ML20031F507
| ML20031F507 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 09/30/1981 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-0798, NUREG-0798-S01, NUREG-798, NUREG-798-S1, NUDOCS 8110200021 | |
| Download: ML20031F507 (40) | |
Text
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NUREG-0798 Supplement No.1 m
UTIS Safety Evaluation Report related to the operation of r Enrico Fermi Atomic Power Plant, Unit No. 2 Docket No. 50-341 Detroit Edison Company, et al.
U.S. Nuclear Regulatory Commissio'n 7
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NUREG-0798 Supplement No.1 Safety Evaluation Report related to the operation of Enrico Fermi Atomic Power Plant, Unit No. 2 Docket No. 50-341 Detroit Edison Company, et al.
U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation September 1981 p.
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ABSTRACT Supplement No.1 to the Safety Evaluation Report related to the operation of the Enrico Fermi Atomic Power Plant, Unit No. 2 provides the staff's evaluation of additional information provided by the applicant regarding outstanding review issues identified in the Safety Evaluation Report issued in July 1981.
It also covers revised designs and the staff's response to the comments in the report by the ACRS.
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CONTENTS Pag jjj Abstract...........................
1 Introduction and General Discussion..............
11 1.1 Introduction.......................
1_1 1.8 Summary of Outstanding Issues 1-2 1.8.1 Prelicensing Issues................
1-2 1.8.2 License Conditions 1-3 2-1 2
Site Characteristics
'5 Geology and Seismology..................
2-1 2.5.4 Stability of Structural Materials.........
2-1 2.5.5 Slope Stability..................
2-1 3
Design Criteria for Structures, Systems, and Components....
3-1 3.7 Seismic Design......................
3-1 3.7.1 Seismic Input...................
3-1 3-2 3.7.2 Seismic Analysis 3.7.3 Seismic Subsystem Analysis 3-2 3-4 3.8 Design of Seismic Category I Structures 3.8.3 Other Seismic Category I Structures........
3-4 3-4 3.8.4 Foundations......
3-5 3.9 Mechanical Systems and Components 3.9.1 Special Topics for Mechanical Components 3-5
-3.9.3 ASME Code Class 1, 2, and 3 Components 3-5 3.9.6 Inservice Testing of Pumps and Valves.......
3-6 3.10 Seismic and Dynamic Qualification of Mechanical and Electrical Equipment Important to Safety.........
3-7 3.11 Environmental Qualification of Safety-Related 3-8 Electrical Equipment...................
4-1 4
Reactor............................
4-1 4.2 Fuel System Design....................
4.2.3 Design Evaluation.................
4-1 v
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CONTENTS (continued)
Page 5
Reactor Coolant Pressure Boundary...............
5-1 5.2 incegrity of the Reactor Coolant Pressure Boundary....
5-1 5.2.1 Compliance with Appendices G and H, 10 CFR Part 50......................
5-1 5.3 Integrity of the Reactor Vessel 5-8 6
Engineered Safety Features 6-1 1
6.2 Containment Functional Design 6-1 6.2.7 Containment Leak Testing 6-1 6.3 Emergency Core Cooling System 6-5 6.3.4 Evaluation Findings................ 5 7
Instrumentation and Control..................
7-1 7.4 Systems Required for Safe Shutdown............
7-1 7.4.2 Specific Findings.................
7-1 7.5 Safety-Related Display Instrumentation..........
7-1 7.5.2 Loss of Power to Instruments and Control Systems 7-1 7.7 Control Systems Not Required for Safety 7-2 7.7.2 Specific Findings.................
7-2 4
11 Radioactive Waste Management 11-1 13 Conduct of Operations.....................
13-1 13.5 Industrial Security...................
13-1 l
13.6 Operating and Maintenance Procedures 13-1 15 Safety Analysis........................
15-1 15.2 Accidents........................
15-1 1
15.2.1 Anticipated Transients Without Scram 15-1 16 Technical Specifications 16-1 18 Repart of the Advisory Committee on Reactor Safeguards 18-1 vi i
CONTENTS (continued)
Page 22 TMI-2 Requirements...................... 22-1 22.2 TMI Action Plan Requirements for Applicants for Operating Licenses................... 22-1 I.
Operational Safety.................
22-1 I.C Operating Procedures 22-1 I.C.1 Guidance for the Evaluation and Development of Procedures for Transients and Accidents 22-1 I.C.5 Procedures for Feedback of Operating Experiences to Plant Staff.......
22-1 I.C.8 Pilot Monitoring of Selected Emergency Procedures for NT0L Applicants..... 22-2 1.D.1 Control Room Design Review.......
22-4 22-6 II. Siting and Design II.B.4 Degraded Core Training...........
22-6 II.F.2 Instrumentation for Detection of Inadequate Core Cooling................
22-6 23 Conclusions...................,.....,
23-1 Appendix A Chronology Appendix B Bibliography ApperCix 2 Report of the Advisory Committee on Reactor Safeguards on Fermi 2 Appendix D Errata to Fermi 2 Safety Evaluation Report i
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1 INTRODUCTION AND GENERAL DISCUSSION 1.1 Introduction The " Safety Evaluation Report related to the operation of Enrico Fernii Atomic Power Plant, Unit No. 2" (NUREG-0798) (SER), prepared by the staff of the Nuclear Regulatory Commission (staff), was issued on July 10, 1981.
The SER provided a summary and results of the staff's radiological safety review of the application by the Detroit Edision Company (applicant) for an operating license for Fermi 2.
The SER concluded that upon favorable resolution of outstanding matters described therein, the plant could be operated without endangering the health and safety of the public.
By Amendments 38 and 39 to the Final Safety Analysis Report (FSAR) and by letters identified in Appendix A to this supplement, the applicant has provided addi-tional information regarding several of the outstanding issues in the SER and has submitted revised designs for the solid and liquid radwaste systems.
The Advisory Committee on Reactor Safeguards (ACRS) completed its review of the appli-cation for an operating license for Fermi 2 during a meeting on August 6,1981 and provided its report to the Commission on August 11, 1981 (see Appendix C).
This supplement (Supplement No. I to the SER) provides (1) the staff's evaluation of additional information provided by the applicant regarding outstanding review issues identified in the SER, (2) the staff's evaluation to date of additional information provided by the applicant regarding revised designs, and (3) the staff's response to the comments in the report by the ACRS.
Except for the appendices, each section of this supplement is numbered and titled l
l the same as the corresponding section of the Safety Evaluation Report that has been affected by the additional evaluation.
Except as noted, each section is supplementary to the corresponding section in the SER.
Appendix A to this supple-ment is a continuation of the chronology of principal actions related to the staff's safety review of the application.
References are listed in Appendix B.
Appendix C is a copy of the report by the ACRS on its review of the application.
Appendix D contains a list of errata for the SER.
One of the outstanding issues in the SER was the applicant's response to the staff's request for verification that Fermi 2 meets regulatory requirements in Title 10 of the Code of Federal Regulations (10 CFR) Parts 20, 53, and 100.
By a l
letter dated September 15, 1981, the applicant provided an indepth comparison of l
the application with the regulations.
The applicant stated that Detroit Edison fully intends to satisfy NRC regulatory requirements or to justify and receive authorization from the staff to use alternative criteria before plant operation.
l L. L. Kintner is the NRC Licensing Project Manager for this project.
Mr. Kintner may be telephoned at (301) 492-703'/.
His address is U.S. Nuclear Regulatory Commission, Washington, D.C.
20555.
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Principal NRC staff contributors to this supplement to the safety evaluation report include:
D. Terao C. Graves A. Cappucci C. Patel D. Rothburg J. Lane C. P. Tam M. Tokar F. Litton R. Schemel A. Lee A. Ramey-Smith J. Kennedy E. Pedersen R. Anaud J. Clifford R. Giardina V. Deliso D. Kunze 1.8 Summary of Outstanding Issues 1.8.1 Prelicensing Issues The partial or complete resolution of some of the outstanding issues identified in the SER is described in appropriate sections of this supplement.
The out-standing issues remaining in the staff operating license review are listed below, with the appropriate section numbers in the SER or this supplement that describe the issue, the status, and plans for resolution.
The staff will complete its review of these items before the operating license is issued.
The resolution of these items will be discussed in a future supplement to the SER.
(1) Seismic reassessment of structures, systems, and components required for a safe shutdown (SER Sections 3.7.1 and 3.7.3 and this supplement Sections 3.7.1, 3.7.3, 3.10, and 18)
(2) Mark I containment analyses (SER Sections 3.8.1, 3.8.2, and 3.9.3, and this supplement Sections 3.8.4, 3.9.3, 3.10, and 18)
(3) Seismic and dynamic qualification of equipment (this supplement Section 3.10)
(4) Seismic and loss-of-coolant accident (LOCA) loading on fuel (SER Section 4.2.3)
(5) Environmental qualification of safety-related equipment (this supplement Sections 3.11 and 6.3.4.1)
(6) Break in control rod scram discharge volume (this supplement Section 6.3.4.1)
(7) Containment leakage tests (this supplement Section 6.2.7)
(8) Fire protection (SER Section 9.5.1 and Appendix E)
(9) Modifications to diesel engines to prevent dry starts (SER Section 9.5.7)
(10) Radwaste system modifications (this supplement Section 11)
(11) Physical security plan (this supplement Section 13.5) 1-2
r (12) TMI Issues (SER Section 22)
(a)
I.C.7 NSSS-vendor review of procedures (b)
I.G.1 Training during low power testing (c)
II.D.1 Testing of safety-relief valves (d)
II.E.4.2 Containment isolation dependability (e) III.A.1.1 Upgraded emergency preparedness III.A.1.2 Emergency response facilities III.A.2 Improved emergency preparedness 1.8.2 License Conditions In its review of outstanding issues identified in the SER, the staff has resolved some of the issues that were identified in the SER as license conditions and has identified additional license conditions.
The license conditions remaining in the staff licensing review are listed below, with the appropriate section numbers in the SER or this supplement that discusses the license condition.
Technical Specifications resulting from the review are listed in Section 16 of P
this supplement.
(1) Modifications to piping and equipment attached to Mark I containment (SER Section 3.8.1 and this supplement Sections 3.9.3, 3.10, and 18) c (2) Inservice testing program for pumps and valves (this supplement Section 3.9.6.1)
(3) Tests to detect fuel channel box deflection (SER Section 4.2.3)
(4) Hydrodynamic stability analysis (SER Section 4.4.1)
(5) Study of multiple control system failures (SER Section 7.2.2)
(6) Low pressure turbine-disc inspection (SER Section 10.2.2)
(7) Retention of persons with BWR operating experience on shift until 100 percent power is achieved (SER Section 13.1 and this supplement Section 18)
(8) Results of turbine trip startup tests (this supplement Section 15.1)
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(9) Station blackout simulator exercise (SER Appendix C and this supplement Section 18)
(10) Design and installation of postaccident sampling equipment (SER Section 22, Item II.B.3)
(11) Instrumentation for detection of inadequate core cooling (SER Section 22, Item II.F.2, and this supplement Sections 18 and 22, Item II.F.2) 1-3
2 SITE CHARACTERISTICS 2.5 Geology and Seismology 2.5.4 Stability of Subsurface Materials 2.5.4.3 Foundations 2.5.4.3.5 Foundations for Pipes and Ducts Seismic Category I piping and electrical ducts between the residual heat removal building and the reactor building are buried in trenches, either in fill or rock.
The SER stated that additional information was needed regarding such construction details as the type of bedding material, backfill around the pipe, and the compaction criteria. The applicant has provided sufficient additional information, as described below.
The safety-related piping and the ducts between the reactor building and the residual heat removal building are buried in trenches which are in the till, except adjacent to buildings where the trenches are in the rockfill backfill.
" Crusher Run Rockfill" (1.5-in. maximum size, described in SER Section 2.5.4.3.2,
" Backfill Materials") is used for the backfill around the buildings and also as bedding and backfill material for pipes and ducts.
This is designated as
" Crushed Rockfill" and was compacted to a minimum relative density of 75 percent (American Society for Testing and Materials (ASTM) Standard D2049). The pipes and the ducts have compacted crushed stone bedding a minimum of 12 in. thick.
By letter dated July 17, 1981, the applicant provided additional information regarding the buried pipe analysis.
For dynamic analyses of buried pipes, the applicant had recommended an apparent shear wave velocity of 2500 fps, based on the work of N. W. Newmark and W. J. Hall.
The staff performed an independent parametric analysis and determined that the stresses in the buried pipes during a safe-shutdown earthquake (SSE) will be within the allowable limits for a value of shear wave velocity as low as 1050 fps.
This value of shear wave velocity j
is very conservative for the compacted crushed rockfill.
Hence, the design parameters used in determining the stresses in the pipes are acceptable.
The bedding and backfill material is compacted to a minimum relative density of 75 percent and is not susceptible to liquefaction.
The staff concludes that the foundation of the buried pipes and ducts is stable and adequate for their j
l intended safety function.
Based on the review of the applicant's design criteria and construction reports, l
and on the review of the results of investigations, laboratory and field tests, and analyses, the staff concludes that the foundations for safety-related pipes l
and ducts meet the requirements of 10 CFR Part 100 and are acceptable.
j 2.5.5 Slope Stability There are no natural or manmada slopes whose failure could adversely affeci, the safe operation of the plant. However, there is a shore barrier (also called a breakwater) on the eastern edge of the plant site bordering Lake Erie.
This 2-1
shore barrier is provided as a protection against the probable maximum surge condition and the wave runup associated with the coincident wind wave activity from Lake Erie (see SER Section 2.4.2.5).
The SER stated that data on the stability of the shore barrier were needed.
By letter dated July 13, 1981, the applicant provided additional information on the slope stability analysis of the shore barrier.
The shore barrier is a rubble mound revetment with a 12.5-ft-thick cover of armor stone (stones ranging in weight from 200 lbs to 5 tons).
The barrier has a toe elevation of 572 ft, a crest elevation of 583 ft, and a lakeward slope of 2:1 (horizontal to vertical);
it is approximately 1050 ft long.
Before the armor stone was installed, the soft and unsuitable soil was removed from the revetment foundation and replaced by compacted clay fill placed to the limits shown in Figure 2.4-22 in the FSAR.
The 12.5-ft-thick armor layer is underlain by a 1-f t-thick crushed stone layer.
Figure 2.4-22 in the FSAR shows the details of the shore barrier cross section.
The barrier holds back and preserves the integrity of the plant site fill, placed to elevation 583 ft, and protects the plant against waves during the probable maximum surge condition.
The applicant investigated the stability of the shore barrier against failure by translation or rotation through the clay stratum.
Bishop's method of slope stability analyses was used to analyze the rotational mode of failure.
The analyses covered the critical combination of water levels for the static condition.
The stability for the SSE was checked by psuedo-static analysis, using an earth-quake coefficient of 0.15 g.
The minimum factor of safety against a slip-circle failure ranged from 1.25 to 2.74 for the several cases investigated.
The sliding-wedge method of analysis was used to analyze the translational mode of failure, and the minimum factor of safety was 8.8.
Hence, the shore barrier is safe against a failure by either the slip-circle or the sliding-wedge mode.
The design of the stone protective system was evaluated and reported in SER Section 2.4.2.5.
The staff concludes that the shore barrier will remain stable during both static and SSE loading conditions.
Based on the review of the applicant's design criteria and construction reports and on the review of the applicant's investigation results, laboratory and field tests, and analyses, the staff concludes that the shore barrier meets the requirements of 10 CFR Part 100 and is acceptable.
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3 DESIGN CRITERIA FOR STRUCTURES, SYSTEMS, AND COMPONENTS 3.7 Seismic Design 3.7.1 Seismic Input As discussed in Sections 2.5.2 and 3.7.1 of the SER, the applicant developed site-specific spectra in response to the staff's inquiries concerning possible character-izations of ground motion for seismic events. The site-specific spectra exceeded the design spectra for the Fermi 2 plant at some frequency ranges.
Staff studies have demonstrated to staff satisfaction that these exceedances are insignificant with regard to the inplace seismic capability of major structures, systems, and components at the Fermi 2 plant.
It was not equally clear, however, that comparable margins existed for all equipment necessary to shut down and cool down the facility in the event of a major earthquake, The staff, therefore, has required that in addition to the original seismic basis for the plant, the applicant evaluate the seismic margins available for equipment necessary for shutdown and continued heat removal at the Fermi 2 plant considering the combination of normal loads and seismic input characterized by the site-specific spectra.
The SER stated that additional information was needed regarding the design para-meters for analysis of buried piping and ducts and the method by which the three components of earthquake motion are combined in seismic analyses.
The staff met with the Fermi 2 design engineers in the offices of Sargent and Lundy, Chicago, on July 21 and 22, 1981 to discuss these outstanding review areas.
As a result of these discussions, the staff's concerns have been satis-l factorily resolved, either through better understanding of the assumptions used l
and the rationale of the acceptance criteria adopted or through the applicant's revision of some of the analyses, as described below.
For the staff's specific requirement that design parameters consistent with the new seismic input be used in the analysis of the buried piping and ducts, the applicant has complied with the staff's regeest and computed the ground t-l particle velocities, the maximum of which is found to be 4.5 in./sec. This S less than the 7.2-in./sec value used in the de' sign.
Therefore, the applicant concluded that the original design remains adequate.
The staff concurs with this conclusion.
By letter dated September 11, 1981, the applicant revised its seismic reassessment report to incorporate the results of this analysis.
In the SER, the staff required that respenses to the three components of earth-quake motion be combined by the square-root-of-the-sum-of-the-squares (SRSS)-
rule or that responses of one horizontal component and one vertical component be combined by the absolute sum (ABS) fole.
However, the applicant has used instead the following response combination:
the two horizontal components are combined algebraically through the use of acceleration time histories which are statistically independent, and the resultant response is combined with*the vertical response by the SRSS rule.
The applicant stated that, from the compari-sons using different methods of response combination, the method used in Fermi 2 is as conservative as the SRSS rule or the ABS rule.
Based on the information provided in meetings on July 21 and 22, 1981, the staff accepts the method of 3-1
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response combination used in Fermi 2.
The applicant has been asked to document its method of response combination in the FSAR.
As part of the seismic reassessment, the applicant has evaluated the seismic and dynamic qualification of equipment required for a safe shutdown using the site-specific response spectra with 7 percent structural damping.
