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Latest revision as of 13:32, 27 February 2020
| ML19326D106 | |
| Person / Time | |
|---|---|
| Site: | Midland |
| Issue date: | 04/13/1971 |
| From: | Kessler W CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | AFFILIATION NOT ASSIGNED |
| Shared Package | |
| ML19326D105 | List: |
| References | |
| NUDOCS 8006060630 | |
| Download: ML19326D106 (450) | |
Text
{{#Wiki_filter:(3 OCCE ".5: X.1 EfiDD. & UT!L. EEC.9-W,MO UNITED STATES OF AMERICA ATCMIC ENERGY CCMMISSION In the Matter of CONSUMERS POWER COMPANY Docket No. 50-329 Docket No. 50-330 (Midland Plant) STATE OF MICEIGAN ) SS. COUNTY OF JACKSON William E. Kessler, being duly svorn, deposes and says that he is Project Manager for the Midland Nuclear Plant and that he has read the following responses to interrogatories submitted by Intervenors Sa61 nav Valley Nuclear Study Group, et al. and that the same are true to his personal knowledge or upon the basis of his infomaticn and belief.
/
William E. Kessler Subscribed and svors to before ne this 13th day of April,1971. sh f. _. _ . _ ]I. ' . x z . Ib ' Aileen R. Nazaruk j Notary Public, Jackscn County, Michigan l My C=nksien Expires April 30, 1973 l l c 1 I i ca,. f Q nun:D y S ELE p;g A?? 131971
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- 2. With respect to each e.periment and its results relied upon in the PSAR to support or justify the design and effective operation :
i of the Iodine Removal Spray System, state: (a) The parameters of each such experiment and what specific factors inhering or governing each such experiment justify reliance in your opinion on such experiment regarding the Iodine Removal Spray Systems in the proposed Midland Units; (b) What factors, if any, in each such experiment, if not in-hering in the operation of the Iodine Removal Spray System in the proposed Midland Units would prevent reliance, in your opinion, upon such experiment as supporting authority for the design and effective operation of the Iodine Re-moval Spray System. If in your answer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth ccupletely the text of each such reference or attach a copy. Answer In accordance with 10 CFR 50 34, the PSAR has a preliminary analysis and evaluation of the design and performance of the Iodine Removal Spray System. In addition, to support the design of the Iodine Re-moval Spray System, Babcock & Wilcox has ccmpleted an extensive re-search and development program along with a thorough evaluation of . the Sodium Thiosulfaf,e Iodine Removal Spray System. W 2-1
The research and developnent program is reported in Babcock & Wilcox Topical Report EAW-10017, Revision 1, " Stability and Compatibility of
, Sodium Thiosulfate Spray Solutions," PROPRI-M. The evaluatica is reported in Babcock & Wilcox Topical Report BAW-10022, " Effectiveness of Sodium Thiosulfate Sprays for Iodine Removal," PROPRIETARY. A copy of each report has been made available to the intervenors.
The above-named reports, along with the PSAR, Sections 6.2, 14.2.2 3 7, 14.2.2.4, and Appendix lhA, support the preliminary design of the Icdine Removal Spray System, and detail the experiments providing the infor na-tion requested. 1 1 2-2 f
,7 3 Give a detailed description of each Iodine Removal Spray System or part thereof engineeringly useable or actually in use, but not planned to be used in the proposed Midland Units. Include within your answer your reasons for discarding such other systems or parts thereof. If in your answer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth completely the text of each such other reference or attach a copy. Answer The Iodine Removal Spray System hardware and arrangement described in PSAR (Reference Section 6.2) is typical of systems being used in, er proposed for other PWR plants. The variation in systems usually lies in the type of chemical additives used. Three different chemical additives have been tested and proven satis-factory for use in PWRs. They are: (a) Base-borate solution. This solution contains about 0 3 M l 1 boric acid and 0.15 M sc dium hydroxide and has a pH of about I 9 5. (b) Base-borate, sodium-thiosulfate solution. This solution con-
'l l
tains about 0 3 M borie acid, 0.15 M sodium hydroxide, and 1% I by vt sodium thiosulfate. The pH is about 9 5 (c) Base-borate solution. This solution contains about 0 3 M - boric acid and sufficient sodium hydroxide to raise the pH -. s to about 7 3-1
.c l
s The base-borate, sodium-thiosulfate solution is proposed for the Midland spray system because it has been ( proven'to have a greater removal capability for methyl iod$de than the other two solutions. The base-borate, sodium-thiosulfate solution also re-moves molecular iodine and particles as effectively or more effec-tively than the other two solutions. Thus, in our opinion, the solution used in the Midland Plant is superior to the others in tems of overall iodine removal. Because this solution requires the injection of two chemical addi-tives (sodium hydroxide and sodium-thiosulfate) the Midland spray system will employ two chemical additive tanks with associated piping, valves, and pumps instead of the one chemical additive tank with associated piping, valves, anc pumps nor= ally employed in systems using one chemical additive (sodium hydroxide). EAW-LOO 22, " Effectiveness of Sodium Thiosulfate Sprays for Iodine Removal," PROPRIETARY. 3-2 z 5.4
- 4. Give a definition of gas film resistance as that term is used in i
Section 14.2.2 3 7 of the PSAR. In connection with your definition and the relie- 3 in the PSAR upon the works of Taylor, Griffith, Ranz and Marshall, as recorded in footnotes 19 through 21 of said Section, state I i whether gas film resistance is the sole controlling factor in the trans- ! fer of elemental iodine into reactive solutions. Include within your 1 answer whether or not the relied-upon assumptions contained in the works l , of Taylor, Griffith, Ranz and Marshall are constant or variable and, if
)
l they vary, state in detail each differing variable, such as, for example, temperature, its relative significance to total and complete iodine re-moval, and how such variable factors affect or may affect your conclu-sions stated in the aforesaid Section of the PSAR. In connection with i your answer, please quantify the relative significance of such variable I fe.ctors regarding total and complete iodine removal by stating each such factor's absolute magnitude. If in your answer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth completely the text of each such other reference or attach a' copy. Answer The difference in the iodine concentration (actually the thermodynamic chemical activity) between the reactor building atmosphere and the inte-rior of each spray drop provides the driving potential for the mass transport of iodine frem the gas phase to the liquid phase. This move-ment of iodine would proceed at an infinite rate if there were no resis-tance to the mass flow. Since the mass transfer rate is finite, the 4-1
~
l movement of iodine is resisted. The resistance encountered in the gas phase is termed the " gas phase resistance" and that encountered in the liquid phase is termed the " liquid phase resistance." As with electrical i current, resistances cause potential drops; large resistances cause large potential drops. For the absorption of iodine into drops containing ! sodium thiosulfate, the liquid phase resistance is extremely small (a very small potential drop); therefore, essentially all of the potential drop occurs across the gas-phase resistance, and thus the gas-phase re- j sistance limits the mass transfer rate. A more complete discussion of gas-film resistances is given in the Mass Transfer Theory Section (Sec- l t tion 51), of Babcock & Wilcox topical report BAW-LOO 22, " Effectiveness of Sodium Thiosulfate Sprays for Iodine Removal," PROPRIETARY (a copy of which has been made available to the intervenors). Section 5 1 also l explains that if the Sherwood number (Sh) is 30 the gas phase resistance is 10% of the total resistance and if it is O.4 the gas phase resistance constitutes 90% of the total resistance. The s aller the Sherwood num-ber the larger the percentage of gas phase resistance. Appendix B of BAW-LOO 22 shews that values for the Sherwood number for a thiosulfate spray system in a typical reactor building range frcm 10" to 10 3, which clearly indicates that all the resistance is in the gas phase re-sistance and that the liquid phase resistance is negligible. Scme of the spray tests analyzed in Sections 5 & 6 of BAW-LOO 22 were performed without thiosulfate and under conditions which resulted in significant liquid phase resistance. As explained in Section 6, the 38W SPEAY-2
)
U 4-2
code calculates and checks the Sherwood number; if it exceeds 0.1, the liquid phase resistance is included in the calettlation of the iodine re-moval rate. Finally, the best proof that the " gas phase resistance con-trols the mass transfer rate is that the experimental data correlates best with calculational techniques that neglect the liquid phase resis-tance (as discussed in Sections 5 & 6 of Ex4-loc 22). When the PSAR was prepared, the contributions of Teylor, Griffith, Ran: and Marshall were an essential part of the design and evaluation of spray systems. Although their work is still valid, the state of the art has reached a point where a slr contributions have either become less essen-tial or have been superseded by new and better data. The adequacy of the iodine removal spray systems for the Midland Plant is supported by the tests and data described in EAW-10022. An explanation of the B&W SFRAY-2 code and a para =eter study shcwing the effects of the various parameters is also presented in Section 6 of BAW-LOO 22. l v 4-3
O , 5 With respect to your assumption set forth at Pages 14-63a of the PSAR "that all the drops are spherical and have the same diameter," state the following: (a) What is the variance of the diameter of the spray droplets for this particular system and the spray nozzles thereof; (b) The optimum size of the diameter of the droplets for the most effective working of the Iodine Removal Spray System. In connection with this answer, state why you have assumed, (PSAR, Page 14-63b), that 1,000 microcs is apparently the most effective diameter size. I.f you rely upon any experi-ment, describe in detail each such experiment including each fact, calculation, assumption and result thereof. O cc) R. Sardine the 1a=, sentence on Pase 1uA-t ef the PSAR, stete each fact, calculation and assumption which you believe sup-ports your assertion that you are " confident" that the Iodine Removal Spray System will perform as predicted. In your an-swer do not merely refer to footnoted textual references, but rather state in detail the basis for your " confidence" in ex-planatory languaBe. If you only rely upon what is contained in the PSAR, then please so state. If in your answe.r you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth completely the text of each such other reference or attach a copy.
- 93. With respect to the conclusion of the adequacy of the proposed Iodine Removal Spray System,. describe in detail each fact, assumption and 5-93-1
calculation by which it is concluded that the SFRAYC0 Model 1713A has demon-strated its ability to generate iodine removing spray. Set forth relevant portions of ORE-4374 and if in your answer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth completely the text of each such other reference or attach a copy. Answer The variance of drop sizes is normally termed the drop size distribution. The drop distribution produced by a given spray nozzle can be conserva-tively characterized by a log-normal distribution. This distribution is defined by the mass median diameter and a geometric standard deviation. The geometric standard deviation used for this type of nozzle is usually about 1.5 This subject is discussed in greater detail on Pa6e 6-9, and in Appendix A of Babcock & Wilcox Topical Report BAW-10022, " Effectiveness of Sodium Thiosulfate Sprays for Iodine Removal," FROPRIETARY, a copy of which has been made available to the intervenors. As shown on Page 14-63b of the PSAR, the removal rate can be calculated by the equation 6V,FH ds " V cdv
)
Where g - removal rate V - deposition velocity F - spray flow rate l H - drop fall height y 5-93-2
. .S ; V - containment free volume d - drop diameter v - drop velocity It can be seen that as the drop diameter, d, becomes smaller, the removal rate,4, increases. Therefore, one may conclude that the most effective drop diameter would be the smallest drop diameter. 1000 microns is ob-viously not the most effective drop diameter. However, 1000 microns is sufficiently small auch that the desired system performance can be achieved, and it is approaching the point of " diminishing return" as explained in Section 3.2 of Babcock & Wilcox Topical Report BAW-10022, " Effectiveness of Sodium Thiosulfate Sprays for Iodine Removal," PROPRIETARY, a copy of
- which has been made available to the intervenors.
.J The SPRAYCO 1713 nozzle was used in many Iodine Removal Spray Syszem tests conducted at Oak Fidge National Laboratory. The attached tables, Pa6es 204 and 205 of CENL 4374, demonstrate the effectiveness (short iodine re-moval half-life) of this nozzle. The SPEAYC01713A nozzle is the same nozzle. This SPRAYC0 nozzle was selected for use in the Midland Units more than two years ago. Since that time, many Iodine Removal Spray tests have been performed in both the Nuclear Safety Pilot Plant and the Contain-ment Systems Experiment. These spray tests used several different spray nozzles, all of which produced rapid iodine removal. The data from forty-three of these tests, which include tests with five different spray nozzles have been correlated with the B&W SPRAY-2 computer program. These re;ults are discussed in Section 6 of BAW-10022, " Effectiveness of Sodium Thiosulfate
/ t 5-93-3
+ , , , - ,
.i
.3-Sprays for Iodine Removal," PROPRIETARY, a copy of which has been made available to the intervenors.
The fact that all tests that used one weight percent sodium thiosulfate spray so:utions exhibited iodine removal rates faster than the rates predicted by the B8N SPRAY-2 computer program, is the basis for our con-fidence that this Iodine Removal Spray System will perform as predicted regardless of whether the SFRAYC01713A nozzle, or for that matter, any other nozzle for which spray drop size data is available is used. The following items further contribute to our confidence that the system will provide the desired iodine removal characteristics:
- 1. The mass transfer theory governing the removal rate of airborne iodine
,y by chemical sprays is well understood, and it is supported by forty-three spray tests in the Nuclear Safety Pilot Plant, and by twenty-eight spray tests in the Contai. snt Systems Experiment.
- 2. Sodium thiosulfate is the most effective reagent tested in these Icdine Removal 3 pray tests.
3 The Iodine Removal Spray System is a redundant system. Even if only one-half of the system operates, the system will perform so that 10 CFR 100 limits are not exceeded during a post-accident period. 4 The removal rate calculated in the PSAR Section 14.2.2.3.7 is suffi- ; l cient to show the adequacy of the preliminary system design. During l the final design phase the removal rate will be reevaluated with the B8M SPRAY-2 computer program to ensure this Iodine Removal Spray Sys-7 tem will perform as necessary to ensure compliance with 10 CFR 100 limits. 5-93 k l
') 5 The doses listed in PSAR Table 11+-11 are 6si nificsntly below the 10 CFR 100 limits.. s i Y 5-93-5
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- 6. State each condition under which the sodium thiosulfate-sodium hy-droxide solution will be unable to retain iodine during the course of the use of the Iodine Removal Spray System. In connection with your answer stating each such condition, also set forth each of the variables involved, explaining each such variable in detail, including each such variable's time history throughout the period of a Maximum Hypothetical Accident (MHA) and thereafter during the time which the possibility of radiation release remains. If in your answer you make reference to other than tex-tual (exclusive of footnote) matter in the PSAR, then set forth completely the text of each such other reference or attach a copy.
15 With reference to your Answer to Regulatory Staff Question 6.7 of Amendment No 5 to the PSAR, as set forth at Page 6.7-1 of the PSAR, state each fact, calculatica and assumption for each of your conclusiens sepa-rately regarding ther=rd stability, radiation stability , materials co= pat-ibility and iodine retention capability of the alkaline sedium thiosulfate spray solution. In your answer, also state in detail each fact, calcula-tion and assumption relied upon and contained in the Thiosulfate Research and Development Program reported in BAW-10017 and attach a copy of such document. If in your answer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth completely the , 1 text of such other reference or attach a copy. 95 With respect to the answers to ACRS Question 6.17, set forth at l Pages 6.17-2 and 6.17-3 of Amendment No 7 to the PSAR, describe in detail each test and experiment specifying each fact, calculation and assumption v 6-15-95-1
A thereof, upon which is based such answer. In addition, for each described test and description, provide the following: (a) When, by whom and for whom these tests were perfor=ed (or will be performed); and s (b) The basis and procedure by which the results of said tests will be incorporated in the final design and operation procedure of the pro-posed Midland Units. If in your answer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth completely the text of each ; 1 such other reference or attach a copy. Answer The Babcock & Wilecx Topical Report EAW-10017 (Rev 1), " Stability and i compatibility of Sodium Thiosulfate Spray Solutiens," FROFRIETARY (a copy
)
of which has been made available to the intervenors) provides most of the detailed information requested in the above interrogatories. This report I presents the results of B&W research and development on the stability and i compatibility of sodium thiosulfate spray solutions. Section 6 of the report presents the results of h8 iodine retentien experiments. The re-sults show that the solution retains the icdine over vide parameter ranges. The solution retains the iodine for all parameter values within ! the envelope of conditicns associated with the postulated MHA. The solu-tion's iodine retention capabilities can be summarized as folicws: l 6-15-95-2 r , l
s 1 (1) as long as any thiosulfate is present, the iodine will be retained by the solution (even if large amounts of oxidizing agents are added). (2) Even attu all of the thiosulfate has been radiclytically oxidized, the solution still does not release iodine. (3) If large amounts of hydrogen peroxide are added to solutions in which all of the thiosulfate has been previously radiolytically ox-idized, only small amounts (<1/2%) of iodine would be volatilized, even if the solution was boiling and being sparged by air. Our conclusions regarding the thermal stability, radiation stability, msterials compatibility, and iodine retention capability of sedium thio-sulfate spray solution are presented in Section 2 of BAW-10017 (Rev 1) (a copy of this proprietary report has been made available to the inter-venors). 'Ihese conclusiens are substantiated by the material in Sections 3, 4, 5, 6 and 7. The requested conclusions can be s=marized as follows: l (1) Thermal Stability - The rate of thermal decc=pesition of sodium thiosulfate spray solutions exposed to the temperatures expected fo11cw-ing a LOCA is 2% of the thiosulfate per 100 hours. l (2) Radiation Stability - The rate of radiclytic decomposition of sodium thiosulfate spray solutions exposed to the radiation doses folicw- l ing a LOCA or !GIA is 0.45 10.02 molecules of thiosulfate per 100 e7. (At this rate about 1 3 x 10 rads are recuired to deplete the thicsulfate 6-15-95-3 1
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7: content of the solution. This would take about 160 days following an , MHA.) The radiaticn also causes the radiolysis of water which produces a net hydrogen generation rate of 0.3 I 0.1 molecules of hydrogen per 100 eV. (3) Materials Compatibility - Alkaline sodium thiosulfate spray solutions are compatible with the materials used in the reactor building, although some precautions are required when using aluminum, copper. or new (untested) protective coatings. (4) Iodine Retention - The absorbed iodine is retained by the sodium thiosulfate spray solution under all conditions within the parameter en-velope associated with a postulated MEA. (The mexi m release will be
<0.1% of the absorbed iodine.)
The information requested on the 5 items listed in answer to IRL Questien 6.17 was to be submitted prior to or during the operating license review. However, all of the testi:ug has been completed and most of the results have already been submitted. The detailed results for Items 1 and 5 have been incorporated into Revision 1 of BAW-10017, which was submitted in May 1970. Item 3, the testing of welded joints has been completed but 1 the results have not been submitted yet. The tests were performed as described in Section 7 of BAW-10017 (Rev 1), except that the specimens were cut from welded plates. The following welded specimens were tested: (1) Stainless steel - 30h welded to stainless steel - 30h (2) Stainless steel - 304 welded to stainless steel - 316 a (3) Stainless steel - 304 welded to Inconel - 600 6-15-95-4
m (4) Inconel - 718 welded to Inconel - 718 (5) Inconel - 600 welded to stainless steel SA-3023 (6) Carbon steel - A-106 welded to carbon steel - A-216 (7) Carbon steel - A-212 welded to carbon steel - A-212 (8) Carbon steel - A-283 welded to carbon steel - A-283 The results show no signs of specimen f ailure and the corrosion rates 1 were similar to those measured on unwelded specimens, l Item 4, the calculation of the solution concentration and pH as a function of time is an engineering detail which is performed during the final de- ) sign stage. Typical values can be seen on the attached figures. All of the research and development reported in BAW-lool 7 (Rev 1) was performed by the Babcock & Wilcox Con:pany during the years 1967 through 1970. All Babcock & Wilcox results which are generally applicable to all thio-sulfate spray systems and to all plants will be submitted as supplements to Topical Reports BAW-10017 (Rev 1) or BAW-10022 and will be incorporated into the Midland application by reference. Results which are specifically applicable to the Midland Units will be submitted as amendments to the Midland application. With respect to containment liner plate protective coating, reference is made to the answer to Interregatory No 110. .d 6-15-95-5
s These hydrogen evolution tests have been performed by an outside inde-pendent research institute, which in this case was Franklin Institute Research Laboratories, Philadelphia. At regular intervals samples of the test chamber atmosphere were with-drawn and analy::ed for 2H content. A hydrogen evolution rate was thus established and when converted to the actual containment atmosphere was found to be acceptable. Sa.mples of the chemical solution were also withdrawn from the chamber and the pH, which is considered the centrolling factor as far as effects on the chemical solution are concerned, was measured. The resulting pH at the conclusion of the test was 9 2 and considered unchanged within the accuracy of the measuring equipment. / 6-15-95-6
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7 With reference to the relied-upon experiments referred to at Page lhA-2 of the PSAR, state each fact, calculation and assumption which you rely upon to conclude that "5 percent removable iodine" is a conservative value for use in the MEA analysis. If your answer relies wholly upon the results of said experiments and not upon any independent analysis, state what re-view or analysis was made of said experiments, and under what conditions, to conclude that the parameters of and factors inhering in the relied-upon experiments will be the same in the operation of the Icdine Removal Spray System in an MHA. If your answer also relies upon independent analysis, state in detail such analysis. If in your answer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth completely the text of each such other reference or attach a copy.
- 8. State a definition of the term " dramatic" as it is used at Page lhA-2 of the PSAR. In connection with your answer, state each fact, calcula-
, tion and assumption, including but not limited to, all numerical values upon which you rely to support your assumption or cenclusien that the removal rate of methyl iodine "is not dramatic." If, after answering this interrogatory, it is still your conclusion that the removal rate of methyl iodine is not dramatic, then state what steps were taken or which you contemplate takin6 to i= prove the efficiency of the removal rate of methyl iodine and if none were taken or are contemplated, then also state whether, in your opinion, consideri:ug the safety of the public and its l l
health and welfare, it is acceptable to subject the population, given an l a MEA, to the consequences of a le:,s than efficient removal rate of methyl 7-8-1
- k. .
3 iodine as that term is used in the PSAR. If in your answer you :ake reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth completely the text of each such other reference or attach a copy. Answer Page li:A-2 of the PSAR, uses the words "5 percent organic iodine" and "5 percent nonremovable iodine" both of which could be stated more accurately by the phrase "5 percent of the initially airborne iodine is 3 present as slow-reacting forms." The apparent confusion in terminology occurred for several reascns: (1) Early experiments were performed with water and not with thiosulfate solutions, so the reaction rates were so i slew that the experimenters thought the slow-reacting forms were actually 1 nonremovable. (2) Most experimental work was done using May-pack sam-l 1 plers which the experimenters knew would absorb the organic iodide in l 1 the section containing a charcoal bed. Thus, they termed the fraction I of iodine on the charcoal bed the organic iodide fraction. More recent experimenta have shown that hypoiodeus acid (HOI) is a volatile form of iodine which is also trapped in the charcoal bed in May-pack samplers. The term " slow-reacting forms" of iodine is meant to include all the organic iod.ide and all the hypoiodous acid. 1 Secticn 3.3 of Babcock & Wilcox Topical Report BAW-10022, " Effectiveness of Sodium Thiosulfate Sprays for Iodine Re= oval," T30FRIETARY (a copy of s which has been made available to the intervenors) discusses the for=s 7-8-2 I
#D ,
w of airborne iodine and evaluates the available experimental data in re-lation with the FHA conditions. This evaluation concludes that, when thiosulfate sprays are used, a conservative value for the amount of iodine present as a slow-reacting form is 2% of the iodine initially airborne. Since hypoioduous acid -- which is a hydrolysis prcduct frem the reaction of iodine and water -- is not formed in thiosulfate solutions. the 2% slow-reacting iodine is essentially all organic iodide, primarily methyl iodide. Thus, the Midland MHA analysis which is based on 5% methyl iodide is a conservative analysis. The use of the term " dramatic" on Pa6e lhA-2 of the PSAR was meant to vividly imply " rapid." Section 5 of BAW-LOO 22 analyzes the results of 43 spray experiments performed in the Nuclear Safety Pilot Plant and in the Containment Systems Experiment. The results show that the removal rate of elemental iodine is generally about 100 times faster than the removal rate of methyl iodide. Section 6 of the BAW-10022 shows that in a typical reactor building the removal rate of elemental iodine is about 1000 times the methyl iodide removal rate. The reason for the differ-ence is that the reactor building is cooled and the pressure reduced much more quickly than in the spray tests. The removal rate of methyl iodide decreases with decreasing temperature, whereas the removal rate of elemental iodine increases. If the removal rate of methyl 16dide were increased, the removal rate of elemental iodine would also increase, thus the elemental iodine would still be removed much faster than the methyl iodide. The i=portant factor V is not the relative removal rate -- it is the resultant thyroid dose. With I I l 7-8-3 l [$'"
the present spray system and assuming 5% of the iodine instead of 2% is present ai methyl iodide, the two-hour thyroid dose is a factor of 3h below the guidelines (10 CFR 100). Thus, a conservative evaluation of the present system indicates that it is extremely effective for limiting the thyroid dose and for protecting the health and safety of the public.
^ .
i l l l 4 V 7-8 h J
3, 9 With respect to the state of the art of iodine removal spray system and the system you have selected for use in the proposed Midland Units, state whether such systems have ever been tested under laboratory con-ditions which would be equivalent to an MHA. If your answer is yes, then describe in detail such testing and its results, including within your answer each fact, calculation and assumption of such testing. In connection with your answer, state whether you rely upon any assumptions or facts which were first developed subsequent to March 23, 1962, and if you do so rely, state in detail each such fact and assumption. If in your answer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth completely the text of each m such other reference or attach a copy. i Answer The conditiens assumed by applicant in analyzing the NHA and the dif- ! ferences in assumptions from TID-lkS4h are set forth in the PSAR in Chapter 14 as amended (14.2.2 3 7 to 14.2.2.h.2). I Iodine removal spray systems have been tested in the. pressure, temperature, humidity, and iodine concentration conditicus which are postulated for the Maximum Rrpothetical Accident. A complete discussion of the h3
)
spray tests perfomed fn the Nuclear Safety Pilot Plant, and the Con-l tainment Systems Experiment is contained in Section 4 of Babcock & Wilcox
-3 9-1 9
topical report BAW-10022, " Effectiveness of Sodium Thiosulfate Sprays I for Iodine Removal," PROPRIETARY, a copy of which has been made avail-able to the intervenors. i t 1 1 t 1 I 4 4 l i I 5 4 9-2 r -r, - ,-- -- -- - e.yy.-- e ewy wy., .,n , 9 ,y-r 7 , ,, - - - . -,-,-., , *,. ,,s.-e- e. =
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- 10. State whether you are relying upon the efficiency of the Iodine Removal Spray System to permit siting on the proposed Midland Units in Midland in light of Technical Infor=ation Document ikS8h. In connection with your answer, state whether if any critical assu=ptions set forth in PSAR concerning your Iodine Removal Spray System are proven unsound or unfounded, then whether in your opinion you would be able to meet the siting criteria set forth in Technical Infor=ation Document ih88k. If in your answer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth completely the text of each such other reference or attach a copy.
i l I Answer 1 Efficient operation of the Iodine Removal Spray System is not needed to '
- .eet the site criteria stated in 10 CFR Part 100 using TID lh8hh assu=ptions. Nomal demonstrated efficiency of the Iodine Removal Spray System vill greatly reduce the iodine concentration to well below that '
l necessary to meet the guidelines of 10 CFR Part 100. The effectiveness of the Iodine Removal Spray System is based upon l proven principles and therefore there are no " critical assumptions" which affect the capability of the spray system to remove iodine during the MHA. 10-1 l l' I I i I l l
- .,-l
._ _ _ _ _ _ ___ _ a
- 11. With reference to the assumptions set forth in Technical Infor=ation Document 1Mkk only, state the geographic area of the exclusion area, low population zone and population of center distance required thereby for the proposed Midland Units. If in your answer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth completely the text of such other reference or attach a copy.
Answer As stated, this question makes reference to TE 1Mkh and asks the geo-graphic area " required thereby" for the exclusion area, low population zone t.nd the population center distance (although the " population of center distance" was specifically asked for). The stated purpose of TE 1M44 in to provide " reference information and guidance en procedures and basic assumptions whereby certain factors pertinent to reactor siting as set forth in Title 10 of Code of Federal Regulations Part 100 (10 CFR 100)( } can be used to calculate distance requirements for re-actor sites which are generally consistent with current siting practices." Therefore, there is obviously no intent of TE 1Mhk to " require" specific siting distances. In addition, the question asks for the " geographic area" ---- " required thereby," which area is not a consideration since the distance is the more pertinent parameter. i 1 Consumers Power Company has evaluated the extent to which it conforms to the intent, as stated, of taese documents. For the purposes of conserv-i
, atism, the exclusion area radius assumed for the MHA dose calculation was 11-1 l
l e
.s 500 meters, although due to physical and geographic limitations it ranges from about 1200 meters to about 2200 meters. The low population zone and the distance from population center, since they are related arithmetica11y by a factor of 1.33 can be considered together in that they provide for the ability to preclude long-term exposure to persons who might occupy a posi-tion near the plant. Insofar as either Consumers Power Company and Dow owns all property out to approximately 1200 meters and in most directions greater than one mile, the protection afforded by controlling the evacua-tion of the remaining area of importance is an asset to this site. In effect, our actual exclusion area and low population zone are larger than the low population zone radius taken credit for. Even with the conserva-tive radii (exclusion area: 500 meters, low population zone: 1600 meters) assumed, the TID lh8 4 type calculation resulted in our conformance with the guidelines.
l 1 b 11-2 o
14 State whether the design of the Iodine Removal Spray System is emplete as set forth in the PSAR or whether, in order to complete the design, you are contemplating, during the course of construction if a pemit is issued, pursuing research and development to improve upon or complete the design Iodine Removal Spray System. If your answer is that you are contemplating such further research and de-velopment, then answer the following: (a) The nature, character and specific details, including re-sults intended, of such research and development, specify-l ing the amount of funds to be allocated to such research and development, and the names and addresses of each per-son, fim and corporation who will participate in such research and development; (b) Why it is your opinion, if it is, that such research and developnent can adequately be carried out prior to the granting of the construction pe mit. If in your answer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth ccmpletely the text of such other reference or attach a copy. Answer The preliminary design of the Iodine Removal Spray System as set forth in the PSAR is ccmplete. (Section 6.2, PSAR) M 14-1
l w No further research and developnent is required to support the design 4 of this Iodine Removal Spray Syc .a. 1 The detailed design will be described in the FSAR. a l I + 14-2
i l
\
- 16. State whether your Iodine Removal Spray System was designed solely with respect to the limits of 10 CFR Part 20 or whether your Iodine Re-moval Spray System, if operating according to design, will result in releases of iodine at levels lower than 10 CFR Part 20 permits. If your answer is that your Iodine Removal Spo'v Syste'm will result in releases at levels lower than permitted by 10 CFR Part 20, then state whether your system is designed to that release of iodine is as low as engineeringly possible and if not, why not. Also in connection with your answer, state <
whether in the course of designing your Iodine Removal Spray System you attempted to design a system to prevent releases of iodine at a specific i level or at any level below the levels permitted by 10 CFR Part 20, whether you chose as your goal the prevention of release of iodine under all situations at " essentially zero." Answer The criteria used in the design of the system are the guidelines of 10 i CFR Part 100 which is applicable to postulated radioactive material re-leases under accident conditions. Even under the maximum hypothetical accident as stated in Technical Information Document Ih8hh, the system will effectively reduce the release of iodine from the containment at-mosphere to a level well below 10 CFR Part 100 limits. The iodine removal spray system of Midland Nuclear Units operates only under loss of coolant accident conditions. The system is not designed with respe:t to the limits ._/ 16-1 1 I 1
-._ ~.
s of 10 CFR Part 20 because 10 CFR Part 20 governs the radioactive material releases during plant normal operation. Therefore, the system is not re-quirei to be designed to reduce during accident the release of iodine to the levels lower than 10 CFR Part 20 limits. l l 16-2 1 e
- 17. With regard to the following two sentences set forth on Page 5 of the Midland ACRS letter dated June 18, 1970:
"Other problems related to large water reactors have been identified by the Regulatory Staff and the ACRS and cited in previous ACRS reports. The Cecnittee believes that resolution of these items should apply equally to the Midland Plant Units 1 & 2."
state the following: (a) Whether "other problems" are related to releases of iodine and/or an iodine removal system; (b) If there are such related problems, state each of them and with respect to each one state what steps you intend to take j or are taking to resolve these problems. Include within your answer whether it is possible to resolve such problems prior to the granting of a construction permit, and if it is l 1 so possible, then state why you have not done so; and ) (c) What "other problems" these sentences refer to, other than l problems related to releases of iodine and/or iodine re-moval system; and with respect to each such "other prob-lems" state, separately for each one, what steps you intend i to take or are taking to resolve such problems. Include l within your answer whether it is possible to resolve these i i problems prior to the granting of the construction pernit, and if it is so possible, then state why you have not dene so. If in your answer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth ccmpletely the text _/ of each such other reference or attach a copy. 17-1 l
3
.e Answer This paragraph from the Midland ACES letter was addressed in Sec-tion 13 9 and in Appendix C of the Sunanary of Atelication dated October 30, 1970. Copies of the Summary of Arplication have been made available to the intervenors.
