ML20052G928
| ML20052G928 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 05/12/1982 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Fiedler P JERSEY CENTRAL POWER & LIGHT CO. |
| References | |
| TASK-03-06, TASK-03-11, TASK-3-11, TASK-3-6, TASK-RR LSO5-82-05-026, LSO5-82-5-26, NUDOCS 8205190162 | |
| Download: ML20052G928 (21) | |
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May 12, 1982 Docket No. 50-219 LS05 05-026 Mr. P. B. Ffedler Vice President and Director - Oyster Creek Oyster Creek Nuclear Generating Station Post Of fice Box 388 Forked River, New Jersey 08731
Dear Mr. Fiedler:
SUBJECT:
SEP SAFETY TOPICS III-6. SEISMIC DESIGN CONSIDERATION AND III-ll, COMPONENT INTEGRITY - OYSTER CREEK NUCLEAR GENERATING STATION We have completed our sels:nic review of the Oyster Creek Nuclear Generating Station.
Enclosed is a copy of our combined safety evaluation report of the two subject topics.
As discussed in 6nis report, the condensate storage tanf was found to
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require upgrading to meet SEP requirements for the postulated SSE. The safety related piping systems were found to be overstressed and therefore, all safety related piping should be reanalyzed and upgraded as required.
In addition, five equipment items (functionability of motor operated valves, and structural integrity of CRD hydraulic contrcl units, reactor vessel in-ternals, motor control centers and switchgear panels) still remain open due to lack of design information. A supplement to this report will be issued after the review of your responses for these five open items are completed.
This evaluation will be a basic input to the integrated safety assessment for your facility unless you identify changes needed to reflect the as-built conditions at your facility. With respect to the potential modifications outlined in the conclusion of this report, a determin& tion of the need to actual?y implement these changes will be made during the same integrated as-sessment. This topic assessment may be revised in the future if your facility design is changed or if NRC criteria relating to this topic are modified before the integrated assessment is completed.
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L Your response is requestad within 30 days of recolpt of this letter.
If no response is received within that time, we will assume that you have no comments or corrections.
Sincerely, s.
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i Dennis M. Crutchfield. Chief Operating Reactors Branch No. 5 Division of Licensing
Enclosure:
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Docket No. 50-219 LS05-82 Hr. P. B. Fiedler Vice President and Director - Oyster Creek Oyster Creek Nuclear Generating Station Post Office Box 388 Forked River, New Jersey 08731
Dear Mr. Fiedler:
SUBJECT:
SEP SAFETY TOPICS III-6. SEISMIC DEStGN CONS DERATION NID I!!-ll, COMPONENT INTEGRITY - OYSTER CREEK NUCLEAR GENERATING STATION We have completed our seismic review of Oyster Creek Nuclear Generating Station. Enclosed is a copy of our combined safety evaluation report of the two subject topics.
As discussed in this report, the condensate storage tank 1; found to be upgraded to meet SEP requirements for the postulated E9'.
In addi-tion, six equipnent items (piping, systems, functionability of motor operated valves, and structural integrity of CRD bydraulic control units, reactor vessel internals, motor control centers and switchgear panels) still renain open due to lack of design information.
A supplement to this report will be issued efter the review of your responses for these six open items are completed.
This evaluation will be a basic input to the integrated safety assess-ment for your facility unless you identify changes needed to reflect the as-built conditions at your facility. With respect to the potential modifications outlined in the conclusion of this report, a determination of the need to actually implement these changes will be made during the same integrated assessment. This topic assessment may be revised in the future if your facility design is changed or if NRC criteria relating to this topic are modified before the integrated assessment is completed.
Your response is requested within 30 days of receipt of this letter. If no response is received within that time, we will ass me that you have no conenents or corrections.
Sincerely, l
AD:SA:DL Dennis M. Crutchfield, Chief bai nneratino peactog Branch No. 5
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Mr. P. B. Fiedler cc G. F. Trowbridge, Esquire Resident Inspector Shaw, Pittman, Potts and Trowbridge c/o U. S. NRC 1800 M Street, N. W.
Post Office Box 445 Washington, D. C.
