10.9% for a thimble cell.
Table 4.1 gives the cell type, the safety analysis DNBR limit, the design DNBR limit, and the amount of available margin.
Table 4.1 Available Thermal Margin Typical Cell Thimble Cell Saf.ty Analysis 1.49 1.47 DNBh limit Design DNBR Limit 1.33 1.31 Available Margin 10.7%
10.9%
For a region average burnup of 33,000 mwd /MTU the applicant calculated a gap closure of 84.0% and DNBR reductions of 11.1% and 13.6% for full flow and loss-of-flow conditions.
Westinghouse does not consider the effects of rod bow for region average burnups greater than 33,000 mwd /MTU since beyond this burnup, N
A H burndown effects preclude the fuel from achieving the limiting value of N
F H 3
Byron SSER #1 4-1 l
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Using the availabe thermal margin and the DNBR reductions calculated at a burnup of 33,000 mwd /MTU, the applicant has proposed an adjusted rod bow penalty of 0.4% for full flow conditions and 2.9% for the loss of-flow transient.
The staff has reviewed the methodology presented by the applicant and concludes that it is an acceptable means of offsetting the effects of rod bow.
- However, verification calculations performed, by the staff, yielded different values of gap closure and DNBR reduction.
For a burnup of 33,000 mwd /MTU the staff calculated a gap closure of 84.36% and a resultant DNBR penalty of 11.2% and 13.74% for the two flow conditions.
Therefore, our adjusted rod bow penalties are 0.5% and 3.04% for the full flow condition and the loss of-flow transient.
The staff will ensure that these margins (i.e. 10.7% and 10.9%) are contained in the Dasis fo the Technical Specifications and are incorporated into plant parameters as specified in the Technical Specifications.
Additional plant specific margins needed to offset the remaining rod bow penalty must also be included.
Based on our review, the staff has concluded that the proposed procedure to offset the DNBR reduction due to rod bow is acceptable with the changes and requirements given above.
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Byron SSER #1 4-2
5 REACTOR COOLANT SYSTEM 5.2 Integrity of Reactor Coolant System 5.2.1 Compliance with Codes and Code Cases 5.2.1.1 Compliance with 10 CFR Part 50.55a The Byron Station Units 1 and 2 control valves are constructed 2 to Section III, Class 1, of the ASME Boiler and Pressure Vessel Code, 1971 Edition, through the Summer 1972 Addenda.
In order to be in compliance with subsection (f)(3) of Section 50.55a, these components should be constructed to ASME Section III, Class 1, 1971 Edition, through the Winter 1972 Aodenda to the code.
The staff has reviewed the differences in these Code Addenda as applicable to the control values, and has identified no major differences.
The staff concludes that updating the control values to meet the requirements of subsection (f)(3) of 10 CFR Part 50, Section 50.55a, would not result in a commensurate increase in the level of safety.
Therefore, the staff finds that the ASME Code and Addenda used in the construction of the control values is acceptable and provides adequate assurance of component quality.
s Constructed, as used herein, is an all-inclusive term comprising materials 1
certification, design, fabrication, examination, testing, inspection, and certification required in the manufacture and installation of components.
Byron SSER #1 5-1 1
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1 6
ENGINEERED SAFETY FEATURES 6.3 Emergency Core Cooling System 6.3.5 Performance Evaluation 6.3.5.1 Emergency Core Cooling System Performance Evaluation The results of the Large Break LOCA analysis for Byron have been modified to reflect the reevaluation of previous single failure assumptions.
In a letter dated January 8,1982, Commonwealth Edison Company informed the NRC that the ECCS analysis for Byron Station, Units 1 and 2, was in error.
The nature of the error is that a failure of an RHR pump is not the most limiting single failure.
It has been determined that for some plant conrigurations, it is more conservative to assume ali ESF equipment operating.
The effects of the applicant's reanalysis with maximum ECCS flow is a 23 F increase in peak clad temperature.
Since the Byron units have a 98 F margin for meeting the peak clad temperature (PCT), acceptancu criteria of 10 CFR Part 50.46, the 23 F increase does not affect the ability to meet PCT requirements.
The staff requires that these calculations be documented in the FSAR as the limiting ECCS analysis.
Byron SSER #1 6-1
i 9 AUXILIARY SYSTEMS 9.5 Other Auxiliary Systems 9.5.6 Emergency Diesel Engine Starting Systems In the Byron Safety Evaluation Report (NUREG-0876), dated February 1982, the staff noted that on each side of the emergency diesel engine a line w;Lh a check valve is connected from the turbocharger air discharge to the starting air header at the front of the engine.
