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CONSUMERS PUBLIC POWER DISTRICT 1 COOPER NUCLEAR STATION DOCKET NUMBER 50-298 v a .. _, .. -! Fi:s of AMENDMENT I
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.:~ 'CNS q AMENDMENT 1 j;
- PART I - INSERT REVISIONS TO PSAR The subject insert revision.pages_ and figures are not included in this bound -
4- volume but are separately attached for convenience in inserting them in the
'PSAR Volumes 1, II, and III.
ERRATA AND ADDENDA SHEET-b COOPER NUCLEAR STATION PRELIMINARY SAFEIT ' ANALYSIS REPORT -
INSERT PAGES AMENDMENT'1, PART I
]
Revision Date 12/1/67
, - 1. Remove the following pages from Volume I..
vii, viii,-. ix, xvii, II-2-3, 11-4-2, 11-4-3, 11-4-4, II-4-5, III-5-7, IV-2-11, IV-2-13, IV-3-3, V-1-2, V-1-3, v-2-1, V-2-3, v-2-4, V-2-5, V-2-6, V-2-7, V-2-5, ,
V-2-10,.V-2-12, V-2-19,.V-2-20, V-2-21, V-2-22, V-2-23, V-2-24, V-2-25, V-3-5, VI-2-3, VI-2-5,VI-2-11, VI-2-23, VI-2-24, and VI-2-27. -l l
- 2. Insert the attached revised and new (12/1/67) pages in Volume I. . -l
, vii, viii, ix, ' xvii, II-2-3, 11-4-2, 11-4-3, II-4-3A, II-4-4, II-4-5, III-5-7, IV-2-11, IV-2-13, IV-3-3, V-1-2, V-1-3, v-2-1, V-2-3, V-2-4, V-2-5, V-2-6, V.2-7, V-2-9, V-2-10, V-2-12, V-2-19, V-2-20, V-2-21, V-2-22, V-2-23, V-2-24 V-2-25, I
V-3-5, VI-2-3, VI-2-5, VI-2-11, VI-2-23, VI-2-24, and VI-2-27.
- 3. Remove the following figures from Volume I. '
1-1-1, IV-2-4, V-2-3, V-2-4, V-2-5,V-2-6, V-2-7, V-3-1, and VI-2-2.
- 4. Insert the attached revised and new (12/1/67) figures in Volume I.
1-1-1, IV-2 , V-2-3, V-2-4, V-2-5, V-2-6, V-2-7, V-3-1, and VI-2-2.
- 5. Remove the following pages from Volume II. 3 VII-4-1, VII-4-2, VII-4-3, VII-6-5, VII-6-6, VII-6-7, VII-6-8, VII-7-1, VII-7-4, i VII-7-5, VII-7-6, VIII-1-1, VLI-3-1, VIII-3-2, VIII-4-1, VIII-4-2, VIII-4-3, VIII-4-6, VIII-4-7, VIII A, VIII-5-2, VIII-5-3, VIII-6-1, VIII-6-2, "Section I-Li.st of Figures", X-2-5, X-2-8, X-2-9, X-2-10, X-3-1, X-3-2, X-3-3, X-3-4, X-3-5, ,
X-3-6, X-3-7, X-3-8, X-3-9, X-3-10, X-3-11, X-3-12, 1-4-1, X-4-2, X-4-3, X-4-4, i XI-2-1, XI-3-1, XI-5-1, XI-6-1, XI-8-1, XI-8-2, "Section XII-List of Figures",
XII-2-1, XII-2-2, XII-2-3, XII-2-4, XII-2-5, XII-2-6, XII-2-7, XII-2-10, XIV-3-15,
, ~XIV-3-17, XIV-3-20, and XIV-3-27.
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PART I - INSERT REVISIONS TO PSAR-(CONT'D)
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- 6. Insert the. attached revised or new (12/1/67) pages in Volume II..
VII-4-1, VII-4-2, VII-4-3, VII-6-5, VII-6-6, VII-6-7, VII-6-8, VII-7-1, VII-7-4, VII-7-5, VII-7-6, VII-7-6A, VIII-1-1, VIII-3-1, VIII-3-2, VIII-3-2A, VIII-4-1, !