20036 Forked River, New Jersey 08731 J. B. Lieberman, Esquire Commissioner Berlack, Israels & Lieberman New Jersey Department of Energy 26 Broadwsy 101 Commerce Street New York, New York 10004 Newark, New Jersey 07102 Ronald C. Haynes, Regional Administrator Nuclear Regulatory Commission, Region I 631 Park Avenue King of Prussia, Pennsyl.vania 19406 J.. Knubel BWR Licensing Manager GPU Nuclear 100 Interplace Parkway Parsippany, New Jersey 07054 Deputy Attorney General State of New Jersey Department of Law and Public Safety 36 West State Street - CN 112 Trenton, New Jersey 08625 Mayor Lacey Township 818 Lacey Road Farked 31ver, New Jersey 08731 U. S. Environmental Protection Agency Region II Office ATTN:
Regional Radiation Representative 26 Federal Plaza
. New York, New York 10007 Licensing Supervisor Oyster Creek Nuclear Generating Station Post Office Box 388 Forked River, New Jersey 08731 9
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SEP SAFETY TOPIC EVALUATION OYSTER CREEK NUCLEAR POWER PLANT TOPICS:
III-6, SEISMIC DESIGN CONSIDERATION III-11, COMPONENT INTEGRITY I.
INTRODUCTION The nuclear pcwer plant facilities under review in the SEP received construc-tion permits between 1956 and 1967.
Seismic design procedures evolve'd The Standard Review Plan (SRP) significantly during and after this period.
first issued in 1975, along with the Regulations 10 CFR Part 50, Appendix A and 10 CFR Part 100. Appendix A constitute current licensing criteria for As a result, the original seismic design of the seismic design reviews.
SEP facilities vary in degree from the Uniform Building Code up through Recognizing this evolution, the staff and approaching current standards.
found that it is necessary to make a reassessment of the seismic safety of these plants.
Under SEP seismic reevaluation, these eleven plants ware categorized into two groups based upon the original seismic design and the availability of Different approaches were used to review seismic design documentation.
the plant facilities in each group. The approaches were:
Detailed NRC review of existing seismic design documents Group I:
with limited reevaluation of the existing facility to confirm judgments on the adequacy of original design with respect to current requirements.
Licensees were required to reanalyze their facilities and Group II:
to upgrade, if necessary, the seismic capacity of their facility. The staff will review the licensee's reanalysis.
methods, scope and results.
Limited independent NRC analysis will be performed to confirm the adequacy of the licensee's method and results.
Based upon the staff's assessment of the original design; the Oyster Creek plant was placed in Group I for review.
The Oyster Creek plant, a Mark I boiling water reactor (BWR), is located on the Atlantic coast, about 35 miles north of Atlantic City, New Jersey.
General Electric Company, the nuclear steam supply system (NSSS) supplier end prime contractor, engaged Burns and Roe, Incorporated for engineering The seismic analysis of plant assistance and construction management.
structures, systems and components was performed by John A. Blume and The plant received its construction permit on Associates. Engineers.
December 15, 1964, and provisional operating license on August 1,1969.
The Jersey Central Power and Light Company (JCPLCo), the owner, filed its application for a full-term operating license on March 6,1972.
. The Oyster Creek plant was originally designed for a design level earth-quake (equivalent to the OBE) with a peak ground acceleration (PGA) of 0.11g and for a safe shutdown earthquake (SSE) with a PGA of 0.229 Siousner ground response spectra scaled to the specified PGAs were used as seismic input for the analyses and design. The vertical component of ground motion was assumed to be two-third of the horizontal components throughout the frequency Far the dynamic a'1alyses of structures (reactor building and control range.
room / turbine building), the buildings were modelled as two-dimensional lumped mass-spring systems with only rotational soil springs attached the corresponding lumped masses to account for the soil-structure interac-tion effects.
Response spectrum analysis approach was applied to generate The turbine building was also member forces for the structural design.
analyzed by time history approach using 1940 El Centro earthquake record scaled to 0.119 (OBE level) as input.
