ML20125B632
| ML20125B632 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 12/07/1979 |
| From: | Finfrock I JERSEY CENTRAL POWER & LIGHT CO. |
| To: | Harold Denton, Ziemann D Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20125B634 | List: |
| References | |
| TASK-03-02, TASK-03-03.A, TASK-03-07.B, TASK-03-07.D, TASK-3-2, TASK-3-3.A, TASK-3-7.B, TASK-3-7.D, TASK-RR NUDOCS 7912190631 | |
| Download: ML20125B632 (18) | |
Text
Jersey Central Power & Light Company k
7 (4" ()j Madison Avenue at Punch Bowl Road Morristown. New Jersey 07960 (201)455-8200 Deceinber 7, 1979 Director of Nuclear Reactor Regulation ATT: Mr. Dennis L. Ziemann Operating Reactors Branch No. 2 Division of Operating Reactors U. S. Nuclear Regulatory Commission Washington, D. C.
20555 Re: Oyster Creek Nuclear Generating Station Docket No. 50-219 SEP Structural Topics
Dear Mr. Ziemann:
Transmitted herewith are our responses to your letter of July 16, 1969 which requested additional information concerning the following SEP structural topics:
TOPIC III - 2 Wind and Tornado Loads TOPIC III - 3A Effects of High Water Level on Structures l
TOPIC III - 7B Design Codes, Design Criteria, Load Combination and Reactor Cavity Design Criteria TOPIC III - 7D Containment Structural Integrity Tests j
As noted in the Attachment, with respect to the requested infor-mation for TOPIC III - 7B, efforts are continuing to address the remaining aspects (Question 4) of your request. The remaining information will be forwarded to you as soon as it is available.
Very truly yours, r
8 Ivan R. Fin ock, r.
Vice President Attachments A through J
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h' sW 90000057 7 91219 0 b 3 / j Jersey Central Power & Light Company is a Member of the General Pubhc Utihties System
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i Docket No. 50-219 ATTACHMENT A OYSTER CREEK NUCLEAR GENERATING STATION SEP STRUCTURAL TOPICS P
90000058 December, 1979
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OYSTER CREEK NUCLEAR GENERATING STATION Additional Information Structural Topics III-2 Wind and Tornado Loads Question 1:
Indicate which standards or codes (including date of edition) were used in the design of each Category I st2ucture for wind loads.
Response
The design bases for structures are the ASA developed irandards for wind loadings. The buildings are designed to withstand at least 30 psf between elevations 30 to 49 fees above grade and 45 psf between elevations 100 to 499 feet.
Those systems necessary for safe shutdown of the reactor and the primary containment system will operate satis-factorily during or following higher winds than the design wind loading.
The reactor coolant system, primary containment, and most support systems are located in the massive reinforced concrete section of the reactor building below the refueling floor.
These systems are easily capable of withstanding extremely high winds.
The control room, occupation of which is required at all times, is a reinforced concrete structure.
The heavy diesel generator with its fuel supply tanks easily withstand high winds.
Power lines from the diesel generator to the reactor building and instrument and control lines from the control room to the reactor building are buried underground.
Question 2:
Provide the information on how the tornado loadings which consist of the translational and tangential wind, the depressurization, and the tornado missile forces were considered in the design of each Category I structure.
Response
Amendment 11 (Page I-1-1-) to the FDSAR for Oyster Creek Nuclear Power Plant provides answer to this question.
A copy of the amendment is attached to this letter (ATTACHMENT B).
III-3A Effects of High Water Level on Structures Question For each of the Category I structures state the water level that was considered in the design.
Response
Our response to Question 9C given in supplement No. 6 (Addendum No. 1) to the application for a full term licenses dated November 21, 1973 provides answer to the question concerning the effects of high water level on structures. The response is attached to this letter (ATTACHMENT C).
90000059
t In addition, flood level studies were performed'for the proposed Forked River Nuclear Station and the results-were reported in the Forked' River PSAR. The Forked River Station site is located next to the Oyster Creek Station site and they share a common intake canal. All of the Class I structures are flood protected to the grade eleva-tion of 23 ft. and 6 inch above mean sea level. The flood level studies reported in the Forked River Station PSAR are attached to this letter (ATTACHMENT D).