VIII-4-1A,-VIII-4-2, VIII-4-2A, VIII-4-3, VIII-4-3A, VIII-4-6, VIII-4-7, VIII-5-1, f VIII-5-2, VIII-5-3, VIII-6-1, VIII-6-2, "Section X-List of Figures", X-2-5, X-2-8, X-2-9, X-2-10, X-3-1, . X-3-2, X-3-3, 1-3-4, X-3-4A, X-3-5, X-3-5A, X-3-6, X-3-7, X-3-8, X-3-8A, X-3-9, X-3-10,'X-3-11, X-3-12, X-4-1,'X-4-2, X-4-3, X-4-3A, X-4-4, ,
XI-2-1,' XI-3-1, XI-5-1, XI-6-1, XI-8-1, XI-8-2, "Section XII-List of Figures".- i' XII-2-l', XII-2-2, XII-2-3, XII-2-4, XII-2-4A, XII-2-5, XII-2-6, XII-2-7, XII-2-10, XIV-3-15,'XIV-3-17, XIV-3-20, and XIV-3-27.
- 7. Remove the following figures from Volume II.
VII-4-5, VII-4-6, VII-5-1, VII-7-2, VIII-2-1,, VIII-3-1, VIII-6-1, X-2-4, X-3-8, X-4-2,:XI-1-1, XI-6-1, XII-2-1, XII-2-2, XII-2-3, XII-2-4, XII-2-5, XII-2-6, XII-2-7, XII-2-8, XII-2-9, XII-2-10, XII-2-11, XII-2-12, XII-2-13, XII-2-14, XII-2-15, XII-2-16, XIV-3-5, and XIV-3-6.
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- 8. In' sert the attached revised or new (12/1/67) figures in Volume II.
VII-4-5A, VII-4-5B, VII-4-6, VII-5-1, VII-7-2A, VII-7-2B, VIII-2-1, VIII-3-1,
' VIII-6-1, X-2-4A, X-2-48, X-2-6, X-3-8, X-4-2, 11-1-1, XI-6-1, XII-2-1, XII-2-2, .j XII-2-3, XII-2-4, XII-2-5, XII-2-6, XII-2-7, XII-2-8, XII-2-9, XII-2-10, XII-2-11, '
XII-2-12, XII-2-13, XII-2-14, XII-2-15, XII-2-16, XII-2-19, XIV-3-5, and XIV-3-6.
- 9. Remove the' following pages f rom Volume III.
" Appendix B-Title Page". D-4-2.
- 10. Insert the attached . revised and new (12/1/67) pages and figures in Volume III.
" Appendix B-Title Page", " Appendix B-Letter", D-4-2, " Appendix' G-Index Tab",
" Appendix G-Title Page", " Appendix G-Table of Contents Page", G-1-1, G-1-2, G-2-1, G-2-2, G-2-3, G-2-4, Gd3-1, G-4-1, G-4-2, G-4-3, G-4-4, G-5-1, and G-6-1.
Revision Date 11/7/67-
- 11. Remove the following pages from Volume III.
1-2, Fig III-2,' Fig IV-7, VI-2, VI-4, Table VI-1, Fig VI-10 VII-6, VII-7, Fig VII-5, Fig VII-6, IX-3, II-A-1, II-A-2, III-A-12.
- 12. Insert the attached revised and new (11/7/67) pages & figures in Volume III.
I-2. Fig III-2, Fig IV-7, VI-2, VI-4, Table VI-1, Fig VI-10, VII-6, VII-7, Fig VII-5, Fig VII-6, IX-3, -II-A-1, II-A-2, III-A-12.
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t CNS AMENDMENT 1 PART III - INSERT REVISIONS TO APPLICATION The subject _ insert. revision pages and figurec are not included in this bound volume but are separately attached for convenience in inserting them in the Application.
' ERRATA AND ADDENDA-SHEET COOPER NUCLEAR STATION APPLICATION INSERT PAGES AMENDMENT 1, PART III Revision Date 12/1/67
- 1. Remove the following pages from Attachment D of the Application.
" Attachment D-3 and D-8".
- 2. Insert the attached revised and new (12/1/67) pages in Attachment D of the Application.
" Attachment D-3", D-3A, D-8, " Exhibit V-Title Page", and " Exhibit V".