The purpose of this line is to con-tinuously purge the piping between the air start valve and the air control valve to prevent the possibility of an explosion due to a leaky air start valve and to prevent any condensation buildup in the air start piping while the engine is running.
This warm turbocharger air vents continuously through the orificed check valves and out the vent port in the starting air control valve.
In response to our concern on moisture in the air starting system due to this design, the applicant responded in a letter dated December 28,1981, that:
"This air is diesel combustion intake air which has been heated and compressed.
This air is filtered low humidity air ranging in temperature between 200 degrees Fahrenheit and 300 degrees Fahrenheit.
When this air is cooled to ambient temperature, the relative humidity would be less than 100% and, there-fore, this system will not introduce condensation into the air start piping."
The staff disagrees.
The staff subsequently indicated that it disagreed with this statement and stated its position that engine starting air shall be dried.
The applicant was informed of this position and in a letter dated February 11, 1982, the applicant provided the results of an analysis which shows that the amount of water that might condense in the short length of piping from the turbocharger to the starting air header under the worst case conditions is insignificant.
The staff has evaluated the results of the analysis and concurs with the applicant that the insignificant amount of moisture entrained in the air (humidity) in the short section of piping in question would have little if any effect on diesel generator operational reliability.
Based on its review, the staff concludes that the emergency diesel engine air starting system meets the requirements of General Design Criteria 2, 4, 5 and 17, meets the guidance of the cited Regulatory Guides and Standard Review Plan 9.5.6, meets the recommendations of NUREG/CR-0660, and industry codes and standards, and it can perform its design safety function and is therefore acceptable.
Byron SSER #1 9-1
12 RADIATION PROTECTION 12.3 Radiation Design Features 12.3.1 Facility Design Features 12.3.1.1 Chemical Decontamination A party to the Byron operating license proceeding has raised an issue concerning the safety impact of a possible chemical decontamination of the primary coolant system at Byron.
It is contended that chemical decontamination may exacerbate safety problems by degradation of the primary coolant system boundary integrity.
A particular concern was expressed about the potential for residual decontamination solution, such as NS-1, being trapped in crevices which could cause corrosion in the reactor system during an extended period between the decontamination and subsequent reactor operaton.
The concern is also raised that chemical decontamination may add to the deposition of radio-active corrosion products.
Finally, it is claimed that decontamination is not discussed in the applicant's FSAR or ER0L (Environmental Report).
There is at present no specific proposal to perform a chemical decontamination at the Byron facility.
Chemical decontaminations of nuclear plant systems and components have been performed in particular cases under 10 CFR 50.59.
This rule permits licensees to make certain changes in the facility or procedures and conduct tests or experiments provided that the change, test, or equipment does not involve a change in plant Technical Specifications or an unreviewed safety question.
In at least one case, an operating license amendment has been requested and granted for a chemical decontamination operation.
Several chemical decontaminations have been performed on nuclear plant systems and components (See Table 12-1) to reduce radiation levels prior to maintenance operations.
No degradation of the primary coolant system pressure boundary was experienced as evidenced by corrosion tests of coupons exposed during the decontamination operations.
Furthermore, these decontaminations were accom-plished as long as five years ago and reactor operation after these decontami-nations has not revealed any degradation of the integrity of the primary system coolant boundary.
Thus, chemical decontamination of the primary coolant system at Byron would not be expected to degrade pressure boundary integrity.
To minimize an increase in the deposition of radioactive corrosion products during plant operation following a decontamination (commonly known as re-contamination) the decontamination process includes a final passivation step.
This passivation step results in a very thin, almost monomolecular protective film on the primary coolant system surfaces which inhibit the deposition of radioactive corrosion products.
There is no evidence based upon decontami-nations that have been performed at Canadian and British reactors to indicate that the recontamination or the rate of radioactive corrosion product deposi-tion on decontaminated surfaces would be accelerated by the decontamination process.