The "other problems" addressed were:
- 1. Once-Through Steam Generator Tests
- 2. Control Rod Drive Line Tests 3 Self-Powered Detector Testa
- k. Thermal and Hydraulic Programs 5 Blowdc'en Forces on Internals
- 6. Fuel Rod Clad Failure
- 7. Xenon oscillations
- 8. Iodine Removal Spray 9 Internals vent valves Each of these items was the sub, Ject of an R&D Program conducted by the Babcock & Wilcox Company. The results of these investigations were re-ported to the DEL staff in various Babcock & Wilcox topical reports as referenced in Append!x C of the Summary of Application. Copies of the referenced topical reports have bee;t made available to Saginaw Inter- i venors.
I l l l 17-2 i 1 m -- . _ . - .
- 18. State each fact, calculations, and assumption which you rely upon
~ to support your assumption that particulate aerosols found in a Loss of Coolant Accident ("LOCA") will be rapidly removed from the reactor build-ing atmosphere. Include within your answer a definition of " rapidly re-moved," includinb :.:merical values with respect to time. If in your answer you make re',erence to other than textual (exclusive of footnote) matter in the PSAR, then set forth coc:pletely the text of each such other reference or attach a copy. Answer During a loss-of-coolant accident, no clad melting occurs, and thus no radioactive particulate aerosols are expected to be formed. However, if a small fraction of the released iodine is assumed to attach to particles and dust, these can be expected to be removed rapidly from the reactor building atmosphere. This is discussed in the PSAR, Appendix 14A, Pages 14A-2 & 3 (Attached) More recent experimental data on the removal of iodine aerosols is avail-able from tests in the Containment Systems Experiment conducted at Battelle - Northwest. This subject is discussed in Section 4.4 (Attached) of Babcock & Wilcox Topical Report BAW-10022, " Effectiveness of Sodium Thiosulfate Sprays for Iodine Removal," PROPRIETARY, a copy of which has been made available to the intervenors. 1 i l 18-1
' G E.
w As show in the attached table, the removal of iodine associated with aerosols is as rapid as elemental iodine removal. The term " rapidly removed" therefore, means removed about as rapidly as elemental iodine. 18-2 t /, .~ -
F ORGANIC ICDEIE The organic icdine consists primarily cf = ethyl icdide bu[ includes a eil, almost insignificant, fraction of other organic iodides . Experi= ental data obtaineti under a vide variety cf egnditiens en the a= cunt cf ' methyl iodide released frc= cverheated fuel are repprted in nu=ercus publica-tions. Icdine release experi=ents using irradiated Zirealcy-clad UO 2 fuel in a PWR accident enviren=ent shev lass than the 5':'ercent crganic iodine assu=eci in Section 14.2.2 3 7 Sixtests(13)wereper'or:edatBattelle-scrthwest Iaboratories. They fcund that 1 percent or ' ss of airborne iodine organic for=. Thirteen cther experi=ents vare perfor=ed in Englanti(vag 1%; inallthe l but two show less than' O.2 percent as =et /1 icdide. The highest result was 3 percent. There are a numbet of other xperiments reported in the literature which deal with the a= cunt of cenreact'j<e iodine released frc= overheate:1 fuel. So=e of these ex; sri =ents have cbserv d greater than 5 percent nenre=ovable iodine; hcVever, these experi=ents ere conducted under conditions that are not applicable to the PWR acciden* enviren=ent. It is en the basis of all the data above that we concluded tra 5 percent nonre=ovable icdine was a conserva-tive value for use in the acci at analysis. Spray tests at Oak Ridge Na ional I4beratory(I' ) and Battelle-Northwest
- Laboratory (16) have de=en rrted that alkaline sodium thicsulfate spray is ef-fective for re= oval of ~ thyl iodide. While the re= oval rate is not as dram tic
- as that for ele = ental ". dine, the = ethyl icdine re= oval rate is sufficient to nake a significant ra uction in the airborne iodine concentratica and thus in ,. the off-site doses.
Spray test A-12 n the Centain=ent Syste=s E:c;eri=ent (CSE)(1b de=custrated I that alkaline .tt.iosulfate spray. re=cves methyl icciine frc= a stea=-air enviren=ent with a half-life of 60 =inutes. Scaling this to Midland Units 1 and 2 predicts a methyl icclide half-life of 52 minutes. PARTICUIATE ICLETE A s=all fracticn of the iodine assu=ed to be released folleving the LCCA =ay attach to particles and dust to fe:: particulate aerosels. Several studies have shewn try , shculd the particulate aeresels be for=ed, they will be rapidly re-moved fre.,,_ the reacter building at=csphere. l Stinchec=be and Gold =ith(II' IO) have shevn that sub=1cren particles are re=cved 4 with 95 to 99 percent efficiency by cendensing stea=. They also shev that the ! ther=al and vapor pressure gradients, which exist in the condensin stes= envi- ! ren=ent, drive the sub=icren particles tcvard the eccler surfaces where they are re=cved frc= the at=caphere with the condensing water va;cr. Particles greater than C.2 =icron size are re=cved efficiently by i=; action with the reacter building spray; those particles less than 0.2 =icren size are not. However, the s= aller particles serve as condensation nuclei which grew until gravitational ae.d inertial forces result in rarid deposition cf thece rar-ticles.(lT,19) The absence of particles, after aging, de=cnstrates th.e effec-tiveness of this re=cval =echanis=. likewise, very s=all the:::al gradients act as driving forces which cause =ig-ation of particles to the spray drop surfaces and thereby enhance re= oval.(lI) lhA-2 Amend =ent ::c. 5 11/3/69 -l-
~ , , , .--s - _
\ High erste concentrations of small(40, 21partic'es are very unstable and rapidly agglo=-
into larger particles. These are effectively removed b:. i= pac-tion with spray drops, by washout from the condensing steam, and by settle =ent. As a result, nearly all particulates are expected to be re=oved from the re-actor building atmosphere. Experiments in the CSE( ) have de=onstrated rapid re= oval of particulates by spraying. These tests indicate re= oval rates in a condensing steam environ-ment - similar to that in the reactor building following an LCCA - of one third to one half as fast as for elemental iodine. HYPOIODCUS ACID Hypoiodous acid is a product of ele = ental iodine hydrolysis by alkaline queous solutions.(22) The addition of sodium thiosulfate to an alkaline sol' ion pre-cludes for=ation of hypoiodous acid in any significant a=ount. Sodi thio-sulfate reacts instantaneously with elemental iodine, reducing the .odine to iodide. Thus, no elemental iodine is available for hydrolysis to the hypo-iodcus form. _REFhRENCES (1) criffith," V., The Removal of Iodine From the Atmos ere by Sprays, AHSB(S)R h5, 1963i pP 7-12. i' (2) Cottrell, W. B., ORNL Nuclear Safety Researe and Development Program, Bimonthly Report for July-August 1967, CRNL M-1986, p 35 (3) Cottrell, W. BI, OPliL Nuclear Safety Re earch and Develop =ent Progra=, Bimonthly Report for-March-April 1968, RNL-TM-2230, p al. (4) Cottrell, W . B., OR7L Nuclear Safe y Research and Development Progra=, Bimonthly Report for May-June 196~, ORNL-TM-2283, pp 64-73 (5) Cottre n , m . B., ORNL Nuclear Safety Research and Develop =ent Progrs=, Bimonthly Report for July-Au tst 1968, ORNL-TM-2368, Section 3 5 (6) Cottren, W. B., ORNL N lear Safety Research and Develop =ent Progra=, Annual Report for 1967, FlIL-4228. , (7) Parsley, L. F. and T. nnreb, J. K., Re= oval of Iodine Vapor From Air and Steau-Air Atmospheres in the Nuclear Safety Pilot Plant by Use of Sprays, ORNL-h253, June 958. (8) Nuclear Safety Quarterly Report; November, Dece=ber 1967, January 1968; for Nuclear Technolo ,/ by Safety Branch the Staff of USAEC Division B of Battelle-Northwest, of.~4L-816. Reactor Development and (9) Nuclea Safety Quarterly Report; February, March, April 1968; for Nuclear Safety Branch of USAIC Division of Reactor Development and Technology, by ' the/ Staff of Battelle-Northwest, FlIWL-885 s
/
14A-3 Amend =ent No. 5 n/3/69 ...
- L .-
- 3. Slower iodine removal in subsequent spray periods after the .odine concentration is re-O duced to <17. of its initial value.
- 4. Only small concentration changes occur dur-ing recirculation periods using borax sprays.
Most of the CSE spray tests were performed with bora. spray solu-tions . In a few experiments thiculfate spray solutions were used in the final spray periqd. Although the thiosulfate spray di markedly in-crease the iodine removal rate (see Figures 4-8 and 9), the rate was relatively slow (27 to 50 minute half-life). The slow removal rate ob-served in these tests was due to the fact that the hree previous spray periods using borax solutions had already rem /oved practically all of
/
the elemental iodine, so that the thiosulfate spray was actually removing the HOI remaining after the borax sprays / . To evaluate the performance of, thiosulfate sprays in removing elemental iodine, thiosulfate sprays /were used ex:1usivel/ in experiment
/
A-12. The results are shown irvFigure 4-12. In this experiment the e sprays were started before the/ iodine.was injected. When the iodine injection was started, the /iodine was absorbed so rapidly that the peak concentration was only 1/100 of that observed without spraying. Within 40 minutes the DF was/almost 1000, and after about 1000 minutes it was
/
about 4 x 105 -two to three orders of magnitude better than any other
/
CSE spray experiment. The me /es of thiosulfate sprays can best be seen in Figure 4-13, which com res the results from two experiments (runs A-10 and A-12) that wer identical except that one used a borax spray and the other a thiosul ate spray. The thiosulfate spray reduced the peak concentration
/
by 2 extra decade, and within 1000 minutes it had reduced the iodine
- 9 ncentration to about 1/300 of that observed in the borax experiment.
- 4. 4. Removal of Iodine Sorbed on Aerosols Data on the effectiveness of sprays for removing iodine associated with airborne particles were obtained from I4 spray tests in the CSE.
The test conditions and the results are summarised in Table 4-5. The rapid removal of iodine aerosols and the high decontamination factors i obtained can also be seen in Figures 4-14 through 4-20,55'" which show ' l 4-7 Babcock & WilcoX
the concentration of iodine sorbed on aerosol particles as a function of time. Two conclusions are evident:
- 1. The iodine aerosols are removed rapidly (t i/a = 2 to 11 minutes). .
- 2. The amount of iodine that remains asso-ciated with airborne particles after rel-atively short spray pe riods is extremely small (DF = 1000).
Table 4-6 compares the half-lives and total DFs obtained for ele-
~
mental iodine with those obtained for iodine aerosols. In an air atmo-sphere, the initial half-life for removal of iodine aerosols is about twice that for elemental iodine; but at longer spray times or at any time in a [ steam-air atmosphere, the half-life for the aerosols is equal to or l' shorter than that for elemental iodine. In a steam-air atmosphere the DF for aerosols is always greater than the DF for elemental iodine
, (except in A-12-2, where the thiosulfate spray reduced the elemental iodine to a very low concentration, yielding a very high DF).
In an air atmosphere, the total DF obtained is smaller for the 4 iodine aerosols, but since iodine aerosols have a mach lower initial concentration, the concentration of elemental iodine and iodine aerosols L approach each other rapidly at a very low value. _ A possible explanation for the similar removal characteristics of elemental iodine and iodine aerosols is as follows: The iodine asso-ciated with airborne particles is not permanently sorbed but is reversibly sorbed by solid particles or liquid drops; thus, when the gas concentra-tion is depleted of elemental iodine, the iodine desorbs from the aerosol particles and is free to be absorbed by the spray solution.
- 4) 5. Removal of Me.thyl lo%de Fifteen methyl fodide r moval em riments have been performed' s <
i - t. J in'the NSPP',/and 21 have bee perfo rme in the CSE..
/
J se experi-ments utilized either sodium thiosulfate s ti containing a borax buf-fer, plain borax solution, or s riatios or modification of these two solutions. _Tw or the 15 NSPP experiments contained sodium thiosul
, and.7 of the 2'l CSE spray tests contained sodium thiosulfate, i
~
.l _'
4-8 Babcock 8. Wilcox
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L 1 . GlDmpari lAoniQff.Spr.a,yf Rgrpgyal i Eff,ggtiveneas; for-Elementa]ilodiba.4nd; Iodine Aerosols 4"S i Spra droll. Spr,4 y,. Elemental Iodine
, Sprays Vhs sell #va: sizei mmdi Spra y, flow. ae ros ols a t mo s phe re- te mpe. W in.p. rato .. ppm, time ._ min. iodine _
Te s t s - l.9 2.0 A\Iv11 $t' daw aff 74 7 VFr@ .' 49 tt _ 36 32 6.0 Xed 2' s t'(aw- a'i'r Ylv P2'l'@ 5@ 2 2.0 X - 7'-- i $ t'Ea m - a'i'r 2'4*f FIP& 49 iD 8.0 48'.S 30 21 X 2' S t'e a m- a'n't 122 WI TO' 6 770 50 3 0,64 0.7 A f Stea m-*f t 247 50 500 220 =162 A-10-2 Steam-air 245 1210 1210 50 400 134 <eo A 2 Sicam-air 246 1
19 State your understanding of the phrase " properly engineered system" as it is used at Page 14A-2 of the PSAR. Include within your answer each fact, calculation and assumption which you re.ly upon for your con-clusion that the Iodine Removal Spray System can be. classified as such a properly engineered system. If in your answer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth completely the text of each such other reference or attach a copy. Answer The Iodine Removal Spray System is a network of components organized to deliver the spray solution to the reactor building spray noz::les. A properly engineered system means not only will the system function =e-chanically properly, but also that the operating envelope of the system is known. A properly engineered system will do what it is designed to do, which means that sufficient knowledge is available, including the completion of any necessary research and development, so that all design parameters of the system vill be met under all expected operating con-ditions throughout the life of the system. Two PROPRIETARY Babcock & Wilcox topical reports have been made available to the intervenors. These are BAW-10017, Rev 1, " Stability and Compati-bility of Sodium Thiosulfate Spray Solutions," and BAW-10022, " Effective-ness of Sodium Thiosulfate Sprays for Iodine Removal." These reports, alon6 with the PSAR, Sections 6.2, 14.2.2 3 7, 14.2.2.4, and Appendix 14A, exhibit the experimental evidence and support the conclusion that the Iodine Removal Spray System is a properly engineered system. 19-1
~.
- 20. With respect to the second paragraph at Pase 3-66 of the PSAR, state:
(a) Each fact, calculation and assumption to support your conclusion that ". . . internal components failure will not occur." Include detailed reference to experiments and their results, if any, which you contend support said conclusion, and also include what basia you rely 1 upon to conclude that you can design a system to pre-vent intemal components failure, if your design in this regard is not yet final or complete; (b) What is the specific nature and :nagnitude of the " dynamic loadings" which you say will occur; (c) Describe in specific detail, including numerical values, the " oscillating differential pressure across the core" i which you say will result from dynamic loadings; (d) Give a definition of " detailed design period;" (e) State in specific detail the scope, outline and intended results of the evaluation contemplated and include within year answer whether such an evaluation should be made prior to the beginning of constructicn of the proposed Midland Units. If your answer is that such an evaluation is not a necessary safety precondition to construction, state why not and what'in detail you will do if during construction your evaluation proves adverse to safety. C 20-1
3 If in your answer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth completely the text of each such other reference or attach a copy. Answer BAW Topical Report 10008, Part 1, Rev 1, " Reactor Internals Stress and Deflection Due to Loss-of-Coolant Accident and Maximum Hypothetical Earthquake," (a copy of which has been made available to Saginaw inter-venors) is a detailed report of all phases of the analysis which has been performed on the reactor internals. Section 3 1 of the Topical Report specifies the nature and magnitudes of the loadings and de-scribes the oscillating differential pressure across the core. The stresses and deflection shown in the Topical Report are for both seismic and LOCA loadings. The controlling load is the LOCA loading and the seismic variation from site to site vill not affect the de-sign since the analysis was done for seismic conditions more severe than those assumed for the Midland site. Generally, the " detailed design period" refers to that period of time between the date when the construction permit is granted and the date when the operating pezuit is granted, since design features must be finalized in that period. V 20-2 s-t - - ,-. . __
.-.g ,,
- 21. With respect to your analysis of LOCA as set forth at Pase 3-66 of the PSAR, state each fact, calculation and assumption upon which you conclude that transient pressure oscillations are dampened out in approximately 0 5 second. Include within your answer all occur-rences, incidents and variables, which are controlling, as well as their time history and uncertainty. Also include within your answer a detailed statement of fuel clad maximum temperature, percentage expected fuel clad perforation and differential pressure fission product letaage, maximum bowing (ie, degree of fuel rod and/or con-trol rod distortion) and the variance of criticality which would occur durirg the approximately 0 5 second. Your answer and anal-ysis regarding the aferesaid core conditions should be weighted
. _ . with respect to each uncertainty, and each such uncertainty iden-tified in specific detail. If in your answer you make reference to other than textual (exclusive of footnote) matter in the P AC , then set forth completely the text of each such other reference or attach a copy.
- 26. With regard to a LOCA, state in detail the cequence of events which would be required for the timely insertien of the control rods, and include within your answer each centrollin6 factor and its uncertainty regarding such timely insertion. If in your an-swer you make reference to other than textual (exclusive of foot-note) matter in the PSAR, then set forth completely the text of i i
each such other reference and attach a copy of each such other reference. 21-26-1
l. Answer The referenced portion of the PSAR deals with the effect of the tran-sient II)CA pressure differential across the core on centrol rod in-sertion time. The analysis to determino fuel rod behavior during a loss-of-coolant accident assumes that the control rods are not used in shutdown of the core, and are not required during the time when pressure oscillations resulting from LOCA might occur. Thus, a minor delay in rod insertion time due to pressure oscillations is unimportant, and would not affect analysis results for breaks where significant pressure oscillations occur. The mmh e'm clad temperature and percenta6e of perforations is given in response to Interrogatory No 3k. There is no criterion for bowing of the fuel rod during LOCA, since ie effect of such bowing would be negligible in ccarparison to the other mechanisms which would affect clad integrity. PSAR Figure 14-33 (attached) shows the reactivity versus time (variance of criticality) for the maHmum possible double-ended break size of a 36-inch ID, hot leg pipe, when the reactor is at ultimate power. This reactivity versus time is based on the moderator density reactivity feedback effects with the maximum possible positive moderator co-efficient. During the first 0.5 second the maximum reactivity addi-tien is 0.1%Ak/k. The peak reactivity addition of 0.2% Ak/k occurs at 0.8 second; thereafter, the reactor is shut down due to continuing moderator density decrease. 21-26-2
3 Density ak ' 2 l n - O . m
- M *
.=
.mt
< g >
h -
~ - -
akTotal} Doppler ak a .
=:
0
-1 i 0.0 0,5 1.0 1,5 2.0 2.5 Time, sec.
REACTIVITY VERSUS TlWE FOR A 35 IN 10 DOUBLE ENDEO. HOT LEG PIPE RUPTURE 'T ULTIMATE POWER xlTHoui *:: q.-
, Fi gu re 14-3
- em .
. .. w-
e D
- 22. With respect to the pressure vessel safety analysis (Section k.3.1.1.1 of the PSAR), state each fact, calculation and assu=ption, other than your apparent total reliance upon the ASE III Code, upon which it is cencluded that the reactor vessel will naintain its integrity despite the potential for limited crack propagation due to themal shock. If in your answer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth completely the text of each such other ref-erence or attach a copy.
- 23. State each fact, calculation and assumption, including the criterien and design philosophy or design basis upon which you conclude that "a sys-tem to mitigate the consequences of a vessel failure due to the=al shock
's following a Loss of Coolant Accident is not justified." (PSAR Pages 4-15)
In addition: (a) If your answer is based in whole or in part upon historical precedent, identify- ' 1 (1) Each document which refers or relates to er decenstrates this precedent; (2) Each oral ccmmunication which refers or relates to or demonstrates this precedent and give, regarding each such ccamunication, the date and place thereof, the identity (by name, address, by whem e= ployed, with what l group or organication affiliated and for whom acting) of each person involved therein, and the cecplete substance of what was said by and to each person. 22-23-52-1
m (b) If your answer is based in whole or in part upon financial, econcmic or engineering factors, identify: (1) Each document which refers or relates to or demonstrates each such factor; (2) Each oral communication which refers or relates to or which demonstrates each such factor and give, regarding each such communication, the date and place thereof, the identity (by name, address, by whom employed, with what group or organization affiliated and for whom acting) of each person involved therein, and the complete substance of what was said by and to each person. s If in your answer you make reference to other than textual (exclusive of e footnote) matter in the PSAR, then set forth completely the text of each such other reference er attach a copy.
- 52. state in detail, including each fact, calculation and assumption thereof, the result of each probability study cade regarding rupture or failure in any form or mode of the proposed pressure vessel. If in your answer you make reference to other than textual (exclusive of fcctnote) matter in the PSAR, then set forth completely the text of each such other teference or attach a copy.
Answer The analysis of the structural integrity of the reactor vessel when sub-jected to a the2nal shock is reported in Topical Report BAU-lool 3. The 22-23-52-2 i
.G results of the investigation indicate that the maximum depth of crack penetration would be 16f,of the reactor vessel's wall thickness. This original design criteria represented what could happen if a reactor ves-sel wall containing a substantial surface flaw were subjected to a themal shock due to actuation of the ECCS at the end of its M-year lifetime. Recent data from the Heavy Sections Steel Technology program (see attach-ment) regarding material properties indicate that our conclusions are i conservative. Accordingly, a system to mitigate the consequences of a vessel rupture is censidered unnecessarf.
^m 1
l v 22-23-52-3 l
( .
)
Figure 5. Toughness Temperature Data (Nominal Case) { I40, I I I g I i l V. Leo D.I., Rgristsct 16-19 j l ' c..... L.... 0... il 1 u2a f l 4 rj 1 y 1,l
/
h'
- A . . ] 4
[ Invalid J j u i n a .o i .i s o }- . 1 Test (Yielded) j I f
.-.n 5 a0 g
) il 'i j .
i i Table #5 L J f
- 5. 7 x 1019 nyt Ileavy Section 7 3 (see ner y ' y Steel Technology '
j Program E ,o _
, Technical Report !
j ,l 0.s=iO'S,,vs)
# # E9 I j /
" / /
/ Referencell 40 -
WCAP 7561'
, Aug 1970
__ - sy ~
# 2x i O ' ' 1' 20 tu E
bx i io n s ,,, ,l 8 w O' ~
,3u0 ,200 E100 0 100 1 200' 300' 400'
[i Tturtaatuar, FI 500 n' o x, Toughness Temperature Data From Reference , Superimposed on Figure 5 - B&W 10018 l v _.
m 24 With respect to Page 4-15 of the PSAR, state what quantitative and calculated influence the fact of proximate population density has on engineering safety systems contained in or proposed to be contained in the proposed Midland Units. Include within your answer what population density would have to be present, in your opinion, for you to change the proposed site, and also include within your answer the differences, if any, between your term, " proximate population density" and each of the terms, " low population cone" and " population center distance" as they are used in Technical Information Document 14844. If in your answer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth cczupletely the text of each such other refer-f ence or attach a copy.
.~~ ,
Answer Specifically with respect to Page 4-15 of the PSAR, " proximate population density" as used therein refers to the fact that Midland had been identi-fied in P. A. Morris' letter to R. D. Allen dated March 28, 1969 as a site comparable in population to the Indian Point and Zion sites. Due to this circumstance, and only from a cualitative standpoint, a system to mitigate a postulated condition of thermal shock induced vessel failure following the MEA was considered for Midland in the same way as it had been in the cases of Indian Point and Zion. To provide a ecxnparable de-gree of protection from the postulated effects of thermal shcck to the C' 24-1 e h
q reactor vessel, provisions were added, as stated en Page h.15, to allow installation if conclusionary R&D programs such as the HSST pro 6 ram at OLTL demonstrate the need for such protection. The term " proximate population density" was used to acknowledge, qual-itatively, the presence of a population center (ie, city with a popula-tion greater than 25,000) nearby whereas TID 14844 uses the tez=s " low population zone" and " population center distance" in ozder to assign protective distances related to the IEA effects. n,
./
V 24-2 x L. m
]J
- 25. Regarding the design and performance of the pressure vessel as described in the PSAR Sections 3 2.2.1.'7, 4.3.1, 4.3 1.1, and 4.3 9.1.1, and the possibility of adding systems to assure continued core cooling, state in detail:
(a) Fcat infomation is expected to become "available in the future to demonstrate" the necessity of such a system? (b) What information, if available, do you consider would re-quire such a system? (c) Why, if such a system will add to the integrity or safety of the en61neering safeguards, you have failed to propose s the inclusion of such a system now? '.) (d) If you were required or future information made it desirable to include such a system, then how would you design such a system? and (e) What provisions are you adding to the building and systems designs to permit the addition, if desirable, of such a system? II in your answer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth completely the te..: of each such other reference or attach a copy. I 25-1 1
3
/
Answer The statement (quoted below) that is pertinent to this question is contained in Paragraph k.3.1.1.1 of the PSAR:
"Due to the proximate population density, however, provisions will be added to the building and systems designs to permit the possible future addition of a system to assure continued core cooling if infor=ation becomes available in the future to demonstrate that such a system is required."
The contingency mentioned in the above statement from the PSAR was not expected to occur and was made to make allowance for additional core cooling systems prior to finalizing calculations on the safety of this plant. Later infomation (see " Response" to Interrogatories No 22 &
- 23) has confined the conservatisms of the data that was the basis of the conclusion that "the addition of nozzles to the reactor coolant outlet piping no longer appears necessary": (Refer to Page 4.8-1&2 of Amendment No 5.) With this as background, the following responses are given to questions asked:
(a) None. (b) Further experimental information which conflicted with, rather than confirmed as conservative, our conclusions in BAW-10018. (c) Such a system does not add to integrity of the enEir.eered safeEuards, nor as indicated above, would it add to the safety of the plant. 25-2
(d) The system, if required, would be designed as described in the answer to AEC Question h.8, which is attached hereto. (e) To alla for installation of such a system, pr W.sions have been made for space for flooding tanks and piping. ( t l l
)
(; ; 25-3 l 9 ---> - - - . r-, ,#%., --w - - --, ,,- ,- y
k.8 Provide a preliminary design of the provisions that will be =ade to accommodate failure of the reactor pressure vessel folleving a less of coolant accident, as discussed in page 4-15 of the PSAR. Answer: As discussed in Page k-15, our evaluatica shows that this accident will not occur; however, considering the postulated failure of the reac*wr vessel following a loss-of-coolant accident, provisions vill be incorporated to allow for the addition of a future system which will provide for rapid and continued flooding of the reactor cavity up to at least the reactor coolant piping nozzles where the water vill overflow and return to the reactor building sump. Such a system as presently envisioned would.censist of the following:
- 1. Reactor cavity ficoding tanks in the reactor bu11dir4 along with associated injection piping and power operated valves (actuated by a combina-tion of core ficoding tank low level and the lev-pressure safety injection signal). These tanks, containing approxi=ately 7200 cu ft of borated water, vill fill the reactor cavity up to the overficv point.
- 2. A gravity drain line frem the lower end of the reac+wr building refueling <-anal, which will allow the water from the reactor building sprays collected within the refueling canal and upper building ficor drains t,o provide additional ficoding capability.
De cavity flooding tank valves will be opened following a sufficient time delay, which is provided to allev the control rec = operator to evaluate the cavity ficoding injection signal and prevent valve opening snould it be determined the signal was false or spurious. Annunciation of this signal vill be providad. If the operator does not override the actuating signal, the cavity will be filled to the reactor coolant piping no::les within 10 =inutes following the loss-of-ccolant accident. Centinued cavity ficoding vill be pro'vided by the low-pressure injec-tion system vater exiting thrcugh the vessel break. This pre 14mina~y design concept is based upon the folleving analysis:
- 1. The time of the rupture of the primary system pipe is censidered at T=0.
- 2. The rupture of the reactor vessel is assumed to occur at 10 minutes.
-This is the approximate time at which the =aximum ther=al stress occurs as pre-dicted by the analysis in 3AW 10018. It is judged that the thermal stress con-ditions analyzed in this report represent the worst case in terJ of crack propagation.
h.8-1 Amendment No. 5 11/3/69
-a ,
3 ne analysis of reactor vessel rupture in conjunct; n with the LOCA clearly indicates that the problem relating to steam bubble for=ation does not increase in severity as long as the reactor vessel cavity is full at the time the vessel failure occurs. The internals vent valves perform the same function of steam venting for this hypothetical vessel rupture accident as they do for a reactor coolant pipin6 cold leg break. The requirement for top flooding of the reactor core is not related to the vessel rupture accident if the cavity is filled by the time the rupture occurs. Therefore, the addi-tion of noc:les to the reactor coolant cutlet piping no longer appears neces-sary. -
- 4. To preclude the possibility of thermal shock during nor=al reagtor operation, a failure analysis of the ficoding system vill be performed to show that no single active ecmponent failure vill allev cooling water to inadvertently reach the hot primary systes cctal.
5 If the water level in the cavity reaches the cavity overflev point prior to the initiation of crack propagation and this level is maintained during the accident, acceptable fuel te=peratures can be maintained. These 9 ' te=peratures would be identical to the values currently presented in Section 14 for the LOCA situation. (.ts \_,,/ 4.8-2 Amend:ent 50. 5 11/3/69 y
- 27. With regard to section 14.1.2.8.2 of the PSAR, calculate in the same
- nanner and detail as required by Interrogatorf 13 the doses following the release of secondary system steam to the atmosphere. Your answer shculd be based upon the highest possible release of radionuclides and should also include the manner or method by which you conclude that such re-leases are the highest possible. If in your answer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth completely the text of each such other reference or attach a copy.
l l l Answer i l This interrogatory asks for doses. Interrogatory 13 deals with con-secuences of doses. However, No 13 refers to No 12, which does ask for doses, so this interrogatorf is answered as if it referred to No 12. The reactor building iodine removal and cooling systems have no effect on the dose from the " blackout" condition. 'Iherefore, the calculated doses from this occurrence remain unchanged from that presented in the referenced PSAR section. 1 i .J 27-1 i
,*I =
1 l 7
- 28. State why you have not included within the PSAR the design or pro-posed design to vent and contain releases of iodine to the atmosphere in the event of a blackout release as defined in Section 14.1.2.8.2 of the PSAR. If in your answer you =ake reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth completely the text of each such other reference or attach a copy.
Answer As stated in the referenced section of the PSAR, the quantity (of iodine) released during this short time (4 minutes while atmospheric dump valves are operating when blackout occurs), is s=all and it would be less than 0.11 MPC at the 500 meters exclusion distance. The description of this event and the bases for this iodine release . is contained in the referenced section of the PSAR with further refer-ence to Section 14.1.2.10 of the PSAR. 1 28-1 g __, ,i-,._ = , - ' ^ ^ -
- 29 With regard to section 14.1.2 9 2 of the PSAR, calculate in the same manner and detail as required by Interrogatory 12 the doses fol-lowing a steam line failure accident. Your answer should be based upon the highest possible release of radionuclides and should also in-clude the manner or method by which you conclude that such releases are the highest possible. If in your answer yoit make reference to other than textual (exclusive of footnote) matter in the PSAR, then set for% ccepletely the text of each such other reference or attach a eupy.
Answer 1 The reactor building iodine removal and cooling systems have no effect on the deses from a steam line failure outside the reactor building. Therefore, the calculated doses frem this occurrence remain unchanged from that in the referenced PSAR section, which is attached hereto. i
~
29-1
- 30. Describe in detail what systems, if any, are proposed to minimize i or prevent release of radionuclides to the atmosphere in the event of a steam line failure accident and in the event of a fuel handling ac-cident. If no systems are proposed, then state why not. If in your answer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth comp?.etely each such other reference and attach a copy of each such other reference.
Answer The steam line failure accident analysis is presented in FSAR Sec-tion II+.1.2 9 2, a copy of which is attached hereto. As the doses calculated therein are wel.1 within the guidelines of 10CFRIOO, no systems are proposed to =inimize or prevent release of steam to the atmosphere from a steam line failure accident. The ventilation exhaust from the fuel handling area is passed tFa ugh charcoal filters during fuel handling as described in Section 9 12.2.2 of the PSAR, a copy of which is attached hereto, thus minimizing the release of radionuclides to the atmosphere in the event of a fuel handling accident.
~
30-1 l
,_. _ - . - , w -- e
- c. No steam generator tube da=a6e vill occur due to the lecs of h3) secondary side pressure and resultant tenperature gracients.
- d. Doses will be within acceptable limita.