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II-0-1
. CNS-1 COOPER NUCLEAR STATION (CNS)
AMENDMENT NO. 1 DOCKET NUMBER 50-298 PART II_
TABLE OF CONTENTS 1 DELETION OF SUPPRESSION CHAMBER BAFFLES 2 DELETION OF EQUALIZER LINE 3 REDUCTION OF VENT AREA 4 REDUCTION IN SUPPRESSION POOL WATER VOLUME 5 REVISED COFIAINMENT PRESSURE RESPONSE l
6 REVISED CONTAINMENT TEMPERATURE RESPONSE 7 EFFECT.OF CONTAINMENT CHANGES ON CALCULATED OFF-SITE DOSES 8 PHYSICAL CHANGES TO PRIMARY CONTAINMENT SYSTEM AND REACTOR BUILDING 9 TOPOGRAPHICAL DESCRIPTION AND MAPS APPENDIX A - ANALYSIS OF DELAYED CONDENSATION--0VIRPRESSURE CONSIDERATIONS l
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'1.- DELETION OF SUPPRESSION CHAMBER BAFFLES
'This section and its Appendix'."A" presents an analysis upon which the .
deletion of the suppression chamber baffles in Cooper Nuclear Station '
.is based.-
The suppression chamber baffics were originally. included in:the design.
to prevent a short term overpressure of.some 6 psig as' observed in a series
., ,of 1/4' scale tests performed at Moss Landing. The basis for their removal-is'three fold
- 1. The suppression chamber design' pressure is 56 psig (based on code allowances for a maximum internal pressure of 62 psig) rather than 35 psig (non-code corrected maximum internal pressure), as first proposed for.Dresden Unit 2 (AEC Docket 950-237). ' Therefore, even if. the observed overpressure were to occur, the design pressure of the suppression chamber would not be exceeded.
2.- Convincing evidence exists that the overpressure would not
' occur in a full. scale geometry.
- 3. The installation of baffles is not required'to prevent azimuthal sloshing, uniform distribution, or other fluid perturbations.
The design pressure for the drywell and suppression chamber for Cooper
- Nuclear Station -(CNS) (AEC Docket f50-298) war established on the basis of the Bodega Bay pressure suppression tests.(1) The applications of these tests in the design of Cooper Nuclear Station containment resulted in a design drywell pressure of 56 psig (based ou code ~ allowances for a maximum internal pressure of 62'psig) and a suppression chamber design ,
pressure of 35 psig. However, to simplify future pneumatic tests of the j primary containment, the suppression chamber design pressure was increased 1 from 35 peig (46-peis) to 56 psig (based on code allowances for a maximum internal pressure of 62 psig). Therefore, the drywell and suppression chamber have identical design pressures, i.e. , 56 psig (based on code allowances for a maximum internal pressure of 62 psig). 1 Item 1, above, is the keypoint in justifying the deletion of the baffles, i.e. , the design pressure is much higher than the maximum pressure observed
_ in any of the tests.
The support of Item 2, above, is based on a detailed study of the quarter scale test data available. In Figure II-1-1 the measured pressure response of the suppression chamber 'is illustrated for the 1/4 scale tests with and l without bafflee and for the full scale tests which were, in essence, fully !
baffled. Note that with no baffles the suppression chamber pressure (at (1) - Bodega Bay Preliminary Hazards Summary Report, Appendix 1, Docket #50-205, December 28, 1962, i
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4 x about 0.7 seconds is greater.than the end'of transient pressure. This
.early overpressure was the-original basis for the baffles. Note, also, ,
that at approximately,the'same time as the overpressure fn'the 1/4 scale ]
C tests,' a pressure peak occurred in the full scale tests. However, in the j full'scalentest.the pressure peak did not exceed.the end of transient.
pressure. - This difference in mean pressure at the.. time of the. peak was ;
caused by a difference in the. pressure rate in the suppression chamber. _)
.In the 1/4 scale tests, the drywell and suppression chamber volumes were ]
- , , reduced by 1/4. The downconer diameter was also reduced.by.1/4,Lwhich reduced the flow area per downcomer to approximately 1/16 of full scale..
- With ' fourteen downcomers the toral vent flow area was approximately the . ;
same for both. tests, the total mass. flow rate into the suppression chamber-was'the same. Therefore, the 1/4 smaller suppression chamber air volume uns pressurized four. times as fast. This results in accentuating the pressure peak for the 1/4 scale tests, since the average pressure at the ,
time it occurred was much higher than for the full scale tests. ;
It'is possible to calculate the suppression chamber mean pressure rise assuming complete condensation in the suppression chamber-pool. The overpressure peak can be simulated by superimposing a mass flow into the suppression chamber air' volume which is >.ondensed-after a,short delay.