Byron SSER #1 12-1 1
1 1
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i Table 12.1 Component decontamination experience Decontaminated Date Decontamination Decontamination Person Rem Plant Component Performed Solution Factor Savings l
l Peach Reactor Water Sept 1977 Dow Chemical 7-10 450 Bottom 2 Cleanup Heat NS-1 Exchanges Peach Reactor Water Apr. 1977 Dow Chemical 7-10 450 Bottom 3 Cleanup Heat NS-1 Exchangers Vermont Reactor Water Oct 1979 London Nuclear 6
700 - 900 Yankee Cleanup System Oct 1981 Can-Decon 4
250 Brunswick 1 Reactor Water Apr 1981 London Nuclear 5-10 900 Cleanup System Can-Decon Brunswick 2 Reactor Water Mar. 1980 London Nuclear 5-10 500 Cleanup System Can-Decon Nine Mile Reactor Coolant Mar 1981 London Nuclear 11 720 Point 1 Recirculation Can-Decon Pumps (Five)
- While Table 12-1 is limited to BWR decontamination experience (PWRs have not chemically decontaminated components at this time) the data presented should be representative for PWR decontamination operations.
There is no specific regulatory requirement to discuss decontamination in an I
operating license applicants FSAR or ER. However, the Byron applicant does discuss chemical decontamination in FSAR Section 12.1, " Ensuring that Occupational Radiation Exposures Are As Low As Reasonable Achievable (ALARA)"
and FSAR Section 12.3, " Radiation Protection Design Features." The applicant has designed the Byron station to facilitate decontamination and has committed to perform decontaminations to implement ALARA during periods of maintenance.
To enhance decontamination capability, the Byron design includes: adequate draining and flushing capability, equipment materials and coatings selected for decontamination, and equipment decontamination facilities.
This design is in conformance with Regulatory Guide 8.8, "Information Relevant to Ensuring that Occupational Radiation Exposure At Nuclear Power Stations Will Be As Low As Reasonably Achievable." Chemical decontaminations have been effective in implementing ALARA as demonstrated in Table 12-1.
Average decontamination factors of 4 to 11 were achieved and 250 to 900 person rem collective radiation exposures were saved during each of maintenance operations listed in Table 12-1 because the chemical decontamination ef fectively removed the radioactive deposits from the contaminated systems.
Byron SSER #1 12-2 t_
The concern that reactor system corrosion will result from residual decontami-nation solution, such as NS-1, being trapped in crevices after the decontamina-tion was raised in a comment on the Draf t Environmental Statement related to the chemical decontamination of Dresden 1 (NUREG-0686) dated May 30, 1980).
The staff response to this comment concluded that a copper rinse followed by three demineralized water rinses after decontamination solution drainage from the system will satisfactorily remove residual NS-1 from the crevices (refer to Final Environmental Statement related to chemical decontamination
- Dresden 1 NUREG-0686 page 8-21, dated October 1980).
The copper rinse is a relatively high pH (9.5) solution which should neutralize any residual acids in the system crevices.
Chemical decontamination was of concern at Dresden 1 because the facility was expected to remain in a " wet layup"* state for an extended period prior to restart.
The staff notes further that other chemicals may be used for decontamination and the NS-1 decontamination solution selected for the Dresden 1 decontamination may not be used on subsequent reactor plants.
Other decon-tamination solutions that have been developed are as effective as NS-1 but also offer the advantage of significantly less waste production because these solutions are less concencentrated than NS-1.
The high cost of low level waste burial is a decided incentive in selecting weakly-concentrated decontamination solutions.
Conclusions In general, chemical decontamination is an effective method for reducing per-sonnel radiation exposures in compliance with 10 CFR 20.1(c).
This rule states that licensees should make every reasonable effort to maintain exposures to radiation as far below the limits specified in 10 CFR 20 as is reasonably achievable.
The process of chemical decontamination is reviewed by the licensee (applicant) under 10 CFR 50.59.
In the event that the licensee (or staff) concludes that an unreviewed safety question is involved or that a technical specification change is required, a license amendment must be obtained to permit chemical decontamination.
The NRC staff would evaluate both the environmental and safety aspects of the decontamination operation to assure, among other things, that primary coolant boundary material integrity is not degraded before decontamination was authorized.
- " Wet layup" consists of protecting the steam generator from corrosive attack during periods of reactor plant shutdown.
I
,L Byron SSER #1 12-3 i
13 CONDUCT OF OPERATIONS 13.1 Organizational Structure of Applicant 13.1.2 Operating Organization 13.1.2.1 Organization In the Safety Evaluation Report, the staff indicated that during the Byron Station initial startup phase, it would require that the applicant staff each shift with an individual who had previous commercial PWR operating experience or experience in startup of a similar Westinghouse plant.