14.1.2 9.2 Analysis and Results . The rate of reactor system cooling fonoving a steam line break accident is a function of the area of the failure and the steam generator water inventcry avaihble for ecoling. The steam generator inventcry increases with pcVer level.
*he inventory at rated pcuer is h6,000 lb and decreases linearly to 20,000 lb at 15 percent power. The steam line break accident is performed at rated power with a minimum tripped rod worth of 3 3 percent ak/K. This tripped rod worth is based on a Doppler deficii; (100 to O percent pcver) of 15 percent t.k/k, a moderator deficit (15 te O percent power) of 0.8 percent ak/k, and a sub-critical margin of 1.0 percent ak/k, corresponding to end-of-life conditions.
The immediate effect of any steam line break accident is a reduction in steam pressure and a reduction in steam flew to the turbine. These effects initially cause the reactor centrol system to act to restore steam pressure and lead generation.
'Ihe steam line failure was analyzed for both a s=all break and the maximum break sice to determine the maxi =um cooling effects.
A steam line rupture of s=all area causes a relatively slev decrease in steam pressure. flev. This places a de=and en the control system for increased feed-vater h ,i eration. In addition, the turbine centrol valves vin open to =aintain pcuer gen-Increased feed-water flev causes the average reactor coolant tempera-ture to decrease, and the resulting te=perature errer calls for centrol red withdrawal. The limiting action in this condition is the 103 percent limit en pcwer de=and to the rod drive centrol system. If the =cderator te=perature coefficient of reactivity is slightly negative, the reactor pcVer win increase when the centrol temperature system reaches the pcVer de=and 11=1 because of centinuing decrease. pressure. The reactor win then trip en high reactor coolant system A reactor trip will initiate a reduction in the feed-water flew tc the steam generators, and cicsing of the turbine step valves and branch valves. With a large, negative, coderator temperature coefficient (-3,0 x 10 -4), (ak/k)/F, the reacter pove:- increases with decreasing average coolant tempera-ture. cent. This win cause control red insertion to limit reacter pcver to 103 per-
-Without autcmatic centrol, additional ecoling causes a high flux reacter trip causing the turbine stop valves and branch valves to close. (See Figures 10-1 and 10-2.) This action assures that the steam generator withcut the steam line break vin be isclated. In the event a break occurs between the steam generatcr and turbine step valves, the steam generatcr with the steam line failure continues to bicv down; however, the second steam generator win be isclated t.fter the branch valves and turbine stop valves are closed. Folleving
~
turbine trip the =ain feed-water vnives are closed after 16 s, but the start-up ' valves (representing a capacity of 7 5 percent of the total flew to each steam generator) control. fonov control system require =ents for steam generater mini =um level
) ,
A steam line failure of large area results in high steam flow with resulting rapid decrease in the reacter coolant system temperature and steam system pres-sure.
' The reactor trips on high flux causing the same turbine and valve action 4
4 14-17 N g..m m- wm ,y 4 y ~ - -- ---e-.
ac discussed above. The steam cenerators are designed to =aintain reactor ()- system integrity upon ics cf accondary-cide precsurn. Therefore, this acci-dent vill not lead to a reactor coolant system failure. Figure 14-19 shows the respcase of the reactor ecolant system for an assu=ed 36-in. double-ended rupture. Initially, both stea generators blev dcun until a high flux reactor trip cccurs (6 seconds after failure) and the turbine s:cp valves and branch valves close. The reacter ecolant cutlet te=perature of the isolated steam generator increases after the respective stop valves close as a result of pressure recovery and a reductica of feed-vater flev. The =ain feed-water valves are closed after 16 seconds, but the start-up feed-water valves re=ain open as stea= blevs thrcugh the turbine bypass valves. Steam centinues to blev through the turbine bypass valve until the isolated stea= generatcr reactor ecolant inlet te=perature decreases to a value corresponding to satu-rated te=perature at the turbine bypass valve pressure set point. After the turbine bypass valve closes, the start-up feed-water valves are closed and no
= ore cooling occurs thrcugh the isolated steam generator.
The steam generator with the assu=ed stea: line failure centinueu to blev devn until 26 s after the failure. The reactor ecolant cutlet te=perature then rises because the secondary ecolant in this steam generater is depleted except for feed-water flew through the start-up feed-water valve. The average core coolant temperature decreases about 30 F during the first 11 s. The rate cf ecoldown then decreases with the taximu= cooldown (37 F) cccurring 110 s after the failure. After this, the reactor ecolant syste= establishes an equilibriu where the decay heat of the reactor is essentially balanced by heat re=cval in
) ,, the steam generator with the assu=ed failure. During these events the =1xi=u=
ther=al pcver is 109 percent and the reacter re=ains approxi=2:ely 1.0 percent ak/k suberitical even when a =ini=us tripped rod verth cf 3 3 percent ak/k is assu=ed. The average tube and shell te=peratures have been calculated for the case where 15 percent feed-water flew is continued. During the first =inute felleving the ; l break, the =axi=us tube-to-shell te=perature difference - where the shell te:- perature is higher - a= cunts to less than 4 F (Figure 14-19a). The peak tube stress for assumed icng-term continuation of feed-vater supply to a steam generater with a stea= pipe break occurs about 67 =inutes after the assu=ed break (Figure 14-20). The peak tensile stress value is approxi=ately 20,000 psi,12,500 psi of which is due to the =axi=u= tube-to-shell te=perature i difference of 105 F and 7,350 psi is frem an assu=ed pressure differential cf ' 2200 psi. This is about 63 percent of the =aterial yield stress. d
'evever, for this postulated accident, pri=ary systes ecoldevn vculd have reduced stess gen-eratien to belev that required to drive the feed-vater turbine before the ti=e of this theoretical peak stress value.
The effect of a steam line rupture inside the reactor building has been evalu-ated by conservatively assu=ing an instantaneous release to the reacter building of the energy associated with this accident. The = ass and energy releases fcr the steam generatcr in this analysis are: 14- 18
-, - _ . _ . n c- ..w ~ - - ~- ---
(] Mass, 15 Enercy, Stu x 10' u Steam Generator 46,000 28.0 Feed-Water Flow (6-s Full Flow Plus Coastdown to 7 5% Flov
@ 16 s) ,
12,800 56 Reactor Coolsnt Systen Energy Transferred - 17.6
'ibtal 58,800 51.2 Based on the above, a single. steam generator release vould result in less than 15 psig pressure rise in the reactor building. This is well below reactor l building design pressure of 67 psig.
The environ = ental consequences from this accident are calculated by assu=ing that the nuclear unit has been operating with a 1-gps steam generator tube leak, and with a steam line break between the reactor building and a turbine stop valve. The reactor ecolant activity assumes prior operation with 1 per-cent failed fuel rods. With these assu=ptions, the steam generators contain a total of .o9 equivalent curies of iodine-131. It is further assu=ed that steam generator leakage continues fcr three hours before the nuclear unit can be cooled down and the leakage terminated. This additional leakage corresponds (#,) to 3.4 equivalent curies of iodine-131. The iodine is assu=ed to be released directly to the at=csphere where it mixes in the wake of the reactor building. With these assumptiens a total integrated dose to the thyroid at the exclusion distance of O'.38 Rem is obtained. The corresponding whole body dose due to krypton and xenon is 0.004 Re=. ' 14.1.2 9 3 Conclusiens This analysis has shown the reactor does trip cnd re=ain suberitical after an assumed steam line failure has occurred. During the resulting transient the max 1=u= ther=al power is 109 percent; there-fore, no fuel da= age vill occur. Also, it has been shown that the environ = ental doses are within acceptable 1 5 1ts. The maxi =um temperature differential that occurs in the steam generator does not produce excessive stresses and steas generator integrity is =aintained. i l p D l 14-19 Amendment No. Ik T/31/70
i w {}( ' Normally, the purge system is not in operation and the purge system isolation valves are closed. When access to the reactor building is desired, the purge fans are started and the isolation valves opened. Supply air is taken through an outside air intake, purge supply fan, roughing filter, and a heating coil if required. The purge supply air is distributed through the reactor building cooler ductwork distribution system into the reactor building. The purge air is exhausted by the purge exhaust fan through a roughing filter and a high efficiency filter to reduce airborne particulate. activity. The purge system discharge to the vent stack is monitored and alarned. The purge fans with associated filters and isolation valves are located outside the reactor build-ing. The purge system is provided with double automatic isolation vaives (or dampers) in both the supply and discharge ducts, These valves are normally closed and are opened only for.the purging operation. The isolation signal and controls are discussed in 5.1.5. Durin8 normal plant conditions, the hydrogen vent system is not in opera-tion. The system for each unit consists of an electric motor driven fan, a roughing filter, a high efficiency particulate air (EEPA) filter, an electric heater and a charcoal filter. Instrwnentation consists of devices , for the indication and recording in the control rocm of the airflow, radio- l activity level and hydrogen centent of the exhaust gases. , Reactor building hydrogen venting after a loss-of-coolant accident is ac-complished by drawing reactor building air frcm a point near the upper region of the reactor building through the filters. The electric heater =aintains [ ') the relative humidity of the air entering the charecal filter at 90% or less. From the fan the gases are discharged to the atmosphere through the main l plant exhaust fans (which need not operate) to the vent stack. Fan motor and electric heater pcVer is supplied trem nor=al and standby sources. The fan, heater and reactor building isolation valves are operable from the control rocm. 9 12.2.2 Auxiliary Building Ventilation l The auxiliary building is served by separate ventilation systems for the fuel ; handlin6 area, the radicactive equipment areas, the nonradioactive equip =ent areas, the offices, the control room, the switchgear rocms, and the cable l spreadic6 rocm. t Outside air is mixed with recirculated air from clean, nonradicactive areas . to conserve heat. Recirculated air is used in the centrol room, switchgear room and cable spreading reces, conference rcce and office areas. Ventilating air.for the fuel handling recm and radwaste areas is designed en a "once-throu6h" basis to control the direction of airflow and to direct all pcten-tially contaminated a'ir to an exhaust. system connected directly to the vent stack. DV l 9-3h , Amend =ent IIo . 2 ' 5/2e/69 .
- . _ _m, - ,
O h' Ventilating air from potentially centa=inated areas is exhausted through high efficiency filters to reduce airborne particulate activity and is =enitored before it is discharged frem the plant through the vent stack. In order that , the air frem these spaces is passed through the filters, the exhaust flev exceeds the supply flow, thereby insurics in-leakage, rather than cut-leakage, frca these spaces. In addition, the exhaust ducting frem the potentially contaminated areas is fitted with charecal filter elements that are bypassed during acr=al cperation of the plant. Upon detection of high activity levels in these areas, re=ote controlled da=pers in the exhaust ductics frem the applicable area are auto-matically aligned to direct the exhaust flow through the charecal filter ele =ent. This align =ent can also be effected by operator action in the control rocm. During spent ibel handling, the exhaust air frem the fuel pool area is pressed through the charecal filters. The exhaust fan =oters on exhaust ducting for these spaces are supplied frcm no: ::al and standby sources. 9 12.2 3 Turbine Building ventilation Turbine building ventilation is provided by means of several fresh air supply systems which conduct cooling air to motors and equipment. The air is re-circulated in cold weather to conserve heat. Outside air is provided in hot weather by means of vall fans with louvers which are located in the areas of maximum cooling loads. roof exhaust fans. Air is exhausted from the turbine building through h,a 1 Ventilation for the e=ergency diesel generater rooms, the intake structure and the service water building are all served by individual fresh air systems in the same manner as the turbine building systems. 9.12.2.h Statien Heating ' Tue station uses lov-pressure extraction steam for station heating. 1 l 9.12.2 5 System Reliability l Ventilatien system and equipment are designed in accordance with the recem-mended practices of the American Society of Heating, Refrigeraticn and Air-Conditioning Engineers Guide, the Air Moving and Conditiening Association and ) I the National Fire Protection Association. Redundant exhaust fans are provided for the potentially centaminated areas. ! 1 1 1 1 I (i ,
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9-3ka Amend =ent No. 6 12/26/69 1
I
- 31. With regard to Section 14.2.2.1.2 of the PSAR, calculate in the same manner and detail as required by Interrogatory 12 the doses fol-lowing a fuel-handling accident. Your answer should be based upon the highest possible release of radionuclides and should also include the manner or method by which you conclude that such releases are . highest possible.
If in your answer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth completely the text of each such other reference or attach a copy. Answer In regard to Interrogatory 12, the reactor building iodine removal and cooling systems have no effect on the dose from a fuel-handli.7g accident. The calculation for the dose from a fuel-handling accident which considers the maximum possible release of radionuclides is based on the following assumptions: 1. The fuel assembly which has been damaged has operated at 2cgt , twice the power level of an average fuel assembly.
- 2. A minimum reactor cool-down period of 24 hours.
- 3. All 208 rods in one assembly are assumed to have been da= aged.
4.* Ten percent of all the contained iodines and 20 percent of all the noble gases are released into the water. 5.* Only ninety percent of the iodines released are retained in the water.
*These conservative assu=ptio sn , used in the calculation, are based on those required by the AEC. PSAR Section 14.2.2.1.2 presents source terms which the applicant believes are justifiable. Therefore, the calculations pre-sented here are considered conservative.
31-1
O.
- 6. The anviliary building vent exhaust charcoal filter iodine removal efficiency is 90%.
7 3 A working breathing rate of 3.47 x 10' m73, Based on these conservative assumptions, the following results were obtained: Excliision Distance - 500 Meters X/Q3 Wh le B dy Thyroid Time (S/M ) Dose - Rem Dose - Rem 8-Hour 1.9 x 10-0 2.40 85.50 24-Hour 3.1 x 10-5 0.39 13.95 30-Day 1.1 x 10 -5 0.14 4.95 Low Population Zone - 1600 Meters X/Qq Whole Body Thyroid Time (S/M~) Dose - Rem Dose - Rem 8-Hour 7 5 x 10' O.95 33.80 24-Hour 1.3 x 10 -5 0.16 5.85 30-Day 0.3 x 10'5 0.04 1.35 31-2
q
- 32. With regard to the accident discussed in Section 14.2.2.2.1 of the PSAR, state each fact, calculation,and assu=ption upon which you con-clude that the maximum hole size resulting from a rod ejection is ap-proximately 1.75 inches. If in your answer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth completely the text of each such other reference or attach a copy.
Answer The value of 1.75 inches for the marimm hole size resulting from a rod ejection is a misprint. The correct value is 2.76 inches. The 2.76-inch value was used in the analysis of Section 14.2.2.2.1. The inside diameter (ID) of the reactor vessel control rod drive no::le is 2.76 inches. Assuming the lead screw assembly of the control rod drive assembly is ecmpletely removed from the control red drive no::le, the mav4-"= size hole would have a diameter of 2.76 inches. { i l i 'v
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- 33. With respect to a LOCA, state each fact, calculation and assumption of Test No 5 % (LCFT semiscale blevdown tests) upon which you rely to conclude that such test and its results suppcrt your conclusion as to the presumed effectiveness of the emergency core cooling system. In-clude within your answer sufficient detailed description of said test, its unknowns and uncertainties so that one can objectively deter.J.ne whether said test has any application to the proposed Mid W d Units, and also state whether any other tests were made or are plarmed to be made related to the purpose or purposes for which said Test :To 5% was made. If in your answer you make reference to other than textual (ex-clusive of footnote) matter in the PSAR, then set forth completely the text of each such other reference or attach a copy.
Answer Test No 5M of the LOFT semiscale blewdevn tests involved the rupture of a small pipe connected to a pressure vessel which had an initial pres-sure of 2150 psig. An orifice was inserted in the line to decrease the sise of the break. The rupture was initiated by pressurizing rupture disks inserted in the line. As stated on Page 14-36 of the pSAR " Test No SM . . . . . is a typical case for the blowdown through a small rupture area." The relationship that this test had to proving the ef- l fectiveness of the emergency core cooling system was to show that for small breaks, the FLASH code conservatively predicted a faster loss of coolant and a smaller mass remaining than that which was observed (see e attached PSAR Figure 14-31). The mass remaining deter =ines the amount of emergency coolant which must be injected to re-cover and cool the core. 33-1
/
( m. This AEC sponsored test series is still underway in Idaho and varicus vessel and piping systems have been used. 0 ' s 33-2 i
t. m. 100 80
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h 225 Measured s 20 ) l l 0 l 0 10 20 30 40 50 ' 60 Time. sec. PREDICTED PER CENT MASS REMAINING
-- VERSUS TIME - LOFT TEST NO. 546 Figure 14-31
- 34. With respect to a LOCA, state under what conditions control rod insertion could be prevented or delayed by higher than normal core pressure during blowdown. For the spectrum of possible delay times, determine:
(a) The -4== clad temperature; (b) The maximum fuel center line temperature; and (c) The percentage of clad perforations considering pressure (internal and ex+ernal) and also considering temperature factors, unless such tesqarature factors are set forth in (a) and (b) above. If in your answer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth completely the text s
, of each such other reference or attach a copy.
Answer
Reference:
BAW-10015, "Multinode Computer Code Analysis of Loss-of- . Coolant Accident," February 1971. (a) With respect to LOCA, all of the cases, both cold leg and hot leg, from the'14.1 ft rupture to the 2 ft rupture appearing in the multinode analysis (referenced above) assume a void shutdown with no rod insertion. If rods l had been inserted, there would have been essentially no difference in the pcwer versus time curve due to the rapid shutdown brought en by the formation of 34-1
N voids in the core. The analysis presented in Section 14 of the PSAR also assumed a void shutdcwn for the larger breaks with no rod insertion. For this analysis, a posi-tive moderator coefficient was used but, again, the rapid void fomation shut the reactor down. The maximum clad temperature calculated in the analysis was 2204 F for double-ended rupture of the 28-inch cold leg pipe. (b) The maximum fuel centerline temperature does not (in the above-referenced document) exceed the steady-state initial fuel centerline temperature. It should be noted that the maximum hot spot temperature occurs for initial end-of-life
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fuel temperatures with a -1.8 x 10 A k/k-F moderator co-efficient. (c) The percentage of clad perforations for the worst case will not exceed 72% of the pina using failure criteria appearing on Page 13 7.2.2-1 of Amendment No 5 to the Midland Plant PSAR and the LCCA analysis in the above-referenced document (EkW-LOO 15). V l 34-2 l 1 i i 1
/
35 Describe in detail.the following with respect to.the. ability of the emergency core cooling system to prevent all of the possible consequences l resulting from a LOCA: (a) What experiments with geometries, if any, representative of the reactor coolant system, have been conducted and i i how do they relate to analytical models and/or other 4 related experiments; (b) What experiments, if any, have been performed to provide ) j detailed information on fluid conditions within a geometry representative of the reactor coolant system; (c) What experimental data has been obtained precisely to. pre-dict heat transfer coefficients from parallel pin arraps that extend over the range of fuel-pin geometries and cool-ant conditions that exist during blowdewn; (d) What experimental data is there to prove that no significant delay of the fuel rod wetting process will occur with regard to the gravity core flooding system; and (e) What testing at high temperatures and at degenerated condi-tions have been conducted with respect to experimental veri-fication of core cooling techniques. Include within your answer the results and each fact, calculation and assumption of your analysis thereof, insofar as it relates to the pro-posed Midland Units, of each such experiment or experimental data. If in your answer you make reference to other than textual (exclusive of (; 35-229-1 i e
O footnote) matter in the PSAR, then set forth completely the text of each such other reference or attach a copy. 229 Describe in detail, stating each fact, calculation and assumption, what experimental verification supported by analysis you have obtained at
% all temperatures related to a LOCA to verify that the situation is con-trollable. If in your answer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth completely the text of each such other reference or attach a copy.
Answer A number of test programs have been run to obtain data for use in the
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evaluation of the emergency core cooling system. The priwy objective of the WR Full Length Einergency Cooling Heat Transfer (FLECHT) test program was to obtain experimental data for use in evaluating the heat transfer capabilities of pressurized water reactor emergency core cooling systems under simulated loss-of-coolant accident conditions. A second series of tests were run by Idaho Nuclear Corp (INC) to establish the effect of the shape of cladding bulging on the heat transfer rates measured in the FLECHT test. In the evaluation of the effectiveness of emergency core cooling, the decompression and cooling prior to the initiation of the emergency core cooling system must be considered. This has been examined in two test facilities: 35-229-2 l
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- 1. The smaiscale vessels and loop at the National Reactor Testing Statien in Idaho.
- 2. The Oak Ridge National Laboratory (ORNL) test loop.
The geareetry of the FLECHT test bundle was chosen to be representative of a typical pressurised water reactor fuel assembly. Tests were per-fonned using stainless steel and sircaloy. rods. The principal di:nen-sions are as follows:
- 1. Heated red length - 12 feet
- 2. Heater rod pitch - 0.563 inch, square pitch 3 Cladding outside diameter - 0.h22 inch
- h. Control red thimble diameter - 0 5h5 inch
- 5. Instrumentation tube diameter - 0.h63 inch
- 6. Rod bundle array - 7 x 7 and 10 x 10 The axial power distribution employed is shown on Figure 1. The heater rods were supported radially within the test bundle by eight "eggerate" grids. One grid is located at the bottcm of the heated length and one at the top. The remaining six are spaced at equal intervals between the top and bottom grids.
Transient heat transfer coefficients and cladding temperature behavior, at different radial test bundle locations and at various axial locations were investigated over the following range of parameters: Peak power density 0.69 to 1.h0 kW/ft Maximum initial cladding temperature 800 to 2320 F Flooding rate 0.6 to 18 in/sec 35-229-3 l l
G , Inlet subcooling 16 to 189 F System pressure 15 to 90 psia All parameters including the flooding rate, were selected to cover the full range of typical PWR conditions. In addition to the above listed parameters, the effects of borated cool-ant, simulated flow blockage, power decay rate, fallback of entrained liquid, and variable flow ficoding rates were investigated. The results can be summarized as follows: A. Constant Flooding Rate Test
- 1. Temperature rise and turnaround time decrease with in-creasig initial cladding temperature.
- 2. Temperature rise and turnaround time decrease with in-creasing inlet flow rate.
3 Temperature rise and turnaround time decrease slightly with increasing pressure.
- 4. Temperature rise and turnaround time increase with in-creasing power density. 3 1
5 Temperature rise and turnaround time increase with slightly I l increasing subcooling. l
- 6. The addition of 2000 ppm of boric acid into the coolant !
resulted in a higher heat transfer coefficient. l I 7 At all times after flood, the heat transfer coefficient l 1 decreased with elevation for low flooding rates of
. l 35-229 4
^ . i i
2in/seeorless. For higher flooding rates, the heat transfer coefficient near the 10 ft elevation exceeded that at 8 ft.
- 8. There is no effect of power decay rate, for the same initia.1 power level, on the heat transfer coefficient for the range of power decay rates tested.
- 3. Variable Floodirs Rate Test
- 1. For some cases during the low flooding rate portion of the test, a second temperature peak was observed; how-ever, the second peak temperature never exceeded the first peak temperature.
- 2. The " fallback" reduced the second peak midplane temper-atures by about 80 F.
C. Flow Blockage Test
- 1. All flow blockage configurations tested, at both 4 in/see and. ; in/see flooding rates, with 75 percent blockage uniform across the bundle or 100 percent blockage it. che sixteen central channels, result in lower midplane temperature rises than the unblocked case for the same condition.
2. The increased heat transfer effectiveness observed with i flow blockage is due to the atomization of large entrained vater droplets and increased turbulence caused by the flow blockage. l 35-229-5 3 G */
3 The increase in heat transfer effectiveness was observed to occur as close as one inch downstream of the sixteen 100 percent blocked channels; therefore, the flow redis-tribution occurs very rapidly downstream of the blockage. 4 There was a small increase in axial pressure drop, on the order of 0.1-0 3 pai for 1 in/see flooding rates, due to the flow blockage. The INC tests were in support of the PWR-FLECE program with the specific objective to detemine by more extensive investigation whether the results of the PWR-FLEC E flow blockage test are valid. The experimental equip-ment utilized consisted of nine electrically heated rods assembled into a 3 x 3 array and enclosed in a transparent tube. These heaters were identical in ccnstruction to those used in the PWR-FLECHT test except they used a 30-inch heated length. The test results led to the following conclusions:
- 1. Severe blockage in small pin arrays does not have serious effect on the capability of emergency coolant to cool the rods under simulated postblowdown loss-of-coolant accident conditions.
- 2. The geometry of the blockage in the blocked regien is not a significant factor in the cooling process.
3 The blocked arrays tended to result in smaller temperature rises. 35-229-6
* .
- 1
m, 4 The higner initial temperatures tended to produce smaller temperature rises. Thus, the more extensive testing of the effects of flow blockage per-formed by DiC does not change the conclusions reached frcm the FIICHT tests. The depressurization test loop at ORNL censisted of a 2igh-pressure closed loop and a low-pressure condenser. A schenstic diagram of the loop is given in Figure 2. The semiscale test loop at DiC consisted of a high-pressure loop which blows down to the atmosphere. A sche-matic diagram of the semiscale loop is given in Figure 3 A comparison of the ORNL test loop and the semiscale test loop with reactor condi-tions is given in Table 1. A number of cases have been run on both test loops. Both loops indicate that the heat transfer is exceptionally good during the first few seconds of blowdown, and would result in much lower pin temperatures at the start of reflooding than is calculated using our analytical methods. For ex- l ample, for the Semiscale Test Number 845, the predicted cladding temper-ature greatly exceeds that measured when we use the same modeling and analytical methods as used on the Midland Plant. The difference is about 100 F at 0.2 sec, 200 F at 1 sec, 300C F at 2 sec, 5000 F at 4 sec, and 6000 F at 6 sec. l l , l V 35-229-7 l l
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') s Table 1 Comparison of Decompression Loop Conditions to Midland Reactor Coolant Loop Conditions Parameter CRNL Semiscale Midland Pressure, Psia 1515 2195 2200 Inlet Temperature, F 500 560 554 outlat Temperature, ? 515 590 603.8 Heated Pin'fength, In 24 9 144 Average Linear Heat Rate, kW/Ft
- 1. Before Rupture 13 12.0 5.66
- 2. After Rupture 13 (3 Sec), 12.0 (until variable 1.25 shut off Follows decay on temp) heat curve 7 ; Geanetry 1 Pin 121 Pins 177 Fuel assem-blies 208 Pins /assem-bly l
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- 36. With respect to a LOCA, what are the effects of a delayed addition of emergency cooling water to the core en the circalcy steam reacticn?
Include within your detailed answer a description cf each computer and experi:nental analysis, specif/ing each fact, calculation and assu=ption thereof, upon which ycu base your conclusion. If in your answer you make reference to other than textual (exclusive of footnote) =atter in the PSAR, then set fcrth ecmpletely the text of each such other refer-ence or attach a copy. Answer The design of the ECCS precludes delays which could result in exceeding the design criteria. To answer the hypothetical question, if a delay occurs the result could be higher core temperatures and a possible oc-currence of zircaloy steam reaction. To assure that significant delay does not occur, the ECCS ccmprises a passive core flooding system plus double the required high- and low-pressure pumped injection systems ' capacities with the systems, initiated by redundant and diverse in-strumentation and power supplies. Additional margin is provided by using conservative design calculational methods as described in Section 14.2.2 3.4 of the PSAR and in the Topical Report BAW-LOO 15 "M11tinode Computer Code Analysis of Loss-of-Coolant Accident." A detailed description of the computer codes and analytical 36-1 n
1
,m techniques used in the PSAR is presented in the proprietary "opical Report BAW-10023, "canputer codes and Methods used in Perfonning LOCA Analyses," which has.been furnished to Saginaw intervenors.
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- 37. Describe in detail what analysis and/or experiments, if any, in-cluding a description of their results and . specific application to the propcsed Midland Units, have been made with regard to the following:
(a) The temperature at which the Zircaloy-clad, UO "E*1l*t 2 fuel rods begin to collapse; (b) The effects of the fuel-clad interactions on the collapse process; (c) The effects of clad-steam reaction on the collapse process. If in your answer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth completely the text of each such other reference or attach a copy. Answer The fuel rod collapse tests are discussed in Section 3 3 3 3 1 of the PSAR and the responses to Interrogatories 56, 57 and 71. The tests show that for reactor operating conditions the rods are stable under external system pressure. Due to a combination of long-tem creep of the fuel cladding and fuel growth, the cladding becanes oval-shaped, and contacts the fuel pellets. This is a nomal fuel design condition and has been investigated in de-tail, as discussed in the PSAR. Clad collapse is of interest only during normal operating conditions and is not a design consideration during LOCA because the external pres-sure is reduced. 37-1 9
T
- 42. With respect to the radioactive waste analysis of Section 11 of the PSAR, give an esti:nate, specifying each fact, calculation and assumption thereof, during normal operation of the proposed Midland Units of the following:
(a) The quantity of each of the principal radionuclides expected to be released annually to unrestricted areis in liquid effluents; (b) The quantity of each of the prir.cipal radionuclides of the gases, halides and particulates expected to be released annu-ally to unrestricted areas in gaseous effluents; (c) The range of maH== potential annual radiation doses to in-tihriduals and suitable samples of population groups in Midland and surrounding population areas or centers resulting frem these releases; and (d) The quantitative percentage contribution to the total present background dose of the dose or doses set forth in (c) above. If in your answer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth completely the text of each such other reference or attach a copy. Answer (a) As noted in Amendment No 20 of the Midland PSAR, the Midland Plant will be employing 100% recycle of liquid wastes. As 42-1 r - _
, y _-
9 such, all liquid wastes, under no2 mal operating ccnditions, will be reused within the Midland Plant and no liquid wastes will be released to the en-vironment except laundrf wastes which are discussed more fully in response to Interrogator /126(h). (b) Under normal operating conditions, as noted in Table 1]-1 of the PSAR, 9,810 3ft of gases will be released to the er.vironment after thirty days holdup in the plant gas decay tanks. The primarf coolant gas con-centration given in Table 11-3 of the PSAR is assumed to be reduced by the standard decay equation (N = Noe ) using 30 days as the normal holdup time. This results in 68,000 curies of krypton-85, 2,500 curies of xenon-131m, 2.2 curies of xenen-133m and 33,000 curies of xenon-133 being released to the environment per year. This amount of radioactive gas represents 6.5% of AEC (lo CFR Part 20) limits at the site boundarf. No halides or particulates are expected to be released with the gaseous effluents. All gases will, however, pass through a roughing and HEPA filter for removal of particulates. (c) The actual dose received by anyone in the Midland area from the Midland Plant is expected to be near ::ero. If, however, someone were to stand at the site beundarf for one entire year, then that persen would recalvr a -%m additional dose of 32m rem from the Midland re-actor. \/ 42-2
R (d) As indicated above, the dose received from the Midland P' ant under nor:nal operating conditions is not expected to be above back-ground. However, under the worst case an additional 32m rem could be received compared to a background of 120m rem. This amounts to 27% increase but the total dose of 152m rem is still belev natural background levels in much tt the world. l i l l l 1 1 i Q/ h2-3 v- w r ,- q~
N, 43 With respect to the testing procedure described in the Answer to Question 6.10 in Amendment No 5 to the PSAR, state whether this testing procedure will be adequate for unknown variables over the entire plant life. If your answer is yes, state each category of variables, includ-ing their uncertainty which you have considered in your answer, as well as each fact, calculation and assumption which leads you to conclude affimatively. If in your answer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth completely the test of each such other reference or attach a copy. Answer i The referenced AEC question, the PSAR page referenced therein, ar;d PSAR, Section 6.2.4, all of which describe the preoperating and postoperating test program for the reactor building spray system, are attached hereto for reference. This program will establish that: (a) The system instrumentation and controls actuate the system. (b) ' Pumps start and valves are properly positioned. (c) Fluid is delivered at the design flow rate up to the reactor building isolation valves. (d) The spray headere and nozzles are not obstructed. h3-1
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6.10
Discuss the ability to test the containment spray systen, indicating provisions made in the design so that tests can be performed to ensure that the system is capable of delivering the spray solution through the spray no::les droplet size. at the proper ficv-rate, concentration of additive, and Answer:
- As indicated en Page 13-23 of Amend ent No. 4, all active ec=ponents of the reactor building spray systen vill be subjected to functional tests. The reactor building spray headers and spray no::les vill be subjected to airficv or smoke tests. Thus, as au ce=penents of the reactor building spray syste vill be rabjected to tests exc$pt for several pipe runs which represent only a small percent of the tota.'. line pressure drop, the spray syste= will have been demonstrated to be capable of providing the design flow to the reactor buildir4 spray no::les.
The spray no::les, Sprayco Model 1713A, have been extensively tested by ORNL(1) and have demonstrated their ability to generate iodine remo The centents of the sodium hydroxide and sodium thiosuHate storage tanks vill be rou,tinely sampled. Thus, as the system flev capability and the i amount of chemicals available vill have been de castrated, the spray additives at the spray no: les vill be at the correct concentration. (1) Parsley, L. F. , and Wantland, J. L. , " Spray Studies at the Nuclear S Pilot Plant," Ending Nuclear Safety Program Annual Progress Report for Period Dece=ber 31, 1968, onNL-h37h, June 1969 6.10-1 Amendment No. 5 n/3/69 m,
Table 1-2 (Centd)b 3
- h. Core Flceding Tanks Functional test for re=ote operation of stop valves, opening of check valves, and test of nitrogen cover-gas pressure control.