However, the staff found, based on its review, that the level of stresses in the main seismic system (primarily structures) could not justify the use of 7 percent structural damping values and that, therefore, a damping value of 5 percent should be used.
The use of the lower structural damping value will not alter the final conclusion of the structural evaluation, but it may affect equipment seismic qualification.
The applicant has agreed to provide response spectra and to evaluate equipment seismic qualification for 5 percent structural damping.
The staff's evaluation of the method of generating the response spectra and of seismic and dynamic qualification of equipment using 5 percent structural damping will be provided in a future supplement to the SER.
(See also Section 3.10 of this supplement.)
Based on its review of the structural reevaluation provided by the applicant the staff concludes, subject to documentation of the method of response combina-tion and of acceptable site-specific spectra with 5 percent damping, that the input values used for Fermi 2 seismic design provide an acceptably conservative basis for designing seismic Category I structures, systems, and components to withstand the operating basis earthquake of 0.08 g and the design basis earth-quake of 0.15 g.
3.7.2 Seismic Analysis In the SER the staff indicated that its conclusion on seismic analysis was subject to the resolution of the concerns indicated in Section 3.7.1.
Because these concerns have been satisfactorily resolved, the staff concludes that the seismic analysis procedures and criteria utilized by the applicant, including those for the evaluation of the interaction of non-Category I structures and piping with Category I structures and piping, provide an acceptable basis for the seismic design.
3.7.3 Seismic Subsystem Analysis In the Safety Evaluation Report, the staff stated that the results of applicant's seismic reassessment of structures, systems, and components required for a safe shutdown were being reviewed.
This review under Standard Review Plan (SRP)
Section 3.7.3 (NUREG-75/087) focused on the seismic reassessment of mechanical components, including reactor pressure vessel internals, piping systems, com-ponent supports, and instrumentation lines.
In the review, the staff considered all essential seismic Category I piping systems designated by the applicant for safe shutdown and cooldown of the reactor.
The staff found several residual heat removal (RHR) Division I piping subsystems that were not computer analyzed in the seismic reassessment; however, those piping subsystems that were excluded in the reassessment were evaluated qualita-tively by inspecting the seismic structural response spectra applicable to those subsystems.
In comparing the design basis structural response spectra (that is,1.875 x OBE) with the new site-specific structural response spectra, the staff concluded that, for the required subsystems, the design basis spectra provide an ample margin of conservatism over the site-specific spectra.
The 3-2
site-specific spectra--where the design-basis spectra were signifi-worst case cantly exceeded by the site-specific spectra--were found to be not applicable to the safety-related mechanical components were considered in the seismic reassessment.
As part of its review, the staff had audited three of the essential bal:nce-of-plant piping subsystems (feedwater, reactor core isolation cooling, and residual heat removal subsystems). The staff found that in the reassessment using the site-specific spectra, most of the piping stresses decreased.
Those stresses that increased were significantly below the limits allowed by the Boiler and Pressure Vessel Code of the American Society of Mechanical Engineers (ASME Code).
All piping stresses remained within the Code-allowable limits. All support loads that were audited (including large-bcre piping and instrumentation supports) were within the design limits.
All equipment nozzle reactions in the audited subsystems decreased with the new site-specific spectra.
As an additional measure of conservatism, the balance-of plant piping stresses were evaluated (in both the design basis and in the reassessment) to Service Level C limits.
For a safe-shutdown earthquake (SSE), the staff position allows the use of Service Level D limits. Thus, there is an additional 33 percent margin in design for which the applicant did not take credit in the balance-of plant piping design.
In selecting the structural response spectra for the NSSS piping system reanalysis, the applicant has used the response spectra corresponding to the center of gravity of the piping system. This methodology for selection of response spectra is not acceptable to the staff.
Accordingly, by letters dated August 25, 1981, and September 9, 1981, the applicant has committed to reanalyze the NSSS piping systems for the final as-built condition as required by IE Bulletin 79-14 (1) by using a multiplied (1.875) operating basis earthquake (OBE) loading to evaluate the SSE or, alternatively, the site-specific earthquake response spectra, and (2) by including all the response spectra applicable to the support and anchor locations (that is, envelope method or multiple support excitation).
The staff's position is that the SSE must be at least equivalent to the site-specific earthquake and that an acceptable methodology for response spectra selection be used (combined envelope or multiple support excitation).
The staff considers the methodologies proposed by the applicant to be conservative and therefore acceptable.
The applicant has committed to submit the results of the reanalysis of the NSSS piping systems by April 26, 1982.
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i Contingent upon the results of the final as-built NSSS piping analyses being 7
within the design limits, the staff finds that the seismic reassessment per-formed by the applicant using the new site-specific spectra, together with the seismic procedures described in Section 3.7.3 of the SER, provides an acceptable t
basis for complying with the applicable portions of General Design Criterion (GDC) 2, " Design Bases for Protection Against Natural Phenomena." The staff will report the evaluation of NSSS piping reanalysis in a future SER supplement.
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3.8 Design of Seismic Category I Structures 3.8.3 Other Seismic Category I Structures 3.8.3.1 Reinforced Concrete and Steel Structures In the SER the staff indicated that its conclusion on the adequacy of reinforced concrete and steel structures was subject to the resolution or the concerns indicated in Section 3.7.1.
Because these concerns have been satisfactorily resolved, the staff concludes that the criteria used in the analysis, design, l
and construction of the Fermi 2 seismic Category I structures to account for anticipated loadings and postulated conditions that may be imposed upon each structure during its service lifetime are in conformance with established criteria, codes, standards, and specifications and are acceptable to the staff.
Therefore, there is reasonable assurance that, in the event of winds, tornados, earthquakes, and various postulated accidents occurring within the structures, the structures will withstand the specified design conditions without impairment of structural integrity or the capability to perform required safety functions.
The basis for staff acceptance has'been the use of criteria as defined by appli-cable codes, standards,_and specifications and the use of applicable design bases as defined in the FSAR including:
the loads and loading combinations; the design analysis procedures; the structural acceptance criteria; the materials, quality control, and special construction techniques; and the testing and inservice surveillance requirements.
Conformance with these criteria and design bases, codes, specifications, and standards constitutes an acceptable basis for satisfying the applicable requirements of GDC 2 and 4.
3.8.3.3 Spent Fuel Pool and Rack In this SER, the staff indicated the rack design was acceptable, subject to the receipt by NP" of an acceptable seismic analysis based on revised floor response spectra.
Such revised floor response spectra were derived from a seismic reassess-ment of the plant (see Section 2.5.2 of the SER).
An acceptable seismic analysis, based on revised floor response spectra, has been received ar.d reviewed.
The seismic analysis resulted in acceptable displacements, deflections, and stresses in the racks for both OBE and SSE conditions, thus providing assurance that the spent fuel racks will perform their it. tended safety functions.
3.8.4 Foundations In the SER, the staff indicated that its conclusion on the adequacy of the design of the foundation was subject to the resolution of the concerns indicated in Section 3.7.1.
Because these concerns have been satisfactorily resolved, the staff concludes that the criteria used in the analysis, design, and construc-tion of all the plant Category I foundations to account for anticipated loadings and postulated conditions that may be imposed upon each foundation during its service lifetime are in conformance with established criteria, codes, standards, and specifications that are acceptabM h the staff.
Therefore, there is reasonable % w e that, in the event of winds, tornados, earthquakes, and various post,a v.)
nts, the seismic Category I foundations will withstand the specified onig;.
Witions without impairment of structural integrity and stability or impd rment of the capability to perform required safety functions.
3-4
The basis for staff acceptance has been the use of criteria as defined by appli-cable codes, standards, and specifications and the use of applicable design bases as defined in the FSAR including:
the loads and loading combinations; the design and analysis procedures; the structural acceptance criteria; the materials, cuality control, and special construction techniques; and the testing and inservice surveillance requirements.
Confermance with these criteria and design bases constitutes an acceptable basis for satisfying the applicable requirements of GDC 2 and 4.
In the structural reevaluation of the reactor building foundation mat for the site-specific earthquake, it was found that the capacity of the foundation mat to resist the torus uplift was reduced from 2000 kips to 1370 kips per torus support.
This value will be confirmed as adequate when the Mark I containment plant-unique analysis is provided. The staff evaluation of the plant unique analysis will be reported in a future supplement to the SER.
3.9 Mechanical Systems and Components 3.9.1 Special Topics for Mechanical Components In the Safety Evaluation Report, the staff stated that results of the independent analysis of a sample piping system performed by Oak Ridge National Laboratory would be reported in this supplement.
This independent analysis is part of the review under SRP Section 3.9.1 pertaining to the computer programs that were used in the analysis of seismic Category I piping systems.
The staff has completed its confirmatory piping analysis of a sample small piping subsystem and has verified that the support loads and piping spans recommended in the Fermi-2 simplified small piping design manual provide an adequate design basis for small-bore piping and instrumentation lines.
3.9.3 ASME Code Class 1, 2, and 3 Components In the SER, the staff stated that certain aspects of design information on the safety / relief valve (SRV) discharge piping system were being reviewed.
The staff has reviewed the applicant's procedures used on the design of SRV systems.
The SRV discharge piping system has been upgraded to ASME Code Class 2.
9 e staff has reviewed the adequacy of the interim structural design of the SRV piping and supports, particularly in the torus wetwell area, including the loads and stresses resulting from SRV actuation.
By letter dated June 8, 1981, the applicant provided supplemental information on the design of the SRV piping subsystem.
The latest SRV transient wave loads propagating through the SRV piping have been calculated using the General Electric Company (GE) RVFOR04 Code.
Although a complete reevaluation of the SRV piping and supports has not been completed at this time, a preliminary assessment of a typical SRV line was found to be acceptable.
Based on an acceptance criterion that all the final loads and stresses must be within the specified acceptable limits, the staff finds the applicant's methodology for the design of the SRV piping system to be acceptable.
By letter dated June 22, 1981, the applicant has committed to perform a supplenental fatigue evaluation of the SRV piping in the torus wetwelI air space.
The applicant proposes to extend the Class 2 fatigue considerations for thermal expansion stress cycles to apply to mechanical load cycles (SRV and seismic).
3-5
The effects of both thermal and mechanical load cycles are evaluated in conjunc-tion with each other to determine the allowable SRV actuations for the piping system.
The actual number of SRV actuations will be monitored during plant operation. When the actual number of SRV actuations approaches the computed allowable number of actuations, the applicant will consider several alternatives to protect against a fatigue failure of the SRV piping in the wetwell airspace.
The possible alternatives include increasing inservice inspe:tions, rotating the valve set points, or replacing the piping in the wetwell airspace. The staff has reviewed the applicant's procedures proposed for the fatigue evaluation of the SRV piping in the wetwell airspace and finds the applicant's methodology acceptable.
Results of the applicant's fatigue analysis will be reported as a part of the Mark I containment plant-unique analysis.
In the SER, the staff noted that the final structural analysis of SRV discharge piping due to hydrodynamic loads from SRV operation and the loss-of-coolant accident (LOCA) have not been submitted.
The applicant will submit results of these final analyses before May 1, 1982, as a part of its report on the Mark I plant-unique analysis.
Any modifications required by the plant-unique analysis will be completed before the operating license is issued.
The staff will review the plant-unique analysis and report the results in a supplement to the Fermi 2 SER.
3.9.6 Inservice Testing of Pumps and Valves 3.9.6.1 Conformance to Section XI of the ASME Code In the SER, the staff stated that the apr'icant's submittal on inservice testing of pumps and valves was being reviewed.
By letter dated August 3, 1981, the applicant has submitted a description of its proposed inservice testing program for pumps and valves, The submittal included a request for relief from the requirements of Section XI of the ASME Boiler and Pressure Vessel Code.
The applicant's proposed program includes both baseline preservice testing and periodic inservice testing.
It provides both for functional testing of compo-nents in the operating state and for visual inspection for leaks and other signs of degradation.
The staff has not completed its detailed review of the applic nt's submittal.
i However, based on a preliminary review, the staff, finds that it is impractical--
J within the limitations of design, geometry, and accessibility--for the applicant to meet certain ASME Code requirements.
In the view of the staff, imposing those requirements would result in hardships or unusual difficulties without a compensating increase in the level of quality or safety.
The relief requested will not endanger life or property and is in the public interest.
Therefore, pursuant to 10 CFR Part 50, Section 50.55a, the relief that the applicant has requested from the pump and valve testing requirements of 10 CFR Part 50, Section 50.55(g)(2) and (g)(4)(i) is granted for that portion of the initial 120-month I
period during which the staff completes its review.
Because the applicant's l
request for relief has been granted and the applicant will comply with ASME Code Section XI and/or the Technical Specifications, the staff finds the Fermi 2 inservice testing program for pumps and valves acceptable.
The staff has scheduled a detailed review of the applicant's program in 1982.
If the review is not completed before the plant is completed, the staff will include a condition in the license requiring that the additional information needed be provided no later than 6 months before the end of the 120-month period.
(
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3.10 Seismic and Dynamic Qualification of Mechanical and Electrical Equipment Important to Safety In the SER, the staff stated that the summary of Fermi 2 equipment seismic and dynamic qualification program provided in the applicant's submittals dated April 8, June 18, and June 29, 1981, had been reviewed with reference to FSAR Sections 3.9.2 and 3.10.
The staff found the applicant's description of the methodology for seismic dynamic qualification of equipment important to safety acceptable, subject to a detailed site audit review of a representative sample of seismic Category I mechanical and electric equipment, both in the nuclear steam supply system and in the balance of the plant.
During th urrent staff review of the seismic input (see SER Sections 2.5.2 and 3.7.1 it was determined that the shape of the original design response spectra f u the Fermi 2 site is not consistent with that currently acceptable to the staff.
The applicant was then required to develop site-specific response spectra for a reevaluation of structural and mechanical component design and equipment qualification.
The site-specific response spectra were reviewed by the staff and found acceptable (see Section 2.5.2 of this supplement).
This change in the input spectrum shape made reassessment of the resulting effect of the loading on the equipment necessary. Only the equipment necessary for safe shutdown and cooldown of the reactor is required for reassessment of seismic qual-ification using the new seismic input based on the site-specific response spectra (see Section 3.7.1 of this supplement). Qualification of other remaining seismic Category I equipment was based on the original ground response spectra as defined in the applicant's FSAR Section 3.7, Figures 3.7-2 and 3.7-3.
The staff seismic qualification review team (SQRT) performed the audit at the Fermi 2 site on July 27-31, 1981, to determine the extent to which the qualifi-cation of the equipment, as installed, meets current licensing criteria in Sections 3.9.2 and 3.10 of the SRP.
The audit included the examination of the equ:pment field installation configuration, as well as the review of the equipment qualification documents, against the original design spectra.
Twenty-four selected pieces of mechanical and electrical equipment were reviewed, encompassing all systems important to safety. Of the 24 selected equipment items,15 items which are in the safe-shutdown systems are part of the applicant's reassessment program.
The audit also included the review of the reassessment of these 15 pieces of equipment, which are treated as being representative of all equipment systems being reassessed.
The applicant was notified at the beginning of the site audit that the 7 percent structure damping volume used in generating floor response spectra for equipment reassessment is not acceptable because the stresses in critical structural elements obtained in the building seismic analysis reassessment are far below yield limit.
For such a case, the staff concluded that an adequate structure damping valve is 5 percent (see Section 3.7.1 of this supplement).
Although the staff proceeded with its review for equipment reassessment based on the spectrum curves of 7-percent structure damping value, during the site audit the staff asked the applicant to perform an updated reassessment of the equipment qualification based on a 5 percent structure damping and submit the results for SQRT review.
As a result of the audit, the staff identified concerns regarding the applicant's original equipment qualification as well as the reassessment due to site-specific response spectra; these were later confirmed in the applicant's letter of August 13, 3-7
1981.
The statt is currently reviewing the applicant's submittal of August 31, 1981, which responds to some of the concerns identified above.
The staff also has reviewed the applicant's September 4, 1981 submittal'and found the applicant has not provided sufficient explanation of his method of generating response spectra for 5 percent structure damping.
The applicant's equipment qualification reassessment for 5 percent structure damping, therefore, remains an open item pending the applicant's submittal of acceptable response spectra and reassessment of the equipment input loading vis-a-vis original qualification programs.
The staff will continue its review of the applicant's qualification program and report its conclusions on the acceptability of the implementation of the program in a future supplement to the SER.
As a part of the Mark I Containment Long-Term Program (see SER Sections 3.8.1 and 3.9.3), which is not included in the above review, the applicant has also committed to complete and submit the Fermi 2 Plant Unique Analysis for staff review before May 1, 1982.
The applicant has also agreed to submit qualifica-tion results for torus-attached equipment under the effects of combined seismic and Mark I hydrodynamic loads for staff review before August 1, 1982, and to have all necessary modification completed before the plant returns to power after the first refueling.
The staff finds the above procedure and commitments acceptable.
3.11 Environmental Qualification of Safety-Related Electrical Equipment In the SER, the staff stated that the results of its review of the applicant's environmental qualification submittal dated June 26, 1981, and of a July 1981 site audit would be reported in a supplement to the SER.
Based on its initial review of this situittal and the audit, the staff concluded that additional information was raeded to complete its review.
Based on a review of the scram discharge volume (Section 6.3.4.1 of this supplement), the staff concluded that additional information was also needed regarding environmental qualification of equipment needed to mitigate a break in this system.
The applicant has agreed to the following schedule f ~ completion of the review, which is acceptable to the staff.
Before December 1, 1981, the applicant will provide complete information, i
including justification for the radiation and temperature environment, to demonstrate compliance with the requirements in Category II of NUREG-0588 for safety related electrical equipment, excluding that required for breaks in the scram discharge system. Where there are deviations, the applicant must commit to corrective action (requalification, replacement, relocation or modification) consistent with the requirements to esttblish qualification.
Before January 1, 1982, the applicant will provide complete information required for environmental qualification of equipment required for breaks in the scram discharge system, as identified in Section 6.3.4.1 of this supplement.
In January 1982, the staff will conduct a site audit if the results of the submittal are complete.
3-8
All correc.tive actions needed to meet the requirements of NUREG-0588 must be completed before the deadline specified in Commission Mernorandum and Order CLI-80-21.
If fuel loading occurs before this deadline, justification must be provided for operation until the corrective actions are completed.