The applicant has subsequently committed (Tramm, February 3, 1982) to have a least one individual with previous operating experience on each operating shift at Byron 1 for at least one year following initial criticality to include the attainment of 100 percent power.
The staff finds the applicant's position to be acceptable, as it provides for substantial operating experience on each operating shift to aid in the diagnosis and mitigation of abnormal plant conditions which could arise during initial plant startup and power ascension.
Additionally, this experience will be provided for at least a year, thereby allowing a reasonable period of time for licensed individuals without previous commercial PWR operating experience to acquire familiarity with plant response under a variety of operating conditions.
13.6 Industrial Security The applicant has submitted security plans, entitled " Byron Nuclear Power Station Physical Security Plan," " Byron Nuclear Power Station Security Force Training and Qualification Plan," and " Byron Nuclear Power Station Plan Safeguards Contingency Plan," for protection against radiological sabotage.
The plans are being reviewed in accordance with Section 13.6 " Physical Security of the July 1981 edition of the " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants." (SRP, NUREG-0800)
As a result of the staff evaluation, certain portions of the above Physical Security Plan were identified as requiring additional information or upgrading to satisfy the requirements of 10 CFR 73.55.
The applicant has been informed of the areas requiring revision and has made commitments contained in Common-wealth Edison's letter, dated March 15, 1982, which are acceptable to the staff to resolve the oustanding issues.
The staff has concluded that the applicant's security plan, when suitably revised to formally incorporate these commitments, I
will be acceptable.
The " Byron Nuclear Power Station Security Training and Qualification Plan" and
" Byron Nuclear Power Station Safeguards Contingency Plan" have been determined to meet the requirements of 10 CFR Part 73 and, therefore, are acceptable.
Byron SSER #1 13-1
An ongoing review of the progress of the implementation of the approved portions of the plans will be performed by the staff to assure conformance with the performance requirements of 10 CFR Part 73.
The identification of vital areas and measures used to control access to these areas, as described in the plan, may be subject to amendments in the future.
The staff has determined that the above-referenced plans contain Safeguards Information which must be protected against unauthorized disclosure in accordance with 10 CFR 73.21.
Byron SSER #1 13-2
19 REVIEW BY THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS On February 26, 1982, a subcommittee of the Advisory Committee on Reactor Safeguards met with the representatives of the applicant and the NRC staff to consider the Commorcaealth Edison Company's application for a license to operate the Byron Station Units 1 and 2.
The.'eeting was held in Rockford, Illinois.
On March 4, 1982, at its 263rd meeting, the full Advisory Committee on Reactor Safeguards met with representatives of the applicant and the staff to consider the application.
The Committee recommended that the staff and applicant continue to resolve the Outstanding items, Confirmatory Issues, and License Conditions as well as certain Unresolved Safety Issues.
As a result of its review and meetings, the Committee reported that, if due cor ideration is given to those items, and subject to satisfactory completion of construction and preoperational testing, there is reasonable assurance that the Byron Station Units 1 and 2 can each be operated at power levels up to 3525 MWT without undue risk to the health and safety of the public.
The Committee tter from P. Shewmon to Nunzio J. Palladino, dated March 9, 1982, i
,ncluded as Appendix G to this supplement to the report.
In response to the ACRS recommendation, the staff will continue its review of the Unresolved Safety Issues, open and confirmatory items and license con-ditions to assure that they are resolved in a satisfactory manner prior to granting operating licenses for the Byron Station Units 1 and 2.
Byron SSER #1 19-1
APPENDIX A CHRONOLOGY OF NRC STAFF RADIOLOGICAL REVIEW 0F BYRON STATION The Safety Evaluation Report, Appendix A, provided a chronology of the NRC staff's radiological safety review for the period from April 30, 1977 to January 22, 1982.
The purpose of this appendix is to update that chronology through March 22, 1982.
January 21, 1982 Letter from applicant transmitting advance FSAR information which will be included in the next amendment.
January 22, 1982 Letter to applicant requesting additional geological information.
January 22, 1982 Letter to applicant concerning the Byron Control Room Design Review.
January 22, 1982 Letter from applicant transmitting advance FSAR information.
January 25, 1982 Letter from applicant concerning Security Plan Revisions.
January 26, 1982 Letter from applicant transmitting information on Instrumentation and Control.
January 26, 1982 Letter to applicant transmitting advance FSAR information.
January 28, 1982 Letter from applicant concerning Geological Study.
February 1,1982 Letter to applicant concerning a request for additional information radwaste volume reduction system.