No floor (or instructure) response Two spectrum was generated for the analyses of systems and components.
approaches (namely, response spectrum analysis approach and equivalent static analysis approach) were applied for the safety related piping systems using Housner ground response spectra as input. All mechanical and elec-trical components were analyzed by the static analysis approach.
Chapter 4 of NRC NUREG/CR-1981 report, " Seismic Review of the Oyster Creek Nuclear Power Plant as Part of the Systematic Evaluation Program," (Ref.1),
sumarizes the details of the original seismic analysis and design.
The SEP seismic review of Oyster Creek facilities addrei:: sed only'the Safe Shutdown Earthquake, since it represents the most severe event that must be considered in the plant design. The scope of the review included three the integrity of the reactor coolant pressure boundary; the major areas:
integrity of fluid and electrical distribution systems related to safe shutdown; and the integrity and functionability of mechanical and electrical equipment and engineered safety features systems (including containment).
A detailed review of the facilities was not conducted by the staff; rather our evaluations relied upon sampling representative structures, systems Confirmatory analyses using a conservative seismic input and components.
The were performed for the sampled structures, systems and components.
results of these analyses served as the principal input for our evaluC ion of the seismic capacity of the facility.
II.
REVIEW CRITERIA Since the SEP plants were not designed to curreni. codes, standards and NRC requirements, it was necessary to perfom "more realistic" or "best estimate" assessments of the seismic capacity of the facility and to consider the conservatisms associated with original analysis methods and design criteria.
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A set of revieri criteria an' guidelines was developed for the SEP plants.
These review criteria and 9.iidelines are described in the fol;owing documents:
NUREG/CR-0098, " Development of Criteria for Seismic Review of Selected l
1.
Nuclear Power Plants," by N. M. Newmark and W. J. Hall, May 1978.
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"SEP Guidelines for Soil-Structure Interaction Review," by SEP Senior Seismic Review Team, December 8,1980.
For the cases that are not covered by the criteria stated above, the follow-ing SRPs and Regulatory Guides were used for the review:
1.
Standard Review Plan, Sections 2.5, 3.7, 3.8, 3.9, and 3.10.
2.
Regulatory Guides 1. 26, 1. 29, 1. 60, 1. 61, 1. 92, 1.100, and 1.122.
III. RELATED TOPICS AND INTERFACES The related SEP topics to the review of seismic design considerations and component integrity are II-4, II-4.A. II-4.B. II-4.C.
These topics relate to specification of seismic hazard at the site, i.e., site specific ground response spectrum for the Oyster Creek site. The seismic input selected for the confirmatory analysis of Oyster Creek facility, namely the Regula-tory Guide 1.60 spectrum scaled to 0.229 peak ground acceleration, envelopes the Oyster Creek site specific ground response spectrum as shown in Fig.
1, therefore, the results for these four safety topic evaluations will not affect the review of seismic design considerations and component integrity.
IV.
EVALUATION A.
General Approach The seismic reevaluation of Oyster Creek Nuclear Power Plant was initiated by conducting a detailed review of the plant seismic documentation.
The results of this review are summarized in the draft report, " Seismic Review of Oyster Creek Nuclear Power Plant - Phase I Report." Then, thq staff and our consultants conducted a site-visit. The purpose of this site-visit were:
(1) to observe the as-built plant specific feature relative to the seismic design of the facility, (2) to obtain seismic design information which was not available to the staff in the docket, (3) to discuss, with the licensee, seismic design information that the staff and our consultants had reviewed, and (4) based on the results of this field inspection, experience and judgment, to identify sample structures, systems and components for which the confirmatory analyses (or audit analyses) would be performed. The results of these analyses, then served as the basis for safety assessment of the plant facility.
When a structure was evaluated, it was judged adequately designed if the results from the structural analysis met one of the following three criteria:
1.
The loads generated from confirmatory analysis were less than original loads; l
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The seismic stresses from confirmatory analysis were low compared to the yield stress of steel or the compressive strength of concrete; and 3.
The seismic stresses from cMfimatory analysis exceeded the steel yield stress or the concrete compressive strength, but estimated reserved capacity (or ductility) of the structure was such that inelastic deformation without failure would be expected.