5 Reactor Building Air Recirculatien Functional test for initiation of valve operators, pu=ps, and fans. Cooling vater flow in each cooler to be verified.
- 6. Reactor Building Spray System
- a. Functional test for initiation and perfor=ance of valve operators and pu=ps.
- b. Operational test for initiation of spray pumps. Establish propor-tional rate of spray water delivery to spray headers (with return through test line). De=custrate flow through spray heads using air or smoke. Demonstrate operation with pu=p suction taken from bor-ated water storage tank.
T. Reactor Building Isolation 1 Functional test for closure of all valves on prescribed reactor buildin6 penetrations. Test to include operation of valves nor= ally open only for specified short-term functions such as in the purge syste= and the emergency su=p drain lines. Building leak rate at desi5n pressure to be established.
- 8. Integrated Engineered Safecuards Test Functional test to demonstrate sequencing of active. cocponents in h16h-pressure injection, low-pressure injection, service water cooling, reactor building air coolers, reactor building sprays, and reactor build-ing air recirculation and ecolin6 in response to actual or simulated signals.
Emergency power and Diesel Leading Test 9 Verify capability of nuclear service buses, power sources, including diesels to start and support safeguards loads. Detailed syste= test require =ents will be prepared for all en61neered safety syste=s when the system designs are finali::ed and the specific system components are identified, and proper reccrds vill be =aintained for all tests.
. ~
13-23 Amendment No. 4 9/26/69
--s_
4 6.2.4 TESTS AND UTS?ECTIONS Components of the reactor building spray system are tested on a regular schedule as follows: Reactor Building These pu=ps are tested singly by closing the Spray Pumps spray header valves and opening the valves in the test line. Each pu=p in turn vill be started by operator action and checked for flow g by recirculating to the storage tank. This test will also verify flow through each of the borated water storage tank outlet valves. Sodius Thiosulfate These pu=ps are tested by recirculating to their Pumps and Sodium respective s1!orage tanks. Hydroxide Pu=ps Borated Water Storage These nor= ally open valves are cycled by re=ote Tank Outlet Valves operator action to insure isolation closure. Reactor Building Spray With the respective pu=ps shut dovn and the nor-Isolatica and Che=ical = ally locked open block valves upstream closed, Injection Valves these valves are each opened and closed by operator action. Spray Nozzles When the unit is shut down, air or s=oke is blevn ( through the test conaections with visual observa-
%j tions of the no::les.
Reactor Building Su=p These valves vill each be cycled open and closed Valves by operator action. t (Paragraph Deleted)
} Provisions e,re made for sampling the contents of the storage tanks to assure that chemical concentrations are within specified limits.
i l l s Amendment No. 2 5/28/69 Amendment No. 4 6-12 9/26/69 ! i
q 4k. With regard to the analysis of a LOCA and the effectiveness of the emergency core coolin6 system, state each fact, calculation and assump-tion upon which are based the use in the PSAR analysis of the ce=puter codes PRIT, SLUMP and FLASH and what uncertainties, if any, are inherent in their use in this analysis. If in your answer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth completely the text of each such other reference or attach a ecpy. Answer Sections 14.2.2.3.3 and 14.2.2.3.4 of the PSAR discuss the use of the above-named codes and the major assumptions which were made in perform-ing the analysis. A discussion of the heat transfer coefficients which were used during the accident is also presented. The proprietary Tepical Report BAW-10023 discusses these codes and shows typical input and cutput. Since the filing of the PSAR, a multinode analysis of the loss-of-coolant accident has been perfomed in answer to an ACES question en this sub-ject. This analysis, along with a descriptien of code revisiens, a list of basic assumptions and a tabulation of a list of uncertainties and cen-servatisms is presented in the Topical Report BAW-10015. C 44-1
l 1 l 1 b-45 With respect to Section 14.2.2 3 7 of the PSAR, state each fact, calculation and assumption upon which the values of Table 14-10a have been determined. If in your answer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth ecmpletely the text' of each such other reference or attach a copy. Answer Table 14-10a lists the fission product activity released to the re-actor building during a " Loss of Coolant Accident." It is assumed all fuel cladding fails releasing all of the plenum gap activity along with the activity contained in the reactor coolant during this accident. Therefore, the total activity released to the reactor building is the sum of the fission product inventory in the reactor coolant and fuel plenum gaps. Thecoolantactivityisbasedonthemaximumactivitylevel(uc/ml) listed in Table 11-3 of the PSAR vhich is attached to this submittal. These values are based on continucus operation du/ing the tair( cycle with one percent defective fuel elements assuming no defects during previous cycles. Multiplying the reactor coolant volume (11,800 ft3 ) by the specific isotopic reactor coolant activities in Table 11-3 (ue/ml) yields the total fission product inventory in the reactor coolant (curies). f"3 u 45-1
O The plenum gap activity in the fuel pins is calculated by using the escape rate coefficients shown in Table 11-2 of the PSAR and fission product yield rates for the specific isotopes. The fission yield rates along with decay constants, branching fractions, and absorption cross sections are incorporated in a digital ccmputer code which solves the differential equation for a five-member radioactive chain for buildup in the fuel matrix and the plenum gap. The discussion of " escape rate coefficients" in the PSAR is attached to this submittal along with Table 11-2 and its references as they appear in the PSAR. A table of fission yield fractions, decay constants, branching fractions,
.m '
and absorption cross sections is also attached to this submittal along with a list of its references. This table lists 209 nuclides that have been grouped into 70 decay chains, the longest of which has 5 chain members.
.~ '~
45-2 4
n The quantity of fission products released to the reacter coolant during
?
steady-state operation is based on the use of " escape rate coefficients" (see4) as determined from experiments involving purposely defected fuel elements. See Reference 1-k. Values of the escape rate coefficients used in the cal-culations are shown in Table 11-2. The calculation of the activity released frem the fuel was perfomed with a digital computer code which solves the differential equation for a five-nenber radioactive chain for buildup in the fuel, release to the coolant, removal from tne coolant by purification and leakage, and collection on a resin or in a holdup tank. The activity levels in the reactor coolant for a nuclear unit containing one percent defective fuel during full (ultimate) power operation at the end of the third core cycle are shown in Table 11-?. Reactor coolant bleed is taken from the devnstream side of the purification demineralizers. It is assumed to have the same activity concentration as the reactor coolant reduced by the decontamination factor of the purification demineralizers. Gaseous activity is generated by the evolution of radioactive gases from the liquids as they are processed through the degassifiers and to a lesser extent as they are stored in tanks throughout the plant. The degassifiers and these tanks are vented to the gaseous radvaste system. The activity of the gases is dependent on the liquid activity. , Table 11-1 Radioactive Waste Quantities (Both Units) Waste Source Quantity ter year Assumutions and Ccmments Liqu3d Waste Reactor Coolant System Start-up Expansion 192,000 gal 4 cold start-ups (each unit) Start-up Dilution 298,000 gal 2 cold atert-ups at be s in-ning of life, and 1 cold start-up (each unit) at 100 and 200 full (ultimate) power days, respectively 147,000 gal 2 het start-ups (each unit) at peak xenon at 100 and 200 full power days, respectively Shia Bleed (1 Cycle) 53k,000 gal Dilution from 1,150 to 17 ppm System Drain (Refueling) 61,h00 gal Drain to level of outlet nozzles ~ _ System Drain (Maintenance) Sh,000 gal Drain 1 steam generator per unit 11-2 Amendment- No. 20 j- f ,. M*
s Table 11-2 Escape Pate Coefficients for Fission Product Release Element Escape Pate Ccefficient, see-1 1.0 x 10'I 1.0 x 10 ~I I# 2.0 x 10' 2.0 x 10" s 2.0 x 10" 2.0 x 10 k.0 x 10~9 4.0 x 10-9 2.0 x 10 -10 2.0 x 10 -10 1.0 x 10 -11 Ce and Cther Pare Earths 1.0 x 10 ~l g' j L
Table 11-3
Reactor Coolant Activity (Reactor Coolant Activity ror :.nc Tuird Core Cycle, Based on 11, Defective Fuel Elements)
(uc/ml) Time, M 1 Power Days lw 200 aco aco 290 300 310 Isotope Kr 85 m 15 ------------------------
- 15 Kr 85 98 9 '.7 6.3 45 33 2.1 1.1 Kr M .84 ----------------------- *
.8L Kr 88 27 ----------------------- +
27 ab 88 27 ----------------------- + 27 sr 89 .Ok1 .041 .040 . Oho .039 .038 .038 sr 90 .0029 .0033 .0034 .003h .0034 .0034 .Cc33 sr 91 .046 .046 .0k6 .046 .046 .045 .0h5 sr 92 .017 - - - - - - - - - - - - - - - - - - - - - - - -- .017 xe 13b 2.0 19 17 15 1.4 1.1 92 Xe 133m 27 27 2.6
- 25 2.4 23 2.1 xe 133 243 238 227 all 205 173 154 xe 135m 94 94 . 94 94 93 92 92 xe 135 56 56 56 55 55 54 53 xe 138 51 ----------------------- . 51 I 131 32 32 32 32 31 30 30
<~
I 132 h.7 47 4.6 4.4 4.2 39 36 I 133 38 38 38 38 37 36 3.6 I 134 50 ----------------------- + 50 I 135 27 27 27 27 2.6 2.6 2.6 cs 136 76 73 .66 58 51 .42 3h Cs 137 26 36 29 22 18 12 69 Cs 138 74 -----------------------
- 7L Mo 99 54 53 52 50 4.8 4.4 4.1 Ba 139 .081 ----------------------- + .081 Ba 1ho .065 .065 .065 .06k .063 .062 .061 IA 1k0 .021 .021 .021 .021 .020 .020 .019 Y 90 .26 57 75 .82 .84 .85 .85 Y 91 .18 .19 .15 .12 .10 .072 .0h9 Ce 14k .0027 .0027 .0027 .0027 .0027 .0027 .C025 BleedRateinvol/sec h.h(-8) 7 5 (-8) 15 (-7) 2.8 (-7) h.h (-7) 7 2 (-7) 9 8 (-7) t u-5 s/-
11.2 3 SOLID WASTES t I Two facilities are provided for processing solid vastes:
., A collection p!. ping system and storage tank received spent resins from the plant ion exchangers. These resins are transferred from the storsge tank into transportable containers for off-site disposal.
A miscellaneous solid vaste station receives all contaminated items such as spent filter ele =ents, rags, clothing, parts and materials. These vastes are compacted into a form suitable for off-site disposal. The shipment and off-site disposal of all radioactive vastes is per-formed by AEC licensed contractors. 11 3 DESIGN EVALUATICN The possibility of an accidental uncontrolled relesse of activity from the radvaste system is minimised by reuse of cuch of the liquid vastes. Liquid and stored gaseous vastes are sampled prior to discharge to the environ =ent. Radiation monitoring equi;=ent and process flov controls automatically alar = and terminate liquid and potentially high activity gaseous discharges to the environment, if the gross activities of the respective discharge strea=s exceed. values which would result in release concentrations in excess of 10 CFR 20 limits. Off-site disposal of solid vastes is accomplished by AEC licensed contractors. 11.h TEST AND INSPECTIONS All system components are tested prior to plant start-up. Operation of the system and equipment in the course of routine processire tasks demonstrates the integrity of those portions of 'the system. That equip =ent not nor: ally g operated. or operated infrequently will be tested periodically.
% 11 5 imM:ac:;dCES Frank, P. W., et al, Radioche=istry of Third PWR Fuel Material Test -
g X-1 Icop NRX~ Reactor, WAPD-TM-29, February 1957 g' ( }Eichenberg, J. D., et al, Effects of Irradiation on Eulk UO , WAPD-183, y October 1957 2 Allison, G. M. and Robertson, R. F. S., The Echavior of Fission Products y in Pressurized-Water Systems. A A Review of Defect Tects on UO2 Fuel Ele-ments at Chalk River, AECL-1338, 1961. Ny (4)Allison, G. M. and Roe, H. K., The Release of Fission Gaces & Iodines
- a. From Defected UO2 Fuel Elements of Different Iengths, AECL-22Co, June 1965
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B&il DIGITAL CODE DATA REFERE17CES e-1. 31cmeke, J. O. , and Todd, M. F. , " Uranium 235 Fissicn Production as a Function of Thermal Neutron Flux, Irradiation Time and Decay Time", CENL 2127, August
~
1957
- 2. Reacter Physics Constants, ANL-5800, July 1963.
- 3. Westinghouse Electric Corporation Periodic Chart, 1960.
4 Nucleanics, Vol.16, No. h, April 1958.
- 5. Nucleonics, Vol. 18, No. 11, P203, Nove=ber 1960.
- 6. Perkins, J. F.,
" Decay of U235 Fission Products", AD h15052, July 1963 i
7 ? Reviews of Modern Physics, Vol. 30, No. 2, Part II, April 1958.
- 8. ~
Neutron Crcss Secticns , 3NL325, Second Edition, Supple =ent No. 2, May 196h. 9
's
- 46. With respect to reactor building leakage during an MHA, as analyzed in the PSAR, state each fact, calculation and assumption upon which it is determined that leakage from the reactor building is assumed to be
.1 percent by weight of the free volume for the first 24-hour period and .05 percent for each 24-hour period beyond the first 2k-hour period following a loss of coolant accident. Your answer should include a con-sideration of all possible variables and uncertainties and an identifi-cation of each such variable and uncertainty. If in your answer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth ccmpletely the text of each such other refer-ence or attach a copy.
\
/
Answer The leakage rates assumed for the Midland Plant reactor building are based upon empirical data available from leak rate tests for similar pressure containing structures. The Palisades Plant report on Con-tainment Building Integrated Leak Rate Test provides a basis for con-fidence in the attainment of leak rate criteria similar to the specified leak rates for Midland. The test measurement showed a leak rate of about 7'% of the adjusted allowable leak rate (0.072 percent) at design pressure and about ko% of the adjusted allowable leak rate (0.0514 per-cent) for a test pressure equal to one-half of the design pressure. l k6-1
~-
M
7 t Significant variables in the actual leak rate are pressure and temper-ature inside the containment and the size of the leak path through the containment boundary. Additional features which are included in the Midland plant leak barrier system but not included in most other containments are: leak chase sys-tems at all liner plate seam welds; double penetrations at all piping, electrical, and access lock liner plate penetrations; and tne capabi 17/ to pressurize these penetrations above the design accident pressure and seal water and air in or between certain isclation valves. The double penetrations and leak chase design provide a continuous moni-toring capability for leakage in these areas. Further, due to the con-
.. tinuous pressurization any leakage at these areas would be noncentaminated air leaking into the reactor building and would provide a positive measure against outward leakage of contaminated gas due to a postulated accident pressure.
t l 46-2
.------v
(
- 47. Set forth in detail each fact, calculation and assumption by which Figure 14-64 of the PSAR was obtained. If in your answer you :nake ref-erence to other than textual (exclusive of footnote) matter in the PSAR, then set forth completely the text of each such other reference or attach a copy.
Answer Figure 14-64 of the Midland Plant PSAR is a plot of the integrated direct
.W iation dose versus distance, integrated for various time periods fol-lowing an MHA.
The :nethod of calculation is to assume a TID-lk8hh release of fission products to the reactor building following an MHA, ccmpute the direct dose rate by the volumetric source method through the reactor building to the whole body, integrate dose per isotope, sum the integrated doses from all isotopes and plot the sum versus distance. The ana ksis assumpticus are the same as those given in the PSAR, Section 14.2.2.4.1.
,8 47-1
m t
- 48. With respect to Section ik.2.2.4.2 of the PSAR, set forth each fact, assumption or procedure, if any, which differs from the facts, assump-tions or procedures outlined in Technical Information Document 1h8hk and new and upon what basis you support or intend to support the use of or reliance upon such different facts, assumptions or procedures. If in your answer you make reference to other than textual (exclusive of foot-note) matter in the PSAR, then set forth completely the text of each such other reference or attach a copy.
Answer
,~, In Section 14.2.2.4.2, the off-site doses resulting frem MHA are calculated based upon a set"of efaumptions. Most assumptions used in that calcula-tion are the samd as that outlined in Technical Information Document ik8hh (TIDlb8 W) except the site meteorology, the containment leakage rate and the iodine removal spray.-
The meteorology assumed in TID 1h8 % is Pasquill's Type F and one meter per second wind speed. This meteorology is used in TID 148kh for the en-tire period of the accident (0-30 days) with the assumption that the wind blows constantly in one direction. This meteorology is not the site me-teorology for Midland Plant and is too conservative to be used for the siting evaluation of Midland Nuclear Units because the stable diffusion regime of Pasquill Type F with one meter per second wind speed cannot invariantly last for a long period of time and the wind direction of blow-ing usually changes. 45-1
O The meteorology used in Section 14.2.2.h.2 of the PSAR is based upon the weather data measured in Dow Chemical Plant which is adjacent to the Midland Plant site and the data measured in Saginaw Tri-City Airport which is about eight miles southeast of the Midland Plant site. It is generally believed that the reliable meteorological information for site evaluation should be based upon the weather data taken from a station which is as close as possible to the proposed plant site. Therefore, the meteorology used in Section 14.2.2.4.2 should be more suitable than the meteorology assumed in TElh84 to the evaluation of the Midland Plant site. A constant containment leakage rate of 0.1% per day is assumed in TIDlkS W for the entire period of the accident. Because the containment pressure decreases as the time after MHA increases, it is more realistic to account for the fact that the containment leakage rate decreases as the contain-ment pressure decays. For Midland Nuclear Units, a conservative approach is made to assume that the containment leakage rate is the rate of 0.1% per day at the design peak pressure for the first 2h hours and then drops to and remains at 0.05% per day for the remainder of the accident. No iodine removal spray system is assumed in TID 1h8hh calculation. For Midland Nuclear Units, however, the containment spray is added with the chemical additives of sodium hydroxide and sodium thiosulfate which will effectively remove the elemental and organic forms of iodine from the con-tainment atmosphere. In addition, the direct gnen dose calculated in TIDlh8M for the off-site h8-2
'D ,
distance is obtained by assuming that no gammas will be attenuated by the containment structure wall, For Midland Units, the containment J building has a three and one-half foot concrete wall which will sub-stantially attenuate the gammas emitting from the containment to the outside atmosphere. t a
',, /
M-3 ( ~i .
m i 49 With respect to each condition affecting the integrity of each part of the primary coolant system in the event of a LOCA, state the following: (a) The predicted design for: (1) Static loads; (2) Dynamic loads; (3) Cyclic loads; and (b) The applicable stress vs frequency curves. State if your answer to any of the above would be different after each successive ten-year period of operation of the proposed Midland Units, if a permit issues, throughout a forty-year period. If in your answer 4 you make reference to other than textual (exclusive of footnote) =atter in the PSAR, then set forth completely the text of each such other ref-() erence or attach a copy. Answer It is assumed that the request for " predicted design" loads ~efers to design loads. Both predicted and design loads, however, are discussed in the Midland Plant PSAR. Section 4.1 of the PSAR defines the appli-cable codes used in the design of each part of the Reactor Coolant System. Sections 4.1.2.3, 4.1.2.4, and 4.1.2.5 of the PSAR define the basis of the design for Reaction Leads, Seismic Loads and Cyclic Loads respectively. The stress vs frequency tables used are those found in Section III of the ASME Code. Table 4-9 of the PSAR lists the number of cycles to be used l t . i (~/' 49-1
in the design and the estimated number of actual cycles expected during the 40-year lifetime of the plant. Appendix A of Topical Report BAW-10008, Part 1, " Reactor Internals Stress and Deflection Due to Loss-of-Coolant Accident and Muimum Hypothetical r,2thquake," defines four loading cases for reactor internals which are also applicable to the re-actor coolar., system. There would be no difference in the answer to this question after each successive ten-year period, because the ent'.re plant is designed to the worst conditions regardless of when they occur in life. s I (
\,.
49-2 i
.: . 1
p
- 50. With respect to the cold leg rupture for rupture si::es dcun to ap-proximately 4 in (0.087 ft ) in diameter, state each analysis, including each fact, calculation and assumption thereof, that " chow that the syste=
pressure will decrease below 600 psig . . ." (PSAR Pages 14-51), so that the core flooding tanks will begin discharging into the reactor vessel. If in your answer you make reference to other than textual (exclusive of footnote) catter in the PSAR, then set forth completely the text of each such other reference or attach a copy. Answer With respect to the cold leg rupture sizes down to appronizately 4 inches (0.087 ft ) in diameter, Figure 1 shows the pressure versus time for a 9.6-inch' diameter ruhture, and Figure 2 shows the pressure versus time for the k-inch diameter rupture. The assumptions used are listed in T61e 1. A multinode analysis was perfor=ed for these calculations using the modi-fied code explained in EAW-10015 but with fewer control volu=es. l Table 1 Basic Assumotions l
)
The basic assumptions used in the analysis are as follows:
- 1. The total energy output of the plant before the accident is 2568 MW 0 t (2.43L6 x 10 stu/s-) .
- 2. The inventory and the enerEy of the primary water are based on 0
a hot leg temperature of 603.2 F and a cold leg temperature of 554.6 F. 1 50-1 1 e' e
~T
- 3. The water volume in the pressurizer before the accident is 1150 ft 3 (high-water level).
4 The initial bulk coolant flow is 36,529.4 lb/s. 5 The initial pressures in each node were calculated assu=ing the pressure at the outlet of the core to be 2200 psia. Mcmentu= pressure losses and gains around the loop were neglected.
- 6. All ruptures open instantaneously. -
- 7. The discharge coefficient (C #* "E ** * *
"* "8 D
Moody leak flows. This is based on correlations between LOFT semiscale test data and predictions made using the Era version of the FLASH-2 code.
- 8. Injection water is supplied by only one high-pressure and one low-pressure pump. No core flood tanks are used.
9 The reactor trips at a reactor coolant system pressure of 2064.7 psia,
- 10. Rod shutdown occurs on trip after a 0.5 second time delay.
- 11. Heat transfer between the pricar/ and secondary sides of the steam generator is a function of the te=perature difference and the primary side mixture height.
- 12. The reactor coolant pu=ps and the turbine trip on reactor trip.
- 13. Auxiliary feedwater begins 36 seconds after the turbine trips.
14 Pricary metal heat was included in the analysis.
- 15. The hot spot analysis utili::ed a hot channel in the modified FLASH-2 code. For this channel, the peak-to-average peaking factor was 3.144, which resulted in a hot spot linear heat rate
'.J
$0-2
n .o of 18.3 W/ft. The initial fuel temperature was 3141 F and the initial cladding te=perature was 742 F.
- 16. The secondary side of the steam generator relief valve is set to open at a pressure of 1040 psig. When the reactor coolant system pressure reaches approximately 1055 psia, both the pri-mary and secondary sides of the steam generator are in equilib-rium. Consequently, heat can no longer be removed through the steam generators.
- 17. For all cases except the high-pressure injection line break, the leak occurs in the 28-inch ID reactor coolant pipe at the bottom of the steam generator.
50-3 l
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(. s Figu re 2 Reactor Coolant System Pressure Vs Time (4.0-Inch-Diamete r Rupture) 2400 2000 1600 ,
.2 E.
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- 51. With respect to the cold leg rupture for rupture sizes below 4 inches (0.087 ft ) state each analysis, including each fact, cal-culation and assumption thereof, upon which you conclude (PSAR Page 14-51) that it would be possible to hold above 600 psig for a " period of time." Include within your answer a definition of
" period of time" as you use that term. Also include within your answer in detail each fact, calculation and assumption which you contend verifies that both core heat transfer and coolant area and primary containment integrating will be maintained under the circumstances of a cold leg rupture for rupture sizes below h inches (0.087 ft ). If in your answer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth completely each such other reference or attach a copy.
Answer In addition to the 3 25 inch diameter rupture analysis described on PSAR Pages 14-51 and 14-52, two-inch and one-inch diameter rupture analyses have been completed. The two-inch diameter (.022 ft ) rupture in the cold leg results in system pressure versus time as shown on Figure 1. If ne operator action is taken, the system pressure will reach 200 psig approximately 5-1/2 hours after rup-ture. The water height in the vessel remains above the nozzles at all times during this transient. 51-1
, w- -w . .
2 The one-inch diameter (.0055 ft ) pressure history is shown on Figure 2. In this case, at approxi:nately 800 seconds, the injection rate is greater than the leak rate causing the reactor coolant system pressure to increase to a quasi-steady state value dependent upon the high-pressure injection pump characteristics. Since the plant will remain subcooled during this accident, controlled shutdown would proceed with natural circulation. In proceeding with an orderly shutdown, the reactor coolant system pres-sure would remain above 600 psig on the order of three hours. The as-sumptions and methods used in these analyses are identical to those listed in the answer to Interrogatory No 50. The resultant containment pressure will be maintained below containment design pressure by the engineered safeguards systems. In PSAR, Sec-tion 1!4.2.2 3.6, it is shown that for the spectrum of break sizes analyzed, the maH== containment pressure results from a reactor coolant system ; 2 rupture of 5.0 ft and the resultant peak pressures to be less than this value for the smaller break sizes. From this comparison it can be l concluded that for the break sizes in question, containment integrity will be maintained. ~ 51-2
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Figure 2 Reactor Coolant System Pressure Vs Time (1.0-Inch-Diameter Rupture) 2400 1 2000 -
~
1600
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2. 1200 -
? l De 800 400 D1
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Q Time, seconds x
q 53 With respect to the memum thermal output (kw/ft) and the ==v4== fuel burnup (Mid/MrU), state in detail each experiment and analysis, including each fact, calculation and assumption thereof, which you con-tend supports the ability to achieve such output and burnup without ex-ceeding fuel integrity. Your answer should include, although should not be limited to, irradiation temperature and pressure effects for each static, dynamic and cyclic condition for Zirconium eladd47 , both as an individual instance and as an accumulative history. If in your answer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth coc:pletely the text of each such other ref-erence or attach a copy. Answer Interregatory No 53 addresses the general question of fuel integrity. Although it requests information specifically in terms of maximum thermal output and ==v4== fuel burnup, it also requests that the answer include other and more general aspects of c1=Ad4M design considerations. This response therefore discusses the general design basis to assure clad integi-ity, as well as the specific design areas questioned. The basic design limits to assure fuel integrity are given in Section 31.2 of the PSAR. Thermal, nuclear, and mechanical limits are all di-rectly related to fuel integrity. One of the primary purposes of estab-lishirg these limits is to assure that core cooling capabilities and 53-1
1 t %. mechanical design considerations are sufficient to assure fuel integ-rity. Section 3 of the PSAR also contains methods of analysis, results of analysis, and reference tc experimental work and operatin6 experience adequate to provide assurance of fuel integrity. Several of the most , I significant factors directly associated with f"c1 integrity are dis-cussed below. 1 l The basic design data used in assessing fuel perfor=ance is given in Table 3-1. This table includes irradiation temperatures (654 F maximum clad surface temperature, 4150 F maximum fuel temperature), and thermal output (16.83 W/ft ==*=).
~
The desien peak burnup is 55,000 wd/m (see Pages 3-1, 3-72, 3-78, 3-80 and 3-82). This is used in evaluating fuel integrity, even though the actual peak burnup expected is less than this value. Average burnup is 27,490 wd/m, (page 3-72); het red av-erage burnup is 38,000 wd/m, (Page 3-78); and peak local burnup is 42,000 wd/m, (Pa6e 3-54). Thus, a significant margin of safety ex-ists which may be utilized to achieve improved core performance in later designs. The heat generation behavior of the reactor (ie, the basis for fuel burnup) is determined by nuclear characteristics of the core (see Section 3 2.2). The heat removal capabilities are discussed in Section 3 2.2,
" Thermal and Hydraulic Design Evaluation."
The mechanical characteristics are discussed in Section 3.2.4.2, " Fuel Assemblies," and the design evaluation of cladding adequacy, (including m 53-2
I I consideration of Pressure Effects, Fuel Irradiation Grcwth and Fuel-Clad Differential Expansion, and Application of Experimental Data to Design Adequacy), is examined in 3 2.k.2.2 therein. Results of tests and expe-rience related to each of these factors is discussed in Section 3 3 I l Further clarification of specific questions related to fuel integrity 1 are contained in response to questions on clad strain (Interrogatories 58 and 83), fatigue (Interrogatories 54, 82, 84 and 218), clad collapse ! I (Interrogatories 56, 57 and 71), and burnup (Interrogatory 59). i 53-3 e
e Q 54 With respect to stress and strain design criterion for fuel assemblies at the most severe abnormal and the most *evere nor=al conditions, state for each such condition and for each component as well as for each of the integrated parts of that component, each instance in which such stress and strain will be relieved, ! in your opinion, by small deformations of the material of which 1 it is fabricated. Also state for each such instance the accumu-lated history of cyclic loadings. If in your answer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth completely the text of each such other reference or attach a copy. 1 i Answer 1 Within the fuel assembly, the fuel rod is the only ecmponent for which stresses are high enough that stress relief resulting in non-
- elastic strain is significant. As discussed in Section 3 1.2.4.2 l
of the PSAR, stresses classified in this category are: ! 1
- 1. Discontinuity stresses in the cladding near the fuel rod end caps,
- 2. Thermal stresses in the cladding due to temperature grad-ients through the cladding thickness, and 3 Fuel cladding stress, relieved by creep, resultira from fuel pellet-cladding differential thermal expansion and from fuel growth.
54-1
3 The evaluation of the cyclic occurrence of the above will be per-formed during the detail design phase to verify that the final fuel rod design requirements are met for the specific Midland operating cyclic characteristics. The results of this evaluation will be incorporated in the FSAR.
)
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1
- 55. With respect to the added effects of irradiation to clad stress levels, state each fact, calculation and assumption upon which you rely to support the statement at Pase 3-76 of the PSAR that a "3:1 margin en stress is more than enou6h to account for decreased stress rupture due to irradia-tion."
If in your answer you make reference to other than textual (exclu-sive of footnote) matter in the PSAR, then set forth completely the ter of each such other reference or attach a copy. Answer The quotation above is incorrect. The PSAR refers to " decreased stress rupture strength" rather than " decreased stress rupture." There has been no problem to date witn stress rupture causing fuel fail-ures in reactor service. Stress rupture is a creep-related phencmenen and the creep data shows that a 3:1 margin should prove a conservative estimate of the effect of irradiation on the stress-rupture properties of Zircaley. Attached are curves showing representative creep data. This data was taken from: (1) Williams, J. A., and Carter, J. W., Handout given at Sixth General Meeting on Irradiation Effects en Reactor Structural Materials, March 1966. (2) Fidleris, V., and Willia =s, C. D., " Influence of Neutron Irradiation on the Creep of Zircalcy-2 at 300 C," 55-1
Electrochemical Technology, Vol !+, Iso 5-6, May-June ic66. (3) Fidleris, V., "In-Reactor Creep of Zirconium Alloys," FD-31+-26. In addition, the 3300 psi internal pressure used as the d.esign condition, and used in for=ulating the 3:1 =arsin statement, is itself conservative, because'the actual rod internal pressure will be less than this design pressure and the margin correspondingly larger. Most fuel rods vill have an internal pressure less than system pressure throughout the core life-time, so that stress rupture is not a factor in their perfor ance. Only near the end-of-life would any of the rods see an internal pressure higher than system pressure, and then only for a short period of time at low-pres-sure differential. 55-2 A
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. ~ Tempera.ture, C 500 450 400 350 300 i i
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-2 l (l ,
20,000 (e6) and 30,000 ($) psi , AECL e4 = In Reactor, O = In Laboratory I; j BNh A = In Reactor, A = In Laboratory O \ ** AECL 2528 1 Constant strain y
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-8 500 C 450 C 400 C 350 C 300 C 250 C 10 ,, ,; ,; ,, , ; ,;
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FIG W 5.22. CREEP OF ZIRCALOY-2 r A
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O
- 56. Describe in detail each test mentioned at Page 3-79 of the PSAR which you contend demonstrates a clad collapsing pressure in excess of 4,000 psi at expansion void mmh m m temperature. Also state the fol-lowing:
(a) By whom and for whom these tests were performed and when they were idrfomed; (b) Each fact, calculation and assumption upon which it is con-tended that each of such tests apply to the Midland fuel cladding under maximum burnup and linear heat rate separately in a LOCA and in the most severe nomal operating conditions. Include within your answer how you arrive at your opinion as to the most severe operating condition. If in your answer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth completely the text of each such other reference or attach a copy.