3-9
-4 REACTOR 4.2 Fuel System Design 4.2.3 Design Evaluation In the SER subsection entitled " Fission Gas Release," the staff stated that the operating license would be conditioned to require a reanalysis of fission gas release at high fuel burnups before exceeding a fuel burnup of 20,000 mwd /tU.
In letters to the staff dated May 6, 1981, and May'28, 1981, GE submitted c generic reanalysis in which it requested that credit for recently approved ECCS evaluation model changes be used to offset any operating penalties due to high burpup fission gas release.
This proposal was found acceptable provided the gereric reanalysis was found applicable to Fermi 2.
By letter dated August 25, 1981, the applicant stated that the generic analysis is applicable to the Fermi 2 aesign. The staff, therefore, concludes that the issue of enhanced fission gas release at high burnup to be satisfactorily resolved for Fermi 2 and that a license condition is not required.
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5 REACTOR COOLANT PRESSURE BOUNDARY
- 5. 2 Integrity of the Reactor Coolant Pressure Boundary 5.2.1 Compliance with Appendices G and H, 10 CFR Part 50 The SER stated that applicant agreed to provide additional information to justify exemptions to Appendices G and H to 10 CFR Part 50.
By letters dated July 15, 1981, and 51y 29, 1981, applicant provided sufficient justification. The staff evaluatiot provided herein supersedes SER Section 5.2.1.
GDC 31, " Fracture Prevention of Reactor Coolant Pressure Boundary," requires that the reactor coolant pressure boundary (RCPB) be designed with suf'icient margin to ensure that when it is stressed under operating, maintenanc, testing, and postulated accident conditions, the boundary behaves in a nonbrittle manner and the probability of rapidly propagating fracture is minimized.
" Inspection of Reactor Coolant Pressure Boundary," requires that the RCPB be designed to permit an appropriate material surveillonce program for the reactor pressure boundary.
The fracture toughness requirements for the ferritic materials of the RCPB are described in Appendices G (" Fracture Toughness Reauirements") and H '" Reactor Vessel Material Surveillance Requirements") of 10 CFR Part 50.
The staff has reviewed the materials selection, toughness requirements, a d extent of materials testing conducted by the applicant to provide assurance that the ferritic materials used for pressure-retaining components of the RCPB have adequate toughness under operating, maintenance, testing, and anticipated transient conditions.
The ferritic pressure boundary materials of the reactor pressure vessel were qualified by impact testing in accordance with the require-ments of Section III of the ~ ASf1E Boib r and Pressure Vessel Code,1968 Edition, including the Summer 1969 Addenda, the construction permit for Fermi 2 was issued in September 1972, and the reactor vessel wat ordered in January 1967.
Pursuant to subparagraph 50.55a(c)(2) of 10 CFR Part 50, the Code applicable to the fabri-cation of the Fermi 2 reactor vessel isSection III of the ASME Code, 1968 Edition, including the Summer 1970 Addenda.
The codes and standards used for the construction of components in the RCPB are identified in Table 3.2-3 of the Fermi 2 FSAR.
The applicable codes and standards were those required at the time the components were purchased..The staff has reviewed and evaluated this area of noncompliance to the regulation and concluded that only minor differences exist between the design and construc-tion requirements of the ASME Code Edition and Addenda applied to the Fermi 2 Class 1(A) components and those required by the regulations.
Further, updating the Fermi 2 design and construction requirements to meet the specific ASME Code Edition and Addenda would not significantly increase the level of quality or safety, and the present requirements for the Class 1(A) components provide an acceptable level of quality and safety.
Therefore, an exemption to the require-ments of subparagraph 50.55a(c)(2) of 10 CFR Part 50 has been granted, and the use of the ASME Code, 1968 Edition, including Summer 1969 Addenda, is acceptable for the fabrication of the reactor pressure vessel.
5'
Because the applicable ASME Code Edition and Addenda defined in paragraph 50.55a of 10 CFR Part 50 as well as the actual Edition and Addenda applied to the construction of the Fermi 2 RCPB components preceded the publication of Appendices G and H of 10 CFR Part 50, some of the mechanical property tests on the ferritic materials in the primary coolant pressure boundary designed to demonstrate compliance with the fracture toughness requirements of these I
appendices were not conducted.
The applicant has proposed using other methods to demonstrate that the alternate methods provide pressure-temperature operating limits that ensure a margin of safety against nonductile failure equivalent to those of a vessel constructed in strict compliance with i
Appendices G and H.
l 5.2.1.1 Compliance with Appendix G, 10 CFR Part 50 l
Section I.B of Appendix G applies the fracture toughness requirements of Appendix G to welds and heat-affected zones of ferritic materials in the pressure retaining components.
The ASME Edition and Addenda to which Fermi 2 was constructed did not require tests to establish the fracture toughness properties of welds and heat-affected zones.
However, paragraph N-541 of the i
applicable code required impact testing nf procedure qualification-test weld deposits and heat-affected zones at a tcoperature selected in accordance with subparagraph N-331.2 of the code. The applicant has estimated an RT f0F NDT for the nonbeltline welds based on the weld qualification tests results and has assumed an RT f r the heat-affected zone equivalent to the RT f the NDT NDT base metal.
The applicant presented technical data, including quality a surance records, to demonstrate that the estimated RT f r the welds and heat-affected NDT zones were conservative.
The staff concurs with the applicant.
Thus the staff I
concludes that an exemption to the requirement of Section I.B of Appendix G is justified and granted.
Section III.A requires that both unirradiated and irradiated ferritic materials should be tested for fracture toughness properties by means of the Charpy V-notch test specified in paragraph NB-2321.2 of the ASME Code.
In addition, when required by the ASME Code, unirradiated ferritic materials shall be tested by means of the dropweight test specified in paragraph NB-2321.1.
Fermi 2 vessel beltline welds, closure head flange, vessel flange, and feedwater-nozzle forgings were not qualified by dropweight tests.
The estimation procedures for the RT NDT values for these components were in alcordance with General Electric procedure Y1006A006 and explained in paragraph 5.2.4.22 of the FSAR.
On July 29, 1981, the applicant submitted technical information and additional data to demonstrate that the estimated RT values were conservative.
The staff concurs with the NDT applicant that estimated RT values were conservative. The staff concludes NDT that an exemption to the requirements of Section III.A of Appendix G is justified and granted.
Section III.B.1 requires that the location and orientation of the impact test specimens comply with the requirements of NB-2322 of the ASME Code, 1971 Edition, including Summer 1972 Addenda.
This code requires that the impact test speci-mens be oriented in a direction normal to the principal rolling direction.
The Charpy V-notch test specimens used for impact test qualification of the I
ferritic materials for Fermi 2 were oriented in the longitudinal direction in 5-2 l
compliance with the ASME Code, 1968 Edition, including Summer 69 Addenda.
The applicant has justified noncompliance to the requirement of Section III.B.1 of Appendix G by adding 30 F to the Charpy V-notch 50 ft-lb longitudinal tempera-ture, as described in Section 5.2.4.2.2 of the FSAR.
This procedura for esti-mating the transverse impact properties by downgrading the longitudinal properties is conservative and acceptable to the staff.
The staff concludes that an exemptior to Section III.B.1 is justified and granted.
Section III.B.3 of Appendix G requires tnat the temperature instruments and Charpy test machines be calibrated in accordance with paragraph NB-2360 of Section III of the ASME Code.
Verification of this required calibration was impossible because the testing organization retained the calibration report only until the next calibration was performed.
However, General Electric has stated that the test instruments and machines were routinely calibrated on a periodic basis.
Based on the standard practice of this period and on past experience with Charpy testing, the staff concludes that it is unlikely that the test instruments and machines were not adequately calibrated.
The staff further concludes that an exemption to the requirement of Section III.B.3 of Appendix G is justified and granted.
Section III.B.4 of Appendix G requires that the testing personnel be qualified by training and experience and that they be able to perform the test in accord-ance with written procedures.
For the Fermi 2 component testing, the written procedures required by the regulation did not exist.
However, the individuals were qualified by on-the-job training and past experience.
Because these tests are relatively routine in nature and are continually being performed in the laboratory that conducted these tests, it is unlikely that the tests were con-ducted improperly.
Consequently, the staff concludes that an exemption from Section III.B.4 of Appendix G is justified and granted.
Section III.C.1 of Appendix G requires that the reactor vescel beltline ferntic materials be Charpy V-notch impact tested at appropriate temperatures over a temperature range sufficient to define the C test curves (including the upper y
shelf levels) in terms of both fracture energy and lateral expansion of speci-mens.
In addition, the location and orientation f the impact test specimens shall comply with the requirements of paragraph NB-2322 of the ASME Code.
The available Charpy V-notch impact and dro9 weight NDT toughness data for the Fermi 2 beltline plates and welds were submitted for staff review on July 24, 1981.
Upper shelf toughness data were not required when the Fermi 2 reactor vessel was constructed.
These data were not available for the welds, but they were submitted for longitudinally oriented specimens for the plates.
The unirradiated RT and upper shelf impact energy values were conservatively NDT estimated by procedures acceptable to the staff, as discussed above regarding Sections III.A and III.B.1.
Te'st data obtained on materials from the Zimmer, LaSalle, and Lena Verde reactor vessels were included in the package submitted on July 29, 1981 to demonstrate that the estimated values for the Fermi 2 reactor vessel were conservative.
The staff concludes that an exemption to the require-ments of Section III.C.1 of Appendix G is justified by the data and is granted.
Section III.C.2 requires materials used to prepare test specimens for the reactor vessel beltline region be takei. directly from excess material and welds in the vessel shell course.
The applicant does not comply with this requirement in 1
5-3
that materials used to prepare weld test specimens for the reactor vessel belt-line region were not prepared using the sa.; metal plate associated with the weld in the reactor vessel.
The weld M t specimens were taken from simulated weldments prepared from excess production plate.
However, the weld wire and flux materials used in tha +est specimens are the same as those used in the reactor vessel beltline.
- ter weld preparation, the weldments were subjected to a heat treatment to obtsin metallurgical effects equivalent to those produced l
l during fabrication of the.eactor vessel.
Based on its evaluation of this l
information, the staff concludes that, although the same base raaterial was not used to prepare the test samples, an exemption from the specific requirements of Section III.C.2 of Appendix G is justified because the same heat treatment, l
weld wire, flux, and welding process used in the vessel welds were used in the test specimens.
Because the weld toughness properties are determined primarily by heat treatment, weld wire, flux, and welding process, and not by differences in similar base materials, the use of weldment test specimens having the same j
weld wire, flux, and heat treatment and made by the same welding process as
{
the vessel welds is sufficient to satisfy the requirements of Section III.C.2 of Appendix G.
Furthermore, this provides acceptable justification for an exemption to the exact requirements of Section III.C.2 of Appendix G.
Therefore, an exemption to Section III.C.2 of Appendix G is granted.
Section IV.A.1 of Appendix G requires that the RCPB ferritic materials meet the acceptance standards of paragraph NB-2330 of the ASME Code.
This paragraph specifies the basis for defining a reference temperature, RTilDT, based on the impact test data. The value of RT is defined by the ASME Code as the higher flDT of either the nil-ductility temperature, as defined by the dropweight test, or a temperature 60 F less than the temperature at which 50 ft-lb' energy or 35 mils lateral expansion is achieved, whichever is higher, by Charpy impact tests.
The Charpy impact tests are to be conducted using specimens oriented in the transverse direction.
Ferni 2 reactor coolant pressure boundary ferritic materials were impact tested to meet the requirements of Section III of the ASME Code, 1968 Edition, including the Summer 1969 Addenda.
This code edition f
required either dropweight tests or three CVN impact tests at a single tempera-J ture equal to 60 F below the lowest service temperature.
Further, there were l
no specific requirements for Charpy specimens for transverse orientation.
The staff has reviewed the impact data and the estimated RTNDT, including the data in the submittal of July 29, 1981, in conjunction with certain General Electric j
correlations and regulatory recommendations of Section IV. A.1 of Appendix G.
The staff concludes that the RT f r e ch of the ferritic components was NDT conservatively estimated and that an exemption to the exact requirements of Section IV.A.1 is justified and granted.
Section IV. A.2.c of Appendix G requires that when the core is critical, the metal temperature of the reactor vessel shall be high enough to provide an adequate margin of protection against Tracture, taking into account such factors as the potential for overstress and thermal shock during anticipated operational occurrences in the control of reactivity.
In no case when the core is critical (other than for the purposes of low-;evel physics tests) shall the temperature of the reactor vessel be less than the minimum permis-sible temperature for inservice system hydrostatic pressure tests nor less than 40 F above the temperature required by Section IV.A.2.a of Appendix G.
General Electric has proposed that 10 CFR Part 50 be amended to make this 5-4
change acceptable in the regulation. The proposed modification to 10 CFR Part 50, Appendix G, Section IV.A.2.c is described in GE Topical Report NED0-21778-A.
As previously reported in a November 13, 1978 memorandum from 0. D. Parr to Dr. G. G. Sherwood, the regulatory staff has reviewed this topical report, has found it acceptable, and concurs that the proposed alternative to the criticality hydrostatic temperature limit is acceptable.
The staff's previous evaluation and acceptance of this GE topical report provides sufficient information for the staff to conclude (1) that the proposed alternate method is equivalent to the current Appendix G requirement and (2) that an exemption to Section IV.A.2.c of Appendix G is justified and granted.
Section IV.A.3 of Appendix G, as amended September 26, 1979 (Federal Register, Volume 44, No. 188, pp. 55328), requires that material for piping, pumps, and valves meet the requirements of ASME Code paragraph NB-2332 and that material for bolting and other fasteners meet the requirements of ASME Code paragraph NB-2333.
However, these specific code paragraphs were not in effect when Fermi 2 was designed, nor were code paragraphs which could be considered to be of com-parable intent to paragraphs NB-2332 and NB-2333 in effect.
The components in question are the main steamline piping, the main steamline isolation valves, and the reactor vessel closure studs.
In the submittal of July 29, 1981, the applicant provided materials property information for the main steamline piping, the main steamline isolation valves, and the reactor vessel closure studs to show adequate fracture toughness properties for these components.
The staff concludes, from its evaluation of the impact data provided, that an exemption to the requirement of Section IV.A.3 of Appendix G is justified and granted.
Section IV.B of Appenoix G requires that the reactor vessel beltline materials have a minimum upper shelf energy (as determined from Charpy V-notch tests on unirradit.ted specimens in accordance with paragraph NB-2322 of the ASME Code) of 75 ft-lbs, unless it can be demonstrated to the Commission by appropriate data anc analyses that % er values of upper shelf energy wi11 provide adequate margin f ar deterioratir.~ 40m irradiation.
The impact tsts are required to be conducted using specimens that are oriented in the transverse '"rection (except for welds).
Charpy V-notch upper shelf impact data wer; not required when the Fermi 2 reactor vessel was fabricated.
However, upper sh'If impact data on longitudinally oriented specimens from the beltline plates were obtained.
The data were submitted on July 29, 1981.
Upper shelf data were not available for welds in the beltline region.
The Materials Engineering Branch Technical Position 5-2, " Fracture Toughness Requirements" (in NUREG-75/087), permits estimation of the transversely oriented upper shelf energy from longitudinally oriented specimens by the reduction of the longitudinally oriented values by a factor of 65 percent.
Further, a value of 70 ft-lbs is considered adequate for material for vessels which are subjected to fluences of less than 1 X 10" n/cm2 (E > 1 MeV).
The staff concludes from its review that the fracture tough-ness of tiie plates in the beltline region is adequate and complies with the requirement of Section IV.B of Appendix G.
Charpy V-notch upper shel, toughness values were not available for welds in the Fermi 2 reactor vessel.
However, impact test results were conducted at 10 F on procedure qualification test welds.
At this test temperature, the impact values were in excess of 75 f t-lbs, except for one weld material.
It was expected that if a higher test temperature had been used, the test results would have exceeded 75 ft-lbs.
Charpy V-notch upper shelf data were presented 5-5
__-_ _-_ a
for welds prepared by the same weld procedures as used for constructing the Fermi 2 reactor vessel, which confirmed this viewpoint.
The staff concludes from its review that an exemption to Section IV.B of Appendix G (which requires that each ferritic beltline material be individually tested to determine a minimum Charpy V notch upper shelf impact energy of 75 f t-lbs) is necessary and justified on the basis of the data presented in the submittal of July 29, 1981.
Based on these conclusions, an exemption to the requirements of Section IV.B of Appendix G is granted.
5.2.1.2 Compliance with Appendix H, 10 CFR Part 50 The staff has reviewed the FSAR and ancillary submittals that detail the extent of the compliance of Fermi 2 to Appendix H of 10 CFR Part 50 and has cetermined that the requirements have been met except for those of Sections II.B and II.C.1.
However, the applicant has submitted sufficient information to justify exemptions to those requirements.
Section II.B of Appendix H requires t'
- he surveillance program for the ferritic materials in the reactor vassal beltl cegion comply with the requirements of ASTM E 185-73, " Standard Recommendem Practice for Surveillance Tests for Nuclear Reactor Vessels."Section II.C.1 of Aopendix H requires that the surveillance specimens be taken from locations alongside the fracture toughness test specimens required by Section III of Appendix G and that these shall meet the testing requirements of Section III.A of Appendix G, including orientation.
Section It.B of Appendix H was not complied with in that the surveillance specimens were not necessarily fabricated from the limiting beltline mterial.
Sections II.B and II.C.1 of Appendix H were not complied with in onat some of the surveillance specimens were longitudinally rather than transversely oriented.
The staff has reviewed the surveillance program for Fermi 2 and concludes that the program will provide adequate informatien for assessing the irradiation damage to the beltline region of the reactor vessel even though limiting mate-rial may not be included in some capsules and transversely oriented specimens are not included in all capsules.
The staff may use empirically derived methods of conservatively estimating the amount of irradiation damage in ferritic belt-line materials on a case-by-case basi. when irradiation test data cannot be obtained for a specific material.
The methods are contained in Regulatory Guide 1.99, "Effect of Residual Elements on Predicting Radiation Damage of Reactor Vessel Materials." Further, the test cata from longitudinally oriented specimens will provide sufficient data to predict the relative shift in RT duetoneutronirradiationbecausespecificmaterialsserviceexperienceib-cates that the relative shift is not greatly sensitive to specimen orientation.
The staff concludes that the Fermi 2 surveillance program will monitor the change in fracture toughness properties of the ferritic materials in the beltline region of the reactor vessel to a degree adequate to determine the press 'e-temperature limits to preserve the integrity of the vessel.