III-7B Design Codes, Design Criteria, Load Combination and Reactor Cavity Design Criteria Question 1:
With regard to the design of the steel containment, provide the design specifications and afpropriate design reports.
This information should include the information requested in items two through six below.
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Response
Primary Containment Design Report is provided in the Amendment 15 to FDSAR for the Oyster Creek Nuclear Power Plant. The report describes the design basis, design evaluation, fabrication methods, inspection, and testing of the primary containment for the Oyster Creek Nuclear Power Plant.
Question 2:
List the codes and standards (including edition date) used for design and construction of all Category 1 structures.
Response
ATTACHMENT E to this letter lists the codes and standards used for design and construction of Category I structures.
Question 3:
List all loads and load combinations considered in the design of each Category I structure, including any missile or pipe break effects.
Define the term " Operating load" listed in load tables 1-A-4 and 1-A-5 of Amendment 22.
Response
Loads and load combinationa are listed in Tables 1-A-1 through 1-A-5 of Amendment 22 to FDSAR for the Oyster Creek Nuclear Power Plant (tables are attached - ATTACHMENT F).
" Operating Load" includes the gravity loads from all equipment and piping supports, the restraint to thermal movement of the structure, and the weight of water over the reactor during refueling and in the storage pools.
Question 4:
Provide the pertinent material properties of the steel and concrete used in the design of all safety related structures (i.e. fy, fe, etc.).
Response : We are in the process of gathering information to answer this question. The information will be forwarded to you when they are available.
90000060
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Question 5:
Describe the method of combining stresses induced by seismic actions with the stresses resulting from non-seismic loads.
Response
- Tables V-3-2 and V-3-3 are provided in.the Oyster Creek Nuclear Plant FDSAR which show loading conditions for seismic and non-seismic stresses for both Primary Contain-ment and Reactor Building
' Copies of.the tables.are attached co this letter (ATTACHMENT G).
Question 6: Provide a summary of stresses or strains at critical locations in all Category I structures for each load combination considered in the design.
Response
- A summary of stresses or strains for Category 1 structures for Oyster Creek Nuclear Generating Station is not available though we maintain the original stress calculations for the plant structures using load combinations acceptable at the time of the plant construction. We also do not have stress for single loading in our records.
Due to a large volume of calculation available, summerizing them j
would require considerable time and effort.
We were told by an NRC staff that the purpose of this question is to combine the non-seismic stresses derived in our original analysis with the seismic stress obtained by the NRC's current analysis in order to evaluate the structures.
It is our understanding that such an evaluation is not proper and simple since the analyses are based on different criteria, loads combination, method of combina-tion, etc.
In order for us to respond properly to this question, we need your clarification as to how the requested information will be used in your evaluation.
III-7D Containment Structural Integrity Tests Question Provide any reports that describe the procedures and results of the primary containment structural integrity test.
Response
For information concerning the containment structure integrity test, please refer to the following enclosures:
ATTACHMENT H:
Expansion of the Drywell Containment Vessel ATTACHMENT I:
Loads on spherical shells. Chicago Bridge G Iron Company (1964)
ATTACHMENT J:
Initial overload tests and leakage rate determination of the pressure suppression vessels. Chicago Bridge 6 Iron Co. (1966).
90000061
Docket No. 50-219 ATTACHMENT B OYSTER CREEK NUCLEAR GENERATING STATION TORNADO LOADINGS December, 1979 90000062
Revised 12/1 OC
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- QUESTION '
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Provide the robability of tornado occurrence in the vicinity of the Oyster Creek Nuclear Power Plant site ikeluding an evaluation of the consequences of a tornado striking Class I struc-t tures and Class I equipment. Consider the effects of high wind velocities, rapid depressuriza-tion, and missiles on a parametric basis and provide an estimate of the maximum wind velocities and pressure drop that the Class I structures can withstand without failure. Provide an analysis of the protective capability for Class I buildings and equipment against various sized missiles that can be caused by tornadoes moving at various velocities.
ANSWER The tornado frequency for Oyster Creek is 2190 years. This means a probcbility of 1.81%
for a 40 year life.