- 57. Describe in detail each test or experiment mentioned at Page 3-79 of the PSAR which you contend demonstrates that the clad meets the long-time (creep-collapse) requirement. Also state the following:
(a) By whom and for whom these tests were performed and when they were performed; (b) Each fact, calculation and assumption upon which it is con-tended that each of such tests apply to the Midland fuel clad-ding under maximum burnup and linear heat rate separately in a LOCA'and in the most severe normal operating conditions. ~ 1 56-57-71-1 m
I Include within your answer how you arrive at your opinion as to the most severe operating condition. If in your answer you make reference to other than textual.(exclusive of footnote) matter in the PSAR, then set forth completely the text of each such other reference or attach a copy.
- 71. With respect to the fuel rod cladding tests described in Section 3 3.3.3.1 at Page 3-104 of the PSAR, state each fact, calculation and as-sumptionbywhichyouconcludethatthesetestsand/orexperimentsrelate and are applicable to the proposed Midland Units. If in your answer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth completely the text of each such other reference 3 or attach a copy.
Answer A description of short time collapse tests which demonstrate a clad col-lapsing pressure in excess of !+,000 psi at expansion void ~ hum tempera-ture is given on Page 3-104, Paragraph 3.3.3.3.l', " Fuel Rod Cladding" of the Midland PSAR. The clad short time collapse tests were performed by B8M for B&W in the time span 1962-1963 The collapse tests were conducted on Zircaloy-4 tubing at temperatures representative of expansion void maximum temperature. Referring to part (b) of Interrogatory 56, there is no requirement to design against clad collapse during a LOCA. During a LOCA, system pressure will drop, thus the pressure differential across ._/ the clad wall will cause tensile stress in cladding. 56-57-71 s.
r% ; , The design requirement for clad creep collaese is that the clad must not collapse (must be free standing or have a(. equate support) on a long-term basis. In the active fuel region, this requirement is met, since the fuel will support the cladding. A series of collapse tests were conducted by BW for BE in 1963 to de-termine whether the end void region of the fuel rod would meet the long-time free standing require' ant. The results of these tests are given on Page 3-104, Paragraph 3.3.3.3.1, " Fuel Rod Cladding," Midland Plant P Ac . The above tests were followed by additional creep-collapse tests conducted by BW for BW in 1968. In the tests, 60 specimens of variable wall thickness were subjected to a pressure of 2,085 psi at 685 F until col-lapse occurred. ThecladdingEallthicknesswas 0.0285, 0.0263, 0.0251 and 0.0240 inch. The cladding thickness included the range of tolerances for production cladding, and the pressure represented the fuel rod maximum pressure differential at operating conditions. The temperature was selected to conservatively approximate in-pile creep rates. It was found that the 0.024-in wall specimens collapsed in less than a month, and several 0.0263-in wall specimens collapsed in less than 3 months. In view of the unknown increase of in-pile creep rates as compared with out-of-pile creep rates, it was decided to provide backup support for the cladding in the upper end void re-gion where cladding temperatures of 650 F occur in hot channels. v 56-57-71-3
^
<g A thin-walled stainless steel tube was selected as a backup spacer. This spacer has the desirable property of providing radia. support without causing large restraint for the axial expansion of the fuel. Developnent tests have been performed to select a spacer that can withstand shipping acceleration of fuel pellets and provide the required backup support for the cladding. The tubes were encap-sulated in 0.Ol6-in wall Zircaloy tubing and subjected to a pres-sure of 2,750 psi at 850 F. This represents a 10 percent margin on system design pressure at the operating temperature for the spacer tube. On sectioning the specimens, there was no measurable deformation of the tube. This test was performed to demonstrate that the spacer is capable of supporting the cladding, as required i in the criteria. V 56-57-71A
=
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(3 f
- 58. Describe in detail each test or experiment mentioned at Pase 3-79 of the PSAR which you contend demonstrates that rods can be safely cperated to the point where total per=anent strain is 1-1/2 percent er higher in the temperature range applicable to PriR claMhg. Also state the follow-ing:
(a) By whom and for whom these tests were performed and when they were performed; (b) Each fact, calculation and assu=ption upon which it is centended that each of such tests apply to the Midland fuel cladding ander ma-a= burnup and linear heat rate separately in a LOCA and in the most severe nor=al operating conditions. Include within ycur answer how you arrive at your opinion as to the most severe i operating condition. If in your answer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth ccmpletely the text of each such other reference or attach a copy.
- 83. With respect to the answer Item C to ACRS questien 3.1, Amendment IIo 5, Appendix Page 3.1-101 of the Appendix to the PSAR, describe in de-tail each fact, calculation and assumption upon which it is asserted that it is reasonable to conclude that "the value 1 percent is well below the lowest failure strain." If in your answer you nake reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth completely the text of each such other reference or attach a copy.
, , 58-83-1 k
Answer Since both of these interrogatories relate to allowable strain, they are discussed together. It is assumed here that Interrogatory 83 refers to Item C in response to DRL C.uestion 31 on Page 3.1-1 of Amendment No 5 of the PSAR. Ref 50 (noted on Page 3-79 of PSAR) is entitled " Fracture of Cylindrical Fuel Rod Cladding Due to Plastic Instability" and is dated April 1967. This work was done in support of the Light Water Bree:ier - Large Geed and Blanket Reactor development program, in which efforts are being made to extend current perfor=ance limits for nuclear reactors. The abstract from this report is given below. This report describes a failure mechanism for the cladding of cylindrical fuel rods whereby fracture occurs as a consequence of plastic instability induced by fuel swelling. The loss of work hardening capacity from irradiation damage reduces the ncminal failure strain by an order of magnitude, resulting in fracture at nominal strains of about 2% with Zircaloy cladding. When recently reported in-pile fuel red failure experi=ents are compared with theoretical predictions, good agreement results. This work provides a basis for setting allowable clad strain design limits. It was carried out in support of the L*.G-L33R fuel element development effort. The work given in the report cites reference to several types of Zircalcy-clad fuel elements including uranium metal fuel as well as UO ' 3 ** f 2 the fuel elements had bonded cladding, and others were nonbonded. Cor-relation of these diverse conditions with theory (as given in the report) is adequate justification for its use in setting design ll=its. The results of this work are illustrated in Figure 2 of the report, which is attached. In response to Interrogatory 58, it may be noted from the 58-83-2
figure, that failure strain is temperature dependent. In regions of high burnup (where fuel growth vill result in clad strain which apprcaches the design limit), the cladding temperatures will also be high (generally above 600 F, and always above coolant inlet temperature of 550 F). The failure strains in this region exceed 1-1/2%. This is the basis for the statement that rods can be safety operated to greater than 1-1/26 strain. In view of the demenstrated capability of Zircaloy to absorb the strain cited, a value of 1% is a conservative design allowable strain. Further discussion related to Part (b) of Interregatory 58 may be found in Response to Interrogatories 53 and 59 With respect to LOCA, (Loss of Coolant Accident), the design strain limit of 1% does not apply, since this is classed as an accident, net an operating condition. The behavior ( , of the fuel rod during LOCA is described in Section 14 of the PSA3. 1 58-83-3 y e
N. . NN (. 7 g g g .- . N,e g _. A 6 2-- ~ EXPERIMENTAL RESULTS _ 0960 STABLE _ O W:, O NECKED OR FAILED y5 I ' W ~ , CLAD THICKNESS
-THEORETICAL FAILURE E' MILS
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o , Ib 200 - 300 400 500 600 700 000 900 1000 110 0 TEMPERATURE *F -> Figure 2. Nominal Circumferential Strain at Instability c = vs Temperature. -
, 0
59 state and describe in detail each of the experiments, including each of their facts, calculations and assumptions, mentioned at Page 3-80 of the PSAR which "you contend supports the varicus individual design parameters and op3raning conditions up to and perhaps beyond the =aximum design burnup of 55 000 .Wd/ICU." If in your answer you =ake reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth completely the text of each such other reference or attach a copy. Answer The experimental work and its application in support of the state =ent / quoted above are given in the text of the PSAR (Pages 3-80 through 3-82) which follows the quoted passage. The nature of the test and its perti-nence to the conclusion is given, ana reference is provided to publically available documents. In su= mary, the tests given on Page 3-80 are presented as examples which illustrate that the fuel. melting criterien is met (note that design fuel-clad gap is 6-1/2 mils, and design maximum heat rate is 16.83 kW/ft - well within the parameters of the test), and that the clad strain criterien is met at beginning of life (no strain was observed for design conditions). The experimental work summarized'en Page 3-81 and its application to the design evaluation is presented as evidence that effects of irradiatien .o 59-1
-m
- i, ) en fuel swelling vill not result in clad strain which exceeds design criteria at the end of life. Conservative assumptions are used in the calculations outlined under " application." For example, clad restraint is neglected, design burnup of 55,000 &dhCJ is used in lieu of maximum expected burnup (k2,000 MidAmJ, see Page 3-54), and the worst conditions with respect to clad and fuel tolerances, fuel density, and pcuer peaking are used. The calculation includes the combined effects and interrela-tionships of thermal expansion, fuel growth, and changes in conductivity at the fuel-clad interface. Additional conservative assumptions are inherent in the selecticn of de-sign allowable strain. See the combined response to InterrcGatories 53
, , and 83.
l l l l t
, ,. 59-2 ;
1
- 60. Describe in detail, including each fact, calculation and assumption, each test or experiment which will be performed in the B & W High Burnup Irradiation Program as set forth at Page 3-80 of the PSAR. Include within your answer a schedule for ccmpletion cf these tests and a de-scription of procedure which will be followed in incorporating the re-sults of these tests into the final design. Your answer should also include a detailed description of the quality assurance program to be followed in this procedure. If in your answer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth completely the text of each such other reference or attach a copy.
- 61. Describe in detail, including each fact, calculation and assumption, each test or experiment which has been performed or will be performed as part of the " Fuel swelling studies at B & W" as stated on Page 3-82 of your PSAR. State in detail how the results of each such test or experi-ment ruate to burnup, heating rate, fuel density, grain size and clad restraint and how such results indicate the effects of fuel swelling.
Include within your answer a description in detail of the reports of the postirradiation examination, including but not limited to, the investigation of dimensional changes, the metallographic examination of fuel and cladding and fission gas relet. :e correlations. If in your answer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth completely the text of each such other reference or attach a copy. 60-61-73-1
73 With respect to the statement in Section 3 3 3 3 3 on Pase 3-105 of the PSAR, state each experiment, test, fact, calculation and as-sumption upon which you conclude that "the infomation is essential for advancement of the art, but is not considered critical in the sense that all of the programs must be completed to insure safe operation." Also state what you mean by " essential for advancement of the art, but not... critical..." to safety. If in your answer you make reference to other than textual (exclusive of footnote) mattar in the PSAR, then set forth completely the text of each such other reference or attach a copy. Answer The B & W High Burnup Irradiation Program is described in detail in Sec-tion 3 3 3 3 3 of the PSAR, including the schedule for the program which is given in Table 3-18. Examination of the test results is now in prog-r.ess.
" Fuel swelling studies at B & W" refers to the B & W High Burnup Irradi-ation Program.
The relationship of this experimental program to the fuel design for the initial Midland Unit cores is clarified in Item D, Page 1.1-5, Amendment 5, of the PSAR, which follows:
"D. High Burnun Fuel Tests The results of the 'High Burnup Fuel Tests' are not needed to
! complete the safety evaluation for Midland Plant.
~
! 60-61-73-2 l
..s "The primary purpose of the High Burnup Program is to demon-strate the capability of the fuel design for future operation at higher power levels and to detemine the swelling rate of UO2 as a function of burmig using fuel rods of the same de-sign as the core in order to advance the state of the art.
In addition to determining the swelling rate, the effect of several other variables, including the density, heat rate, cladding restraint and the resulting clad strain, will be investigated.
"The safe design of the Midland Plant fuel rod is adequately demonstrated by the material presented in Paragraph 3 2.4.2.2 of the PSAR. Thus, this program is not required to substanti-t ")
ate the design of Midland and has been deleted frem Section 1 5." The phrase " advancement of the art" refers to the developnent of the data necessary for the design of future cores which will operate at higher burnup levels than design burnup for the first core of the Midland Units. 60-61-73-3
.s
- 62. With respect to the statement that "the effect of Zircaloy creep on the amount of fuel red growth due to fuel swe77 k g has been investigated,"
state and describe in detail each test and experi=ent, including each fact, calculation and assumption thereof, which describes your investiga:isn and which you contend supports the results or intended use of your investiga-tion. If in your answer you make reference to other than textual (ex-
- clusive of footnote) matter in the PSAR, then set forth completely the text of each such other reference or attach a copy.
Answer The statement quoted above is found on Pa6e 3-85 of the PSAR, in the para-graph entitled "Effect of Zircaloy Creep." The statement is based upon analysis rather than test or experiment. This paragraph cencludes a dis-cussion which is directed towards showing that the strain criteria is cet. The purpose of the paragraph is to illustrate that within the established I bounds of clad strain, contact pressures upon which ther=al conductance ! at the fuel-clad interface is based is valid. This, in turn, provides confidence that design fuel temperatures are applicable (also that assess-ment of fuel melting criteria is valid) even in the presence of the un-realistically high design internal pressure of 3300 psi. This paragraph is therefore added justification that assumptions made in the clad strain analysis are valid. l
. ,. 62-1
i h 63 With respect to the effect of a Departure frem Nucleate Boiling upon the side of a fuel rod adjacent to a guide tube as described at Page 3-E6 of the PSAR, state and describe in detail each test and experi.wnt, in-cluding each fact, calculation and assumption thereof, upon which you con-tend that " insufficient strength would be available to generate a force of sufficient magnitude to cause a significant deflection of the guide tube." If in your answer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth ecmpletely the text of each su6 other reference or attach a copy. Answer The referenced section of the PSAR discusses the possibility of thermal bowing of a fuel rod adjacent to a guide tube resulting frcm a Departure fran Nucleate Boiling. The PSAR points out that if a DNB condition did cause the fuel rod to bow enough to touch the guide tube, any temperature gradient large enough to cause that much bcwing would also reduce the yield strength of the fuel red chmng material, and thus would diminish its ability to deflect the guide tube. The guide tube in this instance would maintain its own structural strength, because of the cooling frem the flow inside the tube, thus assuring that there will be no interference with control red motien in the guide tube. In addition, as the CRDL test program results showed, (see tcpical report BAW 10007) a large lateral deflection of the guide tube itself would be required before there would be any interference with the operation of the control rod. w 63-1 s
,
- a
t ',
- 64. With respect to the flow-induced vibratory amplitude for the fuel assembly and fuel rod, described at Pase 3-86 of the PSAR, state and describe in detail each test and experiment, includir4 each fact, cal-culation and assumption thereof, which will be performed at the Control Rod Drive Line Facility (CRDL) of the akW Research Center, Alliance,
]hio, to ascertain said amplitude. Also state in detail each source which will produce oscillations in the system and its resonant frequency.
If in your answer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth completely the text of each .:uch other reference or attach a copy. Answer During testing of a production type fuel assembly in the CRDL, the vibra-tion amplitude of both the fuel assembly and fuel pin were measured. The results of these measurements for the maximum flow, which exceeded the normal design flow by about 5%, are as follows: Single Amplitude Motion at _ Mid-Scan - Mils Fuel Assembly 7.0 Fuel Pin 03 The major source of pressure fluctuations in the reactor coolant primary system will be from the reactor coolant pumps. The pressure fluctuations from the pump will have a frequency of about 100 Hz. Since the wave 64-1
length of these fluctuatioris is long compared to either the length of the fuel assembly or the fuel pin, the pressure oscilliations will not have significant effects. No other sources of pressure fluctuation which would significantly affect vibratory response of the fuel assembly are known or expected, f a v 64-2
- 65. State and describe in detail each test and experi=ent, including each fact, calculation and assumption thereof, which has been perfor=ed, is being perfomed or vill be perfo=ed to de=enstrate the overall =echan-ical perfemance of the fuel asse=bly as it is stated to be at ?sEe 3-86 of the PSAR. If in your answer you =ake reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth completely the text of each such other reference or attach a copy.
Answer A full size, reference design, fuel assembly was exposed to reactor oper-ating conditions of te=perature, pressure, and coolant flow rate for a ( ; total time period of 3700 hours. The asse=bly was exa=ined in detail. Zio adverse mechanical effects were found. 65-1 L.
, ., -- -~-
t)
- 66. With respect to the control red drive mechanism ("CRIH") described on Page 3-89 of the PSAR, describe in detail, including each fact, cal-culation and assumption, what " extensive analytical, developmental, design, test and manufacturing experience" has been attained. Also, present a detailed descriptien of the operating history of CRIN relied upon by you listing and describing in detail all cases of malfunctions, including within your description, an analysis of the changes, if any, and rear.cas therefor, made in the design or operation of said mechanism as a result of said malfunction disclosed by said operating history. If in your answer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth completely the text of each such other reference or attach a copy.
/] '
d 74 With respect to the statement in Section 3.3.3.k.1 on Page 3-106 of the PSAR, state each experiment and thst, specifying each fact, cal-culation and assumption thereof, upon which you conclude that " material ecmpatibility and structural design of these components will be adequate for the life of the mechanism." If in your answer you make reference to ' other than textual (exclusive of footnote) matter in the PSAR, then set forth completely the text of each such other reference or attach a copy. Answer As stated in Section 3, Page 5, of the Topical Report BAW 10007 entitled
" Control Rod Drive System Test Program," and Question 6, Page 9, of 66-74-1
7
.)
BAW 10007 (Supplement 1), Royal Industries of Santa Ana, California, a prominent supplier of re ner nut drives for other reactor installations, did the design and fabrication work for the prototype of this roHer nut CRIN. It incorporates proven principles and material combinations and is based on extensive analytical, developmental, design, test, and manu-facturing experience gained from a number of years in the nuclear field. The report also states that control rod drive mechanisms based on this roller nut principle and utilizing similar materials have been operating satisfactorily on the Shippingport PWR since December of 1957. In regard to the testing of this CRIN, part of the testing is described in Section 4 of BAW 10007. The prototype mechanism was " life tested" in a full-sized loop at the Babcock & Wilcox Research Center, AHiance, Ohio. With the exception of radiation effects, this facility duplicates l all of the reactor operating conditions, ie, temperature, coolant flow, pressure and water chemistry. A prototype fuel assembly and upper guide structure of the reactor internals were =ounted in the autoclave portien of the loop. The drive mechanism was coupled to a prototype control rod assembly by a bayonet coupling assembly. Within the autoclave it was possible to simulate the guidance, wei6 ht, flow and alignment characteristics of the complete drive line. A " life test" was then run in a 100% misaligned condition to determine the accept-ability of the roHer nut drive mechanism with respect to performance and wear. This ccndition would have the most adverse effect on trip time of the mechanism and produced the most severe component wear during an ~.j l 66-7h-2 l
i N. operational life test, due to increased frictionel drag forces and loads developed in the drive line. This test was run under the most adverse conditions which could exist, based on a study of the design tolerances of the reactor vessel, reactor internals, and fuel assembly. During the " life test," the drive was (a) cycled 23,950 times through various stroke lengths which totaled over 126,000 feet of travel and (b) tripped at least 500 times from the fully withdrawn position. The actual stroke dimensions are tabulated in Table 1, Page 10, of MW 10007. They were designed to duplicate the stroke, travel, and trips expected in a 20-year operational life in regulating red service. The drive was then disassembled and all components inspected for wear 4 - and corrosion. The degree of wear was determined by physical measurements which were compared to the as-fabricated, pretest dimensions. This data is tabulated in Table 2, Page 17 of MW 10007. After the wear measure-ments were compiled, it was noted that the wear on those ecmponents which function in a sliding or rotating manner, or transmitted forces or loads on bearing contact surfaces was satisfactory and well within the design 4 specification requirements. These requirements are that after measure-ments for material loss are taken, these measurements will be multiplied , by a factor of four and the functional and structural adequacy of the component will be reanalyzed using the results. If this analysis, using the same design factors for allowable stress, safety, and wear as used in the original design calculations, results in confirmation of its fune-tional and structural adequacy, the component in question shall be 66-763
O - considered to have successfully passed the life test. For the roller nuts themselves, a factor of two for wear is used rather than four. On the basis of the test results for the prototype roller nut mechanism, the following was concluded: (1) The trip time of this type of mechanism is within the design requirements to ensure safe and dependable reactor operation. (2) The measured wear of the critical components of the mechanism was well within specified design requirements for an operational life of 20 years. (3) In the event of a stuck red condition, the drive can develop an insertion force far in excess of the requirements of the design specification as stated in Question 3, Page 5, of BAW 10007 (Supplement 1). It should also be noted that each individ al production drive is tested before being chipped, as described in response to Question 4, Page 6, of BAW 10007 (Supplement 1). Since the initiation of the prototype roller nut CRIN testir.g at the Babcock & Wilcox Research Center, only one malfunction has even been noted vnich could be related to a question of safety. A guide bushing in the thermal barrier assembly was not properly retained, and it became mis-aligned causing the drive to tri'p improperly. This problem was immediately resolved by the utilization of a welded housing assembly which captures the bushing -- making it physically impossible for the bushing to become J 66-74 k
3 misaligned. Other changes were made for various reasons as stated in Question 4, PaSe 6, of BAW 10007 (Supplement 1). The materials that were used in the prototype roller nut drive have been successfully used in the fabrication of roller nut mechanisms for other reactor installations. Royal Industries established the material selec- , tion for the prototype, and the results of the testing of the mechanism at operating conditions for BE WR have confirmed the adequacy and reliability of the materials, as stated en Page 17 of BAW 10007 (Supple-ment 1). The materials used in the fabrication of the drive are selected on the basis of: (1) Function of the part. (2) The environment in which it will operate. (3) The requirements of the applicable codes. (h) Factors of wear, gall!ng, corrosion, and resistance. (5) The physical conditions to which it will be exposed. As noted in the PSAR, the structual design of the critical components was also dene by Royal Industries for the prototype drive. As stated in ques-tien 1 on Page 2 of BAW 10007 (Supplement 1) the design of the pressure boundary parts complies with Secticn III of the ASME Boiler and Pressure Vessel Code for Class A vessels. The operating transient cycles, which are considered for the stress analysis of the reactor pressure vessel, 66-74-5
O \ are also factored into the ' motor tube design. Materials conform to ASD4 or ASME, Section II, Material Specifications. All welding is performed by personnel qualified under ASME Ccde, Section IX, Welding C,ualifications. Based upon the design criteria, as stated in the preceding paragraphs, and the results of the extensive testing done on this drive, it is felt that the material compatibility and structural design of the critical components will be adequate for safe operation for the life of the mechanism. This Babcock & Wilcox Topical Report on the CRD System Test Prcgrar. was submitted July 1969, and Supplement 1 in June 1970, well after the filing of 'the pSAR in January 1969 (j This report will be incorporated in the application during the final design stage. V 66-7h-6
i i l l l l 67 State and describe in detail each test and experiment, including i each fact, calculation cnd assumption thereof, performed in the research I and development program for fuel assembly heat transfer and fluid flow
" applicable to the design of the reference reactor" as stated in Sec- ,
tion 3 3 2.2 at Page 3-101 of the PSAR. If in your answer you make j l reference to other than textual (exclucive of footnote) matter in the ! PSAR, then set forth completely the text of each such other reference j or attach a copy.
- 68. With respect to each test and experiment mentioned in Section 3 3 2.2 of the PSAR, for which "the results of these tests will be applied to the l
final thez nal design of the reactor and the specification of operating l limits," state and describe in detail each such test, experiment and
\
nnnivsis, specifying each fact and calculation and assumption thereof. l If in your answer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth completely the terc of each such other reference or attach a copy.
)
69 With respect to the tests mentioned in Sections 3 3 2.2.1 and l 3 3 2.2.2 of the PSAR, describe in detail each such test, specifying each fact, calculation and assu=ption thereof. If in your answer you make reference to other that textual (exclusive of footnote) matter in the PSAR, then set forth completely the text of each such other reference or attach a copy. l 1 67-68-69-1
')
Answer Interrogatories 67,68and69areallconcernedwithB&WThermal/ Hydraulic RSD programs and are answered below: The data frcm the experimental programs referenced in the PSAR Sec-tion 3 3 2.2 have been analyzed. The analysis has shown that the W-3 critical heat flux correlation, which was used to design the Midland plant cores, is very conservative. The experimental programs referred to in the PSAR are: A nine-rod, six-foot long bundle with an axial heat flux profile shown by the shaded portion of Figure 1 was tested. The range of conditions and the important gecmetric data are listed below: Heated Length 72.0 Inches Rod CD .420 Inch Pitch .558 Inch Flow Channel Cross Section 1.679 Inches by 1.679 Inches Pressure 2000-2400 Fsia Mass Flux 1.0 - 3 5 x 10613,73773 2 Inlet Enthalpy 415 - 675 Etu/lbm This experimental program confirmed that the F factor correction for non-uniform flux shapes in combination with the W-3 correlation conserva-tively predicts the behavior of a rod bundle with nonurlform axial flux. Nonuniform heat generation is typical of reactor conditions. l 67-68-69-2 j.. n
Similav 9-red bundles were tested at both the Babcock & Wilcox Research Center and Columbia University. The range of conditions and the im-portant geometric data are listed below: Heated Le ch 72.0 Inches Rod OD .43o Inch Pitch 568 Inch Flow Channal Cross Section 1.704 x 1.704 Inches Pressure 2000-2400 Psia 6 14Lss Flux 5 - 3 5 x 10 lbm/Er-Ft2 Inlet EnthalpF 323 - 673 Btu /Hr-Ft2 ! Axial Heat Flux Uniform Profile
~7 These tests were done to insure that there were no systematic errors or !
)
loop effects in the B&W data. The results of these tests were in good : agreement; thus there are no systematic errors or loop effects in B&W { experimental critical heat flux studies. ! Three 25-rod bundles were tested at Columbia University. All bundles had uniform axial heat flux distribution. One bundle had a uniform radial heat flux distribution while the other had nonuniform heat flux distribu-tions with one bundle having a cold center pin to simulcte a control rod. The range of conditions and the important geometric data are listed below: l Heated Len6th 72.0 Inches Rod OD 430 Inch i d Pitch 568 Inch 67-68-69-3
- i. *'[
L m
Flow Channel Cross Section 2.M x 2.% Inches Pressure 2000-2400 Psia Mass Flux .5 - 3 5 x 106 lbm/Hr-Ft2 Inlet Enthalpy 400-675 Btu /lbm These tests showed that the combination of 38M subchannel analysis code and the W-3 correlation would yield very conservative results in large bundles with nonuniform radial heat flux listributions ani thus would result in a conservative core design. Figure 2 is a ecmparison of ex-perimental critical heat fluxes obtained during these investigations with the values predicted by W-3. The experimental values are always greater than the predicted value. 'this fact coupled with the factor of safety used in connection with the W-3 correlation insure adequate core thermal design margin. l I l The pressure drop tests described in the PSAR have been analyzed and-methods based on drag coefficients have been developed to predict the behavior of the spacer grid. These methods have been independently veri-fled by ccuparing the measured pressure drop in our bundle heat transfer tests with the calculated value based on our core hydraulic tests. Fig-ure 3 shows a comparison of measured vs calculated pressure drop for B&W unifom bundle CHF tests. The agreement is very good, especially since the points were in regicns where both two-phase and single-phase flow existed. 1 As in the case of the core heat transfer pregram, BtB recent R&D in the area of core hydraulics, to develop more advanced reactor designs, also l v 67-68-69-4 n
1 verifies the ver/ consertative Midland design. The B&D performed in the core hydraulics area since the PSAR was written is discussed below: A test section consisting of two rod bundles, each having 8 x 15 rods of .430 inch diameter and a .568 inch pitch we built. The flow to each bundle could be adjusted so that a mismatch would occur at the inlet and the resulting cross flow between bundles was found by measuring the inle'c and outlet flows to both bundles. Pressures were also measured at 5 axial planes. The results of this test were used to develop a correlation for 1 resistance to cross flow. The results of calculations using this cor-relation show that the method used to set the hot bundle flow in the ~ .i , Midland design was very con =ervative. l 1 Analysis of B&W extensive critical heat flux bundle data with a subchannel code has shown that the level of turbulent mixing required to correlate the data well is three times as large as the value used in designing the Midland core. This again exhibits the conservatism in the Midland core design. 67-68-69-5 .-- 9
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Figure 3. Comparison of Predicted and Measured AP for B&W Uniform Bundle CHF Tests (System Pressure of 2200 psia) 24 22 - O 18 - O O
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- 70. With respect to the conclusion on Page 3-98, Section 3 2.4.3 5 of the PSAR that " wear of the guide tubes and the CRA will not be of con-cern," state and describe in detail each experiment, test, fact, calcu-lation and assumption upon which you have arrived at this conclusion.
If in your answer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth completely the text of each such other reference or attach a copy. Answer The wear of the control rod assembly (CRA) and control rod guide tubes was monitored as a part of the testing program for the Control Rod Drive Mechanism, discussed in detail in Topical Report BAW 10007, " Control Rod Drive System Test Program," and in response to Interrogatories 66 and ,
- 74. The results given on Page 3-98 of the PSAR are from measurements taken on the fuel assembly included as a part of the test-fixture in the above tests. After the equivalent of three years of operation with the CRA in a deliberately misaligned condition, the wear on the CRA and guide tubes was within acceptable limits.
(.i 70-1
- 72. Describe in detail each test and experi:nent, specifying each fact, calculation and assumption thereof, which will be performed to "detez-nine the structural characteristics of the Nel assembly which are pertinent to loadings resulting from normal operation, handling, earthquake and accident conditions" for the fuel assembly structural components as stated in Section 3 3 3 3 2 on Page 3-105 of the PSAR. Also include within your answer each fact, calculation and assumption upon which you contend that each of the above tests apply to the proposed Midland Units.
If in your answer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth completely the text of each such other reference or attach a copy. I Answer Section 5 and Appendix B of the proprietary E8M Topical Report 10008, i part 2, " Fuel Assembly Stress and Deflection Analysis for Loss-of-Coolant Accident And Seismic Excitation," describe the tests and experiments con- i ducted to determine the structural characteristics of the fuel assembly. ! For normal operational characteristics, see the answer to Interrogatory l 65 I l The fuel assembly to be supplied for the Midland Plant Units will be of a like design to that which was tested and analyzed as reported in the above topical report. Appropriate analyses to confirm seismic acceptabil-ity of the fuel assembly for the Midland 1lant Site using the methods g V tescribed will be ccanpleted during the detailed design period. 72-1
-; ;G
- s. . ..
m. J 75 With respect to the internal vent valve discussed in Section 3 3.4 at Pase 3-109 of the PSAR, state what, if any, relative motion of the vent valve to its seat existed in the vibration test. State at what frequency the motion occurred. If in your answer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth completely the text of each such other reference or attach a copy. Answer As stated in the fifth sentence of the eighth paragraph on Page 3-109 of the PSAR, the test results indicated that there was no relative mo-tion between the valve and its seat for conditions simulating operation. 3, 75-1 i __- _ _ __ t _- . _ - - . , .-
1 1
- 76. Describe in detail each test, including each fact, calculation and assumption thereof, mentioned at Page 12-2 of Amendment No 8 of the PSAR which will be conducted to determine decontamination factors for each radionuclide on each piece of equipment which will be encountered by the process steam to be generated from the proposed Midland Units. If in your answer you make reference to other than textual (exclusive of foot-note) matter in the PSAR, then set forth completely the text of each such other reference or attach a copy.
Answer
, Amendment No 8 concernin6 the process steam to ibw Chemical Company has been superseded by Amendmenta No 17 and 18, copies of which are attached hereto. As explained in Amendment No 17, a tertiary heat exchanger system will be installed to separate the steam in the nuclear secondary system from the process steem to be delivered to Dow Chemical Company.
In the old system, referred to in Amendment No 8, the steam produced in j i the nuclear secondary system was delivered directly to Dow. With the new 1 I system described in Anendment No 17, the steam produced in the nuclear secondary system will be passed through the tube side of tertiary heat exchangers producing process steam on the shell side of the heat ex-changer. Thus, the reactor coolant water containin6 radionuclides must penetrate two barriers, the heat exchan6er in the nuclear system and the tertiary he.t exchanger, before it could enter the process steam. J 76-1 o l
( Because of this added barrier (the tertiary heat exchanger), the proba-bility of radionuclides from the reactor coolant entering the process steam is very small. Therefore, Consumers Power Ccmpany has committed in Amendment No 18 that the process steam delivered to Dow will not contain more radioactivity than the Lake Huron makeup water following any treat-ment it may be given before it enters the tertiary heat exchanger as condensate makeup. In other words, the radioactivity concentration in the process steam delivered to Dow will not exceed the natural radio-activity concentration in Lake Huron. Therefore, the radioactivity in the process steam will be no greater than it would be if it were produced by a fossil fueled plant. ( To provide assurance that radicactivity is not being added to the process steam from the reactor, the radioactivity in the makeup water to the ter-tiary heat exchanger and the steam to Dow from the tertiary heat exchanger will be extensively monitorea. If the radioactivity in the steam, when compared to the radioactivity in the makeup water indicates radicactivity from the reactor is being added to the process steam, the leaking ter-tiary heat exchanger will be isolated from the system. With this new criteria, consideration is not given to the decontamination factors which may or may not be obtained on various pieces of equipment. Therefore, these tests (reference Amendment No 8) are no longer considered necessary.