It has' been determined that a sinusoid.s1 input can reasonably' reproduce the quarter scale
- test pressure response with and without baffles. (See Appendix A)~. Applying the same input to the full scale tests confirms that the pressure' peak would not exceed the end of transient pressure.
It might.be postulated that the time constant of the sinusoidal input should be greater for the full scale geometry. If a time constant four times greater than-the quarter scale tests'is assumed in calculating the full- scale pressure response a pressure in excess of 'the end of transient pressureLis obtained. However, the overpressure is'only a few psi and-would be of no concern with the higher design pressure. Further, the superimposed steam flow rate exceeds the total steam flow rate making delayed condensation to this extent impossible. 'The test data also indicates that the time constant should not be four times greater than for the quarter scale tests. First, there was the pressure peak in the full scale test at about 0.7 second. Second, movies taken of the Humboldt tests showed violent splashing of the; water initially, but after about one second the water in the pool settled down to a relatively calm bubbling type phenomena. The delayed condensation is postulated to occur during the initial violent pool action when the water has been thrown away from.the downcomer exit by the incoming steam-air mixture. The time required for the .I water to rise and fall back to the pool should be approximately the same for both full scale and quarter scale geometry. Hence, the pressure peak should occur at about the same time in both geometries. The test data confirm this' hypothesis. Therefore; it is concluded that the overpressure observed'in the 1/4 scale tests would be manifested as a pressure peak less than the end of transient pressure in a full scale geometry.
The basis for using baffles to control or limit azimuthal sloshing of water in the suppression chamber is based upon two postulated effects, i.e :
- a. Sloshing could result in large waves which might uncover the downcomer; and t
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,~ b. Sloshing could. result in danger to the vent header due'to an
- uplif ting force caused by large waves.
Neither of these postulated effects is of concern. Tests from both:the-
'Humboldt and Bodega series of pressure suppression tests proved that con-
-- densation would be complete even if the downconers were uncovered during the blowdown: therefore, postulated uncovering of the downcomers is not cause for concern. With respect to the second postulated effect, the loading imposed by wave action'is less ~ than 30% of the loading imposed by jet action on.the downcomers during venting. Since the maximum jet action and the wave action will never be. time superimposed one upon the other, the design of-the vent. header to withstand the larger force from the
.immediate jet action aborts any concern with respect to the:later waves-action. Further, it is highly improbable 'that large wave action would be established in the first place. The natural frequency of the suppression chamber pool is quite low-(on the order of 0.1 cps for large waves) and there is no apparent exciting force of a sufficiently low frequency to establish a large wave. ' The venting action would be a uniform, (circum-
, forentially) high frequency loading which, at most, would exit many small' l Lwaves,.but not a large wave. The small waves would not have a large -
enough_ amplitude to be of any concern and would die 'out quite rapidly.
therefore. . large wave action is not likely to be established.
Finally, the vents will blowdown nearly uniformly since the resistance of the vents is much' higher-than the resistance encountered by the steam-mass
' flow traversing; the drywell to any of the vents. This should preclude any inherent tendency to initiate sloshing in the first place.
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-2. DELETION OF EQUALIZER LINE I
In the original Cooper Nuclear Station design, the two jet pump headers l 4
were connected by an equalizer line. The purpose of the equalizer line
- was to improve plant performance and availability with one reactor coolant recirculation system pump inoperative. However, analysis shows that better than 65% of rated power (1548 MWt) is obtainable under this condition with-
- aut the equalizer'line. Therefore, since this is an acceptable operational limitation, the equalizer line has been deleted.
By eliminating .the equalizer line, the maximum break area has been sub-stantially reduced by making blowdown through the equalizer line impossible.
(. , Further,, the logic network 'required to identify the unbroken reactor coolant -
recirculation system line for LPCIS injection purposes has been simplified.
Only a simple differential pressure device (for flow direction determination) is'now needed. Refer to BECO-PNPS Amendment 2, Question-Answer (AEC Docket
- 50-293) for a detailed description of the former LPCIS injection logic '
control system.
The revised CNS-PSAR diagram, Figure IV-2-4, Revised 12/1/67, shows the equalizer line and associated logic network (instrumentation) deleted.