The program will
'erate suffi-cient information to permit the determination of conditions unuer -
ch the reactor
.essel will be operated with an adequate margin of safety against 1.pidly propa-gating fracture throughout its service lifetime.
The staff concludes that exemp-tions to Sections II.B and II.C.1 are justified, and these are granted because the surveillance program will provide the information required to ensure that the Fermi 2 reactor is operated with an adequate margin of safety.
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(
5.2.1.3 Conclusion for Compliance with Appendices G and H, 10 CFR Part 50 The staff's technical evaluation has not identified any practical method by which the Fermi 2 reactor vessel can comply with the specific requirements of Sections I.8, III.A, III.B.1, III.B.3, III.B.4, III.C.1, III.C.2, IV.A.1, IV.A.2.c, IV.A.3, and IV.B of Appendix G and Sections II.B and II.C.1 of Appendix H of 10 CFR Part 50.
However, the alternate methods proposed by the applicant to demonstrate compliance with these sections have been reviewed and evaluated, and the staff has found them adequate to provide the safety margin required by Appendices G and H.
l Compliance with Appendices G and H of 10 CFR Part 50 and with the fracture i
toughness requirements of Section III of the ASME Code ensures that the ferritic components in the primary coolant pressure boundary will behave in a nonbrittle manner, that the probability of rapidly propagating fracture is minimized and that an appropriate material surveillance program exists to monitor radiation damage for the reactor pressure boundary.
Compli:nce with the requirements of the NRC regulations and the specified rodes and standards satisfies the require-ments of GDC 31 and 32.
Based on the foregoing and pursuant to Section 50.12 of 10 CFR Part 50, exemp-tions from the specific requi~ments of Appendices G and H of 10 CFR Part 50 are authorized by law and can be granted without endangering life or property or the common defense and security and are otherwise in the public interest.
The staff concludes that the public is served by not imposing certai, provisions of Appenaices G and H of 10 CFR Part 50 that have been determined te he either impractical or tnet would result in hardship or unusual difficulties ?;hout'a compensating increase in the level of quality and safety.
5.2.1.4 Pressure-Tamperature Limits fp endices G and H of 10 CFR Part 50 describe the Lunditions that require l
pressure-temperature limits for the N and provide the general bases for these limits.
These appendices specit a:lly require that pressure-temperature limits must provide RCPB safety margins at least as great as the safety margins recommended in the ASME Boiler and Pressure Vessel Code,Section III, Appendix G,
" Protection Against Nonductile Failure." Appendix G of 10 CFR Part 50 requires additional safety margins whenever the reactor core is critical, except for low-level physics tests.
The following pressure-temperature limits imposed on the RCPB during operation and tests are reviewed to ensure that they provide adequate safety margins against nonductile behavior or rapidly propagating failure of ferritic components as required by GDC 31:
(1) preservice hydrostatic test l
(2) inservice leak and hydrostatic test l
(3) heatup and cooldown operations (4) core operation Appendices G and H of 10 CFR Part 50 require the applicant to predict the shift in reference temperature due to neutron irradiation. The shift in RTilDT.due to establish the to neutron irradiation is then added to the initial RTilDT adjested reference temperature. The base plate or weld seam having the highest 5-7 t
= _.
I i
4 5
adjusted reference temperature is considered the most limiting material for i
determining the pressure-temperature operating limit.
In the cese of Fermi 2, the most limiting material at end-of-life appears to be weld seams15-308, which contain 0.32% copper and have an unirradiated RT f -44 F.
Once in service, NDT
- the pressure-temperature limits will be revised to reflect the actual r.eutron irradiation dam' age as determined from the results of the reactor vessel materials surveillance program.
The staff has evalaated the proposed operating limits during heatup, cooldown, and core operatioa shown in Figure 5.2-1 of the FSAR and concludes that the-pressure w rperoture limits' are acceptable for the projected end of-life fluences.
The staff requires-that the pressure-temperature limits be revised to reflect the actual neutron irradiation damage received, as determined from the results of the reactor vessel materitis surveillance program.
t The pressure-temperature limits-to be imposed on the reactor coolant system for all operating and testing conditions, to ensure adequate safety margins against nonductile or rapidly propagating failure are in conformance with establishe( criteria, codes, and standards acceptable to the staff.
The use of operati:.s limits based on these ' criteria (as defined by applicable regula-tions, codes, and standards) provides reasonable assurance that nonductile or rapidly propagating failure will not occur; this constitutes an acceptable basis for satisfying the applicable requirements of GCC 31.
5.3 Integrity of the Reactor Vessel The staff has reviewed all the factors contributing to the structural integrity of the' reactor vessel and concludes there are no special considerations that j
make it necessary to consider potential reactor pressure vessel failure for this plant.
The bases for this conclusion are that the design, materials, fabrication, inspection, and quality assurance requirements for Fermi 2 conform to the Commission's regulation and Regulatory Guides and to Section III of the ASME Boiler and Pressure Vessel Code.
Exemption has been granted to certain specific requirements of Appendices G and H and of subparagraph 50.55a(c)(2) of 10 CFR Part 50 where it was demonstrated that alternate provisions assured an equivalent level of qual'ty and safety.
The stringent fracture toughness requirements of the regulations and Section III of the ASME Boiler and Pressure Vessel Code, including requirements for surveillance of material properties, will provide adequate assurance of integrity throughout the plant service life.
Pressure-temperature limits will be established in accordance to Appendix G of I
Section III of the Code and Appendix G of 10 CFR Part 50.
The integrity of the reactor vessel is assured because the vessel (1) is designed and fabricated to the high standards of quality required by the ASME Boiler and Pressure Vessel Code (2) is made from materials of controlled demonstrated high quality i
(3) was subjected to extensive preservice inspection and testing to provide assurance against deficiencies in materials and workmanship (4) will be operated under conditions and procedures and with protective devices that provide assurance that the reactor vessel design conditions 5-8
I will not be exceeded during normal reactor operation, and that'the vessel will not fail under the conditions.of any of the postulated accidents
-(5) will be subjected to periodic inspection to demonstrate that the high initial quality of the reactor vessel has not deteriorated significantly under service conditions (6) may be annealed in the field to restore the material toughness properties should this become necessary.
5-9
l 6 ENGINEERED SAFETY FEATURES 6.2 Containment Functional Design 6.2.7 Containment Leakage Testing The staff re/iewed the applicant's containment leak testing program for compli-ance with the containment leakage testing requirement specified in Appendix J to 10 CFR Part 50.
Such compliance provides adequate assurance tnat the contain-ment leak-tight integrity can be verified throughout the service lifetime of the plant and that the leakage rates will be periodically checked during service on a timely basis to maintain such leakage within the specified limits.. Main-taining containment within such limits provides reasonable assurance that, in the event of any radioactivity release within the containment, the loss of the containment atnosphere through potential leak paths will not be in excess of-the limits spe.ified for the site.
Specifically, the staff reviewed the containment leak testing program to assure that the containment penetrations and system isolation valve arrangements are designed to satisfy the containment integrated leak-rate testing requirements and the local leak testing requirements of Appendix J.
The proposed leak testing practices which differ from the explicit requirements of Appendix J and are acceptable on other defined bases are discussed below.
(1) -Reverse Direction Testing Certain valves will be tested in the direction opposite to the one in w;.ich they perform their containment isolation function.
The results of these tests will be either (a) equivalent to testing in the accident direction because the valves have the same sealing characteristics in either direction, or (b) conserv-ative because test pressure tends to unseat the valve disc.
The staff, therefore, finds this test method acceptable.
(2) Hydrostatic Testing of Containment Isolation Valves Appendix J to 10 CFR Part 50 requires that valves, unless pressurized with fluid I
from a seal system, shall be pressurized with air or nitrogen for leak testing purpose:. (see Paragraph III.C.2).
There are a number of liquia-filled systems, however, that are specifically designed to remain intact following a loss-of-coolant accident and thus provide a water seal for the system isolation valves or assure that only liquid leakage from the containment will occur.
The applicant proposes to perform hydrostatic testing to determine the leak tightness of isolation valves in the following system valves:
(a) Torus pressure and liquid level instrumentation and torus water management system suction and injection.
(b) Residual heat removal (RHR) minimum flow, RHR heat exchanger relief and thermal relief, steam condensing mode header relief, steam condensing mode 6-1
ten. 1 1, RHR heat excharger vent line, liquid sample return, RHR pump suctio,, and pump suction header thermal relief.
(c) High pressure coolant injection (HPCI), reactor core isolation cooling (RCIC) and core spray pump suction, core spray pump suction thermal relief, pump discharge header relief, pump minimum flow, and pump test line, HPCI, and RCIC minimum flow.
For the isolation valves in these system lines, the applicant has shown that (1) the flow paths associated with these lines terminate below the low water level in the suppression pool, (2) a water seal is assured during normal plant operation and for more than 30 days following an accident requiring containment isolation, and (3) it is not credible that these isolation valves will be exposed to the containment atmosphere at any time following the accident.
The combined leakage from all these valves will satisfy the guidelines of 10 CFR Part 100 regarding the site radiological safety analysis and will be included in the plant Technical Specifications. This leakage will be excluded when determining the combined leakage rate for all penetrations and valves as specified in Paragraph III.C.3 of Appendix J.
The staff reviewed the applicant's proposed hydrostatic testing program and concludes that such testing is acceptable for the lines identified above.
(3) Control Rod Drive System Appendix J to 10 CFR Part 50 (see Paragraph III.C.3.(a)) allows exclusion (from combined 0.6La) of leakage from valves that are sealed with fluid from a seal system if the f'Jid leakage rates do not exceed those specified in the Technical Specification Leakage from the CR0 system into the reactor building will be detected for the full spectrum of leakage rates.
Small leaks will be detected by observation during daily inspection rounds of the control unit areas by operators.
Large leaks will be detected by duty-timers on the reactor building floor drain sump pumps.
(The floor drain sump collects leakage, up to a leakage rate limit of 5 gpm, from CRDs, valve flanges, floor drains, the closed cooling water system, drywell cooling unit drains, and other potential sources.) A large leak of reactor coolant from any insert line will-be automatically isolated by the ball check valve in the CRD housing.
Leaks of CRD supply water will be indicated by increased flow as continuously recorded in the control room.
The CRD directional cor. trol valves are normally closed and are automatically closed upon a reactor scram signal.
Excessive leakages through the scram valves will be detected by duty timers on the sump pumps.
Leakages through the eight insert and eight withdrawal spare penetration tubes will be detected during Type A tests.
The staff concludes that the leak testing provisions for the CRD system are an acceptable alternative to the requirements of Appendix J.
(4) Traversing Incore Probe System Testing As discussed in Section 6.2.4, the valve arrangement provided on each guide tube consists of a ball valve and a cable-shearing explosive valve mounted in the guide tube just outside the primary containment.
The guide tube ball valve 6-2
opens only when the traversing incore probe (TIP) is being inserted.
The ball valve position is indicated in the control room.
The shear valve is used only if containment isolation is required when the TIP is beyond the ball valve and when power to the TIP system fails.
In that event the shear valve will cut the cable and close off the guide tube. The shear valves are actuated by detonation squibs.
The continuity of the squib circuits is monitored by indi-cator lights in the control room.
The staff requires the applicant to perform Type C tests on the ball valves, at regular test intervals (that is, at least every 2 years).
Because the shear valve requires testing to oestruction, the applicant need not jerform periodic Type C tests on these valves.
However, identical design shear valves will be shop tested by statistical sampling methods to ensure operability and leak tightness.
To assure that the shear valve will perform its intended function, the staff requires the applicant to include in the Technical Specifications provisions to:
(a) Verify the continuity of the explosive charge at least once every 31 days.
(b) Initiate one of the explosive squibs charge at least once every 18 r.rnths.
The replacement charge for the explosive valve shall be from the same manufactured batch as the one fired or from another batch which has been certified by having one of that batch successfully fired.
(c) Replace all charges according to tM manufacturer's recommended life time.
The staff will report on the resolution of this issue in a supplement to the SER.
(5) Purge Valve Testing Drywell air purge inlet and exhaust containment isolation valves will be Type C tested (Appendix J) at least every 2 years.
However, recent reports have indicated that resilient seats in the large butterfly valves of the purge system may deteriorate unacceptably.
As a result, the staff has determined that more l
frequent periodic leakage tests are required for butterfly valves which have these resilient seats.
The applicant has agreed to-test the butterfly isolation valves in the purge penetrations once every 90 days.
The purpose of the leakage integrity tests is to identify excessive degradation of the resilient seats for thE a valves.
These tests would be performed in addition to the quantitative Type C tests required by Appendix J, and would not relieve the licensee of the responsibility to conform to the requirements of Appendix J.
The applicant has committed to repair valves in which measured leakage increases by 20 percent or more from the similar tests performed at the time of the previous Type C test.
The staff finds these provisions for purge valve testing to be acceptable.
(6) Secondary Containment Leakage Following a postulated LOCA, the pressure in the secondary containment could increase (as discussed in SER Section 6.2.3).
The applicant has committed to perform preoperational and inservice tests of the ability of the secondary 6-3
I containment to draw down to minus 1/4-in. water gage under a design inleakage of 3000 scfm.
The inservice test will be conducted at least once every 18 months, as specified in the Technical Specifications. The staff finds this test procedure for secondary containment leakage acceptable.
(7) Bypass Leakage Bypass leakage is leakage which passes untreated from the primary containment beyond the secondary containment and possibly to the environment.
The applicant has committed to a maximum bypass leakage equal to 4 percent of L the maximum allowableleakrateintheTypeAcontainmentintegratedleakrath, test. Most of the valves identified are primary containment isolation valves located close to the drywell boundary and penetrations and, as such, are tested according to the Type C (Appendix J) provisions.
In some instances, however, only a portion of the flow through a penetration could pass, via a smaller branch line, through the secondary containment.
In these instances, the applicant proposes to designate only the actual bypass i
path and its accompanying valves as being part of the bypass leakage test program, not the entire penetration, because the majority of flow will be contained within the secondary containment.
The applicant has identified four lines in which the bypass leakage valve is not one of the main containment isolation valves:
HPCI steam line drain to main condenser RCIC steam line drain to main condenser HPCI pump discharge to condensate storage RCIC pump discharge to condensate storage Concerning bypass leakage valves in these lines, the applicant has stated that (1) Each line has two valves in series, i
(2) Each valve will be tested according to the specifications of ASME Code Section XI.
(3) All valves are located within the secondary containment.
In addition, the staff requires the applicant to make system modificationc i,o ensure that (1) All valves are snplied with divisional powei.
(2) All valves have diverse isolation signal provisions.
The staff will report on this issue in a supplement to the SER.
Conclusions 1
The staff has reviewed the applicant's proposed leak testing program and concludes that there are two outstanding open items to be resolved:
(1) traversing incore probe system testing, and (2) modifications to the HPCI and RCIC systems bypass leakage valves.
The staff will report on the resolution of these items in a j
supplement to the SER.
6-4
e i
The other aspects of the applicant's proposed leak testing program meet the requirements of Appendix J to 10 CFR Part 50 and are acceptable.
6.3 Emergency Core Cooling System 6.3.4 Evaluation Findings 6.3.4.1 Safety Concerns Associated with Pipe Breaks in the BWR Scram System In the last paragraph of SER Section 6.3.4, it was noted that the staff was reviewing the GE Topical Report NEDO-24342, which responded to a staff concern regarding a break in the control rod drii e scram discharge volume.
The staff has completed the generic re"ies of this concern.
By a letter dated August 31, 1981 to each BWR licensee ano applicant, the staff forwarded its generic safety evaluation on this issue (NUREG-0803) and requested that a plant-specific response be provided to support issuance of an operating license.
NUREG-0803 includes an evaluation of the licensing basis for the BWR scram discharge volume (SDV) piping and an assessment of the potential for the SDV piping to fail while in service.
A discussion of the means available for miti-gating an unlikely SDV system failure is provided.
Generic recommendations are made to improve mitigation capability and ensure that system integrity is maintained in service.
The staff has concluded that the SDV piping system design is acceptable, provided certain conditions are satisfied on a plant-specific basis.
The staff further concluded that the safety concerns associated with a postulated failure of the SDV piping system do not represent a dominant contribu-tion to the risk of core melt, provided certain assumptions used in the risk assessment are validated on a plant-specific basis.
The staff's guidance for implementation of recommended plant-specific actions is provided in NUREG-0803 and summarized for Fermi 2 herein.
The guidance has been grouped into three areas:
piping integrity, mitigation capability, and environmental qualifications.
(1) Piping Integrity To ensure SDV piping integrity, the quality assurance of SDV piping should be verified by providing the results of as-built inspections and seismic reanalysis of the SDV piping and its supports.
If such an inspection has not been previously conducted, the apolicant should conduct an as-built inspection and seismic reanalysis of the SDV piping and its supports and provide a schedule for correct-ing any deficiencies identified.
The SDV hydraulic control unit maintenance, surveillance, and modification procedures should also be reviewed and revised.
Finally, the applicant should propose a program of periodic inservice inspection for the SDV system meeting the requirements for Class 2 piping in the Section XI i
)
ASME Code.
This revised inservice inspection program should be implemented on a schedule consistent with the requirements of the applicant's inservice inspection program for ASME Section XI Class 2 piping.
6-5
l (2) Mitigation Capability (a) Emergency Operating Procedures Guidelines The staff has considered comments from GE and the BWR industry in judging the adequacy of the presently installed mitigation capability for a postu-lated pipe break failure in the BWR scram system following a reactor trip.
The staff.has also considered the perspective provided by the results of its independent quantitative risk assessment, which concluded that the SDV break was not a dominant contributor to core melt risk.
However, this assessment was based on several important assumptions that will require validation on a plant-by plant basis.
Should the plant-specific reviews show that one or more key equipment qualification or operator performance
)
assumptions were not correct (and nonconservative), then the conclusions 1
I stated in this section will be reexamined and may be revised on a plant-specific basis.
The mitigation sequence for the postulated SDV break accident could not, i
for all scram initiators, be based on reliable and early closure of the scram outlet valves for break isolation, followed by an orderly shutdown of the unit.
The mitigation sequences for other potentially large breaks in lines penetrating primary containment are based on closure of qualified and redundant isolation valves.