The 100 year wind storm would be a storm with the greatest intensity that can statistically
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be expected to occur. The wind velocities and wind pressure would be:
Height Zone (ft) 0-50 50 to 150 Velocity (mph) 100 125 Building pressure (psi) 40.3 62.8 i
These building pressures are based on 1.1 gust factor and 1.3 shape factor for a rectang-
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ular building. In the analysis, normal allowable working stress levels were increased by one-third for loading combinations of dead loa'd, live Icid and wind load.
For the concrete system considar the wind acting normal to one wall; the forces distributed between the reactor biological shield wall (37%) and the two exterior walls parallel to the wind 1
(63 %).
I For the steel, horizontal wind loads are transraitted to the column and distributed to the concrete and roof system. The load on the roof is transmitted to vertical cross bracing at the faces of the building and analyzed in tension only.
]
The following table includes the design criteria and stresses the reactor building is subjected to during a 100 year wind storm. The results of this investigation indicate the reac-tor building will be able to safely withstand the 100 year wind storm.
To find the maximum wind force the primary structure system can withstand it is neces-sary to evaluate the most critically loaded structural elements. The results of the investigation indicate the horizontal deflection of the upper steel framework is the limiting factor. Using the maximum deflection of the structure as a basis, the wind forces can be increased to approx-l imately 280 mph.
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Foundation Deflection (in.)
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Conc.
Steel Total Shear Flexural Tension i
Evaluation Criteria 50 450 7350 1.8 35.2 3.0 Biological Shield Wall 6.7 18
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Exterior Walls 13.0 48 Diagonal Bracing 3000 Concrete at EL.119'-3' O.017 Steel at EL.168'-0" 0.252 Total Defl. at EL.168'-0" 0.27 Foundation loading D. L. + 0. 8 LL 8.75 Wind load (max) i 0.44 Total Foundation 9.19 3.84 j
The significance of this wind load should be noted. Because of this extraordinary loading of extremely short duration the design allowable stresses can be assumed to reach and possibly exceed (based on energy con' iderations) the material yield point. If permanent damage to the s
steel superstructure is disregarded the rest of the structure (including the concrete biological shield) can safely withstand the forces caused by a wind with a velocity in excess of 500 mph.
This is true even if the steel structure is assumed to still transmit forces to the concrete portion of the structure.
The tabulation below lists the various Class I structures with their respective maa' mum permissible wind velocity and depressurization values. The allowable stresses do not exceed 90% of yield for reinforcing steel and 85% of the ultimate concrete strength and include the com-bined effect of dead loads plus normal operating loads.
Class I Structures Wind / mph Pressure / psi Reactor Building Exterior Concrete Walls 300 2.0 Reactor Building Insulated Metal Siding 160 0.53 Reactor Building Roof Decking
- 280 0.96 Reactor Building Steel for Craneway Enclosure
- 190 0.68 Control Room - North Wall 160 0,53 Remainder 300
- 2. 0
- Intake Structure 300
- 2. 0 Ventilation Stack 180 2, O Battery Room (interior room)
Diesel Generator and Oil Tank Vaults 300
- 2. 0 Based on siding drag - without siding steelwork can withstand 300 mph J
- Based on maximum suction value = 0.40 p
- See answer to Questions IV-1 and IV-7.
90000064 1
Rsvised 12/19/67
- OC I-2-3 L
Generally Class I equipment is enclosed in the listed Class I structures and is therefore protected w. thin the limits shown. The outdoor Service Water Pumps and Startup Transformer are capabic of withstanding 200 mph winds ano a depressurization 2 psi.
The method of analysis to determine the protective capability of Class I buildings and equipment at;ainst various sized missiles and missile penetration at tornado velocities was based on the Modified Petry Formula (Navy Bureau cf Yards and Docks NP3726).
The missiles assumed were a wood utility pole, 35 feet long by 14 inches in diameter having a velocity of 200 mph and a 1-ton missile, such as a compact-type automobile traveling
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at 100 mph with a contact area of 25 sq. ft.
The results of the analysis indicate that no perforation of the : B-inch thick reactor building walls or the 12-inch thick control room walls will occur, although spalling of the inside concrete face would be expected.