- Detailed procec. res for collecting samples of the condensate and steam, 76-2
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- a. CCNSUMERS POWER CCMPANY APPLICATION FOR REACTOR CCNSTRUCTION PERMIT AND OPERATING LICENSE Locket No. 50-329 Dccket No. 50-330 Ament' ment No.17 This amendment is in response to oral discussions, with the DRL-Staff regarding tertiary heat exchangers and contains revised Fages 11.00-1 and 11.00-2 designated by the notation,
" Amendment No. 17,9/11/70." These pages are to replace Pages 11.00-1 and 11.00-2 transmitted to you in Amendment No. 15 dated August 15, 1970. Pages 16-1 and 16-2 transmitted to you in Amendment No. 16 on September h, 1970 are to be deleted from the application. Also enclosed herewith is Figure 11.00-1 titled "Preliminray Tertiary Heat Exchanger System." Figure 16-1 transmitted in Amendment No. 16 on September k, 1970 is to be deleted from the application.
CONSL11ERS POWER CCMPANY By /s/ B. G. Campbell B. G. Campbell Dated: September 11, 1970 Vice President Sworn and subscribed to before me this lith day of September, 1970.
/s/ Helen R. Lehr (SEAL) Notary Public, Jackson County, Michi6an My Ccmmission Expires December 11, 1973 V
- 11. Radioactivity in the Process Steam (1) General Description The function of the Tertiary Heat Exchanger (THX) syntah la to provide i complete physical separation in the form of closed heat exchangers between secondary system steam from the turbine plant cycle and the tertiary system steam delivered to Dov. The ecmplete system is composed of multiple banks of shell and tube evaporators with attendant drain coolers, feed pumps, etc.
which exchange the heat of turbine steam condensed in the tube side of the exchangers with (tertiary) process steam generated at a lover pressure in the shell side. The THX system is housed in a separate structure as shown on the Amendment 16 revision to PSAR Figure 1-8, " Station Arrangement." The facility is, as shown, within the plant boundary and is operated by Consumers Power Company. The turbine cycle steam supply is essentially as described in Amendment 5, answer to DRL Question 10.1, and referred.to as "High Pressure m Process Steam" and " Low Pressure Process Steam" respectively, and as shown schematically on Figures 10.1-1, 10.1-2 and 10.1-3 vith the exception that *he HP process steam reducing stations vould be eliminated. The attached figure, " Preliminary Tertiary Heat Exchanger System (Fig. 11.00-1)" shows the typical schematic arrangement of one of each of the multiple high pressure and Icv pressure Tertiary Heat Exchangers. (2) Design Standards The THX system components and structures will be designed and con-structed in conformance with the standards established for the turbine generator system, structures and auxiliaries. These standards include, but are not limited to, the following: Structures: Class 2 structures per PSAR Appendix 5A THX's: ASME Section VIII Piping: ANSI B31.1.0 v
, 11.00-1 Amendment No. 17 9/11/To
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- 3. Radiological Consequences of THX Failure The failure of the THX can be classified in two categories; (1) gross failure of the primary side of the THX allowing short-term uncontrolled release of the secondary system steam frcm the turbine plant, and (2) THX tube failure resulting in radioactivity carry-over to the tertiarf steam side of the TEX.
Case (1) is comparable to the steam line break accident analysis described in Section lh of the PSAR vb'ch assumed a 1 gpm steam generator priuary to secondarf leak rate with 1 percent failed fuel radicnuclide inventory in the primarf system. Due to smaller lines and isolation capabilities, the consequences of gross failure of the THX would result in subst.antially less secondarf system steam loss than the steam line break accident and hence is of less radiological consequence. In Case (2), the failure of tubes will be detected by the on-line gross gamma monitoring system as well as the grab sample analysis for gross B. Within the detection limits and turn-around time required to s achieve the detection level, the tertiary stesa delivered to Dow will l not contain more radioactivity than the feedvater supplied to the THX. The grab sample monitoring system vill be designed to distinguish j significant changes in expected feedvater radioactivity levels. The method of detection vill take into account a statistical model, allowing for detection uncertainties, to ascertain the incidence of THX tube leakage. Upon the determination of leakage of radioactivity frem the secondary system to the tertiary steam, the THX associated with the radioactive leakage vill be isolated from the supply of tertiary steam to Dov.' l l s 11.00-2 Amendment No. 17 9/11/70
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('3N0977ERM) COOLER. c N O ' b}' LR ,EED PJMP t'7YR) l Preliminary Tertiary Heat Exchanger System Fig. 11.00-1 l A=end:aent No. 17 9/11/70 i
4 CCNSUMERS POWER CCHPANY APPLICATICN FCR REACTOR CONSI'RUCTION PERMIT AND CPERATING LICENSE Docket No. 50-329 Docket No. 50-330 Amendment No. 18 This amendment is in response to oral discussions with the ACRS subecessittee and the DRL staff regarding tertiary beat ewh=nyrs and ccatains a revised page 11.00-2 designated by the notation " Amendment No. 18,9/15/70." This page replaces page 11.00-2 autaitted to you in Amend:nent No.17, dated September 11, 1970. CCNSUMERS POWER COMPANY By /s/ J. B. Sinreen J. B. Simpson Dated September 15, 1970 Senior Vice President Sworn and subscribed to before me this 15th day of September,1970. (SEAL) q /s/ Helen R. Lehr k/ Notary Public, Jackson County, Michigan My Ccanission Expires December 11, 1973
l i Radiological Consecuences of THX Failure l 3. The failure of the THX can be classified in two categories; (1) gross I failure of the primary side of the THX allowing short-term uncontrolled release of the secondary system steam from the turbine plant, and (2) THX ! tube failure resulting in radioactivity carry-over to the tertiary steam side of the THX. l Case (1) is comparable to the steam line break accident analysis ! described in Section lh of the PSAR vhich assumed a 1 gpm steam generator primary to secondary leak rate with 1 percent failed fuel radionuclide ) inventory in the primary system. Due to smaller lines and isolation capabilities, the consequences of gross failure of the THX vould result ! in substantially less secondary system steam loss than the steam line l break accident and hence is of less radiological consequence. ; In Case (2), the failure of tubes will be detected by the on-line ! gross 6 anna monitoring system as w.11 as the grab sample analysis for gross B. Within the detection limits and turn-around time required to achieve the detection level, the tertiary steam delivered to Dow vill be compared with and vill nct centain more radioactivity than the Lake Huron 1 make-up water following any treatment as supplied to the THX. Due to the l possibility of concentration of naturally occurring radioactivity in the i steam condensate system, scme allowance may be required in the make-up water vs. steam comparison on the basis of suitable evidence of no concurrent radioactivity leaka6e titrough the THI. It is intended that the grab sample monitoring system will be capable of distinguishing sig-nificant changes (e.g., a factor of 2) in the radioactivity levels of ' the make-up water. The method of detection vill take into account a sta-l tistical model, allowing for detection uncertainties, to ascertain the j incidence of THX tube leaka6e. Upon the determinatien of leaka6e of ; radioactivity from the secondary system to the tertiary steam, the THX associated with the radioactive leakage vill be isolated from the supply of tertiary steam to Dow. ) l (/ 1 11.00-2 Amendment No. 18 9/15/70
77 With respect to Item No 12, Pages 12-1 and 12-2 of Amendment No 8 of the PSAR, describe in detail each piece of the machinery, its use and location, whethe: contained or to be contained in a Dow Che.sical Company ("Dow") or Constaaers Power facility, which come in contact with the pro-cess steam proposed to be generated fra the proposed Midland Units. For each such piece of machinery describe in detail the expected rates of decontamination of each of the available radionuclides, their equilib-rium concentration and the total radiation source each machine could be-come. If in your answer you make reference to other than textual (exclu-sive of footnote) matter in the PSAR, then set forth empletely the text of each such other reference or attach a copy. l Anrser Since the activity in the process steam is not pemitted to exceed natural background in the laks water, the question of decontamination factors for Dow process equipment is no longer applicable. l I _ 77-1
m
- 78. With respect to Item No 12, Pages 12-1 and 12-2 of Amendment No 8 ofthePSAR,describeindetailallproductsand/orusesoftheprocess steam which may cause the beholder and/or user to be exposed to radia-tion originating in whole or in part from the precess steam. If in i
your answer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth completely the text of each such other reference or attach a copy. Answer l Because of the installation of the tertiary heat exchanger system (reference answer to Interrogatory 76) the radioactivity concentra-i tion in the process steam will not exceed the natural backgmund concentration in the Lake Huron water after it has been processed. Therefore, the radiation exposure to the beholder or user of the process steam or products produced by the process steam 'till not exceed the exposure he would receive if the process steam were pro-duced by a fossil fueled steam plant. 78-1 _n. \
7
'J 79 With respect to the control rods' freedom of motion discussed at Page 3-85 of the PSAR, state and describe in detail each test, experiment and analysis, specifying each fact, calculation and assumption thereof, which you rely upon to conclude that control rods' freedom of motion will be assured throughout the service life of the reactors of the pro-posed Midland Units. If in your answer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth com-Pletely the text of es,ch such other reference or attach a copy.
Answer The control rods' freedom of motion was verified by the tests conducted
-) by BkW in the Control Rod Drive Line facility at the Alliance (Ohio)
Research Center. These tests are discussed in detail in Topical Report BAW 10007, "Centrol Rod Drive System Test Program." Testing equivalent to several core lifetimes with a deliberately misaligned control rod assembly demonstrated that there is sufficient clearance reliably to insert the rods, even under severely misaligned conditions. , 79-1 l l I
- 80. With respect to the answer to ACRS Question 2.1, Amendment No 5, Page 2.1-5 of the PSAR, state where the ma:cdmum dose to man is expected to occur and state each fact and assumption used in your detemination, stating the =agnitude in rems per year. Include within your answer the exact location of the inner ring in your environmental surveillance plan.
If in your answer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth completely the text of each such other reference or attach a copy. Answer During nor=al operation of the plant, the m 4" = dose to man would be i , expected to be that dose received frem the plant gaseous radwaste re-lease. This dose depends upon the downwind distance because the concen-tration of the radicactive isotopes in air decreases due to atmospheric turbulence as the downwind distance increases. Since no persons except plant personnel are located within the site boundary, the =mcimum dose tc man received frem the plant gaseous radwaste release will be evalu-ated at the plant site boundary of 500 meters. As shown on Page n-3 of Amendment No. 20 to the PSAR, the annual amount of the gaseous radwaste produced is 9810 3ft . This amount of gases is evaluated basing upon k cc H2.per liter of the prd m 7 coolant. Using the mimum cencentratiens of the noble gas isotopes in the primary 80-1
coolant for 1% failed fuel as shoun in Table 11-3, Page n-5 of the PSAR, the release rate of the gaseous radwaste of significant level after 30-day holdup in the gaseous radvaste system for normal opera-tion is shown in the fonowing table: Release Pate Concentration MPC Aci/sec Site Boundary Fractica of MPC Isotope Half-Life pei/ec 30-Day Holdup ,4ci/cc Site Boundary Xe-133 5.27 Days 3 x lo-I 1.04 x 103 6.24 x 10-9 2.08 x 10-2 Kr-85 10.4 Yrs 3 x 10 -7 2.16 x 103 1.30 x lo 4.33 x 10 -2 Total 6.41 x lo In obtaining the concentration of each isotope at the site boundary shown in the above table, the yearly atmospheric dilution factor listed in Table 2A-18 of the PSAR, = o.6 x 10-5 ,,ef ,3 at the site boundary of 500 meters, is used. This f, as stated on Pages 2A-40 and 2A-42 of PSAR, is obtained using the pocrest diffusion condition and the ==v4== frequency (19%) of the wind blowing in one sector. Therefore, this estimate gives the ==v4=in dose received by a person at the site bound-ary regardless of what sector he is in. As shown in the above table, the total concentration at the site bonMw/ released from the plant gaseous redwaste system is only 6.4% of the limit of lo CFR 20. Based upon 500 mrem / year equivalence to one ==*== pemissible concentration (MPC), the maximum dose to man at the site bo*/ (500 meters) would be 32 maem/ year above the background. 'Ihis ==-4== acae to man is the D L 80-2 i
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whole body and skin (gam plus beta) dose resulting frcm the gaseous radwaste release during the plant normal operation with 1% failed fuel and 30-day boldup. As shown in Table 14-11, Page 14-64 of the PSAR, the dose to a person at the site boundary resulting from the "W== Hypothetical Accident" would be about 8.8 rema, thyroid dose and about 0.6 ren to the whole body. The location of the inner ring of sampling stations is described in Table 2.1-3, Pa6e 2.1-6 of the PSAR, and illustrated on the attached map. ., i T With the addition of the modified liquid radwaste system described in Amendment No 20, there would be no dischrge of liquid radwaste under nonnal plant operation and, therefore, practically no additional dose to man above that calculated above. l I l 1 ...A 80-3
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- 81. With respect to the answer to ACES Questien 2.2, Amendment No 5, Page 2.2-1, state the maxi:m2m and mdn%= temperatures of the cooling pond for both operating and shutdown conditions, speciffing each fact, calculation and assumption thereof, including but not limited to all heat transfer assumptions. Also state the min 4 mum and maximum temper-atures of the cooling pccd water which will be released to the Tittabavassee River, stating each fact, calculation and assumption upon which you base your answer. If in your answer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth completely the text of each such other reference or attach a copy.
Answer () Steady-state cooling pond water temperatures, for both operating and shut-down conditions, were calculated utilizing basic heat transfer principles and the results of a physical model test study. The basic heat transfer principles used are described in " Heat Exchange in the Environment" by J. E. Edinger and J. C. Geyer, E2I Publication 65-902, dated June 1, 1965 The model study was performed at Alden Research Laboratories, Worcester Polytechnic Institute, Holden, Massachusetts. When the plant is shut down the water temperature in the pond will tend to follow the variation in equilibrium temperature. The equilibrium temper-ature is defined as the water surface temperature for which the not rate of heat transfer across the water surface is zero. It is a function of the local climatological data. The pond surface temperature will lag be-hind climatic changes, but will constantly shift tewards the equilibrium temperature, v 81-1
When the plant is in operation the average pond surface temperature increases to the level required to dissipate the imposed heat load. The rate of surface heat dissipation is equal to the product of the surface area, the surface heat transfer coefficient, and the differ-ence in temperature between the pond surface and the equilibrium temperature. The surface heat exchange coefficient is a function of local climatological data and the pond water surface temperature. The model studies at Alden Research Laboratories were made to deter-mine the flow and surface temperature distribution throu6hout the pond. The results were used to calculate the predicted prototype cooling pond performance utilizing surface heat transfer coefficients for various prototype conditions. j I i The maM== pond surface temperatures are expected to occur in J: Ly, I and the minimum temperature conditions prevail in January. The per-tinent average climatological data for these two months follows: July January Air temperature ( F) 70 9 21.2 Wind speed (mph) 8.5 11.7 Relative humidity (%) 67.1 65.7 Golar radiation (Langleys) 542 148 1 These data were developed from U.S. Weather Burecu data taken in the Midland area. Air temperature and humidity data were available for
'uJ 81-2
- 3 :
Midland. Solar radiation dats was taken fra measurements at East Lansing, and the wind speeds were taken from Saginaw Airport data. Meyer's evaporation equation, with a coefficient C = 11, was adopted for this study. The mav4== average monthly pond temperature under shutdown conditions is equal to the average July equilibrium temperature which is 73.5 F. The minimum pond temperature will be 32 F (ice cover) under shutdown con-ditions and could occur in December, January and February. The navimum pond temperatures will occur when the plants are operating under full load conditions, that is with valves wide open. The plant heat to be dissipated by the cooling pond under these conditions is 9.43 x ( 9 10 Btu /hr, and the circulating water flow is 558,000 spa. The mavimum and minimum average monthly pond temperatures, and the average surface heat exchange coefficients for the full load condition are listed below: July January
? W rature to condenser ( F) 94.5 64.0 Average surface temperature ( F) 101 5 69 0 Temperature fr a condenrm ( F) 128.3 97.8 Average surface heat exchange coefficient (Btu /ft / F/ Day) 217 118 The cooling pond will experience temperatures higher than the average monthly steady-state values given above due to the diurnal and day-to-day variations in climatic conditions. Studies have been perfomed in t
81-3
, _ . . . _ _ +
m conjunction with the development of a mathematical trsnaient cooling pond model to determine the short-tem cooling perfomance of the pond. Bourly climatological data were obtained for the Saginaw and East Lansing Weather Bureau stations, and a 40-day period in June and July 1966 was detemined to be representative of a hot and variable sunumer period. The cooling pond perfomance and response during this period was detemined using a two-layered numerical pond model developed to simulate the time dependent characteristics of a cooling pond. The results of the mathematical transient cooling pond model investiga-tion were used to detemine the maximunt expected pond temperstures. thder shutdown conditions, the marimum expected avera6a pond surface temperature will be 83 0 F. When operating under full load the mmh =nn expected average pond surface temperature will be 111.0 F, with cold-and hot-water temperatures of about 104.0 F and 137.8 F, respectively. This condition will have a duration of less than 12 hours. l l l l . U 81-4
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- 82. With respect to the answer Item B to ACRS Question 31, Amendment No 5, Appendix at Page 3 1-1 of the Appendix to the PSAR, describe in detail how and in what fashion the accumulation fatigue techniques will be incorporated into the design of the fuel clad. State each fact, cal-culation and assumption upon which you support your analysis, including but not limited to responses to all normal and abnomal cyclic conditions.
If in your answer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth coupletely the text of each such other reference or attach a copy. Answer s Cumulative fatigue techniques as discussed in Item B, Page 3 1-1, Amend-ment No 5, of the Midland PSAR will be used to ensure that the fuel cladding will not experience fatigue failure due to cyclic conditicas. The fatigue analysis for the fuel rod cladding will be performed during the detail design phase. As specified on Pages 3-4, Paragraph 3 1.2.4.2,
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" Fuel Assemblies" and Item B, Page 3 1-1, Amendment No 5, PSAR, the cu-mulative uscge factor will be limited to 0 9 The cumulative fatigue design technique used is essentially the same as that given in Para-graph N 415.2, Section III, of the 1968 ASME Boiler and Pressure Vessel Code. The fatigue design curve used to determine the fatigue damage for each stress / strain cycle is de' scribed on Page 31-1, Amendment No 5 of the Midland PSAR, under the heading " Cyclic Loadings and Strain Limits."
l l 4 l 82-1 1 i
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(..' As is true for fatigue design curves given in Section III of the 1968 ASME Boiler and Pressure vessel Code, the fatigue design curve for the fuel c1=AMeg is conservative in that it has a factor of safety of 20 on cycles or 2 on stress, whichever is more conservative. i V 82-2 l
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%. With respect to the answer Item D to ACRS Question 3 1, Amendment No 5, Appendix Page 3 1-101 of the Appendix to the PSAR, describe in detail each fact, calculation and assumption upon which it is concluded that the reference W. J. O'Donnell and B. F. Langer, " Fatigue Design Basis for Zircaloy Caponents," Nuclear Science and Engineering 20, 1-1/2(19@)supportstheconclusionthatthedesignrequirementswill be adequate to prevent gross cladding collapse. If in your answer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth empletely the text of each such other refer-ence or attach a copy.
Answer We assume that the "ACRS Question 3 1, Amendment No 5, Appendix Page 3 1-101" actually refers to DRL Question 3 1, Amendment No 5, Page 3 1-1. The interrogatory confuses Item D on Page 31-1 which refers to a design criterion that gross collapse shall be prevented, and the last paragraph on that page entitled " Cyclic Loadings and Strain Limits," which includes the reference. These are t.ro different subject areas. Item D elaborates on design limits set to ensure against fuel cladding short-time collapse and given on Pages 3 4, Parsgraph 3 1.2.4.2 of the PSAR. The O'Donnell and Langer paper " Fatigue Design Basis for Zircaloy Ccmponents," is con-cerned with establishing a fatigue design curve for Zircaloy. The re-
- nainder of the paragraph in which the reference occurs notes the basis for concluding that clad fatigue is not a significant problem.
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The results of fuel rod cl=AM ng collapse tests are described on Page 3-ld+, Paragraph 3 3 3 31. Additional information is given by the responses to Interrogativen 56, 57 and 71, which also pertain to clad collapse. I i
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Y 85 Describe in detail each fact, calculation and assumption upon which is based the conclusion at Page 3 6-1 of Amendnent No 5 to the PSAR that " seismic excitation forMidland may be more severe than that discussed in BAW-10008-Part 2, it is not expected to cause structural criteria to be exceeded." Set forth all relevent portions of EAW-10008, Part 2 and in addition, if in your answer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth completely the text of each such other reference or attach a copy. Answer Using the method of analysis given in proprietary Topical Report BAW-10008, Part 2, " Fuel Assembly Stress and Deflection Analysis for Loss-of-Coolant Accident and Seismic Excitation," and the seismic conditions for a site having more severe seismic conditions than those at the Midland Site, the fuel assembly met all the applicable criteria. For confirmation purposes only, the analysis will be redone, using the Midland site seismic conditions, during the detailed design phase. 1
._s 85-1 e _ _- . _ _ _ .
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- 86. With respect to the statement of the results of the comparison set forth at Page 3.6-2 of Amendment no 6 to your PSAR of the analog model and methods stated in BAW-10GO8 with the LOFT semiscale test results, describe in detail each such ec W son, specifying each fact, calcula-tion and. assumption thereof. If in your answer you make reference to otherthantextual(exclusiveoffootnote)matterinthePSAR,thenset forth completely the text of each such other reference or attach a copy.
Answer The statement of Pase 3.6-2 of Amendment No 6 to the PSAR refers to the methods stated in BAW-10008 used to predict results frem the LOFT semi-scale tests. In order to determine the validity of the program used to get LOCA leads en the internals, an analysis was perfomed for the blow-dcwn of the LOFT semiscale vessel. Specific tests which were correlated were Tests 711, 712 and 721. 2he analog ecmputer pregram that was used in the PSAR analysis was used, with the number of regions revised, to represent the semiscale vessel. A 10 region analog model was used to simulate the LCFT sc=iscale vessel as shown in Figure 1. These regions are denoted as circled numbers. The location of the pressure transducers used in the test are also shown en Figure 1. These are shewn by the notations P-1 through P-9 In the analysis, transient pressure curves were obtained for each control volume and compared to the measured pressures. By subtracting the values for s.- 86-1
the pressures, such as P-8 minus P-7, the core pressure drop can be ob-tained. When this was done en Test 711, the predicted peak core pressure drop was 37% higher than the measured peak and the predicted plenum shell pressure drop was approximately 50% higher than the measured peak value. In obtaining these it was assumed that the rupture occurred instantaneously in Region 10. This assumption is consistent with that made in obtaining the loads on the internals in the Midland Plant. Test 721 included eight 1.125-inch diameter holes in the upper plenum assembly to simulate the internals vent valves. The rupture size was 30% of the !+-inch nozzle area, giving a leak size of 0.0062 ft . Typical results for Test 721 are shown in Figures 2 and 3. Figure 2 shows the g ') predicted and measured pressures in the upper plenum of the model and Figure 3 shows the pressure in the half of the annulus between the plenum shell and the vessel opposite the break. In general, the correlation was good but was conservative. w' 86-2 9 9 - -
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- 87. With respect to your proposed system to continue core cooling in the event of a vessel failure, as described at Pages 4.8-1 and 4.8-2 of Amendment No 5 to the PSAR, describe in detail each fact, calculation and assumption upon which is based the assumption that:
(a) " rupture of the reactor vessel is assumed to occur at 10 minutes"; (b) "the probim relating to steam bubble formation does not increase in severity as long as the reactor vessel cavity is full at the time the vessel failure occurs"; and (c) " addition of nozzles to the reactor coolant cutlet piping no longer appears necessary."
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If in your answer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth completely the text of each such other reference or attach a copy. Answer (a) As stated in reply to Interrogatory 22 and as shown in Topical neport BAW-lool 8, 10 minutes represents the time of the most severe thermal shock transient. At this time the analysis used in BAW-lool 8 showed the assumed crack propagated to a mavimum depth of 1(:4 the reactor vessel wall thickness. Ac-cordin617, there is adequate margin of safety against the po-tential of a vessel rupture at any time and such a system is l l not required. m' 87-1
m (b) The Midland Plant reac or vessel has internals vent valves in-corporated into the design. The purpose of the vent valves is to provide a path for venting steam that could possibly get trapped in the top of the reactor vessel, in the event of a break in the reactor vessel inlet piping, thereby forming a so-called " steam bubble." The analysis performed for the PSAR shows that the vent valves prevent the formation of a steam bubble of sufficient pressure to inhibit core ecoling. This i analysis showed that if a water level was maintained in the reactor vessel downcomer annulus, core cooling was insured. This analysis is shewn on Pages 1b 50 and 1L 51 of the PSAR. The statement quoted in Item (b) is an extension of the logic and analysis presented in the PSAR. If the reactor vessel cavity were full of water, and there is a crack in the reactor vessel, then the cavity in effect is just an extensien of the reactor vessel downcomer annulus. The f.nternals vent valves would perform the same function of steam venting for this proposed accident just as they did in the analysis presented in the PSAR and core cooling would be maintained. (c) When it was agreed to provide the capability to install a sys-tem to continue core cooling in the event of a vessel failure, cognizance was taken of the problem in the PSAR with a state-ment made on Page L 15 that design provisions would be added to include het leg nozzles. Subsequent analysis shewed that l 87-2 i i i
3 as long as the reactor vessel building cavity was filled at the time of the postulated vessel failure, core cooling could be maintained 'sith the present emergency core cooling system, which uses nozzles mounted directly on the reactor vessel. Therefore, a clarifying statement was added on Page !+.8-2 via Amendment No 5 as follows: "The addition of nozzles to the reactor vessel outlet piping no longer appears necessary." .m 87-3
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- 88. With respect to the possibility of occurrence of thermal shock during normal operation of the proposed reactors as a result of in-jection of cold water to the reactor primary system from the core flood-ing system, describe in detail each fact, calculation and assumption upon which it is concluded that the as yet incomplete failure analysis contained in the PSAR of the core flooding system will demonstrate that no single active component failure will allow cooling water to inadver-tently reach the hot primary system metal. State when the proposed failure analysis will be completed. If in your answer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth completely the text of each such other reference or attach a Copy.
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- 89. With respect to the consequences of inadvertent addition of the core flooding water to the core, describe in detail each incident and the time of occurrence of all factors, including but not limited to com-ponent failures and emergency system activation, specifying each fact, calculation and assumption thereof, which mitigate and propagate the worst possible chain of events describing such chain of events, includ-ing but not limited to primary vessel rupture. If in your answer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth completely the text of each such other reference or attach a copy.
V 88-89-1
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s Answer There is no possible way that the reactor vessel can be suojected to a thermal shock from the core flooding system during normal operation of the plant. The reactor coolant system is at 2190 pai and the core flood-ing tanks are at 600 psi. This pressure differential definitely prohibits this possibility. During normal cool-down, after the reactor has been made 1% suberitical, as both temperature and pressure are decreased the core flooding system is administratively isolated by the closure of a power-operated valve. If these valves are not closed by the time system pres-sure reaches 650 pai, redundant alarms are sounded in the control room. In the event the operators ignore the alarms, the tank will gradually empty into the reactor coolant system as the system is depressurized. As it is not possible to reduce reactor coolant system pressure rapidly, it is also not possible to drain the co're flooding tanks rapidly. i I 88-89-2 l
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- 91. What is the minimum primary system break size for which (1) the reactor cavity and (2) the reactor building will not maintain structural integrity? Describe in detail each fact, calculation and assumption upon which you base your answer. If in your answer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth completely the text of each such other reference or attach a copy.
Answer (1) While the reactor cavity design pressure is 250 psi, the re-actor cavity has the capacity to withstand a pressure differential of 1000 psi without loss of function. Using ne same input assumption as t) is presented in answer to AEC question 5.1.20, the corresponding break size causing a 1000 psi reacter cavity differential pressure is esti-mated to be 9 2 square feet. (2) The maximum calculated reactor building pressure would occur for a postulated 5.0 square feet reactor coolant piping rupture. The calculated pressure for larger breaks, up to and including 14.1 square feet, was less than for the 5.0 square foot break (PSAR Sec 1k). It is therefore concluded that no postulated primary system break vill produce a reactor building pressure greater than that for the LOCA 5 0 square foot break. Thus, the reactor building design pressure of 67 psig is conservative. Since the structural design of the reactor building
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91-1
m considers a prudent margin before the :naterials reach yield point, and since the reactor buildings vill be safely tested to a pressure in ex-cess of the design accident pressure, the structural integrity of the reactor building will not be violated by any postulated primary system break size. e W 91-2
O 94 With respect to the cenditions within the reactor building fol-lowing a LOCA, describe in detail the pacedure, referred to at Page 6.11-1 of Amendment No 5 to the PSAR, which will be employed for sampling the recirculation water during the long-term mode of core cooling to monitor boron concentration. If in your answer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth completely the text of each such o^her ref-erence and attach a copy of each such other reference. Answer A sample connection will be located immediately downstream of the decay heat cooler on each of the tdo low-pressure safety injection trains. These sample connections will be piped to a sample collec-tion station so located and shielded as to be accessible after a LOCA, similar to that schematically shown on PSAR Figure 9-11. The sample can be transported frca there to a suitably located and shielded laboratory for analysis. Further details regarding the location of the sampling station and laboratory, and the sampling and analysis procedure, will be devel-oped durinc detail design.
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- 06. With respect to the ans,'er to ACRS Question 7.7, set forth at Pages 7 7-1 and 7 7-2 of Amendment No 5 to the PSAR, give a more de-tailed description of each test which has been perfomed and its results or which will be performed regarding equipnent which has been or will be tested at the Palisades Plant. State what alternatives for testing you have planned in the event such tests not yet perfomed can-not be perfomed at the Palisades Plant. If in your answer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth completely the text of each such other reference or attach a copy.
Answer None of the testing referred to in the answer to question 7.7 was or is intended to be accomplished at the Palisades Plant. Testing of Palisades Plant equipnent was seccuplished at the equipnent manufacturers' plants or at separate testing facilities. Similarly, types of Midland Plant equipnent which were not environmentally tested for the Palisades Plant or otherwise qualified will be tested by Midland Plant equipnent manufacturers or separate testing facilities, not at the Midland Plant.
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97 With respect to the answer to ACRS question 712, set forth at Page 7.12-1 of Amendment No 5 to the PSAR, describe in detail "the data available from simile core configurations of emperable size" which will be evaluated "to verify or disqualiff" the stated reliance upon "out-of-core" Justrumentation "for safe and reliable spatial power indication upon the core." Include within your answer your analysis of such data, if already analyzed, specifying each fact, calculation and assumption thereof. If your analysis is not yet complete, state why not and when you propose to complete such analysis. If in your answer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth completely the test of each such other reference or attach a copy. l Answer As explained in Section 13 3 and listed in Table 13-3, the external and in-core nuclear instrumentation systems will be calibrated at various power levels. Several years prior to initial criticality and power escalation testing at the Midland Plant, similar type tests will have been performed and evaluated on several B&W reactors which will be the same size and configuration as the Midland Plant. The first of these < l will be Oconee No 1. The following are direct quotes frem the Duke l 1 Power Capany, Oconee Nuclear' Station, Unita 1, 2 and 3, Final Safety Analysis Report, Docket No 50-269, -270, and -287 filed on June 2, 1969 97-1
Y "Section 13 3 Posteriticality Test Program The Posteriticality Test Prcgram will be performed to provide assurance that the plant is operating in a safe and efficient mann**. Systems and ecmponents which could not he operationally tested prior to initial criticality vill be tested during the Posteriticality Test Program te verify reactor parameters and to obtain information required for plant operation. A stamcary of the posteriticality testing is contained in Table 13-2." w 97-2
-s Table 13-2 Posteriticality Testing Summary Iggt Obiective
- 1. Single CRA To determine worth of the CRA's calculated Worth Measurement to be ther most reactive.
- 2. Control Rod Group
- To determine differential and integral re-Calibrat,1on , activity worth.
- 3. Baron Worth To determine the boren reactivity worth.
Calibration ,
- 4. hNderator Temperature To measure the reactivity effect associnted Coefficient Measurement with a change in reactor coolant temperature..
- 5. Excess Reactivity To measure the excess reactivity of the Measurement core to verify the predicted shutdown margin.
- 6. Reactor Coolant Pump To measure the reactor coolant system flow Flow Test under hot operating conditions.
- 7. Reactor Coolant Flow To verify the flow rate decay for various Coastdown Test reactor coolant pump trip combinat[ons.