If the above criteria were not satisfied, a more comprehensive reanalysis was required to demonstrate its design adequacy.
For piping reevaluation, the results from the audit analysis of each of the sampled piping systems were compared with ASME Code requirements for Class 2 piping systems at appropriate service conditions. This comparison provided the basis for reevaluating the structural adequacy of piping systems.
Because limited documentation exists regarding the original specifica-tions applicable to procurement of equipment, as well as for the qualifica-tion of the equipment, the seismic review of equipment was based on expert experience and judgment. Two levels of qualification were performed, structural integrity and functionability. The results of this reevaluation of equipment served as the basis for modifications or reanalysis to be undertaken by the licensee.
B.
Confirmatory Analysis In order to provide independent analytical results for the reevaluation, a relatively complete seismic confimatory analysis, which started with a definition of seismic input ground motion and ended with responses of -
the safety related structures and selected systems and components, during the postulated earthquake event, was performed. The analysis procedures and results are briefly discussed on the following sections.
1.
Seismic Input When seismic review of Oyster Creek plant started in mid 1979, the site specific ground response spectra were not available.
In order to perform the review on a sampling basis that could be applied with confidence, a more conservative ground motion, namely Regulatory Guide 1.60 horizontal ground response spectrum (R. G.1.60 spectra) scaled to 0.229, the original design peak ground acceleration (PGA),
was used as the horizontal component of postulated ground motion for an:1ysis. The input motion in the vertical direction was taken as 2/3 of the value in horizontal direction across the entire frequency range.
. Recently, the site specific spectra development program was completed, and the spectrum developed for the Oyster Creek site was issued to the licensee on June 17,1981 (Ref. 2) for any future work that may be required. The basis for the development of site specific spectra was documented in NRC NUREG/CR-1582 Report, " Seismic Hazard Analysis,"
(Ref. 3). This site specific spectrum is appropriate for assessing the actual safety margins present for any structures, systems and components that have been identified as open items.
In Figure 1, a comparison is made for the ground response spectra that were used for the original plant design and for SEP seismic reevaluation (Reg.
Guide 1.60 spectrum and the site specific spectrum).
2.
Acceptance Criteria and Scope The specific SEP reevaluation criteria are documented in NUREG/CR '
0098 and SEP Guidelines for Soil-Structures Interaction Review.
These documents provide guidance for:
a) selection of the earthquake hazard; b) design seismic leadings; c) soil-structure interaction; d) damping and energy absorption; e) methods of dynamic analysis; f) review analysis and design procedures; and g) special topics such as under ground piping, tanks and valuts, equipment qualification, etc.
These criteria are felt to more accurately represent the actual stress level in structures, systems and components during a postulated earthquake event and consider, to certain extent, nonlinear behavior of the systems.
The SEP seismic reevaluation of Oyster Creek facility was a limited review centering on:
Assesi,ent of the general integrity of the reactor coolant pressure bo udsry.
Evaluation of the capability of essential structures, systems and components required to shutdown the reactor safely and to maintain it in a safe shutdown condition (including the capability for removal of residual heat) during and after a postulated seismic event.
V
. A total of two (2) structures, four (4) piping systems, eighteen (18) equipment components (mechanical and electrical) were fully evaluated and several others samples were evaluated on a limited basis in this work. They are:
Structures - Reactor Building and Control Room / Turbine Building.
Piping Systems - Main steam, feedwater, isolation condenser, and CRD return lines.
Equipment - 8 mechanical equipment items and 10 electrical equipment items.
Others - Ventilation stack and condensate storage tank.
Additional samples will be selected if any open items cannot be resolved by analysis.
3.
Analysis of Structures Analytical procedures and methods conforming with the current state of the art were used. These procedures considered the three-dimensional dynamic response of buildings, soil-structure inter- -
tion effects, a wide range of dynamic properties for the soil foundation, structural damping in accordance with calculated stress levels, equipment masses, and so forth.
(a) Analysis of Reactor Building The reactor building (reactor building structures, portion of office building extent. ion, drywell, shield wall, reactor vessel /
support pedestal, and foundation mat) wa modelled as four lumped mass-spring closely coupled systems supported by the foundation mat.