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- 3. REDUCTION OF VENT AREA As a direct result of the reduction in maximum break area from the dele-tion of: the equalizer line, the required size of the drywell/wetwell vents has been reduced. The reduction in vent area represents no change in u basic containment design criteria. The same ratio of. break area / vent area that has been.used in all previous designs starting with Oyster Creek, Unit (AEC Docket MO-219 ) is maintained. The ratio of break area / vent ares is a parameter in establishing peak drywell pressure in a. pressure suppression containment, and has been established by tests in the Moss Landing facility. -
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, - 4. REDUCTION IN SUPPRESSION POOL WATER VOLUME On CNS-PSAR page VI-2-29 the mode of. operation for.the Reactor Core Isolation Cooling System (RCICS) is correctly described. There has_been no change in the system nor in the system operation as described, but on original'CNS-PSAR page X-2-10 temperatures; applicable to an earlier RCICS
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' mode of ~ operation similar to ' that ; of. Dresden 2 wereL quoted. ' With the _
present design ~ using the RHR heat exchanger as a condenser. . the ' mode of operation as described on CNS-PSAR page VI-2-29, results.in a suppression chamber poo1' temperature af ter four hours of RCICS; operation of only 1100F (assuming an' initial pool temperature and RERS-service water temperature 1
of 90 0F) instead of.130*F as originally quoted on CNS-PSAR page X-2-10.
Therefore,-the difference between maximum allowed pool temperature from (
RCICS operation during plant " hot standby" operation, and the. design basis j
' maximum temperature (170 F) af ter a design basis loss-of-coolant accident blowdown'is now 600F. 'A design limit of 1700F for the suppression chamber pool temperature af ter,the design basis loss-of-coolant accident blowdown is ' justified as described in the answers to questions F4 and F5 of Amend-
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ment' 2 -to the Quad Cities 16. 20PDAR.- (AEC Dockets 150-254 and-50-265). 1 Only 500F of. the available 60 F is actually utilized, as discussed below.
To take advantage of the increased temperature differential available af ter -
four hours of'RCICS operation during plant " hot standby" operation, the mingmum volume of the suppression chasher water has been reduced to 87,660
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ft. (based on ~ utilizing only 500F of the available 600F temperature dif-ferential, thus; resulting in an end of accident blowdown maximum tempera-ture of 1600F.) The further reduction in' water volume needed to give an end of accident blowdown temperature of 1700F was offset by the increased air volume required at the higher temperature and the resulting increase in the partial pressure of the water vapor. In addition, certain geometric limits established to insure applicability of the test data. available make further water volume reductions impractical. Therefore, the suppression chamber. pool temperature following a design basis loss-of-coolant accident
. blowdown immediately after four hours of RCICS operation has been reduced to 160 F on'CNS rather than the 1700F originally quoted on CNS-PSAR page X-2-10. CNS-PSAR page X-2-10 has been revised in this Amendment 12/1/67 to indicate the new design temperatures.
During the detailed design phase of the CNS, thermal shock considerations with respect to cyclic operation of the system will be thoroughly analyzed.
Preliminary investigations indicate that no cyclic thermal shock problem exists.
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-II-5 t CNS-1 S. REVISED CONTAINMENT PRESSURE RESPONSE The changes in containment design discussed above in sections 1 through !-
have little effect on the containment pressure response. However, sub-sequent to release of the CNS-PSAR a slight modification has been made in the analytical model used to calculate. the containment pressure respecse i during the first 30 seconds. The new model yields peak drywell pressures '
that are about 3 to 4 psi higher than the old analytical model due to the considerations of effects of liquid carryover into the vents. Examination of the calculations show the change in peak drywell pressure is caused mostly by.the change in cal:ulational technique and.only slightly by.the above (sections 1 through 4) changes in containment design.
The reduced water volume in the suppression chamber pool does result in a slight increase in pressure late (>200 seconds) in the long term re- l sponse. The later pressure peak increases to about 16 psig from 14 psig,.
but this is of no concern since it is well below the design pressure of 56 psig. ,
(See revised CNS-PSAR Figure XIV-3-5, 12/1/67.)
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'6. . REVISED CONTAINMENT TEMPERATURE RESPONSE-J. -
-The changes in containment design discussed above-(Section 1 through 4)-
have'little (negligible) effect on the containment temperature response.