February 1, 1982 Letter from applicant concerning preserivce inspection program.
February 2, 1982 Letter from applicant concerning Security Plan for Special Nuclear Materials Revision.
February 3, 1982 Letter to applicant concerning a request for additional information compliance with regulations.
February 3, 1982 Letter from applicant concerning advance FSAR information.
February 4, 1982 Letter from applicant transmitting Amendment No. 36 to the FSAR.
February 5, 1982 Letter to applicant transmitting 2 copies of the L
SER - Byron.
f Byron SSER #1 A-1
February 5, 1982 Letter from applicant concerning control of heavy loads.
February 10, 1982 Letter to applicant and Westinghouse withholding material on loop blowdown force computation from public disclosure as proprietary.
(Material submitted via CAW-81 Westinghouse).
February 11, 1982 Letter to applicant concerning comments on Byron DES.
February 12, 1982 Letter to applicant concerning resolution of Byron SER open items.
February 17, 1982 Letter to applicant transmitting 20 copies of the SER related to operation of Byron (NUREG-0876).
March 9, 1982 Report from the Advisory Committee on Reactor Safeguards.
March 10, 1982 Letter from applicant forwarding responses to comments on DES.
March 11, 1982 Letter to applicant requesting additional information regarding evacuation time estimate study.
March 11, 1982 Letter to applicant transmitting ACRS report.
l March 15, 1982 Letter from applicant on revised Security Plan.
March 16, 1982 Letter from applicant forwarding responses to Q 5.3, Q 40.120, Q 321.15-321.41 and revised information on Volume Reduction System.
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Byron SSER #1 A-2
APPENDIX F NRC STAFF CONTRIBUTORS AND CONSULTAf!TS This Supplement No. 1 to the SER is a product of the NRC staff.
The following NRC staff members were principal contributers to this report.
NAME TITLE REVIEW BRANCH R. Giardina Reactor System Power Systems G. Alberthal Nuclear Engineer Reactor Systems F. Witt Chemical Engineer Chemical Engineering J. Rajan Mechanical Engineer Mechanical Engineering R. Lipinski Sr. Structural Engineer Structural Engineering R. Skelton Safeguards Analyst Safeguard Licensing B. Elliot Materials Engineer Materials Engineering K. Connaughton Inspector, Management Region III Programs J. Holonich Nuclear Engineer Core Performance R. Kirkwood Principal Mechanical Mechar.ical Engineering Engineer i
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Byron SSER #1 F-1
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UNITED STATES
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A_PPENDIX G March 9, 1982 Honorable Nunzio J. Palladino Chairman U. S. Nuclear Regulatory Commission Washington, DC 20555
SUBJECT:
REPORT ON THE BYRON STATION UNITS 1 AND 2
Dear Dr. Palladino:
During its 263rd meeting, March 4-6, 1982, the Advisory Committee on Reactor Safeguards reviewed the application of the Commonwealth Edison Company ( Applicant) for a license to operate. Byron Station Units 1 and 2.
A tour of the facility was made by members of the Subcommittee on Febru-ary 25,1982, and a Subcommittee meeting was held in Rockford, Illinois on February 26, 1982 to consider this project.
During its review, the Commit-tee had the benefit of discussions with representatives of the Applicant and the NRC Staff.
The Committee a130 had the benefit of the documents listed. The Committee commented on the construction permit application for this Station in.its report dated May 13, 1975.
The Byron Station is located in Ogle County, Illinois, about 17 miles southwest of Rockford.
Rockford is the nearest densely populated center and had a 1980 population of about 140,000 peopl e.
The Byron Station uses two Westinghouse four-loop pressurized water re-actors, each having a rated power level of 3425 MWt.
Each is housed in a steel-lined, reinforced concrete containment building with a design pres-sure of 50 psig.
Construction of Unit 1 is about 82% complete and Unit 2 is about 70% complete.
The Applicant now has seven operating reactors and has accumulated over 80 reactor years of operating experience.
We reviewed the Applicant's staffing, training, and technical support capabilities for the Byron Station and believe that these capabilities are satisfactory.
The NRC Staff has identified in its Safety Evaluation Report dated Feb-ruary 1982 certain Unresolved Safety Issues as being applicable to the Byron Station as well as a number of Outstanding Items, Confirmatory Issues, and License Conditions; these include some TMI Action Plan re-
, e believe that these issues can be resolved in a manner W
quirements.
satisfactory to the NRC Staff and recommend that this be done.
