ML19256D251
| ML19256D251 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 04/21/1971 |
| From: | Case E US ATOMIC ENERGY COMMISSION (AEC) |
| To: | US ATOMIC ENERGY COMMISSION (AEC) |
| References | |
| NUDOCS 7910170763 | |
| Download: ML19256D251 (7) | |
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4 APR 211971 i
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Peter A. Morris, Director Division of Reactor Licensing METROPOLITAN EDISON CEMFANT/JERSET CENTRAL POWER & LIGHT CQ!PANY -
THREE MILE ISLAND MUCLEAR STATION UNIT NO.1, DOCEE" UO. 50-289 I
l The FSAR information subunitted *oy the applicant has been reviewed and j
evaluated by the DRS Structural Engineering Branch. Enclosed is an j
valuation of the information submitted to date which includes Amend-ment No.18, received March 22, 1971. Tentative conclusions, for
.which confirmatica is still required, are enclosed in parentheses; the material,in brackets provides a summary of actions to be taken to resolve issues still open.
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OriginalSigned By E. G. Case Edson G. Case, Director j
Division of Reactor Standards i
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THREE MILE ISLAND UNIT NO.1 Docket No. 50-289 STRUCTURAL EVALUATION CLASS I STRUCTURES For reinforced concrete Class I structures, the principal methods of analysis have been the Working Stress and Ultimate Strength design methods as defined in ACI 318-63 "ACI Standard Building Code Requirements for Reinforced Concrete."
All Class I structures vital to protection of the reactor coolant pressure boundary, safe shutdown of the plant, and/or which contain radioactive materials, as listed in 5.1.3 of the FSAR, are designed for aircraf t impact and associated loadings as described in Appendix SA of the FSAR. The aircraf t for which these structures are designed to withstand the impact is identified as weighing 200,000 lbs, and impacting at a velocity of 200 knots with an effective impact area 19 feet in diameter.
In addition, two smaller aircraf t sections of 4,000 and 6,000 lbs were assumed to impact at 200 knots with respective effective impact areas of 3 ft. and 5 ft. in diameter The control building floors have been separated from the exterior walls exposed to an aircraf t impact by a 2" vide joint, so that i450 LOS
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- personnel and control equipment would be protected against the shock effects of an airplane impact. The concrete floor slabs are supported on steel beams which rest on vibration dampening elastomeric pads.
The analysis of the reactor building indicated the most critical areas for impact ~to be at the dome apex, dome to girder transition, girder to cylinder transition and impact at grade.
[The applicant's conten-tion that the containment and those other portions of the plant essential for safety have been designed with the postulated impact load as a limit of structural functional integrity is still under review by us and our consultants -- Mr. J. Proctor of the Naval Ordnance Laboratory and Dr. N. Newmark of N. M. Newmark Consulting Engineering Services.
It is our belief that the impact design criteria vill be met when the appropriate documentation is submitted by the applicant for our final reviev.]
(We have reviewed the design criteria for the Class I structures as defined and listed by the applicant in the FSAR, and found them to be adequate.)
FOUNDATION AND ENVIR0!NENTAL CONSIDERATIONS All Class I structures, except the diesel generator building, are founded on bedrock of Gettysburg shale. The diesel generator building is founded on compaated backfill. Foundation mats have been used for H 50 106
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. these structures. No unusual foundation conditions have been reported during excavation which would af fect the finding of acceptability made during the construction permit review.
Station Class I design wind loads are derived from the ASCE Paper 3269 and are based on a 100-year recurrence. Tornado loads on Class I structures are applied as a 300 mph tangential wind with a 1.3 gust factor, a 3 psi pressure drop, and impingement of a tornado generated missile similar to missilee postulated on other plants (e.g., Beaver Valley).
Seismic loads are based on horizontal ground accelerations of 0.06g and 0.12g with vertical accelerations equal to two-thirds the horizontal ground acceleration. Other environmental loads such as snow, ice, and floods have also been considered in the structural design. The foundations and environmental loading crite-ia are considered to be acceptable for this plant.
CONTAINMENT DESCRIPTION, DESIGN CRITERIA, AND LOADS The primary containment is a prestressed concrete reactor building with a cylindrical portion of 130 feet inside diameter and 3 foot thick walls, and a height of 157 feet from the top of the foundation slab (9 feet thick with a 2 foot thick concrete flo r over the liner) to the spring line. The shallow dome roof is 3 feet thick. The cylinder 1450 107
valls are post-tensioned horizontally and vertically by means of BBRV tendoas of 169-1/4-inch wires each. Each wire is " button-headed" at both ends of the tendon to anchorage hardware. This tendon system, which is the first application using such a large BBRV system, was reviewed and found acceptable during the construction permit review.
The dome is also prestressed using the same tendon system, while the foundation slab is conventionally reinforced concrete. A steel p' Ate lines the interior of the reactor building to ensure a high degree of leaktightness during operation and accident conditions.
The cylinder sad dome liner material is 3/8 inch thick steel, while the base liner material is 1/4 inch thick. The liner material conforms to ASDi-A-283-67 Grade C.
Around the contain=ent penetrations, ASTM-A-516-67, Grade 55 material has been used.
The reactor building has an internal free volume of two million cubic feet and $s designed to withstand accident pressurization loads of 55 psig internal pressure coincident with 281*F temperature.
It is also designed for an external pressure loading under normal operations of 2.5 psig as well as the loadings previously described for other Class I structures. [With the resolution of the aircraf t impact loading discussed earlier, dhe reactor building structural design loads and design criteria are considered to be acceptable.]
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. The design pressure differential across walls of enclosed compartments in the conteinment internal structure are: primary shield wall (reactor cavity), 200 psf; secondary shield wall (steam generator compartments), 15 psf. For the pipe break assumptions outlined in section 5.2.5 of the FSAR, the primary shield loading will be such that the wall reinforcement will be in the yield strength range.
[The stated design limits are considered adequate, provided the structure's ductility factor and deformation characteristics, which information is still to be submitted by the applicant, are shown to be acceptable.]
CONTAIhMENT DESIGN ANALYSIS Static load stresses, natural frequencies of vibration, and the corresponding mode shapes of the respective free vibrational modes produced in the shell and foundation were calculated by means of a Kalnins* computer program based on equations developed by Reissner for thin, elastic shells of revolution.
Stress concentration factors were applied to analyze the membrane stresses around the containment penetrations due to thermal loads and pipe reactions and moments.
This method of design analysis employed for the reactor building is considered applicable and acceptable for this type of structure.
A Kalnins--Lehigh University 1450 109
TESTING AND SURVEILIANCE Structural proof testing will be accomplished by pressurization at 115%
of 55 psig (or 63.3 psig). The tests and acceptance criteria as described in the FSAR, Appendix SE are acceptable.
(Tendon surveillance will be performed at intervals of 1 year, 5 years, and 20 years af ter the initial structural proof test.
Six vertical and nine hoop tendons are to be inspected for broken wires.
Two of the six vertical tendons will have lif t-off tests performed to verify the prestress levels. Five wires will also be removed from five of the fifteen inspection tendons and subjected to visual, physical, and chemical testing.)
(The statistical basis for this program is still under discussion with the applicant, and will be reviewed and evaluated upon receipt of a proposed submittal.]
Seismic instrumentation will consist of a strong motion recording accelerograph mounted in the basement of the fuel handling building.
This seismic instru=entation is considered to be acceptable for this plant.
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