Because of the high degree of asymetry of this building, a fully three dimensional model was developed to present the structure.
In order to calculate the soil spring constants to take account of the soil-structure interaction effects, the structural foundation mat was considered as an embeded rigid plate on an elastic half space. The input ground motion, R. G.
1.60 spectrum scaled to 0.229, was defined at the free field ground surface. As required by SEP soil-structure interaction guidelines, this input ground motion was applied directly at foundation of structures without considering any reduction from the foundation embedment. The response spet rum analysis approach conformed with the SRP requirement in that a combina-tion of modal and directional responses, etc., was used to generate the structural responses. The final analysis results (dynamic moments, shears and axial forces) used for the evalua-tion of the structure were envelopes of the three sets responses 1
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. generated by considering three levels of soil shear moduli for the purpose of accounting uncertainty of soil properties.
The details of analysis and final results are sunt.arized in Chapter 5 of Oyster Creek NUREG report (Ref.1).
The time-history analysis approach together with an artificial time history record (acceleration) was used for generating in-structure (or floor) response spectra. Again, smoothed envelopes of the three sets of in-structure response spectra corresponding to three soil conditions were used as input motions for the evaluation of piping systems and equipment.
Appendix B to the Oyster Creek NUREG report contains a summary of all the generated in-structure response spectra.
The results of evaluation showed that reactor building is capable of withstanding the postulated seismic event.
(b) Analysis of Turbine Building The same acceptance criteria and analytical approaches used for the reactor building were applied to the turbine building. The details of modelling techniques, analysis procedures and analysis results (dynamic forces used for structural evaluation and in-structure re-sponse spectra used for equipment and piping evaluation) are found in Chapter 5 and Appendix B of the Oyster Creek NUREG report. The results of evaluation showed that the turbine building is capable of withstanding the postulated seismic event.
4.
Analysis of Piping Systems As discussed in the Section 2 above, four piping lines were selected and analyzed to verify the adequacy of the original design. The piping selected were portions of the main steam, feedwater, isolation condenser, and CRD return lines. The selections were based on:
(1) the expert's judgment and observations during the walkdown of the facility, (2) review of the original analyses and design, and (3) a desire to provide a range of piping sizes. Audit analysis which incorporated current ASME Code and Regulatory Guide Criteria and used the floor response spectra as input motion were performed for each portion of piping system selected.
The results from these analyses were compared to ASME Code requirements for Class 2 piping systems at the appropriate service conditions.
This comparison i
provided the bases for assessing the structural adequacy of the piping under the postulated seismic loading condition.
Assumptions I
made for the analysis, methodology employed and detailed preliminary results are found in the INEL report (Ref. 4).
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. The preliminary results of the confirmatory analysis showed that some locations of sampled piping systems were found to be overstressed under postulated S$E loading. Since the design adequacy of piping systems has not been demonstrated, all safety related piping should be analyzed and upgraded as required.
5.
Analysis of Condensate Storage Tank The integrity of the condensate storage tank was evaluated for:
(1) uplift and overturning, (2) anchor bolt forces, (3) buckling of side wall, (4) hoop stresses due to static and dynamic pressures, i
and (5) sloshing effect. An equivalent static analysis with 0.22g R. G.1.60 spectrum as input motion was performed for this tank.
The detailed evaluation was described in the Oyster Creek NUREG report. The results of this evaluation showed that the anchor bolts would need to be upgraded. This tank is the only safety related tank at this plant. No additional sample is needed for tank evaluation.
6.
Analysis of Ventilation Stacks The 368 ft ventilation stack was modelled as a two-dimensional lumped mass-spring system with soil springs attached to the founda-tion to account for the soil-structure interaction effects and was analyzed by time-history approach. The details of modelling techni-ques, analysis procedures and evaluation results were summarized in Chapter 5 of Oyster Creek NUREG report.
The results of this evalua-tion demonstrated that the stack was adequately designed.
7.
Analysis of Selected Mechanical and Electrical Eouioment The evaluation of equipment was done on sampling basis.