(See revised CNS-PSAR Figure XIV-3-6, 12/1/67.) -
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- 8" g . .. . 7. EFFECT OF CONTAINMEKf' CHANGES ON CALCUIJLTED OFFSITE DOSES
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' The slight. increase in peak containment pressure in the first 10 seconds-
.. and again.later'around.4 hoars after the accident (CNS-PSAR Figure XIV-3-5)
- @$, does not. change the calculated offsit'e radiation doses for'the design basis loss-ogy coolant accident'(CNS-PSAR Table.XIV-3-9) within'the first-signi-
. ficanthfigure accuracy reported. The leakage rate response curve (CNS-PSAR Figure XIV-3-7) has not been revised since the effects of the con-tainmetu design changes . (secticns 1 through 4) would have negligible effect on thishtrve. ' Thus CNS-PSAR Figure XIV-3-7 will remain an applicable,-
1 conservation leakage rate response curve.--
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< CNS-1 8.0 PHYSICAL CHANCES TO PRIMARY CONTAINMENT SYSTEM AND REACTOR BUILDING As a result of the elimination of the equalizer line and the correction of the suppression pool temperatures associated with the RCIC system operation as described in paragraphs 2, 3 and' 4 above, the size of the vent pipes between the drywell and torus has been reduced from 6'9" I.D. to 5'11" I.D.,
the number of downcomers has been reduced from 96.to 80 (see page V-2-4 revised 12/1/67); the water volume in the torus has been reduced thereby allowing a reduction in the major and minor diameters of the torus (see page V-2-4 revised 12/1/67).
The reduction in the major and minor diameters of the torus has in turn allowed a reduction in the overall dimensions of the reactor building.
These dimensional changes are shown in revised figures XII-2-4 through XII-2-11, revised 12/1/67.
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-a . ;x b ,9. -TOPOGRAPHICAL DESCRIPTION AND MAPS
- W m M![ A topographical description of the.-site' and the. surrounding area is' shown
, on. Figures 9-1 and 9-2. . These Figurer. are copies of topographic maps published by the? United States Department of Interior, Geological Survey.
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Figure 9-1 is a 'large scale topographic map included primarily to'.'show '
. elevations"and the. extent of the Missouri; River flood plain. Since the map was; printed in 1949 much of the information shown (other than the elevations) is not accurate. .. i.e. , nost: of the country schoolsL shown have been abandoned, the' levee as it now exists .is not shown and the river' channel -
has been relocated by the Corps of Engineers stabilization program.
' Figure 9-2 is a topographic map covering a larger area than that shown in .
p, Figure 9-1 but to a smaller scale. Since this map was printed in.1958 it gives a more accurate description of the area as-it now exists; however,'
there is an inaccuracy in this map also, namely that many of the country, schools shown have been abandoned.-
The Cooper Nuclear Station is located on the Missouri River flood plain on the west side of the river. The relatively flat flood plain extends
.approximately one mile to the west and more than five miles to the east i from the plant sita.
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.CNS-11 APPENDIX A - ANALYSIS OF DELAYED CONDENSATION--0 OVERPRESSURE CONSIDERATIONS In order to obtain a' quantitative feel for the. degree of delayed conden-sation required to cause the suppression chamber overpressure observed in the 1-3/4 scale pressure suppression phenomena Moss Landing-Bodega Bay tests, an equivalent but simplified model ns analyzed. The model analyzed can be out-
, lined as follows:
- - Assumptions
- 1. . Constant total mass flow rate from the drywell'is assumed and the fraction of steam and air is determined assuming ideal mixing and constant pressure in the'drywell.=
?
- 2. Assuming; complete condensation of the steam, the pressure response of the suppression. chamber is calculated.- The suppression chamber air space is assumed to be isothermal and constant volume.
- 3. To simulate steam added to the suppression chamber, but not condensed, an additional' mass addition is made to the air volume. .An equal nega-tive mass addition later simulates the delayed condensation of.this st steam. One complete sine wave is used to accomplish the mass addition and removal.
- 4. The total pressure response is calculated for the quarter scale tests with and without baffles. :The magnitude of the sinusoidal' mass flow of' item (3) above is selected to agree'with the experimental data.
- 5. Using the'same constants derived in item (4) above the pressure re-sponse for the fu11' scale test is calculated and compared with the data'.
- 6. The effects of changes in some of the key constants are discussed.