The potential for noncleared trip conditions, loss of control air, or other factors that could prevent successful closure of the scram valves requires that acceptable mitigation be based on demon-strating adequate long-term core cooling capability, which involves reactor coolant system (RCS) depressurization with the break unisolated, followed by entry into the reactor building to terminate the leak using the manual isolation valves associated with each control rod drive (CRD) hydraulic l
control unit.
Reactor system depressurization with adequate core cooling involves control room operator recognition of the need to depressurize and the ability of the depressurization and core cooling systems to perform their required functions in the adverse environment created by the break.
For a plant adequately following the guidance summarized in Table 6-1, the staff would consider the presently available break-detection instrumen-tation to be adequate to detect an SDV break or other equipment-threatening leak outside primary containment, thus enabling the timely and correct operator action and equipment performance required for acceptable mitigation.
This would require that the BWR Owners' Group Emergency Procedure Guidelines be evised to lead the control room operator into a controlled blowdown whtn plant conditions indicate a reactor system leak outside the drywell.
Implementation of procedures based on such revised guidelines would allow the staff to conclude with reasonable assurance that an operator using
(
these revised emergency procedures would take the appropriate actions necessary to successfully mitigate a leak or break in the SDV or elsewhere in the secondary containment.
Specifically, the presently installed detec-tion equipment--in conjunction with procedures that require a prompt blow-down (in excess of the 100 F/hr cooldown rate) whenever a trip condition that cannot be reset occurs coincident with indication of a leak in the reactor building or a leak that cannot otherwise be promptly isolated--would be acceptable.
The rate of depressurization must be compatible with the 6-6
T
)
qualification of equipment in the reactor building to withstand an unisol-able liquid break.
It is expected that the revised guidelines will direct a complete RCS depressurization within 1 hr of break indication in secondary containment.
An early blowdown with a minimum number of SRVs would be preferable to a full but delayed automatic depressurization system (ADS) blowdown.
Even so, final staff acceptance of the modified, improved proce-dures will eventually require the support of sufficient equipment qualifi-cation information documentation, as summarized in Table 6.1, to allow the staff to conclude that the depressurization system and other required emergency core cooling system equipment are capable of performing the functions required for successful mitigation.
Table 6.1 Schedule for Implementation of Recommended Plant-Specific Actions for Fermi 2 Section in NUREG-0803 Guidance Schedule 2.1 Periodic inservice inspection and September 1, 1982 3.1.2 surveillance for the SDV system 3.1.1 Threaded joint integrity (if applicable)
January 1, 1982 3.2.1 2 Seismic design verification January 1,1982 3.2.1.8 HCU-SDV equipment procedures review January 1,1982
- 2. 3 Environmental qualification of January 1, 1982 prompt depressurization function 3.1.1 As-built inspection of SDV piping September 1, 1982 and supports 4.2.3 Improvement of procedures January 1,1982 4.3.1.1 Verification of equipment designed January 1, 1982 l
for water impingement 4.3.1.1 Verification of equipment qualified January 1, 1982 for wetdown by 212 F water 4.3.1.3 Verification of feedwater and condensate January 1, 1982 system operation independent of the reactor building environment 4.3.1.3 Evaluation of availability of HPCI-January 1, 1982 LPCI turbines due to high ambient temperature trips 4.3.2.3 Verification of essential components January 1, 1902 4.4.1 qualified for service at 212 F and 100% humidity 6-7
If the applicant intends to show acceptable mitigation on this basis and to follow the specific guidance provided above for plant-specific submittals, the applicant should provide a commitment to implement the required revised emergency procedures in the plant-specific responses.
As an alternative acceptable basis for showing SDV break-mitigation, the applicant may provide procedures that would permit a complete RCS depres-surization 1 hr or more after break detection.
However, these procedures must be supported by the documentation of acceptable ECCS equipment quali-fication which bounds the conditions predicted by an adequately conservative plant-specific ~ analysis (and model) of the reactor building for the maximum credible SDV leak rate.
(b) Reactor Water Specific-Activity Limits GE provided data demonstrating that actual reactor water concentrations are lower than the Standard Technical Specification (STS) values by about 1 order of magnitude.
From this fuel performance data, the staff has con-cluded that the STS provides a reasonably conservative upper limit of reactor water iodine concentration.
Fermi 2 will use STS limits.
The staff has concluded that, if the plant operates within STS coolant-activity limits, the probability of requiring operator access to the reactor building is consistent with the staff's quantitative risk assessment (pipe break plus failure to reset scram).
The staff's analysis also showed that the offsite doses would not exceed 10 CFR Part 100 guideline values if the plant operates within STS coolant-activity limits.
(c) Environmental Qualification Preliminary results from studies modeling a typical reactor building response to the postulated SDV failure indicate that the expected environ-mental conditions would be less severe than the environmental conditions I
postulated for a steamline break in the reactor building.
However, the staff cannot make a definite generic conclusion because specific equipment in the SDV area has not necessarily been qualified for the local conditions that would exist.
The staff recognizes that the SDV piping system is excluded from the scope of the Commission's Order CLI-80-21, regarding environmental qualification reviews; however, the Order was issued before the j
recognition of this concern.
Based on its evaluation of this concern, the staff requires that the adequacy of the qualification of the required miti-gation equipment be verified.
Therefore, applicant's plant-specific responses should Identify the equipment that would be used to detect a break and/or leak in the SDV system and include the qualification of this equipment in the Fermi 2 equipment qualification (EQ) program to show that it would perform the identification function.
Identify the equipment needed to nitigate an unisolable break in the SDV system and include the qual Mication of this equipment in the 6-8
l Fermi 2 EQ. program to show that it would perform the mitigation function, paying particular attention to the guidance summarized in Table 6.1.
When the qualification of this equipment is provided, qualification for service at a given temperature and humidity shall mean that the equipment is capable of remaining on standby and of being energized and operated in the presence of the actual SDV break environmental transient (including specified level of adverse environment as a maximum condition) without any immediate or subsequent loss of required 5
SDV break-mitigating capability.
The SDV break environmental profile assumed for qualification purposes shall be prescribed as to temperature and humidity levels and time duration at each level; it shall be a conservative replica of the postulated transient.
For any equipment required for identification and/or mitigation that is not qualified for service at 212 F and 100-percent humidity, provide the plant-specific SDV break environment and a commitment to qualify the equipment in the Fermi 2 EQ program.
The schedule for applicant's submittal of plant-specific responses to areas of concern identified in NUREG-0803 is provided in Table 6-1.
This issue must be satisfactorily resolved prior to the issuance of an operating license.
The staff will report its evaluation of the Fermi 2 plant-specific responses in a future supplement to the SER.
)
i 6-9 J
7 INSTRUMENTATION AND CONTROL 7.4 Systems Required for Safe Shutdown 7.4.2 Specific Findings The SER subsection entitled " Residual Heat Removal System - Reactor Shutdown Cooling Mode" stated that acceptable procedures for testing interlocks of pres-sure isolation valves during plant operation must be provided to make the Fermi 2
~
design an acceptable alternative to the recommendation in the Instrumentation and Control Systems Branch (ICSB) Technical Position ICSB-3 (NUREG-75/087) for-diverse pressure sensors to the interlocks.
By letter dated July 2,1981, the applicant identified test procedures that will be used to verify operability of the Fermi 2 interlocks during plant opera-tion.
The staff has reviewed these procedures and finds them acceptable and will verify that the procedures are implemented before the operating license is issued.
7.5 Safety-Related Display Instrumentation 7.5.2 Loss of Power to Instruments and Control Systems The staff asked the applicant to review the adequacy of emergency operating procedures to be used to attain safe shutdown upon loss of any Class IE or nonclass IE buses supplying power to safety related or nonsafety-related instruments and control systems.
(This issue was addressed for operating reactors thrcugh IE Bulletin 79-27.) The applicant responded to this concern in FSAR Amendment 35, in which the applicant stated that no modifications are required at Fermi 2 to ensure cold shutdown in the event of the loss of power occurrences discussed in Bulletin 79-27.
However, the staff was concerned whether the applicant had addressed all of the applicable systems that are pertinent to this issue.
The applicant responded by documenting an expansion of the scope of review that included 28 systems or subsystems.
Each system was reviewed for redundant safety functions accomplished by Division I and Division II to determine electrical independence of the power distribution to redundant devices.
The review also verified that the balance-of plant loads could not degrade both divisions of the Class IE instrumentation power supplies.
The results of this review con-firmed the applicant's earlier conclusions that a loss of a safety-related or nonsafety-related instrumentation and control bus would not jeopardize the ability to safely achieve cold shutdown.
The staff also asked the applicant to list procedures for safe shutdown if a particular Class IE or nonclass IE bus supplving power to safety-related or nonsafety-related instrument and control systems is lost.
The applicant responded by supplying additional information regarding the operating procedures. A loss of a safety-related or nonsafety-related instrument bus is interpreted by the operator to be a loss of part or all of any system powered by that bus.
This issue is addressed in the event-based Abnormal Operating 7-1
Procedure specific to the particular component or system.
In the event a safety-related instrument bus is lost, the procedure may instruct the operator to shut down, consistent with the requirements of the Technical Specifications.
For the loss of nonsafety-related instrumentation, the procedures direct the operator to reestablish system operation through the use of alternate available equipment or to reduce the plant load, as necessary, to compensate for the loss of the nonsafety-related equipment.
In the event the nonsafety-related instru-ment bus losses are compounded through multiple failures, the symptom-based Emergency Operating Procedures would be utilized to mitigate the failure and ensure a safe shutdown of the reactor.
In all cases, the operator is given direction to mitigate the effects of instrument bus losses.
t These additional responses satisfied the staff's concerns regarding the appli-cant's original response to IE Bulletin 79-27.
The staff, therefore, concludes that the applicant has demonstrated that sufficient equipment for safe shutdown would remain available after the loss of any IE or non-IE electrical bus and that the applicant has adequate emergency operating procedures to deal with the resulting plant conditions.
7.7 Control Systems Not Required for Safety 7.7.2 Specific Findings As stated in the subsection in the SER entitled "High Energy Line Breaks and Cansequential Control System Failures," before startup after the first Fermi 2 refueling, the applicant is required to provide results of a review to resolve concerns that high energy line breaks might cause control system malfunctions with consequences more severe than those analyzed in FSAR Chapter 15.
The staff stated that the operating license would be conditioned to reflect this requirement.
The applicant performed a review of high-energy line breaks using a format established by the BWR Owners' Group.
From this review, the applicant concluded that no identified safety action would be negated by the failure of nonsafety equipment resulting from the environmental effects of a high-energy line break.
This conclusion is also supported by the plant analysis for pipe breaks contained in Appendix C of the FSAR.
The only area of concern was the temperature effects of the pipe break on the level instrumentation sensing lines; this has been addressed and resolved in the GE Generic BWR report NED0-24708.
The air-operated head vent valves on Fermi 2 have been replaced with qualified electrically operated valves, thus eliminating a concern identified by previous reviews.
The applicant's review of high-energy line breaks and consequential control system failures has satisfactorily relieved this concern; therefore, the staff concludes that a license condition is not required.
7-2
i 11 RADI0 ACTIVE WASTE MANAGEMENT The staff's detailed evaluation of the radwaste systems is in Chapter 11 of the SER.
Subsequent to the issuance of the SER, some changes in the liquid and solid radwaste systems have been submitted by the applicant-in FSAR Amendment 38 (dated These changes involve significant changes in the design of the July 1981).
solid radwaste system and the modification of the liquid waste system, including revised estimates of radioactivity releases and corresponding doses to an indi-vidual and the population surrounding the site.
The applicant has chosen to utilize the volume reduction and solidification system designed by the Werner-Pfleiderer Corporation (WPC). Tnis system uses asphalt as a solidification binder for the wet waste.
Asphalt and the waste are mixed in an extruder / evaporator heated process that simultaneously removes the moisture from the waste while producing a homogeneous mixture.
This mixture converts into a solid product when it cools.
The system, on a generic oasis, was described in a Topical Report (WPC-VRS-1) prepared by WPC in November 1976.
This report has been approved by the staff for reference by users of similar The applicant has provided specific information regarding the solid systems.
radwaste system for Fermi 2 in Chapter 11 of the FSAR.
The applicant has also provided information about the modified liquid waste system to show the compli-ance with the requirements of 10 CFR Part 20, 10 CFR Part 50, Appendix I to 10 CFR Part 50, and other applicable regulations.
The staff will review the specific information about the liquid and solid radwaste systems provided by the applicant and report its findings in a future supplement to the SER.
=
11-1
13 CONDUCT OF OPERATIONS 13.5 Industrial Security In the SER, the staff inkated that certain areas of the security plan, contin-gency plan, and guard tro.oing and qua?ification plan required revision.
The applicant submitted Amendment 2 to the Physical Security Plan for the protec-tion of the plant against potential acts of radiological sabotage on July 20, 1981.
The staff has this document under review.
The staff will report its conclusion on the acceptability of the revised plan in a future supplement to the SER.
Additionally, the applicant has submitted a revised Safeguards Contingency Plan and a Security Force Training and Qualification Plan.
These plans have been reviewed by the staff and have been found to be acceptable.
The identification of vital areas and measures used to control access to these areas, as described in the Physical Security Plan, may he subject to amendments based on a confirmatory evaluation of the plant to determine those areas where acts of sabotage might cause a alease of radionuclides in suf'icient quantities to result in dose rates equal te or exceeding 10 CFR Part 100 limits.
The applicant's plans are being withheld from public disclosure in accordance with Section 2.790(d)(1) of 10 CFR Part 2.
13.6 Operating and Maintenance Procedures The steif has reviewed the applicant's plan for development and implementation of operating and maintenance procedures as described in the FSAR.
The review was conducted to determine the adequacy of the applicant's program for ensuring that routine operating, offnormal, and emergency activities are conducted in a safe manner.
The following description and evaluation are based on information contained in the FSAR, as well as on supplemental information obtained during the review.
10 CFR Part 50.34 and Regulatory Guide 1.33 were the criteria used to determine the acceptability of the applicant's program.
In the FSAR, the applicant has committed to a program in which all activities affecting safety-related structures, systems, and components are to be conducted in acurdance with detailed written and approved procedures which meet the requirements of Regulatory Guide 1.33 and ANSI 18.7-1976/ANS 3.2.
The applicant uses the following categories of procedures for those operations performed by licensed operators in the control room:
general operations system operations, alarm response, abnormal operations, emergency operations, and surveillance.
Other operating and maintenance procedures include the following areas:
instrumentation and control, reactor engineering, health physics, chemistry control, radioactive materials handling, fire protection, radiological emergency plan, security, fuel handling, radwaste handling, and environment.
Based on its review, the staff found that the applicant's program for developing operating and maintenance procedures meets the relevant requirements of 10 CFR 13-1
Part 34 and is consistent with the guidance provided in Regulatory Guide 1.33 and ANSI 18.7-1976/ANS 3.2.
Therefore, the staff concludes that the applicant's program is acceptable.
The staff's evaluation of the applicant's programs for reanalysis of transients and accidents and for development of upgraded emergency operating procedures pursuant to TMI-2 Action Plan requirements (NUREG-0737) are in Section 22, Items I.C.1 and I.C.8, of this supplement.
13-2
P 15 SAFETY ANALYSIS 15.1 Abnormal Operational Occurrences In the SER (page 15-5), the staff stated it was reviewing applicant's safety analysis of postulated turbine trip and generator load rejections without turbine bypass.
This analysis takes credit for steam flow to the reheater.
The staff now has completed its review of this analysis.
The ODYN* calculations for turbine trip without bypass were made with a specific flowrate-versus-time relation for steam flow to the reheaters.
Flow to the reheaters from the main steam header was also simulated Ly modifying the flowrate-versus-time relation for the main turbine.
In order to demonstrate that the-treatment of the reheaters in the ODYN calculations was conservative, the appli-cant submitted the results of several RETRAN calculations. The RETRAN calcula-tions included simulation of the reheater lines, reheaters, and associated flow paths to a seal tank and feedwater heaters, as well as the reactor and main steam system.
The calculations indicate that the ODYN simulation of the effect of the reheaters on minimum critical power ration (MCPR) should be conservative.
However, the staff believes that there are significant uncertainties in the heat-transfer relations for the reheater and in the simulation of water separation in the reheaters.
In view of these uncertainties, the staff requires experimental confirmation that the reheater steam-flow relation used in the ODYN analysis is conservative.
Therefore, the staff will include a condition in the operating _
license requiring that reheater steam flow measurements be made for turbine-trip and load-rejection tests during plant startup.
An acceptable steam flowrate from the main steam header to the reheaters is a value greater than or equal to that used in the ODYN calcuations during the first 2 sec after turbine trip.
s 15.2 Accidents 15.2.1 Anticipated Transients Without Scram In the SER, the staff stated that the applicant had committed to develop a plant-specific procedure for anticipated transients without scram (ATWS) using reactivity control guidelines being developed by a GE owners' group.
Ihe Fermi 2 ATWS procedure was submitted by the applicant.
It was reviewed by the staff, who also discussed it with Fermi 2 operations personnel during the review of selected emergency operating procedures (see Item I.C.8, Section 22.2 of this supplement).
Necessary changes were made to the entry conditions to ensure that e
f the operators are provided with the information necessary. to recognize an ATWS.
5 The staff has recommended to the Commission that rulemaking be used to determine I
any future modifications necessarv to resolve ATWS concerns and the required
[
schedule for implementation of st.d modifications, d
1 f
- 0DYN Computer Code; see GE Topical Report NE00-24154.
15-1
Based on :ss evaluation, the staff concludes that the Fermi 2 ATWS procedure provides an acceptable basis for licensing and interim operation of Fermi 2, pending ihe outcome of the proposed rulemaking on ATWS, and is in accordance with GDC 10, 15, 26, 27, and 29 in Appendix A to 10 CFR 50.
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16 TECHNICAL SPECIFICATIONS In the SER, the staff identified certain conditions which must be included in the Fermi 2 Technical Specifications in addition to Standard Technical Specifica-tions.
In this supplement, the staff has identified the need for additional plant-specific Technical Specifications.
The complete list of plant-specific Technical Specifications is given below; they are discussed further in sections of the SER or this supplement as indicated.
Technical Specifications Identified in the SER (1) Shore barrier annual inspection (2.4.2.5)
(2) Leakage tests for pressure isolation valves (3.9.6.2)
(3) Core performance (4.4.1)
(a) Prohibition of natural circulation and single-loop operation.