Safety related components required for safe shutdown, the primary pressure boundary, and engineered safeguard features were categorized as active or passive and as rigid or flexible according to the criteria in R. G. 1.45 and SRP 3.9.3.
A representative sample (or samples) from each group was selected and evaluated to determine the seismic design margin or adequacy of each group.
In this way, groups of similar components were evaluated without the need for detailed reevaluations of all individual components. The sampled mechanical and electrical equipment items and the basis for this sampling are described in Table 1 below:
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nochanical and electrical components selected by the review team for seismic evaluation, and the basis for selection.
Item No.
Description Reason for selection 1.
Emergency service water pump This item has a long, vertical unsup-ported intake section that was originally statically analyzed for seismic effects.
2.
Emergency isolation condenser This item is a horizontally mounted component supported by three saddles that do not appear to be seismically restrained. Concern was expressed about the saddles' ability to carry required seismic loads, particularly in the longitudinal direction.
3.
Containment spray heat his item is unique in that the heat exchanger exchanger is vertically oriented and supported by four brackets. Concern was expressed about the exchanger's ability to withstand overturning effects.
4.
Recirculation pump support This item is a vertical component sup-ported by hangers and critical to en-suring reactor coolant system integrity.
5.
Emergency diesel oil Anchor bolt system for in-structure storage tank flat-bottom tanks that are flexible may be overstressed if tank and fluid contents were assumed rigid in the original analysis.
6.
Motor operated valves A general concern with respect to motor operated valves, particularly for lines 4 in, or less in diameter, is that the relatively large eccentric mass of the motor will cause excessive stresses in piping attached to valves not externally supported.
7.
CRD hydraulic control system Item is particularly critical to including tubing and support insuring reactor coolant system i
system integrity.
8.
Reactor vessel supports and Same as Item 7.
internals
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(Continued.)
l Item No.
Description Reason for selection 9.
Battery racks ne bracing required to develop lateral load capacity may not be sufficient to carry the seismic load.
10.
Instrument racks The racks consist of channel and angle members that may be overstressed due to seismic loads. Anchorage to floor may not be adequate.
11.
Motor control centers Typical seismic qualified electrical equipment. Functional design ade-quacy may not have been demonstrated.
In addition, anchorage to floor structure may not be adequate.
12.
Transformers Same as Item 12.
13.
Switchgear panels Same as Item 12.
14.
Emergency generator Adequacy of anchorage was questionable. Functionality is important for safe shutdown.
15.
Control rocza electrical The control panels appear adequately panels anchored at the base. However, there appear to be many components canti-levered off the front panel, and the lack of front panel st,iffness may permit significant seismic response of the panel, resulting in high accel-eration of the attached components.
16.
Battery room distribution Same as Item 15.
psnels 17.
Isolation phase ductwork The ductwork support system does not supports appear to have positive lateral restraint and load carrying capacity.
18.
Electrical cabic raceways The cable tray support system does not appear to have positive lateral restraint and load carrying capacity.
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The licensee was asked to provide seismic qualifications data for each sampled component including design drawings, specifications, and design calculations. After a detailed evaluation of each com-ponent was completed, conclusions were drawn as to the overall seismic capacity of the safety related equipment at the Oyster Creek facility. The description of analytical procedures and eval-uations are found in Chapter 6 of the Oyster Creek NUREG report.
As discussed in the NUREG report, a total of 16 open items (structural) and/or functional integrity) out of 18 sampled equipment Were addressed as a result of the evaluation.
Some of these items remain open due to lack of design information. After the review and incorporation of additional infonnation submitted by the licensee (Refs. S thru 9 and Attachment I), the results are summarized below:
a) Five mechanical equipment items and 2 electrical equipment items were found to be adequately designed.
b) The structural integrity of motor operated valvec, has been de-monstrated.
However, functionability of the valves still remains open due to lack of design information.
c) The structural integrity of the following mechanical equipment items remains open due to lack of design information:
(i)
CRD hydraulic control unit (ii) Reactor vessel internals d) The review of safety related electrical equipment showed that the structural integrity of sampled transformers, control room electrical panels, and battery room distribution panels would be maintained under the postulated SSE loading condition.