Analysis By assuming ideal mixing in the drywell the ratio of steam flow to total flow can be approximated by:
. 1.e4t (1)
"t The constant is a function of the total mass flow rate and the air volume of the drywell and suppression chamber. For.the 1/4 scale tests, = 4.61, while for the full scale tests, = 0.77.
Assuming an ideal gas in the suppression chamber M asRT p,
V,,
L For variable mass T y(g) (2)
P(t) =,
V as 4
- 9
.+
!. _ ~ 'II-A-2 s-CNS-l' Where M(t) s' M, sydt (3)
- Assuming' complete. condensation of the steam, the flow into the suppression chamber free volume is: j
- f. k ) 'k
.A ne *
~
g" T I~
- p
, Substituting from (1) -
t s cc =& T e
-At (4) 1
' k T ,
(5)
. , p(t) = p, +' (1-edt) 1 Equation (5) is compared'with the test data in Figure 1-1 for both the full I scale and quarter scale' tests. For easy comparison,- equation (5). has been
~
nondimensionalized, and tha test' data normalized to agree 'with equation (5) '
at the end'of' transient.
-In order to simulate the effects of delayed condensation, an additional' ;
mass flow into the suppression chamber was superimposed on the above flow. A- l
.sinusoidal input was arbitrarily assumed, i.e. I
, ' A* = a sTsinwt 05 U w !
's* = 0 t >M i w
~ ~
Where: a is'a constant represent 1ng the maximum frac. tion of the total flow rate -
which is not condensed in addition to the flow given by equation (4). .
ak g*----
, t .U.
m* w j t ;
' The constant w was selected to give the peak overpressure at the time observed !
in the tests. Specifically 2 so that the steam is completely condensed at t = 1 second. =d In Figure II-A-2, the calculated responses for 4
a = 0.136 a = 0.264 ]1 are illustrated along with test data with and without baffles. Also shown in l Figure II-A-2 is the suppression chamber pressure response under the assumption j of no condensation. As can be seen, the above analysis can reasonably repro- l duce the data if the correct constants are used. d 0
Y
)
i II-A-3 CNS-1 1
Unfortunately, there is no way to a priori calculate the constants "a" and "w" for use in calculating the full scale response. As far as "w" is concerned, it seems reasonable to assume the same value as for the quarter scale tests, as discussed in the text of the ' cover letter. In Figure II-A-3, the pressure response for the fully baffled full scale test is illustrated.
The calculated responses assume a = 0.3 and a = 0.9. Two points should be noted.
A. The test pressure rate is faster than the calculated pressure rate for no condensation.
B. Assuming a 2 0.3 results in bne + $* >I"T , i.e. , air flow plus uncondensed steam flow is greater than the totc1 flow rate, which is not physically i posrible.
Item (A) is due to a combination of effects. The most important are:
(a) A piston like action of the pool water which compresses the air in tthe free suppression chamber. This piston action apparently did not occur in the 1/4 scale tests, probably because the smaller water volume covering the downcomer exit broke up when hurled upward by the initial air blast. [
-(b) Initial flow rates in excess of the average value used in analysis.
I In spite of the disparities between the calculated response and the measured response, one conclusion can still be reached. With w> 2w overpressure in excess of the design pressure are not possible without some kind of piston action by the pool water. Such a piston action is more likely (though not probable) to be promoted by baffling rather than prevented.
In Figures II-A-4 and 11-4-5, the calculated pressure responses for w = r and r/2 are shown even though these values of w are not considered reasonable.
For w = 7r , a pressure peak in excess of the end of transient pressure is obtained only for a 2 0.9, which requires uncondensed flow greater than the total flow.
Only when w = 7/2 is an overpressure obtained for a reasonable value of "a".
Even then, with a = 0.3, the overpressure only exceeds the end of transient pressure by about 3 psi. Recall that for the quarter scale tests, a = 0.264 correlated the overpressure reasonably well for the unbaffled case. Note, how-ever, that for w a r/2, the full scale data cannot he approximated for any value of a. This fact further confirms the argument that w must equal approxi-mately 2 T. .