(b) Surveillance of core flow and flow-biased scram.
(4) Augmented surveillance tests of control rod drives (4.5)
(5) Periodic venting of ECCS and RCIC discharge lines (5.4.1, 6.3.4)
(6) Surveillance tests of low pressure coolant injection system (5.4.2)
(7) Surveillance tests of torus-to-drywell vacuum breakers and drywell-to-torus vents (6.2.1)
(8) Surveillance tests of secondary containment inleakage and drawdown time (6.2.3)
(9) Limiting containment purge to 90 hr/yr (6.2.4)
(10) Surveillance tests of mainsteam isolation valve leakage control system (6.2.5)
(11) Surveillance tests of selected control systems (7.1.2)
(12) Surveillance tests of high reactor vessel level trip (7.7.2)
(13) Surveillance tests of swing bus transfer (8.3.1)
(14) Limitation on battery charger operation (8.3.2)
(15) Reactor building crane inspection and maintenance (9.1.4)
(16) Monitoring of spent fuel racks material (9.1.5)
(17) Surveillance tests and inspection of turbine stop valves, turbine control valves, and turbine bypass valves (7.7.2, 10.2.1) 16-1
1 1
(18) Response time tests nf thermal power monitor (15.1)
(19) Prohibition of operation with partial feedwater heating (15.1)
(20) Annual reports of failures of relief and safety valves to close and of outages of ECCS equipment (Section 22.2, Items II.K.3.3 and II.K.3.17)
Additional Technical Specification Identified in this Supplement (1) Leakage testing of containment isolation valves and valve operators (this supplement Section 6.2.7).
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18 REPORT OF THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS s
During its 256th meetii.g August 6 to 8, 1981, the Advisory Committee on Reactor Safeguards (ACRS) completed its review of the application by the Detroit Edison Company for a license to operate Fermi 2.
A subcommittee meeting was held on July 24, 1981, and a tour of the facility was made on July 15, 1981.
A copy of the ACRS report on its review of Fermi 2 appears as Appendix C to this supplement.
In its report, the ACRS made recommendations for the resolution and completion of outstanding issues in the review.
The staff has considered the recommenda-tions as described below.
(1) Modifiea Mark I Containment The ACRS noted that applicant will confirm by analysis, including inplant tests of SRV operation, that the modified Mark I containment meets staff's acceptance criteria for LOCA and SRV hydrodynamic loads (NUREG-0661).
The ACRS recommended that this matter be resolved in a manner satisfactory to the staff before full-power operation and asked to be kept informed of the resolution.
As described in SER Sections 3.8.1, 3.8.2 and 3.9.3, the applicant has committed to resolve this matter on the following schedule that is acceptable to the staff:
Before May 1, '982, applicant will submit the plant-unique containment analysis, including a structural analysis of the torus internal structures and piping, and a fatigue evaluation of the discharge piping for SRVs. As noted in Section 3.8.4 of this supplement, the torus uplift load for a LOCA must be confirmed to be less than 1370 kips per torus support, based on the applicant's seismic reassessment.
f 1
Before August 1, 1982, the applicant will submit analyses of torus-attached piping systems and components, including information needed for seismic and dynamic qualification of equipment important to safety (see Section 3.10 of this supplement).
Before fuel loading, the applicant will have completed all (major and minor) containment modifications based on the containment plant-unique analysis.
Before the plant returns to power after the first refueling, the applicant will have completed all modifications based on the torus-attached piping analyses.
The staff will condition the operating license to require modifications to be completed at this time.
The staff will conduct a moderately detailed, postimplementation audit review of the plant-unique containment analysis and the torus-attached piping analysis.
Franklin Institute will perform the structural audit review and Brookhaven National Laboratory will perform the load audit review under contracts with the NRC.
The audit reviews will include evaluation of any alternate methods of analysis or exceptions to acceptance criteria identified in NilREG-0661.
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i The staff will report its conclusions on the acceptability of the Fermi 2 con-tainment and associated piping systems in a supplement to this report and verify that modifications have been completed at the times specified herein before the operating license is issued.
Copies of the report will be sent to the ACRS.
(2) Control Room Staffing Augmentation The ACRS noted the NRC staff's requirements (SER Section 3.1(c)) that Fermi 2 control room staff be augmented with vendor personnel who have commercial BWR operating experience, and recommended that the staff assure these personnel will remain on site until applicant's staff has obtained necessary operating experience.
Since the ACRS meeting, staff has discussed this requirement further with the applicant.
The applicant expressed concern regarding the requirement in the SER that these personnel "... remain on shift until the plant has attained a 100 percent power level." The applicant said this requirement could result in an unncessarily long retention of these personnel should the applicant have to operate the plant for an extended period of time at a' power level less than 100 percent (for example at 97 percent) or to place the plant in cold shutdown j
for modifications for an extended period of time.
In the case of Diablo Canyon Unit 1, these potential concerns have been satis-factorily alleviated by a license condition for this requirement as follows:
r The Pacific Gas and Electric Company shall augment the plant staff to provide on each shift an individual experienced in comparable size pressurized water reactor operation.
These individuals shall have at least one year of experience in operation of large pressurized water reactors or shall have participated in the startup of at least three pressurized water reactors.
At least one such experienced individcal shall be on duty on each shift during the initial fuel loading and through the startup test program whenever the reactor is i
not in a cold shutdown condition for at least the first year of i
operation or until the plant has attained a nominal 100% power level, whichever occurs first.
The applicant agreed that this wording would satisfactorily alleviate its cor' cerns.
The staff will include a license condition similar to the Diablo Canyon Unit 1 condition in the Fermi 2 license to assure that vendor personnel experienced in BWR operation will remain on site until the applicant's staff has obtained the necessary operating experience.
(3) Simulator Training in the Use of ATWS Procedures The ACRS noted that a Fermi 2 training simulator will be used for training operators and other plant personnel, including managers and supervisors, and recommended that the staff assure that anticipated transient without scram (ATWS) procedures and the associated simulatcr training are well coordinated.
As described in Section 15.2.1 and Section 22.2, Item I.C.8, of this supplement, the staff has reviewed the Fermi 2 ATWS procedure during the review of selected emergency operating procedures.
This review included observing a simulator 18-2
exercise by Fermi 2 operating personnel in which the emergency operating proce-dures, including the ATWS procedure, were used to mitigate simulated transients and accidents.
The staff will assure that the ATWS procedure is included in Fermi 2 emergency operating procedures before an operating license is licensed.
The staff also will administer simulator examinations of transients and accidents, including.
a simulated ATWS,'as a part of the applicant's nuclear plant staff training program (see SER Section 22.2, Item I.A.3.1).
(4) Training in the Use of Station Blackout Procedures The ACRS recommended that the staff assure that procedures exist to address a station blackout (loss of all onsite and offsite alternating current) and that operating personnel are adequately trained in the use of these procedures.
The ACRS asked to be kept informed on this training.
By a letter to all licensees and applicants dated February 25, 1981, the staff requested that station blackout procedures and training be implemented and that an assessment of procedures and training be provided.
By a letter dated May 18, 1981, the applicant provided an assessment of existing and planned procedures and training for station blackout events.
The present operator training program includes procedures and training for loss of all onsite alternating current power.
The applicant will develop emergency operating procedures and include simulator exercises for the postulated loss of all alternating current power (onsite and offsite) in its requalification training program, which will include the use of the Fermi 2 onsite simulator.
These procedures and training and the installation of the Fermi 2 simulator will be completed during the first fuel cycle.
The staff will include a condition in the operating license to assure that the station blackout procedures are incorporated into the plant procedures and that operating personnel are adequately trained in their use before plant startup after the first refueling.
(5) Seismic Reassessment of Structures, Systems, and Components Required for Safe Shutdown The ACRS recommended that the applicant resolve to the satisfaction of the staff the outstanding review areas in the seismic reassessment of the structures, systems, and components required for safe shutdown based on currently accepted NRC design response spectra.
As described in Sections 3.7.1, 3.7.2, 3.7.3, 3.8.3.1, 3.8.3.3, 3.8.4, 3.9.3, and 3.10 of this supplement, the applicant has provided additional information regarding seismic reassessment, and staff has concluded that the Fermi 2 seismic design methods provide an acceptably conservative basis _for designing seismic Category I structures, systems, and components.
The completion dates for outstanding reas resulting from the seismic reassessment are Before May 1, 1982, the applicant will submit the seismic reassessment of NSSS piping (Section 3.7.3 of this supplement).
_Before May 1, 1982, the applicant will complene data in the summary sheets needed to demonstrate equipment seismic qualification for the site-specific 18-3
earthquake used in the seismic reassessment based on 5 percent structural damping (Section 3.10 of this supplement).
Before September 1, 1982, the applicant will install modifications resulting frem the seismic reassessment.
The staff will report =its evaluation of the analysis of the NSSS piping and of the equipment seismic qualification in a supplement to the SER, and will verify installation of equipment modifications before the operating license is issued.
(6) Use of Thermocouples To Monitor Inadequate Core Cooling The ACRS noted that the applicant has not yet' agreed to_the staff requirement that thermocouples be installed in the core to measure coolant temperature at the core exit under inadequate core cooling conditions during and following an accident.
The ACRS recommended that staff reevaluate its requirement to deter-mine if adequate information can be obtained from a few thermocouples in a more accessible location.
The staff nas revised its requirement for thermocouples to detect inadequate core cooling to allow applicant to propose an alternate location and/or number of thermocouples compared to the formerly proposed incore thermocouples-(see Section 22., Item II.F.2 of this supplement).
(7) Simulated Control Room Fire Test The ACRS noted that the staff had concluded there were deficiencies in the applicant's simulated control room fire test. This test was run to demonstrate that a fire external to the control panels would not result in loss of redundant shutdown functions.
The ACRS recommended that this outstanding item be resolved 4
in a manner satisfactory to the staff.
By letter dated July 31, 1981, the applicant provided a report documenting the fire test procedure, simulation of control panels, test data, conclusions, and a discussion of deficiencies in the test identified in the SER.
The applicant ~
concluded that the test demonstrated that safe-shutdown components within the control panel would maintain their state and circuit integrity during and after a fire external to the control panels.
The staff is reviewing the applicant's fire test, including the July 3,1981 test report. Tne staff will report the results of its review in a later SER suppl ement,.
(8) Outstanding Issues The ACRS recommended that other outstanding issues identified in the Safety Evaluation Report be resolved in a manner satisfactory to the staff.
This supplement describes the resolution of some of these outstanding issues; the remainder are listed in Section 1.8.
The staff will report the resolution of these remaining outstanding issues in supplements'to the SER before the operating license is issued.
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The ACRS Report concluded The Committee believes that if due consideration is given to the recom-mendations above, and subject to satisfactory completion of construction, staffing, and preoperation testing, there is reasonable assurance that the Enrico Fermi Atomic Power Plant Unit No. 2 can be operated at power levels up to 3292 MWt without undue risk to the health and safety of the public.
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22 TMI-2 REQUIREMENTS 22.2 TMI-2 Action Plan Requirements for Applicants for Operating Licenses
- I.
Operational Safety I.C Operating Procedures I.C.1 Guidance for the Evaluation and Development of Procedures for Transients and Accidents Discussion and Conclusions In the SER, the staff stated that it was reviewing the Fermi 2 emergency opera-ting procedures to assure that BWR Owners' Group Guidelines have been acceptably incorporated into Fermi 2 procedures.
As described in Item I.C.8 of Section 22.2 of this supplement (below), the staff has completed its review of these proce-dures. The review included observing the procedures being implemented on a simulator and a walk-through of the control room.
Based on this review, the staff has concluded that the BWR Owners' Group Guidelines have been adequately incorporated into the Fermi 2 procedures.
The staff review of the BWR Owners' Group Guidelines is continuing; however, the Fermi 2 procedures based on the January 31, 1981 generic guidelines are acceptable for interim full power operation of Fermi 2.
I.C.5 Procedures for Feedback of Operating Experiences to Plant Staff Discussion and Conclusion In the SER, the stafi concluded that the applicant had taken the right steps to meet the staff position on this item.
The staff found, however, that the proce-dure concerned itself only with licensee event reports (LERs); it did not include the review of other operating information from publications such as IE Bulletins, Circular, Notices, and pertinent NRC or industry assessments of operating exper-ience. Moreover, the procedure dio not indicate how operating experie,1ce is incorporated into training and retraining programs.
By letter dated July 29, 1981, the applicant submitted the following summary of its position on Action Plan Item I.C.5:
The Detroit Edison Company will review and assess both internal and external operating experience to assure that information pertinent to plant safety is continually supplied to operators and other personnel as appropriate.and is utilized to effect design and procedural changes to correct generic or specific deficiencies and to enhance plant safety when warranted.
- See NUREG-0737, 22-1
l The review of externally generated operating experience will be carried l
out primarily by individuals within the Nuclear Saf;ty and Plant Engi-neering Group (NSPE).
This experience will include but not be limited to General Electric NSSS reports, NSAC reports, LERs forwarded from the INPO program "Significant Event Evaluation and Information Network" (SEE-IN), and NRC Bulletins, Circulars and Notices.
Information which is considered to be of a nature such that urgent disposition is neces-sary will be forwarded to the appropriate plant production staff imme-diately.
Operating experience which is considered to warrant further evaluation is assigned to individuals within the NSPE group or plant staff as appropriate for evaluation and recommendations.
The conclu-sions and recommendations are returned to the NSPE group for approval and then assigned for implementation to the Plant Modification Group for design changes and/or to the appropriate plant staff and training organizations for procedural modifications.
Procedural changes will be reviewed and approved by OSRO.
Operators and STAS are notified of implemented changes as well as the Nuclear Training Department.
The review of internally generated operating experience (primarily LERs) begins with the Technical Engineer who prepares plant LERs.
The Technical Engineer submits the LER to the On-Site Review Organiza-tion (OSRO) which is made up of the section leaders of the Nuclear Production staff and the plant superintendent.
Shift supervisors and STAS will be notified of all plant LERs.
The OSR0 will evaluate the LER to determire whether procedural and/or design changes are called for or whether further analysis is necessary.
For straight-forward procedural matters, the OSR0 will recommend and approve the changes and assign them to plant staff for implementation including training proaram modifications.
For LERs which relate to plant design or require indepth analysis, OSR0 will transfer the information to the Nuclear Safety and Plant Engineering group along with their recom-mendation, if any.
This group will determine the appropriate response and assign the change for implementation to the Plant Modification group or plant staff as appropriate.
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The Nuclear Safety and Plant Engineering group will receive all plant
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LERs and internal operating experience as a matter of course and there-l fore may choose to take action on an LER even if OSR0 elected not to 1
refer it to them.
I Any change which constitutes an unreviewed safety question will be reviewed and approved by the Independent Review and Audit Group (IRAG) prior to implementation.
The staff has reviewed the above additional information and finds that it meets the requirements of Action Plan Item I.C.5; therefore, it is acceptable.
I.C.8 Pilot Monitoring of Selected Emergency Procedures for NT0L Applicants Discussion and Conclusion In the SER, the staff stated that it was reviewing selected emergency operating procedures.
The following discussion and conclusion are based on the staff review since the SER was published.
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The staff and personnel from Battelle Pacific Northwest Laboratories reviewed the procedures forwarded by the applicant to the NRC to ensure that these pro-cedures were consistent with the plant design and the BWR Owners' Group Guide-lines, and iniorporated applicable human factors considerations.
The review resulted in two pages of general comments and numerous specific detailed com-ments on the procedures.
The general comments included human factors considera-tion on the transition among procedures and the concurrent use of procedures, clarification of the use of cautions, procedure identification, and comments related to consistency among the procedures.
The specific commente included concerns about the lack of detail in the procedures, procedural compatibility with the control indicatioris and controls, resolution of ambiguous wording and logic statements, inclusion of action steps in cautions and noter usability of figures and tables, and the interface with nonaal and abnormal operating procedures.
The sta'f met with the applicant on June 17, 1981, to discuss the results of the review.
During the meeting many of the comments were resolved by considera-tion of the control room design, plant characteristics, administrative controls, and operator training.
As a result of these discussions and an independent review of the procedures by the applicant, changes were made to the procedures.
On July 25 and 26, 1981, the revised procedures were amployed to direct operator responses to simulated transients and accidents on the Brown's Ferry-TVA simulator. The simulated transients and accidents were designed to test all the procedures, with special emphasis on events that required transitions among precedures, difficult decision points, and operator interactions.
A team of NRC and contractor personnel observed the simulator exercises and discussed the effectiveness of the procedures with the operations personnel after each simulation.
The transients and accidents observed included a wide range of simulations, from minor transients to major accidents with multiple system failures.
Tne simulated transients and accidents involving multiple system failures included:
(1) turbine trips resulting in scrams, (2) loss of feed-water from various system failures with and without other injection systems available, (3) loss-of-coolant accidents (LOCAs) due to steamline, feed line, retirculating line, and injection line breaks over a wide range of break sizes, and (4) recovery from inadequate core cooling with various cooling mechanisms unavailable.
Other simulated transients and accidents included:
(1) loss of offsite power and failure of part of the diesel generators with subsequent loss I.
of all diesel generators, (2) failure of individual components of the emergency core cooling system in conjunction with a small LOCA, loss of level indication, and loss of temperature indication.
Between simulations, the review team and operating crew critiqued the activities and discussed use of the procedures, the technical content of the procedures, the sequences followed in the procedures, and the methods used to follow two or more procedures concurrently.
Several changes were made to the procedures based on these discussions, including sequencing of steps, clarifications for the operator, and modifications to reflect the priorities identified by the operators.
On July 28, 1981, the team of reviewers who had participated in the simulator exercises participated in a control room walk-through of the emergency opera-ting procedures.
The control room walk-through consisted of a small-break LOCA which further developed into a large-break LOCA and inadequate core cooling.
The simulated sequence of events began with a loss of coolant which 22-3
could be controlled us'ng the emergency core cooling system.
The level in the vessel was stabilized above the top of the active fuel.
This was followed by the failure of a number of systems and a concurrent increase in the leak rate.
The sequence of events concluded with a failure of the suppression pool torus with the resultant inadequate core cooling.
The entire sequence was discussad in detail with the control room operators at the conclusion of the simulated event.
Some changes were made to the procedures based on the control room wal k-through.
The effective manner in which the operators used the emergency operating procedures indicates that they are clear, properly sequenced, and compatible with the control room and its equipment.