The control room, battery room, emergency diesel generator building, emergency switch gear and related wiring has been designed for tornado protection and will be beorporated in the final design.
There is essentially no missile protection of the magnitude discussed above in the metal siding walls of the reactor building above the refueling floor and the equipment access opening.
There is also no missile protection for the Class I pumps at the intake structure.
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90000065 4
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Docket No. 50-219 ATTACHMENT C OYSTER CREEK NUCLEAR GENERATING STATION EFFECTS OF HIGH WATER LEVEL ON STRUCTURES 90000066 December, 1979
QUESTION
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Either show that all plant features outside of the barrier provided by the reactor-turbine building complex are not necessary for cooldown, or that all features outside the reactor-turbine building complex can withstand very severe wave action, or provide a detailed wave analysis that will demonstrate that during a postulated PMH wave action on these exterior facilities will net impair their functionality.
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RESPONSE
The plant features outside of the barrier provided by the reactor-turbine building complex are as follows:
Circulating water system pumps Emergency service water and service water pumps Standby diesel electrical power building The locations of these plant features are shown in Figure C-1.
The circu-lating water and service water pumps are mounted on a deck at -Elevation +6 feet MSL above the intake structure. An elevation section of the intake structure is shown in Figure C-2.
The standby diesels are located on the plant grade at Elevation +23 feet MSL alongside the discharge canal.
These features, being located in the lee of the reactor-turbine building complex are exposed only to wave action which can be generated over flooded h
terrain and within the intake / discharge canals themselves during the postulated PMH. A detailed analysis of PMH wave action-in the lee was undertaken beginning f
II with an evaluation of the open coant surge, routing of the open coast surge through the bay into the intake / discharge canals and estimation of the extent of
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i flooded terrain.
Simultaneously, the PMH wind field was transmitted inland along a projection of its overwater path and the component winds are determined along the appropriate fetches for wave generation.
Consideration of the topography of the intake / discharge canal area and the h
l postulated component winds show that vave action at the plant features under 90000067 9-3-1 d
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study must be generated on a northerly fetch which extends along the intake l
canal for about.1200 feet and then continues an additional effective distance j
of 4300 feet over flooded terrain. Wave action cannot enter the final 1200 feet j
of intake canal from any other direction as previous studies of wave action in I
the canal have shown.
(reference Docket No. 50-363, Supplement No. 2 for the Forked River Nuclear Power Station.)
)
The detailed study of wave, action revealed the following sequence of events which could be attributed to the postulated PME.
The PMH surge hydrograph and time history of wave action in the intake canal in front of the intake structure are shown in Figure C-3.
Time zero corresponds to the arrival of the peak surge at the open coastline. As the terrain to the north of the curving intake canal is roughly at Elevation +15 feet MSL, the waves in the intake canal can be in-fluenced by the overland fetch from time -0.3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> to +2.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> only, corres-ponding to surge levels in excess of +15 feet MSL.
The waves troving directly down the intake canal could break and dissipate energy by turbulence on the embankment at the end of the canal, or could be reflected to produce standing waves by interference with the incident waves, or could be transmitted over the embankment with attendant energy loss. The waves which are reflected, either by the canal embanknent or by the intake structure for the Forked River Nuclear Plant on the opposite side of the canal from the intake structure under study, could reach the circulating water and service water pumps.
Generally, wave reflection from embankment slopes and irregular structures results in a decrease in reflected wave height when compared to incident wave height.
In this study, the assumption was made that the reflected wave is reduced to approximately 75 percent of the incident wave height.
In addition, waves moving along the intakecanalcanalsodiffractintbtheintakestructure. The diffracted wave will be reduced in the area of the pumps to approximately 25 percent of the in-cident wave height.
The result of the superposition of the diffracted waves 90000068 9-3-2
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and the reflected waves in the vicinity of the pumps cohld. produce wave heights equal to the incident wave heights in the canal.
The refraction of the waves as they move over the deck of the intake structure towards the pumps has little or no effect on the significant wave, while the maximum wave will undergo a slight reduction in height. The vsve heights in the vicinity of the intake ~
structure are considered conservative estimates as no credit was taken for the presence of the numerous pipes and obstructions on.the intake structure deck in front of the pumps which would dissipate the waves due to friction and tur-bulence.
