ML20062G005
| ML20062G005 | |
| Person / Time | |
|---|---|
| Site: | Shoreham File:Long Island Lighting Company icon.png |
| Issue date: | 12/18/1978 |
| From: | Novarro J LONG ISLAND LIGHTING CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20062G009 | List: |
| References | |
| SNRC-347, NUDOCS 7812260126 | |
| Download: ML20062G005 (7) | |
Text
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LONG !SLAND LIGHTING COM PANY
? fit'O J a m,w SHOREHAM NUCLEAR POWER STATION
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,_ P.O. BOX Sie, NORTH COUNTRY RO AO + WADING RIVER, N.Y.11792 December 18, 1978 SNRC-347 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C.
20555 Shoreham Nuclear Power Station - Unit 1 Docket No. 50-322 1
Dear Mr. Denten:
Enclosed are fifteen (15) copies of Revision 3 to the Shoreham Plant Design Assessment Report (DAR) for SRV and LOCA loads.
This report is submitted in accordance with our Mark II Containment Closure Program as outlined in our letter (SNRC-298) dated June 15, 1978.
The DAR Revision 3 contains
- 1.he follcwing:
1.
Incorporation of information previously submitted in our letter (SNRC-309), dated July 28, 1978.
2.
Information relating to load definition, load combination methods, acceptance criteria and Shoreham plant capability to accept the hydrodynamic loads (including annulus
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pressurization effects) as required by the NRC load evaluation report, NUREG-0487, dated October 1978.
3.
The final revision to the DAR, scheduled for submittal in June 1979, will contain the confirmatory analyses
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and demonstrate the capability of all plant structures and equipment to accommodate the Mark II loads.
We have also included an Executive Summary of the Mark II Loads Closure Program for the Shoreham Nuclear Power Station.
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Mr. Harold R. Denton December 18, 1978 Re: Shoreham Plant Design Page
~c Assessment Report (DAR)
A proprietary supplement to Revision 3 to the Design Assessment Report is being sent under separate cover.
Very truly yours, J. P. Novarro, Project Manager Shoreham Nuclear Power Station I
BRM/cl FORUCLOSL7Ir To tomam RMUEST BGn? IIC FILE;ttATIRIAL
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EXECUTIVE SUMPlaY l
MARK II LOADS CLOSURE PROGRAM SHOREHAM NUCLEAR POWER STATION i
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f The Shoreham Mark Il loads Closure Program has utilized the design assessment report (DAR-Revision 3,
November 1978) to provide the NRC Staff with the documented information necessary to continue and complete the licensing of SNPS-1.
The pertinent information relating to load definition, load combination i
methods, acceptance criteria, and SNPS-1 glant capability to i
sustain the effects of the hydrodynamic loads (including annulus l
pressurization effects),
addressed by the NRC in the load evaluation
- report, NUREG-0487 (October 1978), has been compiled within this document.
The following " Executive Summary" lists the major component parts to this " Closure" report.
h I.
General l
l A design assessment has been performed on the containment and major structures subjected to loading conditions which include the effects of SRV discharge and LOCA blowdown loads.
Design assessments of the seismic Category I balance-of-plant (BOP) and NSSS piping, components and equipment, have been performed on representative essential systems.
The design assessment of the containment, major structures, I
and BOP /NSSS piping, components, and equipment, has shown i
that SNPS-1 has sufficient structural integrity to withstand the additional effects of the hydrodynamic loads (including annulus pressurization).
Plant modifications that have been I
made or will be required (see Paragraph V.B) to meet the sj acceptable stress / load levels are delineated in the report.
A final DAR (June 1979) will document the confirmatory analyses, which will demonstrate the suitability of plant structures and equipment to accept these loads.
That report will present the design assessments of plant structures and components not sufficiently avaluated for inclusion in this report.
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i' II.
Load Combinations and Acceptance Criteria The load combinations for the SNPS-1 containment and i
internal concrete structures are the same as in the original issue of the DAR/DFFR and are consistent with the NRC
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positions.
l.'jl The design bases of load combinations and acceptance I
criteria for the SNPS-1 BOP and NSSS piping systems and 4
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components are pre;ented.
The load combination methods are described in Appendix D (SRSS application criteria as applied to MK II) and is consistent with Attachment II of the NRC acceptance criteria.
The applicable load cases and 4
the dynamic analysis procedures used for SNPS-1 are consistent with Attachments III and V of the NRC acceptance s'
criteria as follows: (1) Use 115 percent peak broadening 'of Amplified Response Spectra puus), (2) Use Regulatory Guide i
1.92 to combine modal responses, (3) Use OBE damping for normal / upset
, conditions and SSE damping for j
emergency / faulted conditions, (4)
Annulus pressurization 1
effects are combined with SSE, (5)-
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SRV "
loading j
condition is assessed, and (6) criteria to assure functional i
capability for all essential components in Appendix E are in
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conformance with NUREG/CR-0261 and consistent with j
Attachment V-B of the NRC acceptance criteria.
4 III. Load Definition 4
A.