During the review it was noted that some plant-specific data were not available; these were designated by "(Later)"..The applicant has committed to incorporate the data when they are available and before the issuance of an operating license.
The staff will verify that these data have beea included in the procedures.
The staff concludes, subject to inclusion of plant-specific data, that the applicant's emergency operating procedures are adequate for full power operation.
The requirements of Item I.C.8 have been acceptably met.
I.D.1 Control Room Design Review Discussion and Conclusions In the SER the staff identified five open items related to control room design.
By letter dated July 31, 1981, the applicant submitted additional information on three of the five open items:
(1) the acceptability of sound level measure-ments with the HVAC operational; (2) a description of procedures for making permanent modifications to control boards and panels; and (3) a description of the color code standard in use in the control room.
The applicant's reponse to these three open items and the Human Factors Engineering Branch (HFEB) position on each are described below.
(1) SoLnd Level Measurements The applicant has indicated that background noise with the HVAC operating was measured at 15 locations in the control room.
The average level was 54 dB(A),
and the highest level was 56.5 dB(A).
These levels are well below the guide-line value of 65 dB(A) and permit intelligible verbal communications at normal 7
voice levels.
Tests of the audible annunciators showed that the signal is such that the operators can reliably discern the signals above the control room background noise.
Thus, the staff finds the applicant's response acceptable.
) Procedures for Permanent Modifications The applicant submitted a copy of Administrative.'.ocedure 12.000.11, " Plant Design Changes," for implementing design changes after the plant is operational.
The applicant established this procedure to ensure that (a) The impact of the change is carefully considered, required actions are documented, and information concerning the change is transmitted to all affected persons and organizations.
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1 (b) Design changes are subjected to the control measure commensurate with those applied to the original design.
(c) Qualified individuals make a determination of the applicability of the requirements of 10 CFR 50.59 to all proposed changes.
(d) Design changes carried out pursuant to 10 CFR 50.59(a) do not involve an unreviewed Safety Question and/or a change in the Technical Specifications.
(e) Quality requirements set forth in the FSAR and the Edison Operational Quality Assurance Manual EF-2, applicable to design changes which are nuclear safety related, are properly implemented.
The staff finds that the use of this procedure for implementing design changes
.fter the plant is operational will ensure that the human factors practiced in developing the original control room panels will be included in all future modifications. Therefore, the staff finds the applicant's response to this item acceptable.
(3) Color Coding in the Control Room The applicant has provided information regarding the control room color coding used on both electrical and mechanical equipment.
Colors used are Red (flow of energy) - indicates continuity of energy flow through a pipe, wire, valve, switch, and so forth.
Green (nonflow of energy) - indicates discontinuity of energy flow through a pipe, wire, valve, or switch.
White (abnormal operating conditions) - indicates abnormal interruption of energy flow through a pipe, wire, valve, switch, and so forth, or operation of emergency standby equipment.
Amber (caution) - indicates abnormal flow of energy through a pipe, wire, valve, or switch.
Blue (limit) - indicates that a preset limit (not alarm condition) has been reached.
til illuminated pushbuttons have black mounting barriers.
All illuminated displays have white mounting barriers.
The staff finds that the applicant's color coding standard provides consistent meaning across all applications in the control room.
The color coding standard does not use the "greenboard" concept.
However, the information provided by the ut'lity and its commitment to further study the color coding scheme as part of the NUREG-0700 review are satisfactory.
The applicant's Julv 31, 1981 response to the three open items is acceptable to the staff.
Baseu on a review of the information submitted, tne staff concludes that the actions specified resolve these items.
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The applicant has not yet addressed the remaining two open items--the adequacy of the control room emergency evacuation alarm and the availability of the process computer software.
The staff will verify the installation and acceptability of these two items before an operating license is issued.
II.
Siting and Design II.B.4 Degraded Core Training Discussion and Conclusions In the SER, the staff stated that additional information was required to complete the review of this item.
The applicant has committed (in a letter dated September 4, IM1) to an l
acceptable degraded core training course.- The course covers 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> of training to coniJul or mitigate an accident in which tne core is severely damaged.
This course will be given to all shift technical advisors and opera-tions personnel, from the plant manager to and including licensed uperators.
This training program will be in general agreement with the " Guidelines for Training for Core Damage" from tha Institute of Nuclear Power Operations, Document Number STG-01, Revision 1, dated January 15, 1981.
The applicant has committed to complete the training of all operating personnel in the use of installed systems to monitor and control accidents in which the core may be severely damaged before fuel loading.
Managers and technicians in instrumentation and controls, health physics, and chemistry will be given training commensurate with their responsibilities during accidents which involve severe core damage.
Based on its review of applicant's submittal, the staff has concluded that the applicant has met the staff requirements for a program to train personnel-in the use of installed plant systems to control or mitigate an accident in which the core is severely damaged.
The NRC will verif3 that training has been completed before fuel loading.
II.F.2 Instrumentation for Detection Inadequate Core Cooling Discussion and Conclusion In the SER, the staff required the applicant to incorporate incore thermocouples into the design of the inadequate core cooling (ICC) monitoring instrumentation.
In its review, the ACRS recommended that the placement and required number of thermocouples be given further study (see Appendix B of this supplement).
The staff has revised its requirement, based on this-recommendation, as follows:
(1)
Incorporate thermocouples into the ICC monitoring system prior to June 1983 in accordance with Regulatory Guide 1.97.
(2) Provide the documentation required by Section II.F.2 of NUREG-0737, addressing the inclusion of thermocouples in the final.ICC monitoring system, on a schedule acceptable to the staff.
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(3) Based upon the BWR Owners' Group Appeal of the requirement for incore thermocouples during the ACRS review of Fermi 2 on August 6, 1981, the ACRS recommended that further study be given to placement of a small number of thermocouples in a more accessible location.
Therefore, the staff requires the applicant to perform a study to confirm that the LPRM assemblies are the most suitable location or propose an alternate location and/or number of thermocouples to detect inadequate core cooling.
The staff will condition the proposed operating license for Fermi 2 accordingly.
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23 CONCLUSIONS Based on the evaluation of the application es set forth in the SER issued on July 10, 1981 and on the evaluation set forth in this supplement, the staff concludes that the operating license can be issued to allow full power opera-tions (3292 megawatts thermal) of Fermi 2, subject to favorable resolution of outstanding matters described in this supplement.
The staff further concludes that Fermi 2 will operate in conformity with the provisions of the Atomic Energy Act and the rules and regulations of the Commission, and that there is reasonable assurance that the activities that would be authorized by the operating licenses for these plants can be conducted without endangering the health and safety of the public.
The staff reaffirms the conclusions as stated in the Safety Evaluation Report.
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APPENDIX A CHRON0 LOGY OF THE RADIOLOGICAL SAFETY REVIEW 0F THE DETROIT EDIS0N COMPANY APPLICATION FOR AN OPERATING LICENSE FOR THE ENRICO FERMI ATOMIC POWER PLANT UNIT 2 September 2, 1980 Letter fron applicant answering NRC Mechanical Engineering Branch Question 110.16.
February 26, 1981 Letter from applicant discussing the status of the Mark I Containment Long-Term Program and the time schedule for the Plant Unique Analysis.
April 8, 1981 Letter from applicant summarizing information on seismic qualifications of equipment.
April 10, 1981 Letter from applicant regarding the application schedule of safety-related electrical equipment.
April 15, 1981 Letter from applicant concerning development of site-specific spectra for seismic reassessment.
April 20, 1981 Letter from applicant concerning the use of Category I masonry walls in plants under construction.
May 1, 1981 Letter from applicant regarding Chemical Engineering Branch questions on reactor water purity.
May 4, 1981 Letter from applicant describing the Equipment Environ-mental Qualification program and schedule.
May 5, 1981 Letter to the applicant discussing safety concerns associated with pipe breaks in the BWR Scram system and requesting additional information.
May 8, 1981 Letter from applicant transmitting additional information on the design and use of concrete block walls in plants under construction.
May 12, 1981 Letter to applicant summarizing April 24, 1981, OL review meeting regarding fracture toughness of materials used in the containment pressure boundary, feedwater pipes, and main steam pipes.
May 12, 1981 Letter to applicant summarizing April 28, 1981, OL review meeting regarding reassessment of seismic structural margin.
A-1
May 14, 1981 Letter to applicant summarizing March 27, 1981, OL review meeting regarding revised seismic response spectra and reassessment of structural and equipment design margin.
May 14, 1981 Letter to applicant summarizing April 23, 1981, OL review meeting regarding operation and open review items.
May 14, 1981 Letter from applicant concerning the plan and execution of TMI Task Action Item I.G.1, Special Low Power Test Program for BWRs.
May 18, 1981 Letter to applicant summarizing May 13, 1981, OL review meeting regarding spent fuel pool cooling system.
May 18, 1981 Letter from applicant responding to NRC questions regarding TMI issues.
May 19, 1981 Letter from applicant providing information on Appendix G and H of Pressure Vessel Weld material data.
May 19, 1981 Letter from applicant concerning questions on the High Density Spent Fuel Sto Ige System.
May 26, 1981 Letter from applicant responding to several NRC requests for additional information on Unit 2, NRC Docket No. 50-341.
May 26, 1981 Letter from applicant concerning the requirements of Appendix 2, III.A.2, NUREG-0654.
May 26, 1981 Letter from applicant transmitting information in support of GDC 51 compliance.
May 29, 1981 Letter from applicant demonstrating compliance with 10 CFR 50, Appendix I, in a cost-benefit analysis of Section 11.0 and design objectives of RM-50-2.
May 29, 1981 Letter from applicant authorizing NRC to act on certain emergency licensing issues in order to ensure short turnaround times.
June 2, 1981 Letter to applicant giving the results of the staff's preliminary control room design review, asking the applicant's plans to correct the deficiencies noted, and transmitting requests for additional information from the Mechanical Engineering, Reactor Systems, and Power Systems Branches (MEB, RSB, PSB).
June 3, 1981 Letter to applicant summarizing May 19, 1981, OL review meeting regarding exemptions to tne requirements of Appendices G and H to 10 CFR Part 50.
A-2
June 3, 1981 Letter to applicant summarizing May 5-8, 1981, site visit by Instrumentation and Control Systems Branch (ICSB) staff for OL review.
June 4, 1981 Letter to applicant summarizing May 20, 1981, OL review meeting regarding Mark I containment structural analysis.
June 4, 1981 Letter to applicant summarizing May 27, 1981, OL review meeting regarding the Fermi 2, fire protection review.
June 8, 1981 Letter from applicant regarding documents submitted on treatment, residual heat removal, and cracking.
June 12, 1981 Letter from the applicant forwarding insert to FSAR Appendix A which reflects the installation of the loose parts monitoring system.
June 15, 1981 Letter from applicant transmitting response to Part 4 of Question 021.32.
June 16, 1981 Letter from applicant regarding the torus attached piping modifications program.
June 17, 1981 Letter to the applicant asking for additional information on the Fermi 2 Security and Personnel Training and Qualification Plan and the Safeguards Contingency Plan.
June 24, 1981 Letter to applicant asking for additional information for the Quality Assurance Branch (QAB) to complete its safety review.
June 26, 1981 Letter from applicant transmitting the Fermi-2 equipment environmental qualificatior, submittal.
June 29, 1981 Letter to applicant requesting additional information in 33 areas as a result of the staff safety review of the Fermi 2 Final Safety Analysis Report.
V June 29, 1981 Letter from applicant transmitting the response to Question 021.32.
June 30, 1981 Letter from enolicant transmitting revised response to the draft Plant Systems Branch (PSB) position on reactor containment electrical penetrations.
June 30, 1981 Letter from applicant committing to perform an inservice inspection program and stipulating details of the program.
July 1, 1981 Letter from the applicant transmitting a copy of the calculations for the shore barrier slope stability analysis.
A-3
July 1, 1981 Letter from the applicant transmitting the dynamic rigid wall lateral nressure computations relating to Geotechnical Branch Position 1, " Lateral Pressure Computations."
July 1, 1981 Letter from the applicant providing the results of a preliminary review of the BWR safety / relief valve test program to confire the adequacy of the Fermi 2 safety /
relief valves.
July 2, 1981 Letter fr a the applicant providing additional information on bypass leakage testing.
July 2, 1981 Letter from the applicant transmitting draft residual heat removal procedures.
July 2, 1981 Letter from the applicant providing the status of responses to information requested in an NRC letter of June 29, 1981.
July 2, 1981 Letter from applicant concerning the Emergency Notifica-tion System.
July 6, 1981 Letter to the applicant commenting on the applicant's radiological emergency response plan and requesting additional information on the plan.
July 6, 1981 Letter fram the applicant committing to the installation of modifications to the Fermi 2 containment long-term program.
July 6, 1981 Letter from counsel for the applicant forwarding copies of the 1980 Annual Reports for Detroit Edison Co.,
Wolverine Electric Cooperative, and Northern Michigan Electric Cooperative.
July 7, 1981 Letter to the applicant asking for additional information on the applicant's degraded core training program.
July 7, 1981 Letter to the applicant requesting information on equipment qualification for seismic and hydrodynamic loads for Fermi 2.
July 8, 1981 Letter from applicant listing Mark I Owners Group documents transmitted to NRC by GE and determined by the applicant to be applicable to the Fermi 2 docket.
July 10, 1981 Letter from applicant regarding safety concerns associated k
with pipe breaks in the BWR Scram system.
Jtly 13, 1981 Letter from applicant transmitting the shore barrier slope stability analysis, inccrporating three revisions to the original calculations.
A-4
July 13, 1981 Letter from applicant transmitting the equipment qualification summary for balance of plant and nuclear steam supply system items identified in an NRC letter of July 7,1981.
July 14, 1981 Letter from applicant transmitting revision 1 of the
" Supplementary Seismic Evaluation Report."
July 15, 1981 Letter from applicant transmitting information on ferritic materials used in pressure-retaining components in response to Question 121.17.
July 15, 1981 Letter from applicant transmitting information on Appendix G.
July 16, 1981 Letter from applicant submitting supplemeit.a1 materials in response to NRC request for additional information on post-turbine trip reheater steam flow.
July 17, 1981 Letter from applicant regarding the design shear wave velocity for buried pipe used at Fermi 2.
July 20, 1981 Letter from applicant transmitting amendment 2 to the Fermi 2 Physical Security Plan, revision 1 to the Safe-guards Contingency Plan, and revision 1 to the Security Personnel Training and Qualification Plan.
July 20, 1981 Letter to the applicant requesting confirmation of a prompt emergency notification system, a schedule for system installation, a description of the system, and a description or listing of the problems that may retard implementation.
July 21, 1981 Letter to the applicant summarizing the May 11-15, 1981, fire protection review at Fermi 2.
July 21, 1981 Letter to the applicant summarizing the May 18-21, 1981, audit of Detroit Edison corporate and Fermi 2 plant personnel who will manage the operation of the Fermi 2 plant.
July 21, 1981 Letter to the applicant summarizing the May 28, 1981, site visit to review and discuss the preservice inspection program.
July 21, 1981 Letter to the applicant summarizing the June 15, 1981, meeting regarding mechanical design aspects.
July 22, 1981 Letter to the applicant summarizing the June 3, 1981, meeting regarding applicant's plans for correcting deficiencies identified in the control room design audit.
A-5
July 22, 1981 Letter to the applicant summarizing June 30, 1981, meeting to discuss the review of the Fermi 2 emergency prepared-i.ess plans.
July 28, 1981 Letter to the applicant acknowledging receipt of the application for extension of the completion date of Construction Permit CPPR-87 and requesting a Class II fee of $1200.
i July 29, 1981 Letter from applicant submitting information on training for mitigating core damage, in response to NRC letter of July 7,1981.
July 29, 1981 Letter from applicant responding to Questions 121.16 through 121.27 regarding ferritic materials as well as l
the materials surveillance program.
J July 29, 1981 Letter from the applicant transmitting summary of appli-cant's proposed procedure to response to TMI Issue I.C.5 l
and indicating that the FSAR will be amended appropriately.
i July 30, 1981 Letter from applicant transmitting drawings to supplement information relative to the seismic qualification review audit.
July 31, 1981 Letter from applicant responding to three open items resulting from the Control Room Design Review Audit.
July 31, 1981 Letter from applicant's counsel transmitting FSAR Amendment 38.
July 31, 1981 Letter from applicant responding to concerns raised about containment leakage testing.
July 31, 1981 Letter from applicant transmitting the report on the control panel fire test.
July 31, 1981 Letter from applicant transmitting the updated fire hazards analysis (FSAR Appendix B), including the results of the response to Question 021.32.
July 31, 1981 Letter frei applicant transmitting testimony by Mr. Terence M. McKelvey at the ACRS Subcommittee Meeting, jay 24, 1981, Washington, DC.
July 31, 1981 Letter from applicant responding to three open items in the HFEB-1 Control Room Design Review / Audit.
August 3, 1981 Letter from applicant transmitting a report on the revised pump and valve inservice testing program.
August 4, 1981 Letter from applicant summarizing the reheater bypass flow calculations the applicant will perform, as agreed to at an informal meeting July 30, 1981.
A-6
August 13, 1981 Letter from the applicant concerning the SQRT July 27-31 site audit and submitting a list of the resulting action items assigned to the applicant in the exit interview, August 14, 1981 Letter from applicant submitting requested material on the conservatism of reverse testing of butterfly valves.
August 18, 1981 Letter from applicant transmitting revised emergency J
operating procedures.
August 21, 1981 Letter from applicant regarding methods to be used for prompt notification in the event of an emergency.
August 21, 1981 Letter from applicant regarding the antitrust review, and informing NRC that the applicant did not receive the results of a study on alternate energy sources for the City of Wyandotte.
August 24, 1981 Letter to applicant requesting information about the applicant's deep draft pump long-term operability program.
August 25, 1981 Letter from applicant responding to SER questions on fission gas pressure in fuel.
August 25, 1981 Letter from applicant committing to perform the "as-built" analysis of NSS piping using the envelope or multiple-support-excitation method of seismic response spectrum analysis.
August 28, 1981 Letter from the applicant submitting the $1200 fee for the Class II fee requested in the July 28, 1981 letter-to the applicant.
August 31, 1981 Letter from the applicant transmitting the results of the reassessment of high-density-fuel racks for site-specific earthquakes.