I Conclusion ,
j In conclusion, it condensation of the steam flowing into the suppression chamber is complete af ter about one second, pressure peak in excess of the end of transient, pressure cannot occur. Comparison of test data and analysis indicate that any delay in condensation will occur during the first one second of the ;
transient. Therefore, baffles are not required in any design having suppression 1 chamber pressure rates comparable to or less than the full scale Bodega Bay tests. )
(
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- -* Rav'. '12/1/67 y
[ +
- - ,D_
s Summary of pertinent provisions, terms and conditions
- of BURNS AND ROE, INC. ~ Contract for architect
- engineering services is as f ollows:
g
' The. District.hae entered into an agreement with Burns and Roe, Inc. to
- provide the architect engineering services for the 800 MW nuclear plant.
The' contractor- has prepared site surveys, engineering descriptions,; cost estimates and certain bidding specifications. It will prepare other necessary
- plans and specifications, technical and economic evaluation of bids, will do
- engineering liaison with manuf acturers of major equipment during fabrication, and other duties normally associated with architect engineering for such plant. The District is paying a fixed fee plus the engineer's costs as defined in the Contract.
.* Summary of pertinent provisions, terms and conditions of BURNS AND ROE, INC. Contract for construction
" management services is as follows:
- The District has an agteement with Burns and Roe, Inc. to manage the construction of the 800 MW nuclear plant and the related 345 KV switchyard.
The contractor will manage, the construction contracts, coordinate super-vision of the work and receipt and control of equipment and materials at the site, maintain necessary time and material records, provide for safety and' site security and furnish other services normally associated with con-struction management. The District will pay a fixed f ee plus the contractor's costs as defined in the Contract.
- Sunnary of pertinent provisions, terms and conditions of STEARNS-ROGER CORPORATION l i
contract is as follows:
- The District has entered into an agreement with Stearns-Roger Corporation to provide quality assurance for the 800 MW nuclear plant. This will in-clude witnessing of factory tests and engineering inspection of all equipment and materials and on-site engineering and technical inspection of all con-struction in progress as well as other work normally associated with Quality Assurance. . The contract specifically gives the District an option to employ Woodward-Clyde-Sherard and Associates to take responsibility for quality L ^ assurance With respect to the excavation and structural fill. The District is paying the payrcil costs for this work plus a percentage override for L
. overhead and services of the firm.
L j
~Y .
e
~
, l K
e - 3A - Rzv. 12/1/67
,g I' j L' '
r -la j
.D.
-l
- Summary of pertinent provisions, terme and conditions of CENERAL ELECTRIC CO.
P Nuclear Fuel Contract.is as follows:
- The District has entered into a contract with General Electric Company to design; furnish materials (including special nuclear material) for; 2 fabricate; aand deliver.to the District the initial core-load of fuel
' '] '
, bundles and the second core load of fuel bundles. The contract fixes the price and the buyback price with formulas for adjusting both prices. There is. a perf ormance warranty in the contract.
- b. Separately' Identified Costs FPC- 4 Account Description Amount -
Number 322 thru ,
.325,
-incl. Startup Costs (net) not included in I. A.1. a. but including pre-operational testing 750,000 320 thru 325 Fayroll costs for in-house personnel and miscellaneous other costs 700,000 Contingencies and Escalation 3,775,100 Administrative, legal and financing costs 378,000 Refunding Costs prorated to Facility 6,389,000
. Working Capital 1,150,000 Supplying Replacement and Contingency Fund 2,000,000 h- Net Interest during construction
- -39.950,000 15.492.00_0_
24.458,000 Funding Debt Service Reserve '7,990,000 Bond Discount 3,760,000 t
'j
O Rev. 12/1/67
- D The ratings accorded the District's outstanding revenue bonds are as follows
- ,
l-
' Moody's. Standard and Poor's
. Bonds Maturing on or bef ore 1972 A A Bond Maturing After 1972 Baa A The District's plans for future bond offerings and effect on capitalization are contained in the Outline of Smith, Barney & Co. attached hereto as Exhibit 1.
Funds to be generated internally are shown in Table III of Exhibit 1.
- b. The District is not an instrumentality as agent i; for anyone else. As stated in Exhibit 2, the District will sell one-half of the plant's output to Iowa Power and Light Company. A copy of the Facilities Agreement between the District and Iowa Power and Light Company is attached hereto as Exhibit 2. A copy of the latest' Annual Report of Iowa Power and Light Company is attached hereto '
as Exhibit 4.
3
-
- A summary of the District's Power Sales Contract with i Iowa Power and Light Company is attached hereto as /
Exhibit 5.
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FIGURE 9-1
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