However, the integrity of anchorage and support system of motor control centers and switch-gear panels have still not been demonstrated by the licensee due to
'ack of design information.
e) The functionability of all safety related electrical equipment as well as the structural integrity of internal components of all safety related electrical equipment is being evaluated through SEP Owner Group program. This program is scheduled for the completion by the end of 1982.
f) Qualification of electrical cable trays is being evaluated by testing through SEP Owners Group p ogram. This program is scheduled for completion by June of 1982.
- 1. CONCLUSION Based on ti e review of the original design analyses, the results of confirmatory l
analyses performed by the staff and its consultants, and the licensee's responses i
to the SEP seismic related safety issues, the following conclusions can be drawn:
Structure:- All safety related structures and structural elements of the Oyster Creek facility are adequately designed to resist the postulated seismic event (Ref.1).
y...
. Piping Systems - According to the preliminary results of SEP piping audit analyses (Ref. 4), some locations (including snubbers) of the sampled piping
- ystems were found to be overstressed under the postulated seismic loading.
In addition, the design adequacy of pipe supports has not been reviewed because the implementation of NRC IE Bulletin 79-14 was being continued and the design information regarding pipe supports were not available to the staff during SEP seismic review. The staff recomends that the safety re-lated systems at the Oyster Creek facility be reanalyzed to the SEP require-ments and upgraded as required.
Mechanical Equipment - A total of 8 mechanical equipment items were samhled.
From the 8 items, 5 have been detennined to be adequate. The remaining open itens (structural integrity and/or functionability) are due to lack of design information. This does not necessarily imply that safety deficiencies exist.
The reevaluation of design adequacy of the 3 remaining open items is being conducted by the licensee and the final results will be reviewed during integrated assessment.
Electrical Equipment - As discussed above, the structural integrity of 7 out of 10 sampled safety related electrical equipment items was found to be adequate under postulated SSE and the remaining 3 open items are due to lack of design information.
Since the anchorage and support system of all safety related electrical equipment was upgraded based on the same criteria, it is the staff's judgment that the structural integrity of motor control centers and switchgear panels will be demonstrated when the evaluation program conducted by the licensee is complete. As far as the functionability of electrical equipment and the design adequacy of cable trays, two (2) activities are being conducted by the licensee:
(a) a program has been initiated for the documenta-tion of seismic qualification (functionability of the equipment and structural integrity of internal components) of all safety related electrical equipment, namely the SEP Owners Group program, and (b) a program for seismic qualifica- -
tion of electrical cable trays based upon testing by the SEP Owners has been implemented. These latter two programs are intended to confim the adequacy of existing designs and equipment.
Recently, NRC has initiated a generic program to develop criteria for the seismic qualifications of equipment in operating plants; Unresolved Safety Issue (USI) A-46.
This program is scheduled for the completion in March 1983 Under this program, an explicit set of guidelines (or criteria) that could be used to judge the adequacy of the seismic qualifications (both functional capability and structural integrity) of safety related mechanical and electrical equipment at all operating plants will be developed. Considering that:
F
. (1) All safety related electrital equipment has been properly anchored; (2) Past experience and testinq results (from both nuclear and nonnuclear facilities) indicate in geteral that electrical equipment will continue to operate under dynamic loading conditions with only limited transient behavior, if the equipment is ad2quately anchored; and (3) The SEP Owners Group programs from which a set of general analytical methodologies is being developed for the seismic qualifications of cable trays and for documentation of other safety related electrical equipment (functionability);
it is our judgment that for the interim period until a technical resolution of USI A-46 is reached regarding methods for assessing seismic qualification of equipment in operating plants, the safety related electrical equipment at Oyster Creek plant will function during and after an earthquake up to and including the postulated SSE.
If additional requirements are imposed, as a result of USI A-46, regarding functional capability of safety related electrical equipment, the Oyster Creek facility will be required to address these new requirements along with other operating reactors.