- 8. Unit Startup That To determine performance characteristics of the unit during. the inicial period of opern-e tion at low power.
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- 9. Unic Heat Balance ' To determine the heat balance at various Test power levels to provide the necessary daga for calorimetric calibration of the nuclear instrumentation and verification of the reactor coolant system flow rate instrumen-tation.
- 10. Biological Shield Survey To verify the adequacy of radiation shield-ing.
- 11. Natural Circulation To demonstrate natural circulation removal
- of decay heat in the event of 3oss of all Test reactor coolant pumping capability.
- 12. Integrated Control To identify unit response to guide adjustment System Test of the integrated control systen to maintain
. stability during steady state and load trans-fent conditions.
- 13. Turbine Generator To verify cperation of the turbine-generator Operation and Test and its associated controls and trips.
- 14. Xenon Rasetivity Worth To determine reactivity associated with xenon as a function of power and during xenon depiction. ;
13-8
. , , Table 13 Continued i Posteriticality Testing Summarv -
- 15. Reactivity Coefficients To determine reactivity coefficients asso-at Power Measurement cisted with baron concentration, moderator temperature, and control rod assemblies at various power levels.
- 16. Power Coefficient To measure the power coefficient at Measurement various power levels.
- 17. Core Power Distribution To measure flux and power distribution Measurement ,
utilizing incore instrumentation.
- 18. Unic Load Steady State '
To determine unit response characteristics and Transient Test to step and ramp load changes.
- 19. Unic Loss of Elactrical To verify the response of the reactor and Load Test auxiliary systems on separation of the generator from the transmission system.
- 20. Turbine / Reactor To verif, unit response during trip of tur-Trip Test bLne or reactor.
- 21. Unit Power Shutdown To verify the adequacy of the operating Test procedures used in shutting down the unit g from power.
- 22. Induced Power Oscillation To measure the core power distribution Test following a deliberate perturbation by con-trol rods.
19 v 9 0 13-9 -
- 98. With respect to the answer to ACRS Question 7.20, set forth at Page 7.20-1 of Amendment No 11 to the PSAR, describe in detail each fact, calculation and assumption upon which it is concluded that a i
diverse backup to the low reactor coolant system pressure trip will be provided by either the void shutdown mechanism or the power / flow 4 comparator. If in your answer you make reference to other than tex-tual (exclusive of footnote) matter in the PSAR, then set forth com-pletely the text of each such other reference or attach a copy. Answer At the time the statement on Pa6e 7.20-1 of Amendment No 11 was made on 1 May 1JTO, it was anticipated (not concluded) that the diverse backup to the low-reactor-coolant-system-pressure trip would be pro-vided by the void shutdown mechanism or the power / flow comparator. Since that time, and as a result of discussions with the ACES and the DEL staff (see Section 13.3 of the .Sumary of Application and Section , 8.1 of the Safety Evaluation), a decision has been made to provide a high-reactor-building-pressure trip as the backup to the icw-reactor-coolant-system-pressure trip. I 1 l l 98-1
- 99. Describe separately and in detail each consequence, specifying each fact, calculation and asstnuption thereof, resulting from (a) inadvertently opening the single valve between supply lines of pro--
posed Midland Units 1 and 2 and -(b) inadvertently opming the valve between the main steam headers of propossi Midland Unit 1 and Midland Unit 2 when the units are operating at different power levels. DO NOT consider an answer ctamplete by stating as was done in response to an identical question by ACRS that such conditions are administratively impossible to occur. However, regarding said response to ACRS, state , in detail the a44nistrative procedure referred to and analyze each i 1 possible inadvertent malperformance of such procedure. If in your an- j swer you s he reference to other than textual (exclusive of footnote) ! s matter in the PSAR, then set forth completely the text of each such other reference or attach a copy. Answer As stated in the answer to DRL Question 7.10, Amendment No 5, "The detailed development of the system has not been finalized; however, the design bases include the requirement of separation between NSS systems as well as single failure analyses." There will also be an analysis. of the effects of inadvertant opening of the valve between (a) the (feedwater) supply lines and (b) the main steam headers, between Units 1 and 2.
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I l l 1 1 O l This analysis will consider the various modes of operation including ' operation of units at different power levels. i It is asstaned that the "... response to an identical ACRS question" referred to is the response to Parts k and 5 of DEL Question 10.1, PSAR Amendment- No 5, which states: "'the valves between the main steam and feedwater supply lines of Units 1 and 2 are normally locked closed and require administrative procedures for their operation." This answer also refers to the answer to Question 710, the last sen-tence of which states "The transfer will normally be accomplished in the manual mode with interlocks to assure proper sequence of the pro-cess steam transfer." s' The " interlocks" referred to here are control interlocks so that, while noministrative action, including the use of a key to unlock the cross-connecting valves, is required to initiate transfer, a nonfail-safe failure of a control interlock must occur to allow inadvertant opening I of a cross-connecting valve. t 99-2 l l l
T 100. Describe in detail the proposed location and design of the pro-posed process steam lines and state what industry or other codes, if any, shall be adhered to in their design construction. If in your an-swer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth completely the text of each such other reference or attach a copy. Answer The proposed location of the process steam lines is shown on PSAR Fig-ure 1-1, Section 1, Amendment No 5, " Site Plan"; Figure 1-8, Section 1, Amendment No 16, " Station Arrangement"; and Figure 10.1-3, " Isolation and Reducing Valves in Process Steam System to Dow Chemical," included with answer to DRL Questica 10.1, Amendment No 5. The following is a preliminav description of these lines, which may be subject to minor changes as the design proceeds. (a) Ten-inch high-pressure steam lines branch off a main steam line of each unit at a point on the roof of the auxiliary building, and l join into a single header (Figure 10.1-3) which mns north from the , l plant to the tertiary heat exchanger system (or reboilers). The tertiary heat exchanger system is described in reply to DRL Ques-l tion 11.00-1, criginally answered in PSAR Amendment No 9 and revised I in Amendment No 17. 100-1
.-3 i. High-pressure process steam produced in the tertiary heat exchanger system is piped in a header north to the site boundary as shown on Fig- , ures 1-1 and 1-8. 1 These lines are designed for a ==W-= steam flow of 600 3000 pounds per hour (400,000 pounds per hour normal) and for the following m=H== con- ) ditions: Supply to tertiary heat exchangers: 1050 psig and 6000F. Process steam from tertiary heat exchangers: 660 psig, saturated. (b) tro 46-inch low-pressure lines, one from each moisture sepa-rator on Unit 1 turbine, run vest along the turbine deck and then ncrth out of the turbine building (Figure 10.1-3) to the tertiarf heat ex-changer system. Low-pressure process steam produced in the tertiary heat exchanger sys-tem is piped in 46-inch headers north to the site boundary as shown in Figures 1-1 and 1-8. These lines are designed for a mmW== steam flow of approximately 1,800,000 pds per hour each, and for the following ==*=m con-ditions: Supply to tertiarf heat exchangers: 368 psig at 505 F. Process steem from tertiary heat exchangers: 155 psis, ' saturated. V 100-2
(c) 28-inch branch lines frem each unit main steam line pass through backup low-pressure reducing stations on the am414av build-ing roof. Downstream of these stations these lines enlarge to 142 inches and tie into the low-pressure supply lines ((b) above) as shown on Figure 10.1-3 The "HP Reducing Stations" referred to on this figure were deleted as stated on Page 11.00-1, Amendment No 17. All process steam piping vill be in accordance with the Code for Power Piping ANSI B31.1.0, applicable portiens of the ASTM standards, and the Pipe Fabrication Institute (PFI) Standards. Frm the site boundary, the pro:ess steam lines will be piped by Dow north across a pipe bridge 3panni.ng the river to Dcw's distribution system. Provisions for expansion loops and anchorages have not been designed at this time, pending more detailed development of the tertiary heat exchanger system. v 100-3
l l 101. Foreachradienuclidewhichwillbedepositedand/orconcentrated ' in the cooling pond water, state your estimate of its mari mmi and normal i 1 levels, in curies per milliliter, of deposit and/or concentration both I for normal and abnormal operating conditions of the proposed reactors. Describe in detail the method which you employ to determine these esti-mates. If in your answer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth completely the text of each such other reference or attach a ecpy. Answer There will be no radioactivity released to the cooling pond at Midland, as noted in response to AEC Question 2.10, a copy of which is attached hereto. i
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s 2.10 We understand from oral statements made at our meeting of December 5,1969, that the activity in the cooling pond will be limited to the 10 CFR 20 Appendix B Table 2 Column 2 values. Confits this understanding. If the cooling pond were operated continuously with activity levels corresponding to Table 2, Column 2, of Appendix 3 to 10 CFR 20, what amount of Cs-137
. would settle to the bottom of the pond throughout plant life? Estimate the potential Cs' 137 and tritium concentrations in the aquifers and at the nearest wells in both aquifers at the end of plant life.
Answer: A decision has been made to discharge routine liquid radwaste at 10 CFR 20 MPC to the Tittabawassee River and not to the cooling pond. Under extreme or emergency conditions, excess liquid radwaste could be discharged to the pond and at a level which will assure the cooling pond concentration to be well below 10 CPR 20 MPC levels for drinking water. If used in this way, a detailed analysis of the effect of radionuclide concentrations and potential aquifer contamination vill be available and reviewed for acceptance with the DRL staff prior to discharge into the pond. In addition, several wells will be' drilled in locations suitable to allow conitoring of local aquifers throughout plant life. 1 i 1 l 1 I l 2.10-1 Amendment No. 7 1/19/70 m
s 102. Describe in detail each fact, calculation and assumption which forms the basis for the conclusion that " pond seepage is estimated to make a negligible contribution to the plant radwaste discharge to the Tittabawassee River" (Page 11.2-1 of Amendment No 6 to the PSAR). If in your answer you make reference to other than textual (exclusive of 1 footnote) matter in the PSAR, then set forth completely the text of each such other reference or attach a copy. 1 l Answer As noted in the response to Interrogatory No 101, there will be no re-lease of any radioactivity frcza the Midland Plant in the Midland cooling pond. l 102-1 i m
/O, 103 With respect to operation of the proposed Midland Units, when makeup water cannot be taken from the Tittabawassee due to low flow ;
conditions, describe in detail the evaluations which form the basis for the limits on liquid radwaste release which would be employed. Assume in your answer that askeup water cannot be taken for a period of 100 consecutive days. If in your answer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth ccupletely the text of each such other reference or attach a copy. Answer As noted in Amendment No 20 to the Midland PSAR and as also noted in the response to Interro6atories No h2, 101 and 102, liquid radioactive vastes, with the exception of laundry wastes, will be recycled within the Midland Plant and will not be released to the Tittabawassee River, under normal operating conditions. 103-1 l l l t ( L ~
10h. With respect to the occurrence of a radwaste system failure, state: (a) What are each of the possible modes of failure (eg, missile generation and tank fracture) for this- system which would result in a release of the radioactive contents; ) i (b) With respect to each such possible mode described in (a) above, ) 1 state in detail what safety measures will be taken; and 1 (c) ' Separately, in the event of such a failure due to each possible mode described in (a) above, state what procedures will be followed to limit exposure to the population from the radio-active releases and to contain them. If in your answer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth ecmpletely the text of each such other reference or attach a copy. Answer (a) As described in the answer to AEC Question 113 and elsewhere j l in the PSAR, the radwaste system is located within a Seismic Class I, I tornado- and flood-protected structure, and therefore only failures of portions of the system containing gaseous radwaste would result in release of radioactivity to the atmosphere (via the auxiliar*y building ventilation l l 10h-1
system and vent stack).* Possible modes of failure for these portions of the system are piping or tank rupture. (b) The radwaste system is desi6 ned in accordance with applicable nuclear codes. In addition, in accordance with Appendix 5A of the PSAR, portions of the system whose failure would result in doses at the site exclusion radius in excess of 10CFR20 limits are designed and constructed in accordance with Seismic Class I criteria; the vaste gas surge tanks and decay tanks are presently so designated as shown on PSAR Figure 11-3 (c) As described in PSAR Section 9 12.2.2, exhaust ventilation air from potentially contaminated areas is directed through charcoal filters upon detection of high activity levels in these areas.
\
l
*In addition, the clean liquid wastes holdup tanks c.re located in the reactor buildin6; due to the gaseous centainment afforded by this build-ing, any gaseous release to the at:nosphere from tank failure would be negligible.
0 l 10h-2 i s . - . _ _ _ . , ,
105 Describe in detail each consequence, specifying each fact, cal-culation and assumption thereof, of an inadvertent control rod removal when all reactor coolant pumps are'not operating (ie, startup accident when coolant pumps are not operating), regardless of whether you believe such a circumstance possible. Your answer should include but is not limited to a description of clad damage which could result if adequate cooling is unavailable to remove energy generated by the transient. If in your answer you make reference to other than textual (exclusive of footnote) matter in the PSAR, then set forth ccanpletely the text of each such other reference or attach a copy. Answer Calculations of the startup accident indicate that the reactor thermal power during the transient never exceeds the trip value for the neutren flux. The neutron flux trip value produced by the neutren power / reactor coolant flow trip di'scussed in Section 7.1.2.4 (attached) is zero because no reactor coolant pumps are in operation. Thus, the maximum possible themal power for a single control rod group startup accident when no reactor ecolant pumps are operating would be less than the power at which trip occurs, which is essentially cero (less than 1% of rated power). This neglects the trip that would be initiated by hi6h startup rate well belcw the power range. Sufficient natural circulatien capability is l available to remove at least 5% of rated thermal power if none of the ' reactor coolant pumps are operating. Thus, there will be no cladding damage as a result of this accident. 105-1
- \
n
+e 4 .n The operation of the engineered safeguards channels and the trip relays fonn-() ing the system logic is described in 71.2.2.
The high order of system redundancy assures compliance with the single failure criteria of 7 1.1.2.1. T.1.2.4 Summary of Protective Actions The abnormal conditions that initiate a reactor trip are as follows: Steady-State Trip Value or Trip Variable No. of Sensors IDmal Fange Condition for Trip Neutron Flux 4 0-100% 107 5% of full (rated) power. NeutronFlux/ Reactor 4 Flux 2 to 4 Pumps (1) Loss of one oper-
' Coolant Flov 16 Reactor Coolant -
ating coolant pump Pump Monitors motor and reactor neu-2 Flow Tubes tron power exceeds pre-detemined level. (2) Loss of one oper-
, ating reactor coolant pump motor in each icop
. ) and reactor neutron power exceeds 50% FP.
(3) Loss of two over-ating reactor coolant pumps in ene loop. (4) Ratio of reactor
' neutron power to total reactor coolant ficv exceeds 1.07 Start-Up Rate 2 0-2 Decades / 5 Decades / Min Min Reactor Coolant 4 2,120-2,250 2,350 Psig Pressure Psig 2,050 Psig Reactor Outlet 4 520-603 F 610 F Temperature The reactor trip functiens of the power / flow conitor logic are summrized es fellows:
l ( i l l 7-lh ]
.- l
.o - - . _ . . ,
.i l Trip Variable No. of Sensors Neutrori Flux = $ 12; 4 Flux Channels Reactor Coolant Flow = IF 2 Flow Tubes; 8 AP; h IF No. of Operating Pumps = P, 16; 4 Pump Monitors Reactor Trip e
(a) Loss of One Pump and & > Xf (b) Loss of One Pump in Each Icop and & > 50% (c) Loss of Two Pumps in One Loop (d) ($ > 1.07 IF) . Predetermined neutron pcver level to be specified during detail design. Actions initiated by the engineered safeguards actuation system are as follows: 3 Steady-State - Action Trip Conditien Normal Value Trip Point High-Pressure Low Reactor 2,120 - 2,250 psig 1,500 psig ) Injection Coolant Pressure l or - i
, High Reactor Atmospheric Building Pres-sure Low-Pressure Very Low Reactor 2,120 - 2,250 psig 200 psig Injection Pressure !
or High Reactor Atmospheric i Bailding Pres- I sure Start Reactor High Reactor Atmospheric Building Emer- Building Pres-gency Cooling sure Unit and Reactor Building Iso-lation Reactor Building Hi 6h React ~or Atmospheric Spray Building Pres-sure Reactor Building High Reactor Background Isolation Building Ra-(Type II) diation T-15 Amend =ent No. 5 . , 11 l't /40 l e - . , ----l
O 106. With respect to the answer to ACRS Question 13 3 7 set forth at Page 13 3 7-1 of Amendment No 5 to the PSAR, describe in detail each development and refinement of the turbine which " prohibits specific missile parameters at this time." State when these parameters wi'1 be designated. In addition, when the final critical structure design is complete, what changes, if any, in the " developments and refinements" of the proposed turbine would result in inadequacy of the aforesaid design. Set forth au relevant portions of GE Report TR67AI211 "An Analysis of Turbine Missiles Resulting From Last Stage Wheel Failure" and, in addition, if in your answer you make reference to other than textual (exclusive of foctnote) matter in the PSAR, then set forth ccan-pletely the text of each such other reference or attach a copy. Answer The referenced Report TR675SI2n (not TR67AI211) is attached. The parameters referred to in answer to Question 13 3.7 were nore detailed covering the actual turbines, which will have 43-inch last stage blad-ing instead of 38-inch blades, upon which the reference report was based; although it is noted in Section 1-A of the referenced report that the concrete slab thickness penetrated is increased by only 1% for the longer blades. Subsequent information received from General Electric included the re-quired details for the 143-inch blades and the probability of occurrence
/
1c6-1
,~., of external missiles; together with turbine design development, this canpletes the scoping of turbine missiles for the Midland plcnt. The only critical structural design question r - 'nina is whether the probability of a potentiany damaging missile enterin6 the fuel pool is negligible or whether additional protection is needed over the pool. If required, this additional protection will be provided. The detail design of other critical structures for the postulated " worst case" missile is routine and is not expected to introduce any significant changes. 106-2 i
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GENERAL (&3 ELECTRIC SchENECTAD't, N' 7; YOR5 N e g
- FOR USE OF G-E EMPLOYEES ONLY GENER AL h ELECTRIC TECHNICAL INFORMATION SERIES Title Page AursoR sualcci No. TR675L211 E.E. Zwicky,Jr. Turbines, Impact
""' AN ANALYSIS OF TURBINE MISSILES c.c. Cuss i RESULTING FROM LAST-STAGE WHEEL FAILURE covr. cuss None REPRODUCIBLE COPY FIL(Q Ar Technical Report Library, Building 155, Room 219 NO, PACES 72
SUMMARY
- The report describes calculations made to evaluate the potential damage which could be done by hypothetical turbina missiles in the o
unlikely event of the fo!!ure of the last-stage wheel of a nuclear machine. It Is concluded that the missile kinetic energy absorbed by the turbine casing is es large er larger than previously assumed, and that a 120' wheel fragment is potentially more damaging than either a 90' or 180 fragment. The potential damage increases slightly with last-stage wheel size. The concrete slab thicknesses penetrated for - the 38", 43" and 52" wheals are in the approx! mate rotics (1):(1.01): (1. 08).
- KEY WOR 05 INFORMATION PREPARED FOR Turbine Ennineerino, LST-G Dept.
TESTS MADE BY AUTHOR E. E 1 1ck y. A COMr0NENT Motorials & Processes Laboratory I APPROVED l _
e (~ s DISCLAIMER OF RESPONSIBILITY This report was prepared as an account of research and develop-ment work performed by General Electric Compcny. It is being made available by General Electric Company without considera. sinn in the interest of promoting the spread of sevbnical knowl. ealgr. Neither General Electric Company nor the individualausbar: A. blakes any warranty or representation, expressed cr implied, with respect to the accuracy, completeness, or usefulness of the information contained in this report, or that the use of any information disclosed in this report may not infringe privately owned rights; or B. Assumes .my responsibility for liability or damage which may result from the use of any information disclosed in this report. O s t.N 2 i c- :
GENE 2AL ELECTRIC CcMPA :V
' mtothA u.su TECHNICAL INFORMATION SERIES CONTENTS PAos TR67SL211 CONTENTS OP REPORT O NC. PAG 4S TEXT NO. CHARTS I4 CRawlNo Nos. None >
PHOTO NOS. None s CISTRf 8UTION . M & P LAB. LG & M Technical Report Library 219 K. Seeliger 5 C. W. Elston 115 E. J. Flynn 201 SST l LST-G A. W. Rankin - Fitchburg , l R. E. Brandon - 273-363 MSTG 1 J. E. Downs - 273-328 L. H. Estebrook - 273-362 C. E. Evans - Lynn River C. Schebtach - 273-298
- 0. P. Timo - 273-362 OTHERS V.R. - 273-259 S.Vandenburgh Mu Ick - 273-259 APED J. F. Proctor Novel Ordnance Leberatcry J. E. Corr -
- Son Jose White Ook Silver Spring, Merylcnd 20910 l
l e' v
1
. l TR675L211 Page li Table of Contents Page
! Purpose and Background of Turbine Missile Study 1 A. Summary I
- 8. Overall Description of the Missile Problem 1 11 Idealization of impact Problem and Description of Analysis 4 Ill Results of Calculations and Conclusions 9 A. Frogment Properties 9 B. Fragment Energies Outside Turbine Casing 9 C. Comparison of Ccst Iron and Steel Diophragms 9 D. Significance of the "No. 2nd Phase impact" Cases 10 E. C wnporison of Various Wheel Fragments 10 F. Portition of Energies After the Inner Casing Impact 10 G. Comparison of 2nd Phase impact Models 11 H. Air Friction Effects 11
- 1. Effect of Non-Zero P.estitution Coefficient - 12 J. Secondcry Missiles from Ring Fragments 12 K. Summary of Data to be Furnished to Customers 12 IV Analyses A. Wheel Frogment Motion Before impact 13 B. Wheel Fragment Position at impact on Inner Ccsing 14 C. Impact of Wheel Fregment on inner Casing, First Phase 18 D. Impact of Wheel Fragment on Inner Casing, Second Phase 24 E. Energy Relations in Terms of impulse Values Relative Velocities; Deformation Energy Limit 31 F. Energy losses During inner Casing impact 34 G. Outer Casing Penetration 36 H. Energy Losses Due to Air Drag 38
- 1. Penetration of Concrete Slch 41 J. Wheel Fregment Properties 43 K. Wheel Bursting Speed 45 L. Numerical Dato Used for Calculations 46 M. Computer Program Description 48 V. Ac'<nowledgements 55 VI References 56
o . TR675L211 Page 111 LIST OF TABLES TABLE Page I Wheel Fragment Properties 57 11 Fragment Translational Energies Outside Turbine Casings '58 Ill Summary of Data to be Furnished to a Customer 59 D e r ( D I l t .
l
. TR675L211 Page i
- l. Purpose and Background of Turbine Missile Study A. Summary The report describes calculations made to evaluate the potential damage which could be done by hypothetical turbine missiles in the unlikely event of the failure of the last stage wheel of a nuclear machine.
,The previous work on this sub[ect, reported mainly in letters and memerenda (by J. E. Corr, APED and E. E. Zwicky, LST-G), has been utended in two principal ways, From this, it is concluded that the missile kinetic energy absorbed by the turbine casing is as large or larger than previously assumed, and that a 120* wheel fragment is potentially more damaging than either a 90* or 180' fragment. Finally, the potential
. damage increases slightly with last stage wheel size. The concrete slab thicknesses penetrated for the 38", 43" and 52" wheels are in the approximate ratias (1):(1.01):
(1. 08). . Although the analysis.Is fairly extensive, o great many approximations and simplifica-tions were made, it has been the Intent to find the " worst case"; to the extent this has been done, the results should be Interpreted as upper bounds. B. Overall Description of the Missile Problem An extensive safety cycluation is made of many aspects of a nuclear power plant during its i.1itial design. Much of this has to do with the engineered safnauerds and safety systems required to protect the reacter core and/or to prevent leu'<cqn of radio-active material which could endanger the public. Part of thme studies orn concerned with possible physical damage to the containment which wrrounds the reacter vessel, as well as to the centrol and safety systems themselves. Eccthquakes, tornades and ether natural phenomena, as well as power plant equipment foilure, are censidered in this review. - The main turbine-generator, since it is the largest rotating equinment, must be considered as a potential source of damaging missiles. Both turbine and generator failures have occurred in the past, none however in nuclear plants (R eference (2)* Includes a fairly cornplete !!st with references to their description). Some of these occurred at er near rated speed; through progress in design, better materials and quality control, mcre rigorous acceptance criteria and improved machine operatien, the likelihood of burst failures near running speed has been substanticily reduced. Other failures have cccurred between 125% and 170% of running speed; careful analyses of these failures have led to design, inspection and testin9 Frecedures to substantially reduce the possibility of overspeed causing a failure of modern units.
' Numbers in parentheses refer to references.
O
. ~
s TR675L211 Page 2 Hypothetically, if the turbine reached 170-180% speed, failure of a last-stage wheel.could produce fragments having significant mass and velocity. Such an occur-rence would require the following consecutive events:
- 1. Opening of the generater breaker to disconnect the machine from the power system.
- 2. Failure of the normal speed governing system to close the steam admission volves.
- 3. Failure of the independent overspeed governing system to close the emergency stop valves.
- 4. Failure of the high vibration and loose parts to damage the turbine and its steam path sufficiently at 130-150% speed to prevent further acceleration.
- 5. Failure of the severe generator damage through thrown windings and probable retaining ring failure at 150-170% speed to broke the turbine.
Postulating that the extreme overspeed is reached,'ond the last stage wheels failed, the fragments would tend to be thrown radially from the turbine. Those moving below the machine centerline would be Intercepted by the icundation structura and posa no f.czard to the reactor containment. Depending on the plant layout, fragments traveling at a small angle above the horizontal may strike the side of the centsinment provided that they are sufficiently energetic to penetrate intervening cbstacles such as personnel shleid walls and building structures. At higher angles, the fragments wculd fly herm-lessly over the top, Fragments released almost vertically, if directed precisely, might fall on top of the contcInment. The low angle missiles are potentially the mere dangerous, sinco seme of the initial energy of the high ongle missiles is obscrbed by air drog in their Icng flight. The containment design, however, either for radiction shielding or strength, may result in different impact resistances on the sides than on the top. Therefere, both misilles should be evaluated. The above brief descript!cn leads to the questions which this repcrt tries te answer. Given that a wheel failure occurst
- 1. How much energy will the frcgments lose in possing through the casing and exhaust hecd cf the low pressure turbine?
- 2. Which shape of fragments will have the greatest penetrating effect citer having passed thrcush the casing? j 1
9 4-
, TR675L211 Page 3
- 3. What are the relative penetration effects of fragments from turbines having 38", 43" and S2" last-stage buckets?
Note that for each of these questions we cre seeking a " worst case" result. Namely, it is presumed that wheel failure actually occurs at the maximum speed which the wheel is capable of reaching. The results should therefore be viewed in a statistical sense; of all the accidents which could happen, this is among the least likely. Failures at lower speed would produce frcgments'with lower energy and therefore les capability for perforation. The calculation method presented in this document is to define the size and energy of missiles escaping from the turbine. This information is provided so that others con provide suitable protection from these missiles where necessary. h l l i l b
. TR6751.211 Page 4 p
II. Idealization of Immon:t Problem and Description of Analysis it Is clear that the answers to the questions posed in the previous section should properly be obtained in a probability sense. Many factors could Influence the result, depending on the exact orientation of the parts, the amount of prior damage, the strength of the parts, etc. This approach was not attempted. Instead, a deterministic model was synthesized to preserve the mc[or features of the physical problem but to emphasize the " worst case". The principal features of the idealization are described below. The process is conceived to have five sequential steps, each cf which entalls certain assumptions. The assumptions and discussion of them are listed below fer each of the steps. Refer to Fig.15, pg. 73, for a sketch identifying the major parts. A. Before initial Imooct
- 1. Only last-stage wheels considered.
Based on the fragment sizes and the nature of the surrounding structure, the last-stage wheels are considered the most dangerous. Typically they are also the most highly stressed and hence the most probable candidates for failure.
- 2. Wheel fractures into 2, 3 or 4 equal pie-shaped fragments.
These cases are not equally likely. Most of the ovellable data suggests either the 2 or 4 fragment case. Three 120' fragmerjts, however, give a " worst-case" result. The wheel failure speed was taken as that obtained if there were no undiscovered bcre defects. However, the failure speed was based on the minimum specifica- l tion tensile strength. The expected scatter In fa!!ure speeds for sound wheels ! would not Increase the assumed value by more than a few percent. If un- I discovered defects were present, failure could occur at lower values. I
- 3. Wheel fragments Include bucket dovetall, but not the bucket vanes.
Failure speed was assumed to be high enough that all the last-stage vanes hed failed near the vane roots before wheel burst occurred. No significant damage to the surrounding structure was escribed to these failures. This is based in part on field reports of previous partial vene failures and also on a qualitative review of the proboble vane motion. l l
, Note, however, that the failure speed in item 2 is on estimate based on wheel stresses which include the effects of vanes. While this is inconsistent, one could argue that if the vanes failed (and their predicted failure speed is less than that for the wheel), wheel failure could not occur even at theoretical maximum l runaway speed because of the decrease in centrifugal stresses.
l l
# , p
. TR675L211 Page S s
- 4. There Is no wheel fragment interference on failure.
Calculations show that the 180* fragments, but not the 90* or 120' fragments, will strike the turbine shaft or one another. Since such secondary Impacts can only reduce the predicted fragment energies, and since the 180' fragment Is not the most dangerous, r:o attempt was made to analyze the effects.
- 5. Diaphragm Inner web was Ignored.
The diaphragm web, being cost-Iron, was casumed to fracture without changing the energy or motion of the wheel frogment. The Impact is therefore assumed to concentrate on the outer ring of the dlophragm. l B. First phase of the inner c'esing Impact (before initial ring failure) 1
, 6. Plane impact.
l The v< heel rim is assumed to strike the outer diaphragm ring in the original i i plane cf the wheel. Forces and reactions in the plane of the wheel were Included in the analysis. Axial forces and reactions were ignored. This greatly simplifies the analysis and should include the principal effects. Approximate calculations were made to verify that the wheel fragment would not move axially and "miss" the Inner casing.
- 7. Wheel fragments Impact on inner casing simultaneously. !
To preserve the symmetry of the analysis, the 2, 3 or 4 fragments were assumed ! to behave Identically. It is believed that small deviations would not alter the ! results more than a few percent. 1 i
- 8. Inner casing ring treated as a rigid body for tangential forces.
The tangential or frictional forces, assumed to be symmetrically applied to the ring, were resolved into a torque about the machine center-line. Elastic waves and distortlens were neglected because the velocit!es are relatively low. The mass treated as the ring was Isolated from the remaining part of the turbine Inner casing. This reduced the resistence to tangential motion somewhat, but the overall effect In small.
- 9. Effective ring mass for radial ferees taken as 1/4 of the total ring mass for each fragment.
The radial fcrces acting on the ring are applied too quickly to accelerate the entire ring in a " breathing" mode. Mobility analysis of a ring acted upon by i. sinusoldaily varying point loads Indicates that approximately 25% of the total ? ..
TR675L211 Page 6 ring mass is effective at the load point. This obviously would not cpply if more than 4 fragments impacted the ring.
- 10. Local elastic distortions were neglected.
The elastic forces due to bending and compression of the ring were considered negligible compared to the inertia forces.
- 11. Available deformation energy in the ring was calculated from: (volume of steel parts) x (minimum specification tensile strength) x (minimum specification elongation). For each fragment, 25% of the ring parts plus 10% of the dicphragm portitions was considered effective. Corner squashing of the wheel fragment was added. Note that more total deformation energy is cvoilable with 4 fragments than with 2, since the necesscry distortions can be "spreed" over more material.
- 12. First phase ended when deformation energy capability is reached.
, This is an cpproximate method for predicting when the inner ccsing ring fails.
It was used in place of the detailed analysis of stresses and deformations. C. Second phase of the inner casing impact (ofter ring failure)
- 13. Three independent cases wero considercil fer comperison, each using en assumud
' mode of ring failure following the first phase impact. The cclculations gave similar results, indicating that the model need not be too precise. The models are:
- a. Ring failure near the impact point, so the wheel fragment acts on the end of a ring segment.
- b. Ring failure between impact points, so the wheel fregment acts near the center of a ring segment.
- c. Ring breaks into a " cloud" of small, independent particles.
- 14. Motions during the first phose impact cre small.
This allows the use of the scme geometric relations (crienterions, positions of CG's, etc.) as employed fcr the first-phase. A rcugh estimate of the possible motion een be obtained from the tensile strain of the ring et failure, if this is approximately 20%, the radius of the ring will have increased cpproximately 20%; local bending could increcse this somewhat more. The effect of an overage increcse of 10% was Icoked et in a limited way by simply increcsing the centcct radius on the ring for several cases, keeping the ring properties and wheel fragment properties the same. The resulting concreta slab penetrations decreased 10-15%. On this bcsis, it is believed that igncring the geometricci changes provides a conservative result (predicted penetrations too Icrge). l
O TR6751.211
, Page 7
- 15. Deformation energy not limited.