The critical wave action in the intake canal will occur at time +0.5 hoer which corresponds to a surge elevation of +20.5 feet MSL in the lee of the structure, a significant wave height of 2.0 feet with a period of 2.4 seconds, and a maximum wave height of 3.3 feet with a period of 3.9 seconds. The maximum
. wave height in t'ne vicinity of the pumps will be reduced to 3.0 feet due to the effects of refraction as mentioned previously.
The SWL of 20.5 feet was determined from the PMH surge hydrograph by eliminating 1,0 feet of water level risc which was attributed to wave setup.
The local wind setup and storm surge components remain undiminished.
The justi -
fication for eliminating the wave setup component in these calculations is that wave action and the resulting setup on the lee side of the plant are negligible when compared to the 1.0 feet postulated for the exposed side of the plant.
At time +0.5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> the embankment at the end of the canal would be overtopped by i
roughly 10 feet of water and most of the wave action in the canal can be expected to be transmitted over the embankment into the discharge canal.
The height of the transmitted significant wave is 1.6 feet and the transmitted maximum wave, i
2.6 feet.
The crest elevation of the transmitted maximum waves should not exceed Elevation +22.0 feet MSL and hence, they should not affect the diesel _ building i
F 90000069 9-3-3
located along the discharge canal at Elevation +23 feet MSL. As the wave crest should be normal to the canal side slopes at thia point, wave runup may be neglected.
Partofthewaveactionintheintakecanal)however,willbedirected towards the intake. structure.
Both the circulating water and service water pumps will be completely submerged at this time.
Assuming that each of the
- i larger circulating water pumps may be approximated by a vertical cylinder, 8 feet in diameter and 14 feet high, the maximum horizontal thrust force due to a maximum wave 3.0 feet high and 3.9 seconds in period is about 2400 pounds.
The corresponding maximum moment is 66,500 foot-pounds. These maximum values occur when the water surf ace is at the stillwater level (i. e., midway between wave crest and trough) and are directed alternately towardt and against the wave travel. Drag forces and transverse forces were negligible.
These reflected /
diffracted waves will run over the intake structure and the lower grade at Elevation +15 feet MSL and eventually runup the reactor-turbine building grade.
Considering the runup on a smooth, impermeable composite slope with its crest i
at Elevation +23 feet MSL and 2.5 feet of freeboard, a maximum wave height of j
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3.0 feet and period of 3.9 seconds, the maximum distance the runup may be ex-I pected to reach on the plant grade is 55-65 feet from the edge of the grade.
i The reactor-turbine building complex is located at least 75 feet from the edge of the grade and hence, the waves should not affect these structures. This assumes that the grade will have proper drainage towards the canal.
At the time of maximum stillwater level at the site, +1.0 hour0 days <br />0 hours <br />0 weeks <br />0 months <br />, the signi-ficant wave height in the canal would be reduced to 1.7 feet.
This case, as well as the remaining cases of wave action in the canal system during the rising and falling surge, are less critical than the situation described previously in detail 90000070 9-3-4
The wave runup elevations and wave forces are considered conservative estimates as no credit was taken for the presence of obstructions on the intake structure deck or the lower plant grade, which are clearly _ indicated on Figure C-1 and which certainly would contribute to decreased runup and forces du to friction -
and turbulence.
In conclusion, the wave analysis demonstrates that there will be no impairment of the functionality 'of the diesel generators in the event of a PE i
flood.- With regard to the emergency service water pumps, the service water pumps and the circulating water pumps, the storm surge during a PE would submerge the
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intake structure and consequently the pumps located at this elevation.
In this i
event, the plant would shutdown and' decay heat removal could be facilitated by use of the emergency condensers.
The condensate storage tank and the plant ventilation systems would not be affected.
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TIME (HOURS SINCE PEAK SURGE AT OPEN COASTLINE) a PMH SURGE HYDROGRAPH AND WAVE CHARACTERISTICS o
5 gi IN THE INTAKE' CANAL 90000074 IN FRONT OF THE INTAKE STRUCTURE OYSTER CREEK NUCLEAR GENERATING STATION is
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