August 31, 1981 Letter from the applicant transmitting the response to six open items that resulted from the Seismic Qualifica-tion Rev ew Team audit.
i August 31, 1981 Letter from applicant's counsel transmitting FSAR-Amendment 39.
August 31, 1981 Letter from applicant submitting information regarding diesel engine lubrication in response to lice.nsing question 222.48.
August 31, 1981 Letter from the applicant presenting additional informa-tion regarding the systems review to meet the requirements of IE Bulletin 79-27, " Loss of Instrumentation and Control Power."
August 31, 1981 Letter from the applicant submitting the results of calculations on reheater bypass flow.
A-7
August 31, 1981 Letter from the applicant responding to the staff's July 6, 1981 comments on the emergency plan description of meteorological instrumentation and stating that a revised radiological emergency response plan will be available by December 15, 1981.
September 2, 1981 Letter to the applicant summarizing July 21-22, 1981, limited audit of the Fermi 2 seismic reassessment of Category I structures.
September 2, 1981 Letter to the applicant summarizing June 23-25, 1981 PSB site visit to review and resolve open items identified in the June 12, 1981, draft Safety Evaluation Report and to perform a site audit.
September 2, 1981 Letter from the applicant committing to complete the as-built seismic reanalysis of NSSS piping by September 1, 1982.
September 2, 1981 Letter from the applicant regarding limiting conditions for operation and surveillance of the suppression chamber-drywall vacuum breakers.
September 3, 1981 Letter to the applicant summarizing July 22, 1981, MEB audit of the seismic reassessment of Fermi 2 mechanical components.
September 3, 1981 Letter to the applicant summarizing August 17, 1981, meeting regarding the potential merger of two cooperatives which are co-owners of Fermi 2.
September 4,1981 Letter from the applicant stat',ng that the procedure for computing combined responses will be incorporated in the next F5aR amendment.
September 4, 1981 Letter from the applicant providing additional informa-tion in connection with open item 3a of the SQRT audit, stating that ground response spectra provided 5-and 7-percent structural damping for 5-and 7-percent equipment damping for the north-south and east-west directions.
September 4,1981 Letter from the applicant stating that the procedure for combining horizontal and vertical seismic components will be included in the next FSAR amendment, Section 3.7.2.1.2.4.
Sep' ember 4, 1981 Letter from the applicant submitting the present Fermi 2 operational radiological monitoring program plans.
September 4, 1981 Letter from the applicant providing the predicted drywell abnormal temperature and pressure response for the whole break spectrum, from the smallest steam break to the largest liquid leak.
A-8
September 4, 1981 Letter from the applicant providing additional informa-tion on training for mitigating core damage,
September 4, 1981 Letter from the applicant committing to complete operating and administrative procedures sufficiently before fuel loading to allow adequate familiarization with them.
September 9, 1981 Letter from the applicant comuitting to provide the seismic reassessment of NSSS piping by April 26, 1981.
September 10, 1981 Letter to the applicant summarizing September 2, 1981, meeting to discuss the preparation and review of the Technical Specifiations of the Fermi 2 operating license.
September 11, 1981 Letter frcm the applicant amending the supplementary seismic evaluation report providing data on the reassess-ment of buried pipes and duct runs.
September 14, 1981 Letter to the applicant notifying of staff acceptance of the Plant Contingency Plan submitted July 20, 1981.
September 14, 1981 Letter from the applicant transmitting copies of annual report of the preoperational environmental radiological monitoring program for 1978, 1979, and 1980, and the first quarter report for 1981.
J September 15, 1981 Letter from the applicant submitting 63 copies of the Fermi 2 Regulatory Requirements Review Report.
September 17, 1981 Letter from the applicant submitting six copies of
" Seismic Analysis of High Density Fuel Racks."
1 A-9
APPENDIX G BIBLIOGRAPHY Except as discussed below, docume.its referenced in or used to prepare this Safety Evaluation Report may be obtained through most major public libraries.
Correspondence between the Commission and the applicant (including the Final Safety Analysis Report, the Environmental Report and the application), as well as the Commission's rules and Regulatory Guides, vendor documents, and industry codes and standards, may be inspected at the NRC Public Document Room, 1717 H Street, NW., Washington, D.C.
Correspondence between the Commission and the applicant also may be inspected at the Monroe County Library, 3700 South Custer Road, Monroe, h:chigan.
NUREG Reports may be purchased from the NRC/GP0 Sales Program, Washington, D.C. 20555.
General Electric Company Documents Code RVFOR04 Procedure Y1096A006 Topical Report NEDO 21778-A Topical Report NEDO 24154 Topical Report NEDO 24342 Topical Report NED0 24708 Industry Codes and Standards American National Standards Institute Standard 18.7-1976/American Nuclear Society Standard 3.2 American Society for Testing and Materi.ls, Standard 02049 American Society for Testing and Materials, Standard E 185-73, " Surveillance Tests on Structural Materials in Nuclear Reactor Vessels," Annual Book of ASTM Standards Part 30 American Society of Mechanical Engineers, Boiler and Pressure Vessel Code 70 1968, 1970 editions including 1969, 1970 summer addenda Institute of Nuclear. Power Operations, " Guidelines for Training for Core Damage," Document SiG-01, Revision 1, January 1981.
NUREG Reports NUREG-75/087 Standard Review Plan (now published as NUREG-0800)
NUREG-0588 Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment, July 1981 NUREG-0661 Safety Evaluation Report on Mark I Containment Long-Term Drogram, July 1980.
i B-1
P NUREG-0737 Clarification of TMI Action Items, November 1980 NUREG-0798 Safety Evaluation Report related to the operation of Enrico Fermi Atomic Power Plant, Unit No. 2, July 1981 NUREG-0803 Generic Safety Evaluation Report Regarding the Integrity of BWR Scram System Piping, August 1981 Other Material i
Hall, W. J. and N. W. Newmark, " Seismic Design Criteria for Pipelines and Facilities," Journal of the Technical Councils of ASCE, November 1978.
Werner-Pfleiderer Corporation, " Topical Report WCP-VRS-1," November 1976.
U.S. Nuclear Regulatory Commission General Design Criteria (from Appendix A to 10 CFR Part 50) 2 Design bases for protection against natural phenomena 4
Environmental and missile design bases 10 Reactor design 15 Reactor coolant system # sign 26 Reactivity control systems capability 29 Protection against anticipated operational occurrences 31 Fracture prevention of reactor coolant pressure boundary 32 Inspection of reactor coolant pressure boundary U.S. Nuclear Regulatory Commission Regulatory Guides 1.33 Quality Assurance Program Requirements (Operations) 1.97 Instrumentation for Light Water Cooled Nuclear Power. Plants to Assess Plant and Environs Conditions During and Following an Accident 1.99 Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials B-2
APPENDIX C Report by the Advisory Committee on Reactor Safeguards on Fermi 2
/ a nouq'o UNITED STATES Ig NUCLEAR REGULATORY COMMISSION o
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E ADVISORY COMMITTEE ON REACTOR SAFEGUARDS o
g WASHINGTON, D, C. 20555
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August 11, 1981 The Honorable Nunzio J. Palladino Chairman U.S. Nuclear Regulatory Commission Washington, D.C. 20555
SUBJECT:
REPORT ON ENRICO FERMI ATOMIC POWER PLANT UNIT NO. 2 s
Dear Dr. Palladino:
During its 256th meeting, August 6-8, 1981, the ACRS completed its review of the application of the Detroit Edison Company (Applicant) for a license to operate the Enrico fermi Atomic Power Plant Unit No. 2 (Fermi-2). A Subcom-mittee meeting was held in Washington, DC, on July 24, 1981 to consider this project.
A tour of the facility was made on July 15, 1981.
During its re-view, the Committee had the benefit of discussions with representatives of the Applicant and the NRC Staff.
The Committee also had the benefit of the documents listed. The Committee reported on the construction pemit applica-tion for this unit in its report dated March 9,1971.
The Enrico Femi pl ant is located in Frenchtown Township, Monroe County, Michigan.
The nearest pop:11ation center is the city of Monroe, Michigan about 5.5 miles west-southwest of the site.
Fermi-2 is equipped with a General Electric BWR-4 nuclear steam supply system with a rated power level of 3292 MWt and has a Mark I pressure suppression containment with a design pressure of 62 psig. The Applicant has perfomed a detailed evaluation of the containment's ability to withstand LOCA and relief valve hydrodynamic loads as required by the NRC for the Mark I Containment Program.
As a result of this evaluation, extensive modifications were required and are underway. However, since the evaluation was perfomed prior to the issuance of the NRC report delineating the Staff's acceptance criteria (NUREG-0661 - Safety Evaluation Report, Mark I Containment Long-lum Pro-gram - Resolution of Generic Technical Activity A-7), the design has not yet been shown to be completely in conformance with this report.
The Applicant has made a commitment to perfom a plant unique analysis on the basis of the NUREG-0661 criteria and other requirements established by the Long-Tem Program, including in-plant confirmatory tests to assess loads resulting from safety relief cive operation. The Applicant will submit this analysis to the Staff for audit review upon its completion.
Subject to the results of this analysis, the NRC finds the Applicant's evaluation generally acceptable.
This matter should be resolved in a manner satisfactory to tne NRC Staff prior to full power operation. We wish to be kept infomed.
C-1
Honorable Nunizio J. Palladino August 11, 1981 We note that Detroit Edison has acted as its own architect-engineer for this proj ect.
The Applicant stated that this arrangement will result in a valu-able carry-over of knowledge as people transfer from construction to plant operation activities.
The NRC Staff has reviewed the Applicant's organi-zation and management structure and has expressed some concern about the personnel transition. The Staff recommends that care be taken to assure that quality of construction and safety of operations are not compromised during the transition. We concur in this recommendation. To address a concern over a lack of commercial nuclear power plant operating experience, the NRC Staff is requiring that the control room staff be augmented with veidor personnel during startup. We recommend that the NRC a'sure that these personnel remain s
on site for a period of time which permits the necessary operating experience to be obtained by the Applicant's Staff.
The Applicant described the program and the philosophy for training of personnel. Training has a high priority and a training simulator has been ordered to aid in this effort.
The simulator will be used.for operator training and will al so be used to train other plant personn'el including managers and supervisors.
It will also be used to test ATWS operating procedures.
The NRC has reviewed tne Applicant's ATWS procedures and finds them generally acceptable.
The NRC should assure that the ATWS procedures and the associated simulator training are well coordinated.
The Applicant discussed provisions to address station blackout.
In the event of a loss of all offsite AC power and loss of all onsite emergency diesel generators, the Applicant can call on a ' sel f-starti ng turbine-generator located onsite. While we recognize that this additional power source further lowers the probability of a station blackout, we recommend that the NRC Staff assure that procedures exist to address a station blackout event and that operating personnel are adequately trained in the use of these procedures.
We wish to be kept infonned.
Construction of this unit has taken a longer than usual time owing to - fi-nancial difficulties and the impact of the TMI-2 accident.
As a result, the l
Applicant has been required to perfonn a seismic reassessment of the struc-tures, systems, and components required for safe shutdown based on currently accepted NRC design response spectra.
This reassessment is still under way.
Preliminary results indicate that there is sufficient margin in the original design to meet the NRC requirements and that only minor equipment changes will be required.
This matter should be resolved to the satisfaction of the NRC Staff.
The NRC has begun review -of the Applicant's emergency pl anning.
Because of the plant's location, interaction with Canadian authorities is neces-sary.
Responsibility for this interaction rests with the offices of the Federal Emergency Management Agency.
\\
t N
C-2
l Honorable Nunizio J. Palladino August 11, 1981 The NRC Staff proposes to require the installation of core thermocouples in Fermi-2 as specified by Regulatory Guide 1.97, Revision 2, "Instrumenta-tion for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident."
The Applicant has not yet agreed to this requirement.
The ACRS supported use of core thermocouples in BWRs in its lette; of November 10, 1980 to the NRC Executive Director for Operations, but called attention to the need for further study to detemine the appropriate vertical location of such thermoccuples.
Since most of the information of interest from thermocouples may be obtainable from a small number of themocouples placed in a more accessible location, we recommend that this requirement be reevaluated.
The Applicant's security plan was discussed.
We note with approval that security guards will be Detroit Edison employees.
As part of the NRC Staff review of plant fire protection provisions, the Applicant simulated a control room fire to demonstrate that a fire external to the :ontrol panels will not result in a loss of redundant shutdown func-tions.
The NRC Staff has identified what it believes to be deficiencies in the test and the Applicant has responded in a recent submittal.
We believe this item should be resolved in a manner satisfactory to the NRC Staff.
Other issues have been identified as Outstanding Issues in the NRC Staff's Safety Evaluation Report dated July 1981. These include some TMI Action P1an requi rements.
We believe these issues can be resolved in a manner satisfac-tory to the NRC Staff and recommend that this be done.
The Committee believes that if due consideration is given to the recommenda-
^
tions above, and subject to satisfactory completion of construction, staff-ing, and preoperational testing, there is reasonable assurance that the Enrico Fermi Atomic Power Plant Unit No. 2 can be operated at power levels up to 3292 MWt without undue risk to the health and safety of the public.
Sincerely, J. Carson Mark Chairman
References:
1.
Detroit Edison Company, "Enrico Femi Atomic Power Plant Unit 2 Final Safety Analysis Report," Volumes 1 - 11 and Amendments 1-37.
2.
U.S. Nuclear Regulatory Commission, " Safety Evaluation Report Related to the Operation of Enrico Fermi Atomic Power Plant Unit No. 2," USNRC Report, NUREG-0798, dated July 1981.
3.
U.S. Nuclear Regulatory Commission, " Safety Evaluation Report, Mark I Containment Long-Term Program - Resolution of Generic Technical Activity A-7," USNRC Report, NUREG-0661, rated July 1980.
1 C-3
APPENDIX D ERRATA TO FERMI 2 SAFETY EVALUATION REPORT Page 1-9, Line 22:
After this line add "17.
Procedures for testing interlocks on RHR pressure isolation valves (7.4.2) 18.
Analysis of turbine trip including effect of reheater (15.1)"
Page 2-26, Line 6:
Change Section number from "2.5.4.5" to "2.5.4.3.7".
Page 3-16, Line 9:
Delete paragraph beginning "We have contracted...."
Page 5-10, Line 15:
Change " mode" to "model".
Page 5-11, Line 6:
Change "110 psig" to "1101 psig".
Page 5-24, Line 45:
Change " valve" to "line".
Page 6-23, Line 45:
Change "Section 5.4.7" to "Section 5.4.2".
Page 13-3, Line 35:
After this line, insert a new paragraph:
"We have reviewed the qualification, education, and experience requirements for the corparate technical staff specifically supporting the nuclear plant operations, as contained in a June 15, 1981 letter from Detroit Edison, as well as the r(sum (s of key personnel filling these positions, and teund these ecceptable.
In our review, we compared their qualifications against the guidance given in NUREG-0731, " Guidelines for Utility Management Structure and Technical Resources.
Page 13-6, Line 29:
Delete paragraph beginning "We have reviewed...."
Page 13-8, Line 24:
After this line add the paragraph that was deleted from Page 13-6.
Par 16-2, Line 9:
After this line add "18.
Leakage tests for pressure isolation valves (3.9.6.2) 19.
Periodic venting of ECCS and RCIC discharge lines (5.4.1,6.3.4) 20.
Limiting conditions for operation with one battery section inoperable (8.3.2)".
Pa;;e 22-35, Line 23:
Change " full power" to " exceeding 5 percent power".
Page 22-48, Line 48:
Change " Table II.E.4.2-1" to " Note a of FSAR Table H.II.E.4.2-1".
D-1 1
V Page 22-88, Line 25:
Delete the last two sentences and add:
"These systems have been found acceptable by the staff because the are redundant and safety grade (see Sections 5.4.2 and 6.3.4 of this SER). Therefore leakage from accumulators need not be considered for Fermi 2 and the accumulator design is acceptable."
Page E-5, Line 28:
Change "first" to " fire".
Page E-7, Line 31:
After this line insert text from Page E-10, Line 36 through Page E-11, Line 9.
Page E-10, Line 35:
AfteribslineinserttextfromPageE-7,Line32-through Page E-9, Line 38.
Page E-12, Line 19:
Delete entire Section X.
I i
i D-2
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[
l U.S. NUCLEAR REGULATORY COMMISSION
[.n I
BIBLIOGRAPHIC DATA SHEET No. 1 pp 4 TlTLE AND SUBTITLE (Add vosume Na, of epicanare)
- 2. (Line Nanaf Safety Evaluation Report related to the ooeration of Enrico Fermi Atomic Power Plant,
- a. RECIPIENT S ACCESSION NO.
Unit No. 2
- 7. AUTHOI SI
- 5. DATE REPORT COMPLETED
"$"e'n t em b e r I"
1981
- 9. PE RFORMING ORGANIZATION NAME AND MAILING ADDRESS (include 2,0 Codel DATE REPORT ISSUED U. S. Nuclear Regulatory Commission
= NTH, lvEAR Cffice of Nuclear Reactor Reaulation e P. ein b e r 1981 Washington, D. C. 20555 6
'L'* ' **a*>
- 8. ILeave Nank)
- 12. SPONSORING ORGAN 12 ATION NAME AND M AILING ADDRESS (incluctr 20 Codel p
Same as 9, above.
- 11. CONTRACT NO.
- 13. TYPE OF REPORT PE RIOD COV E RE D (/nclusive dalesJ Technical Report - Safety Evaluation July 1981 - September 1981
- 15. SUPPLEMENTARY NOTES l 14. (Leave danal Docket No. 50-341 l
16 ABSTR ACT 200 evords or less/
Supplement N'o.
I to the Safety Evaluation Report related to the operation of the Enrico Fermi Atomic Power Plant, Unit No. 2 provides the staff's evaluation of additional information provided by the applicant regarding outstandina review issues identified in the Safety Evaluation Report issued in July 1981.
It also covers revised designs and the staff's response to the comments in the report by the ACRS.
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17 KE Y WORDS ANO DOCUMENT ANALYSIS 17a. DESCRIPTORS I
i 17b IDEN TIFIE RS OPEN ENDE D TERMS 18 AV AILABILITY STATEMENT 19 SECURITY CLASS (Th,s report /
21 NO OF PAGES IINrtARR1FIFO Un1imited 20 sE CURITY CLASS (Th,s paget 22 PRICE llN C t_ A S S I F I E D s
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