G-1 1
Honorable Nunzio J. Palladino March 9,
1982 The Committee believes that, if due consideration is given to the recom-mendation above, and subject to satisfactory completion of construction I
and preoperational testing, there is reasonable assurance that the Byron Station Units 1 and 2 can be operated at power levels up to 3425 MWt without undue risk to the health and safety of the public.
Sincerely,
\\.
P. Shewmon Chainnan References 1.
Commonwealth Edison Company, " Final Safety Analysis Report for the Byron /Braidwood Stations," including Amendments 1-36.
2.
U. S. Nuclear Regulatory Commisson " Safety Evaluation Report Related to the Operation of Byron Station, Units 1 and 2," NUREG-0876, dated February 1982.
3.
Letter from League of Women Voters of Rockford, Illinois regarding the agenda of the ACRS Subcommittee neting on Cyron on Februa.ry 26, 1982 in Rockford, Illinois dated February 26, 1982.
4.
Letter from Elizabeth McKay to P. Shewmon regarding grouting of plant foundations with bentonite, dated February 26, 1982.
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APPENDIX H ERRATA TO BYRON SAFETY EVALUATION REPr Page Line Change 1-11 20 ForOpenItem(4), Change Section 3.9.5.2" to "Section 3.9.3.2", add " assurance" af ter " operability".
1-12 18 For Confirmatory Item (8), Change " Pump and Valve Operability Assurance (Section 3.9.3.2)" to " Inservice Testing of Pumps and Valves (Section 3.9.6)."
1-13 2
For Confirmatory Item (23) Change "7.3.2.2" to "7.3.2.3".
1-13 Last For License Condition (4) Change "S.2.2.5" to "7.2.2.5".
1-14 1
For License Condition (5) Change "7.5.5.2" to "7.5.2.2".
1-14 2 and 3 DeleteLicenseCondition(6).
2-11 10 Change "33 m" to "933 mm".
2-33 32 In last line of Section 2.5.4.3.4.1 insert " based" between " Valves" and "on".
3-1, 2 553.2.1, See corrected versions in SSER.
3.2.2 3-1 30 In first paragraph of Section 3.5.2 Change last sentence to "The guidelines of..."
3-16 17 Change "WCAP 8002A" to "WCAP 8082A".
10-2 45 Change "Section 14.1" to "Section 14".
10-3 14 Insert "and" between "FSAR" and " concludes".
10-4 8
Change "Section 10.3.3" to "Section 10.3.2".
10-9 16 Change "Section 14.1" to "Section 14".
10-11 4
Change "Section 14.1" to "Section 14".
I H-1
RNMN U S NUCLE A~i CEGUL ATO]Y COMMISSION f
BIBLIOGRAPHIC DATA SHEET Supplement tJo. 1 4 TITLE AND SUBTITLE (Add Volume No4.f eprmrestel 2 (Leave 0/ mal Safety Evaluation Report related to the operation of Byron Station, Units 1 and 2 3 REclPIENT'S ACCESSION NO
, AUTHOR (Si
- 5. D ATE REPORT COMPLE TE D l YEAR MONTH March 1982 9 PE HF ORMING OHGANilATION N AVE AND MAILING ADDRESS (Inctuce Esp Codel DATE REPORT ISSUED U. S. fluclear Regulatory Commission MON m l YEAR March 1982 Office of fluclear Reactor Regulation Washington, D. C. 20555 6 (t >** **aal 8 (Leave Nanti 12 SPONSOHING OHGANil ATION N AME AND M AILING ADD RE SS (Incluor 10 Codel 10 PROJECT <T ASK/ WORK UN'T NO Same as 9 above ii CoNTR ACT NO 13 i Y PE OF HE POH T PE MIOD C OV E RE D (inclusive dJffs)
Safety Evaluatio. Report 15 5UPPLE YE N T A H Y N OTE S 14 (L e8ve D/e* /
Pertgine tg n g'g n t r,!ge. ST" 50 454 =d STM 50-455 e
16 Alb iH ACT (200 +oras or sessJ Supplement tio. I to the Safety Evaluation Report of Commonwealth Edison Company's application for licenses to operate the Byron Station, Units 1 and 2, located in Rockvale Township, Ogle County, Illinois, has been prepared by the Office of I
fiuclear Reactor Regulation of the U. S. fluclear Regulatory Commission.
This supplement reports the status of certain items that had not been resolved at the time of publication of the Safety Evaluation Report.
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