SRV Loads i
A quencher device has been selected to obviate
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temperature instability problems associated with the
- ramshead, and will be. installed in the Shoreham Plant.
The device selected is the T-Quencher designed for Mark 4
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II plants by KWU.
This particular quencher device was 4
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- selected, based upon the f ull-scale test results 4:
performed by KNU at Karlstein, which provides a
I confirmatory data base of improved load definition for j-a MK II plant such as Shoreham.
i To expedite the NRC review, and to comply with the NRC j
acceptance criteria report, the containment wall load j
definition for
- ramshead, based on the methodology i
prov' t -J in DFTR-Rev. 2, is used for the plant design j
asse:- v.- nt presented herein for containment, and 4
BOP /Fl.i;S piping, components and equipment.
i Since the containment wall load detinition for the i
quencher device is available, it is presented in the DAR for information.
This quencher load definition is based on the quencher data most suitable for the actual j
device to be Anstalled in the plant, and is expected to a
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i provide realistic and bounding data for the anticip'ated plant operating conditions.
The report presents comparisons of ARS between the ramshead and guencher discharge
- devices, and the resulting effect on stress / load for BOP
- piping, components, and equipment.
The comparison demonstrates qualitatively and quantitatively that the Shoreham ramshead load definition is bounding with respect to the quencher load definition.
B.
LOCA Loads
_The LOCA load definitions, including the methodology for vent-clearing jet loads, bubble formation, and bulk pool swell velocity and acceleration calculations, determination of swell
- height, and maximum drywell floor uplift load, impact and drag loads, air bubble, condensation oscillation and chugging boundary
- loads, and lateral loads on downcomers are presented.
The load definition and methodology are based upon the DFFR l
Rev.
2 and 4T test data, with provisions as noted, in response to the NRC Staff requests for additional claritication of information.
Plant unique responses to NhC MK II generic questions are incorporated where necessary or a reference is provided to the appropriate generic response.
IV.
Dynamic Responses to SRV and LOCA Loads The dynamic responses of the reactor building and its contents resulting from SRV and LOCA related loads are presented.
These dynamic-responses are included in the appropriate load combinations to perform design assessments
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of the containment, and LOP /NSSS piping, components and equipment.
In addition to containment forces and
- moments, amplified response spectra (NRS) are presented at representative building elevations for the various SkV and n
l LOCA events.
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ARS data are presented for both the design assessment basis ramshead and the T-Quencher "to-be-installed".
One purpose of including the T-Quencher load definition is to demonstrate that the SNPS-1 ramshead load will bound the responses expected from the T-Quencher, and to show l
quantitatively the design conservatism in using the ramshead for design assessment basis.
The NRC acceptance criteria allow for an increased data base for quencher loads and it is also intended that the inclusion of T-Quencher responses is in that direction.
Since the NRC Staff review incorporates the T-Quencher load definition, it is intended
_ that the information presented herein will provide a
quantitative input to that program.
V.
Plant Design Assessment A plant design assessment has been performed for the i
containment structure, and representative essential EOP/ESSS
- piping, components, and equipment.
Although the assessment is based upon the ramshead as the design
- basis, sufficient
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information is available for the T-Quencher for NRC review and comment.
A summary of design assessment follows.
A.
Containment Structure An assessment of the structural capacity of the foundation
- basemat, reactor support pedestal, and primary containment, to accommodate the load combinations, including SRV discharge and LOCA transient
- loads, indicates an adequate design margin.
The SRV load case, used for this assessment, although highly conservative and unrealistic, is based upon a simultaneous and in-phase entry of air bubbles into the pool.
Zurthermore, as described in detail in Appendix B, a positive design margin is maintained even if the hydrodynamic SRV and LOCA loads were arbitrarily increased by more than 100 percent.
B.
BOP Pipino and Equipment The current piping and component re-evaluation results indicate that a significant E rtion of piping design will be modified in order to accommodate the SkV and LOCA loads.
The modifications ~ include pipe routing
- changes, addition of new snubbers and restraints, j
upgrading of existing snubbers and restraints, and modifications to support steel structural members and anchorage.
Although the design assessment is based upon the ramshead discharge devices, the T-Quencher device loads have also been evaluated for representative piping and components.
The result of the evaluation demonstrates that the T-Quencher device has better performance than the ramshead.
It is clear that the T-Quencher device does reduce the pipe stress and pipe support loads.
The load reductions are even more prominent if the loads from ramshead and from T-Quencher are compared alone.
The assessment of equipment has indicated that floor-mounted duct systems, _ conduit and instrument lines, and cable tray systems require only selected minor change +
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l to a_cconmodate the SRV and LOCA loads.
_The largest potential impact occurs with pipe-mounted valve operators (MOV 's).
Based upon the design basis ramshead
- load, 30 percent of the operators are not presently qualified.
The evaluation also indicates a
reduction to 12 percent not qualified, based upon T-i Quencher loads.
C.
NSSS Pipino and Equipment As demonstrated in the
- report, the NSSS
- piping, components, and equipment will meet the loading combinations and acceptance criteria delineated.
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