Furthermore, since the ground response spectrum (0.2g R. G.1.60 spectrum) used for Oyster Creek seismic reevaluation envelopes the Oyster Creek site specific ground response spectrum, additional safety margins in the struc-tures, systems and components do exist for resisting seismic loadings. Thus, the staff concludes that the Oyster Creek plant can continue to operate with reasonable assurance that the operation of the facility will not be inimical to the health and safety.of the public until a resolution is reached for the items identified.
i
. REFERENCES 1.
NUREG/CR-1981 Report, " Seismic Review of the Oyster Creek Nuclear Generating Station as Part of the Systenatic Evaluation Program,"
November 1980.
2.
NRC Letter, " Site Specific Ground Response Spectra for SEP Plants Located in the Eastern United States," June 17, 1981.
3.
NUREG/CR-1582 Report, " Seismic Hazard Analysis," Volumes 2-4, October 1981.
4.
EGG-EA-5211 Report, " Summary of the Oyster Creek Unit 1 Piping Cal-culations Perfonned for the Systematic Evaluation Program," July 1980.
5.
GPU Service Corporation, " Seismic Analysis of Oyster Creek Emergency Service Water (ESW) Pumps," Revision 1 dated 12/8/81, transmitted by GPU Service Corporation letter dated Decenber 10, 1981.
6.
GPU Service Corporation letter fran J. T. Carroll to D. M. Crutchfield dated September 2,1981.
Subject:
Seismic Evaluation Program, Seismic Considerations - Evaluation of Containment Spray Heat Ex-changer Bolts.
7.
Oyster Creek Nuclear Generating Station letter dated September 2, 1981 from J. T. Carroll to D. M. Crutchfield.
8.
GPU Service Corporation, " Effects of Eccentric Mass of Value Operators on Piping Seisnic Stresses," dated November 30, 1981 transmitted by letter from Y. Nagai to T. Cheng, NRC dated December 10, 1981.
9.
GPU Service Corporation letter from Y. Nagai to T. Cheng dated November 24, 1981 transmitting October 19, 1981 meeting minutes and enclosures.
i
~
j TABLE 19. Conclusions regarding equipment review for seismic design f
adequacy of Oyster Creek.
Item Description Conclusion and recommendation 1.
Emergency service 0.K.
water pump 2.
Emergency isolation 0.K.
condenser 3.
Containment spray 0.K.
heat exchanger 4.
Recirculation pump 0.K.
support 5.
Emergency diesel oil 0.K.
storage tank 6.
Motor operated 0.K. for structural integrity. Functional valves adequacy of motor control valves has not been demonstrated.
7.
CRD hydraulic control 0.K. if detailed evaluation of stress in units limiting support elements are within ASME Service Condition D stress limits for supports when considered dead weight, axial-bending interaction effects and the effects of element curvature.
8.
Reactor vessel Design of core internals shroud appears internals adequate but detailed design calculations are not available to evaluate design adequacy.
9.
Reactor vessel and 0.K. for current site specific reduced supports spectrum if original analysis was adequate.
l e -~ -
TABLE 19. continued - Conclusions regarding equipment review for seismic design adequacy of Oyster Creek.
Item Description Conclusion and recommendation 10.
Battery racks 0.K.
11.
Instrument racks 0.K. for structural integrity. No information on function.
12.
Motor control Additional analysis demonstrating that MCC centers will not tip backwards is required. For MCC's identified as " vital" anchorage adequacy has been established.
13.
Transformers 0.K. for structural integrity.
Functionality has not been demonstrated.
14.
Switchgear panels Additional analysis is required. This analysis shall either demonstrate the rigidity of the component or should be based on a maximum acceleration of 1.09 In addition, shear load calculations must account for the simultaneous occurrence of two mutually orthogonal horizontal components.
15.
Emergency generator 0.K. for structural integrity.
Functionality has not been demonstrated.
16.
Control room 0.K. for structural integrity.
electrical panels Functionality has not been demonstrated.
It should be confirmed that main control room panel has been evaluated.
17.
Battery room 0.K. for structural integrity.
distribution panels Functionality has not been demonstrated.
18.
Isolation phase 0.K.
ductwork supports 19.
Electrical cable No evaluation has been made since no drawing raceways or design calcuations are currently ava lable.
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