In contrast to the first-phase, no limit was placed on the defermation-energy absorbed in the second phase. This is more realistic fer the "cicud of particles" model, where deformation is prine!po!!y in squashing. For the other models, if additional deformation creates further failures, they tend to approach the particle model. The particle model reduces the wheel fragment energy the least. D.' Outer hood perforation and air drag
- 16. Perforation energy loss found from " Stanford-formulo" (see sect. IV-G). The effective diameter of the wheel fragment was taken to be the hydraulic diameter
. of the minimum cross-section. (Fig.14)
- 17. A plug of exhaust hood material equal to the minimum wheel fragment cross-section is accelerated by inelastic impact.
- 18. Fragment rotetion is ignored.
None of the perforation fermulas er experiments found in the literature treat a rotating missilo. Fortunately, rotational velocities are relatively low after the inner casing impact in most cases.
- 19. Damage: to the cuter hood by ring fragments is ignored.
This is en unconservative cssumption, since it is conceivable that the ring parts could " clear a hole" fer the wheel parts. However, the overall effect is rather small, since the predi.;ted trem!ation kinetic energy losses due to hood penetration are 10% er less. For the velocity ranges being treated, the concreto slab penetrations would increase only 7-9% if this loss were neglected.
- 20. Average projected cross-section crea used in drog calculation.
For missiles with vertical flight paths, air friction effects were taken into account by using a drog coefficient of unity en a " blunt" body using the overcge of the minimum and maximum cross-sectional areas of the wheel fragment (Fig.14). This approximates the effect of the slow rotation during the flight. The fragment is assumed to be criented In its original plane, rather than tumbling so the larger " flat-side" arec is not used in the drog calculation. . E. Concrete penetration
- 21. Average projected cross-section area (Fig.14) used in "Am!rkien-formula" (section IV-l).
e
o TR6751.211 Page 8 Little guidance is provided by the literature for "non-standard" misille shapes. It was considered unrealistic to use the minimum projected crea because of the remote chance of striking with this orientation or of the missile remaining in in this orientation during the period of penetration. Furthermore, most impacts will be soniewhat cblique, which tends to decrease the penetration capability. S 9 4 s 4
-*a
O TR6751.211 Page 9 111. Results of Calculations and Conclusions The results of this study cre discussed below, as well as in several tables end figures. Calculations were made for severci prototype turbines; numerical results for a 38" last-stege bucket machine are given as an Illustration. Because of continuing developments and refinements in design, the specific data fer a given machine cannot be generated in advance. These specific data will be ebtained by the procedures of this report. A. Fragment Properties , Table I lists the significant wheel fragment properties. The notation "38 inch wheel" refers to the wheel used for the 38" last-stage bucket. The projected areas are those shown on view "A" and view "B" of Fig.14. The initial velocity is the peripheral speed of the fragment center cf gravity at the instant of fcllure. The trcnslational kinetic energy of the 120* frcgments is seen to be higher than that for 90* or 180* fragments. A 134' fragment has the highest pessible trans-lational kinetic energy, but the 120* fragment provides a close approximation and is simpler to treat in the analysis. B. Freament Energies outside the Turbine Casing Tcble 11 lists the translottenai k!netic energies of all the fregments citer pene-trating the turbine casing. The data should not be viewed cs the mest probcble values resulting from a turbine wheel failure. Rcther, they represent a reasonable upper bound, since th2 pcrameters used for friction and restitution, end the .nodel used for the second phase impact tend to maximize the final energies. Thus, a low friction coefficient (p = .15), zero restitution, and the " cloud of particles" ring model in the second phase were used fer all the results ilsted. The 120* fragment is seen to have the forgest residual translational kinetic energy. The rotational energies cre not listed, but cre generally negligible for the 90* and 120* fragments, cnd cre in the range of 100 - 200% of the translattenci energies for the 180* frcgments. . M!ssile rotation is believed not to be especially 1 Importent in penetration of concrete barriers, although no stud!as of this were found in the literature, l e
q '
' TR675L211 Page 10 C. Significance of the "No. 2nd Phase Impact" Cases Certain combinations of the pcrameters give results in which the inner ccsing Impact " terminates" before all the ring hos been defermed to the point of fracture.
" Termination" means that the normal velecity of the ring and the wheel fragment at the Impact point are the some, or that the relative compression velocity is zero.
- The two parts are still moving at different velocities, and both in the radially outward direct!on. A careful look at the resultant motions shows that the wheel fragment, which is moving faster, will " overtake" the ring causing another Impcet.
This impact will be et enother ring !ccation, and probchly near the opposite outer
. corner of the wheel.
. Because of the additional Impact, both the concrete slab penetrations, shown on Fig.1, and the wheel fragment vanslation kinetic energies, shown on Fig. 2, will be smaller then shown for the "no 2nd phase Impact cases". It should not be concluded that the fragments cre centained In the casing, necesscrily, although it is obviously pcssible to provide sufficient deformchle material to ensure this.
~
- No study was made of the cdd!tional Impact for the "no 2nd phase case", since none of these corrcspend ta current turbine censiuction, or reasoncble values for friction coefficients.
D. Compcrison of Various Wheel Fregments Fig.1 Indicates that the 120' fragment is almost always the n.ost dangerous In terms of the max! mum thickness of a concrete slob which can be perforcted. (The exceptions are fer friction eceffielents which are unrealistically hlgh). This might have been prodleted from the kinetic energies of Table 1, but the missile crea os well as Its weight end velocity influence the penetration. Fig.1 includes the Interaction of these factors. E. Partition of Energies After the Inner Casing Impact Before impacting the Inner cosing, each wheel fragment has k:netic energies both in translotlen and rotation. This total energy per frcgment splits into six p=rts as a
~
l
l l TR675L211 , Fog'e 1I l result of the impact:
- 1. Wheel fragment trcnsiction
- 2. Wheel fragment rotation
- 3. Ring trcnslation 4 Ring rotation
- 5. Friction loss
- 6. Deformation loss The energy balance for all wheel fragments is shown on Fig. 2. All these results are for the case C, " cloud of particles" model of the second phase inner casing impact.
- The total deformation losses are lcrger then that required to " fracture" the inner cosing during the first phase of the impact. The additional loss is considered to come from " squashing" the ring pcrticles as the wheel fragment posses through the " cloud".
Notice, however, that the deformation loss doesn't change much as the friction coefficient is increased. Rother, the larger frictional losses are Eclanced by decreases in ring and wheel fragment energies. The selection of a friction coefficient is seen to have some impertence in the results. Friction is fer from constont during the imp'ect. Initially, because of the very mgged surfaces, friction must be very high (coefficient of unity cr Icrger). However, due to the high pressures and localized heting, lecci creos of very low friction een be produced. It is unrecscnoble to disregard friction cornpletely, although calculctions for e few trial cases indicnte the resultant missile energy would increase only 10 - 15%. An average value of 0.15 was used as representative of the conditions during the whole impact process. As menticned ecrlier, most of the initial .eheel frcgment rotational energi is converted to other ferms. This effect is probcbly unique to the perticuler see ietry of the Innct cosing impact. It is ecsy to visualizo impcct problems where most o the initial translational energy is converted to rotation. F. Comocrison of 2nd Phase Impcct Modeis The three modeis used to predict the 2nd phase of the inner ccsing impact are compared on Fig. 3. Even with the quite different models, the predicted concrete sleb perforations are Hmile . This lends some confidence to the results, since rather significont changes in the model have little effect. Case "C", vehich gives the greatest predicted perforations will be used for determining the missile energies to be furnished to customers. G. Air Frictica Effects _ For fragments which leave the turbine necrly vertically, cdd:tional ene rgy is obtarbed by cir drog during the up-ond-down flight. Fig. ; litustrater tt mcgnitude TT
TR67S't.211 s Page 12 of this effect on penetration of eencrete sicbs. The upper lines are those for low angle miselles which have shcrt filaht pcths. The lower lines Include the otr-drog for the vertical missiles. The total Influence is not large, but it is readily predictable and is therefore included. , H. Effect of Non-zero Restitution Coefficient Although the enclysis has provision for varying the coefficient of restitution during the Inner casing Impcet, all the reported results are for zero restitution. This is realistic if the deformation energy at the end cf the approach period in the Impact is compared with the capacity for storing elastic energy. It is found that the elastic energy is negligibly small. As confirmed by a number of calculctions, non-zero restitution tends to increase ring fragment energies while decrecsing wheel
~
fragment energies. e
- 1. Secondcry Missiles from Ring Fregments -
The possage of the wheel frcement thru the turbine casing has the potential for creating secondcry missiles from fragments of the stationcry pcrts. These fragments, however, will not be as penetrating cs the wheel fragment. In generci, the calcula-
! r tions show that the velocifics of the ring pcrts have approximately half the wheel fragment velocity (approx. 200 ft/;ec vs. 400 ft/sec). From the A.neriklen formula (sec !cn IV - 1), it can then be shown that spbrical frcr,r.acnts wculd have to be 10 ft. In diameter, or prismatic fregments would have to be nearly 7 ft. long to be more domcging than the wheel. The spherical size is not possible; seven-fcot prismatic secticns cre most unlikely. Furthermere, segments of rings, even striking in the most favorable crlentation, mcke "pocr" missiles beccuse cf the tendency to bend and deform on impcet. Based on this, it is evident that plant designs which are safe ogcInst wheel fragment missiles will also be scfe from secondcry fragments.
J. Summary of Data to be Furnished to Customers The results discussed chove suggest that the data to be used by customers in evaluating nuclear plants fer protection against missiles should be bcsed on a 120* wheel fragment, 0.15 friction coefficient, 0.0 restitution coefficient, and the l Case C second-phase Impact model. Each of the chove choices lacreases the l penetration effects of the wheel fregment. Tcble ill glves a list cf typical data fer a 38" last-stage turbine; most of the data is also included in Tables I cnd 11. Penetrations or perferctions in concrete cre not included since these depend on the plant design; their prediction must be left to the picnt designer. I l 1
l l TR6751.211
} IV. Analyses Page 13 A. Wheel Fregment Motion Before Impact in the idealized Impact assu.ned, the wheel is censidered to fracture into n(=2, 3 or 4) equal fragments, the fracture faces being radial plcnes. At the instant of fracture, the wheel is rotating at a speed W. At the instant efter fracture, each fragment has two motions. The first is a translation of its center of gravity. The second is a rotation cbout the center of gravity. Fig. 4 shcws, for n = 3, the metion of the wheel fragments in three successive time intervals efter frccture. Althcugh not shown on the figure, the fragments cre not complete sectors but have on inner radius where the shaft runs thru. It is observed that the fragments sepcrote clearly, with no tendency for mutual interference. (However, see the later discussion). .
To conserve momentum, both in trcnslction and rotation, the linear end engular
. velocity of each fregment cbout its C.G. are:
A-1) V=d - l
~ Q=w 1 where R is the radius of the fragment center of gravity es shown in Fig. 4.
wis the wheel speeJ or the instant of fracture Since these velocities cre constant, thn lincer and engulcr dispinmements of each j fragment are: A-2) d = Vt = R wt i 6 = Gt = wt lt is <een, therefore, that the distence and rotatinn of the fregment cre related, so that the angular velocity does not directly offect the geometry depicted in Fig. 4: A-3) d=R6 Although the fragments, os noted chove, initially move epcrt without interference, there are possibilities for mutucl collision as the corners marked I, 2, cnd 3 cn Fig. 4 rotate towcrd the cdjacent segments. Explicit expressions were not cbtained, but graphical plots were made for cach of the cases studied. It wcs found that only for 180* frogments was there any danger of collision. In this case, the "jecding corners" (those with numbers in Fig. 4) tend to pass necr the original shcft center when the fragment has rotated to about 6 = 130*-140*. The results show that the 180* frcgment is not the most dangerous missile, so the effect of mutual fragment collisicn was net pursued further. Another potentici interfere sce con occur between the fregment and the shaft. For the 180* fragment, but only fcr this one, the "leeding corner" will strike the shaft at s about the some time the fragment strikes the inner casing. This case was not inv :stigated since it was felt that the resultant penetrations would be reduced. A second in' .rference exists between the inner corncrs of the wheel cnd the shaft, again caly for the ic0* segment. This also was ignored. . O
TR675L211 Page 14 i B. Wheel Fragment Position at Impact on inner Casing When the wheel fragment strikes the Inner casing ring, there will be an Interaction between the linear and rotational motions. Forces will be established at the contact point, verying with the position of this point with respect to the freument C.G. Depending on the radius of the ring compcred to the frcoment dimeratons, there are several possibilities. For the geometries considered in this report, the Impccts were either necr, or at, a corner of the wheel fregment. Two ccses cover these possib!!!- ties. I. Tangent Collision For this case, the outer are of the wheel fregment contacts the ring so that the two circles are tangent, cs shown In Fig. 5. When this occurs, we must have: 8-1) A- B - V X,.* + 'f?
*n where X, and Yccre the cecrdinates of the .: enter, point C.
Frem Fig. 5, and eq. A-3), the center coordinates cre:
' X " 3 (0 - 0}
B-2)
'f, - R (1 - = 0)
The only verloble abcva is the rotation cr.gle, 6. Tc.igency l> ebtcIned if: B-3) h(s) = y(e-Aef & (1- c= 0) -
=0 We can find the 0g which sottsfles this equation by a simple iteration as follows:
Let O gg = A-8. Then use the recursion: h (Got) = f(Geg -A Ocl[ + (I- c"' O ci N -
#} h(Gai) 4;,,) = Got i i = 0, I , 2 ' -
9 This converges very rapidly (! = 4 or 5 gives en 8-digit result). Now, the value cf 9 g ound f chove mcy not be a valid solutloo, since it only ensures tcngency of the two circles ci radius A and B. It must be checked het the point Q actually lies on the frcoment arc. The conditten on 6, the engle tc. the contact point O is:
O TR675L211 . m Pege 15 1 B-5) {-B- a z 4 S*k~h*
~
where ' T = la - Actually, only the upper limit is significant for the cases being considered; the test is made in the form: B-6) r + e - f/2 -% <o If this is satisfied, the impact point coordinates cre:
. x, = A = f .
B-7)
. y, . A 4 I ' And the fragment rotation angle at impact is:
B-8) e, = Og if inequality B-6) is not satisfied, we must check whether a corner impact is ob-toined, the second case.
- 2. Corner impact This case is Illustrated by Fig. 6, where contact is modo ct the point P. Itis observed thnt tho fragment has rotated ferther (Olcrger) than required to ebrain tangency '
between the ring and the circle about the fragment center. Eq. B-6) would not be satisfied. The coordinates of the impact point, based on the fragment position, are: Xp = Xe + B 4 (9 -fh) B-9) y, , Y, & B c= (8 - f/2 ) with X, & Y,, the center coordinates given by eg. B-2) l I For contact on the ring, we must have: B-10) J( ,* + F,' = M The only variable in this equatien is 0, so the condition for contact is: B-1;) G(9) < A'-ErNo]^ho)*O l 1
Tg Call Pg je 16 This equation cannot be solved explicitly for 9, but on iteration is possibe. The cbove equation clways has et leest one root, but it may cerrespond to an unr ali. able case, and there may be as many as three roots. The only case treated in the anciysis
- is the one shown in Fig. 6, where the contact is made as the point P on the ficgment J
first approaches the ring. To ensure this, it is necessary that the circle of rc i':s O through the C.G. intersect the ring. The limiting cose, when this circle is ,
- tangent to the circle indicates grazing contact. This end larger fragment rotations er- not considered in the enolysis. For grazing contact, we must have:
R' + d 2 = RM1 + eh,d = @- # B-12) where If = R'* B* 2RScm $2) B-13) So, from B-12): 6% : (A-D'* -1 B-14) (gj For the porticuler conteet b2ing considered, it is then required that: l Vle,,, ~)
>- R I !
_ = B-15) A Om X, (G,,, ) Failure to settsfy this inequality is treated as on error, since the cerner of . erest misses the ring on its first cppreach. If B-15) is satisfied, we use the followind erotion to find the 9pwhich satisfies'eq. B-11). Set 8p, = 0,;n, then use the recursion: Xg " W- W(%) B-16) . Xe - X,(ei,-) .
\
0,cg,,) = 67z &
, c = o,s. 2 , . - l with X (6) & Y (6) being determined from eq. B-9). This procedure ah yields ahut 8 sighficant figures in 4 or S iterations.
Corresponding to eg. B-7), the impact point ecordinates fer this case cr. - p P X ~ X (b) . n
TR675L211 i ' Page 17 The angle Sis: (for both cases) B-18) I" x
- 3. Both Cases For the analysis of the impact, the locatien of the fragment C.G. must be known relative to local coordinates through the impact point, T. (This is either the P or Q treated above). Refer to Fig. 7.
The coordinates of the impact point, I, relative to the C.G. cre: fx,=X,-d.=AemE~R0 r "I f y, , yz - R = h f - R where 0; is either egor Og and Gis determined either by B-5) or B-18). The angle Als then: B-20) X* C -E From this, the coordinates of the C.G. in the tangential and normal coordinates : x & y at the impact point, are: a,= D A A j B-21)
. bs
'D*1 where D' X[+( , or given by eg. B-13).
Finally, we need the angle of the velocity of the C.G. in the x y coordinates. This velocity is parallel to the X - axis; the angle from the x-axis is: B-22) h=[-f
- r o a e
A
. TR675L211 Page 18 -s C. Impact of V/ heel Fragment on inner Casing - First Phase The inner casing impact is treated in two stages. In the first phase, the casing is treated as a solid ring. This phase is terminated when the total kinetic energy lost due to deformation equals the energy which can be absorbed by the deformation processes described in section IV-F. At this point, the inner casing is assumed to fail, and the second phase begins. The calculation of deformation energy losses is described in section~ IV-E.
The impact being considered is on idealized one. Specifically we will treat the case where 2, 3 or 4 equal wheel fragments strike simultaneously the ring representing
- the composite of the diaphragm outer ring and parts of the end of the inner casing.
Only plane motions are considered, since *his greatly simplifies the analysis and appears to be a reasonable approximation. Fur;her, the problem is an=lyzed by elementary
- impulse-momentum methods, rather than decling with the details of local and general elastic and plastic deformations. The purpose of the analysis is to assess the energy losses associated with the impact, separate the-losses into frictional and deformational components, and to include the exchange of kinetic energy between rotational and translational components.
Geometry and nomenclature for one of the n equal fragments is shown on Fig. 8. Tangential and normal coordinates x & y are constructed at the impact point. The J fragment CG is located at the point (aj , - bj ) in these coordinates. At the instant of striking, the fragment hcs a linear velocity V o anu a retational velocity wo . Because of the impact, ner nal and rangential forces, FN & FTare created at the impact point. The detailed history cf Fy &FT re n t determined by this elementary . analysis, rather we deal with the normal and tangential impulses: t N = f. F, dl : C-1) t 7 , f. F7 dt Positive directicns for N & T are shown on Fig. 8. Positive directions for FN &F T cre the same as these for N & T. Then, censervation cf momentum principles require that the change in momentum, either in translation or rotation, must equal the applied impulse. Then, for each fragment, we can write:
- T en, ( u, - N w 6 ]
C#2) % [^4 ' ( '; 3 L,[e,-w]=- P M where m, is the fragment mass k, is the square of the fragment radius of gyration u, is the instontaneous x-direction velocity of the wheel fragment C.G. 1
TR675L211 Page 19
, ~. , v, is the instanteneous y-direction velocity of the wheel fregment C.G.
w, is the instanteneous enculcr velocity, clockwise The effect of the impact on the ring is considered to produce a local radici dis-placement, and to rotate the entire ring as a rigid bcdy about the shaft center. For the first, we consider a fraction of the entire ring mcss to be effective ct each impact point. For the second, all n fregments tend to rotate the ring in the same direction. Note thct local distortions and bending of the ring caused by the forces ccting on the inner surface are neglected except for the mass fraction used for trcn?lction. We obtain, then, two further equations: fng 7 3Vs
- C-3) , n A T-mz is the entire ring mass Is is the fraction of the ring mass ' considered effective at each impcet point for radial motion (.25 was used for the numerical results)
Vsis the instantaneous y-direction velocity at the impcet point Its is the squcre of the radius of gyration of the entire ring n is the number cf wheel fragments A is the inner radius of the ring Following Golds.nith (1), we write the relative velocities of sliding, S, end compression, C, as:
' $= U, 4 h, u), b Ldz C-4)
Cs V, + d,W, - V2 At the instant of impact, these have the values: 5, = V, u a h + b, W e C-5) c, = % 4 Y + a>w. Combining eq. C-2) thru C-51, we can obtcin: (S: f, - G, T - G, Al C-6) { C' . C. - G, T - G, 4 j I
- i b* nA*
where: 6,= K +u,L,+ p, m, i a .' t l
. G: = , + u, m, + Am z l l
C. b.
' G, = u , ,,, ,
1 l i n
TR675L211 Page 20 5 and C are importent because they determine various aspects of the impact process. The frogment and ring approach one ar oiber as leng as C >0. C = 0 signals the end of approach and the beginning of the reboun'd, if there is any. 5 determines whether or not sliding is occurring and its direction. This, in turn determines the direction of the frictional force. So long as there is sliding, the normal and tangential forces cre related by the friction coefficient, p: F C-7) F7 = *f a with the sign determined by the sign of S. 3
. Integrating this with respect to time, we find a relation between the impulses:
C-8) T= *y A * 'u where the constant C, depends on the past history of sliding, or not. For the initial part of the impact, C; = 0. N and T ere always linearly related. When there is sliding (S /0), eg. C-8) . governs. When sliding stops (S = 0), the first of eq. C-6) is used in the ferm G,T + G,N = S,. The N-T plane, as Goldsmith points out, is conve'nient fcr portraying the impact, since trajectories er state-lines are eesily, defined. Fig. 9 -hows the fcur kinds of state-lines which nre possible for this problem. Since, if there is to be contact, FN must tw ys be positivo, N must increase monotonically (see eq. C-1). Therofere, as shewn en Fig. 9, the motion en the state-line must always be upward. The dashed parts of the stch-lines indicato sliding is taking place. The solid parts (cases 2 & 4) indicate sliding ';as stepped. This depends, as shewa, on the r-f ativa magnitude of the friction coefficient, p, and the ratio G3/G .1For smcil values of p, as S ees S to zero and chenges sign, the friction force is not large encugh to prohibit sliding, but the direction of FT reverses. This causes T to decrease. (see eq. C-1) 4 The state-line for all four cases may be represented by two line segments intersecting at the point (T3 , N 3). Since the first segment must start at (0,0), we can write these lines as: (N=G,T osd8Ns C-9) N 5 Al ()Al=6,T+G c , i The classification of the lines and values for 4G , G5 , G 6, N3 and T3 con be
. conveniently displayed in a decisio~ n table.
** d
- w. - .- .. ,y-,,, , . , , , , , , , , - , . - , --,,,, , _ ,, , ,- m,, - --r-w---
TR675L211
. Poge 21 p p <0 <0
. So g G3 G3 G3
<g p < >- g >' K Cose @ h b @
1 1 I 1
~~
G4 - p P P - p 1 Gj 1 Gj
- G "T 5 ~f 3 P 3 So So G6 2N G 3
3 G%4 where N3 = g, g , g, T=3 G, G, + G, Disregarding, for the moment, the ending of the first impact phase by energy limitations, the normal impoet end con be determined. Fer this, we first find when the approach period ends, or when C = 0. This line is shown on Fig. 9. Note that from eq. C-6) it con be shown that both lines C = 0 and S = 0 slope downward to the right, but S = 0 is the steepest. We must always, then, have on intersection of a state-line with C = 0. The intersection, or end-of-opproach has the coordinates (T, , N,), where: C. G, _ ( ) N' G, GJG, \ o s +I, 5 Al, i T, = g, g , g, ) C-10) b Ge c
- 6d 1 G, G, + Gs , >;,
- G, Ge
\
l
- T. = c,c s G, .
l
- TR675L211 l Page 22 The end of impact is then a function of the coefficient of restitutien, e. Among the forms in which this is used is the following, where N2, the normal impulse at the end of impact is related to Nj, the normal impulse at the end of approach:
C-11) M = (1+8)d i i For inelastic impact, e = 0, and N2= N 9 For completely elestic impact, e = 1, and N2= 2N . jWe ignore cny restitution for the tangential impulse and assume the original state-line is followed. The termination point, (T2 , N 2) is:
~
N, = (it e) d i
% i C-12) 1
- g. . A-4 , N, a #2 1
. n Gs o a 4, 4 # , T's 3 h , 2 The fragment and ring velocities at cny tirne during the impact process are defined by the values of N cnd T. The expressions are found by solving eq. C-2) and C-3):
f u, = Vcad-kT o 4 v, s V. 4 4 - $ d C-13) e, = w. - , T- k N 4, , M
\ e,L nA -r m,m.
From these, we con find various kinetic energies. First, that due to fragment translation is: . EE,=ym,[u,tw,') = i sn, $ - V. Te h + Nd h
- TbN C-14) _
For fragment rotetten: . ; - 1 KE, = I m,k. w? ' $'" k We
- Wa .\, T + a, M +
g y ,, _b, M , % C-15) The translation of the nth part of the ring gives: 1 1 1 C-16) ME, = I % Es vs - 2 ,,,g, Finally, the rotation of the nth pcrt of the ring gives: C-17) k'E, c f % = Y{ A' T * - I l l l m _ _ _
TR675L211 - Pc,qe 23 .- From these, the total kinetic energy per fragment, at any time during the Impact
'" ge.,, = y m.T + t mA w,* - Tr (re.e 4 e it .a., $)
C-18) i
. a. y , 2 m, k,
_t_ m, o.2 m, k. i rnt,/ g 9 It is seen that the first two terms are the initial fragment energy. The remaining terms con be simplified by making use of eq. C-5) and C-6). The loss in kinetic ene,rgy then becomes-4 ,= KE ,,- KEm 2 . = 5,' 7 + C, N -- p G, T* - G, N T - j- G, N This equation represents a family of ellipses in the T-N plane. A typical family is shown on Fig.10. ,This case Is for a 120* fragment of the 38" wheel. The ellipses h;ve a ecmmon center at the Intersection of the lines C = 0 and 5 = 0. If the impact were terminated here, the total energy loss would be maximized. The stnte-lina for the 120* fragment with a frictlan coefficient of 0.1 is also shown. The points (Tj, N ) where the approach period ends, and (T 3
, N3) where s!! ding stops ere marked. "wo possible end points 2(T , N 2 ) fer re:titution coefficients of .5 and 1 are also shown. For zero restitution, (Tj, N j) is the end point.
Since the Inner ensing has a limited capability to absorb energy through deferma-tien, the procedura outlined in section IV-E is used to determine when the first phase ends. This makes use of a form which splits KELOSS f eq. C-19) into two pcrts- one contributed by fricrlonal effects and one by defermetton effects. In some cases, the inner casing con abscrb more energy than required so the first phase ends according to the relations given in eg. C-12). There is no second phase in this case. However, there will be on additional Impact, or impacts, between the wheel fragment and the ring with Increased energy losses. Since these "no second phase" cases do not occur for parameters typical cf actual situations, they were not treated further.
.f e' -
. TR6751.211 Page 24
- D. Impact of Wheel Fragment on Inner Casing, Second Phase if the first phase impact is ended by a limit on the deformation energy loss, the It is assumed second phase impact of the wheel fragment on the inner casing begins.
that reaching the energy limit produces failure of the ring, but the nature of that failure is unpredictable. Whatever has happened, the wheel fragment and ring fragments are still in contact and have different velocities. It is therefore possible to transfer additional kinetic energy from the fragment to the ring mass. Admittine that the ring failure mode is uncertain leads to the consideration of three separate models as illustrative of the second phase impact. The first two of these are
' illustrated schematically on FIG.11. The ring is assumed to fracture symmetrically into n equal frcgments, one fragment per wheel fragment. For cose A, the fracture occurs near the Impact point. For case B, the impact point is assumed to be midway between two fractures.
In Cese C (not illustrated) assume that the ring fractures into meny very small fragments. Fcr visualization, the ring may be considered to be o " cloud" of particles occupying the region where the ring was befcre failure, but not interacting with one another. All three cases can be analyzed with a single set of equations by making :uitable definitions for the perererers. Tha general geometry is shown on FIG.12. The x y cocrdinates of the wheel and ring frcgment C.G.'s cre (a,, -5) and (a,, -h). The impulse-mementum eqvctions for this are similcr to eq. C-2) und C-3) of the previous section, except that the: velocity components of the :econd mass (ring) are now referred to the CG. For the wheel fregment: in, (u, - u, ) = - T D-1) l % ( 7 - #'8) * ~ d ( ,,k , ( w, - w,, ) - - h, T - 4 For the ring fragment: l
. D-2)
(" * ~"2* d
$ se (s - 4.)
I ( pp, (%-%,) - a,4 * !'J
TR6751.211 Page 25 where N & T ere the normni & tangential impulses defined by eq. C-1) m,, k,, u, , v, , and w, ere as defined following eq. C-2) ma is the mass of the entire ring n is the number of wheel (end ring) fragments k,is the radius of gyration of the ring frogment about its CG u, & y, are the velocities of the ring fragment C.G. in the x & y directions en, is the angulcr velocity of the ring fregment about its CG, clockwise The initial velocities, u,,, v.,% and u,,, v,,, w,o are those obicined at the end
. of the first phase of the impact. Explicitly, using eq. C-13), we obtain:
u,, T{ = d - f;- Tr
% ' W h - b- As
%- *, - M, rf .& e4 D-3) u,, . rf - 6, q, A f ~ *n Uso f, *
^
5*
#2a ' m y, lt should be noted thct kg is the squcre of the radius of gyrction for the ccmplete ring, while k, is that for o single frecment. Further, in the expressions for vu ond v,,,
the negative terms cppear since tho velocities of eq. D-3) cre those of the fragment CG, while v, in eq. C-13) is the y-direction velocity at the Impact point. Finally, it is assumed that the impact time hos been very rapid, so the reictive positions of the pieces is the some as that found in section IV-3. The relative sliding and compressien, S & C are: 5- u, # 6, 4 - u, - f, 4 D-4 C - %
- a.uk - % - aA They have initial values:
& = u,, r b, % - U ,
2 baQ C, = Ma + Q % - An - Os % u l l
. TRo75L211
- Pogs 26 Combining eq. D-1), D-2), D-4) and D-5), we obtain: :
g, 5, - G, T- G, N 3 ) D-6) C . C, - G, T - G, 2 l where 6:
- n .
m , *2' s 4,' ,, 6 ' ,+ ,,,4, m, + g,, a "J o s. .a. This relation is exactly the same form as eq. C-6), but with different definitions
- for G, , G, and G, . The construction of the impact st=te-line therefore uses the same relations in the second phase as those used for the first phase of the impact
[ eq. C-9), C-10), C-11) and C-12) , The instantaneous velocities of the wheel and ring fragments can again be expressed in terms of the normal and tangential impulses by rearrcnging eg. D-1) and D-2): ( u, . u,, ,',' r .- tt, = %, - M
~
W '
#:* ~ m, h, a.
- a. . s +tr ,
Y*g Y, g f M,= W,, & T+- A Using these relations, the instantaneous kinetic energies are: Wheel fragment translation: D-e) es, . p.9,',,c) + ~.k..'rs.:) - (u..r+ s.")
- i ,,. (a' r'}
Wheel fragment rotation: D-9) WE, = } m, k, w,' i m,A,w,l - w.. (b,Te a,4) + ,fg., 0. Y+ ,$ Ring fragment translation: D-1 C) k'E,={Qu,',q,')={f(u,,'rd')*{UwTr u ew N) + Q {M'e T')
~~,9 en O 4
"* ~' =
TR575L211
.., Page 27 Ring fragment rotation:
d'I ** $D #' # D-11) %'En , R '
- The first term in each of the four preceeding equations is the energy component at the beginning of the second phase impact. The kinetic energy lost during the second phase therefore becomes:
0 12) gg - K %,,- KE7 ,, = (U o + 5. "1. ~ "u ~ b "sa) T~ n
+ (s. + a. m. - s. - a, w.) N d, *
- rn, * [,
L NT
( ",h
, .. r. -
,/8 4,* ,: ri da i &
~ 1 ( ';i~ & Q * ,n,
- q)N Making use of eq. D-5) and D-6), this may be simplified to:
, D-13) . k'f,,,
- 5, 7
- q N - ] G Y '- G, Ni~ - } C, N '
This form is exactly the same as that obtained for the first phase impact, eq. C-19), provided C, , S, , G, , G and G, cru given their proper definitions. There is a strong possibility that the [orm giving tofal energy loss in terms of impulse values can be generalized, but this was not pursued. As shown in the following section IV-E, KE t of eq. D-13) con be split into separate parts, one representing friction losses an t e other representing deformation losses. The separate components are added to those accumulated during the first phase impact to obtain the total losses in the inner casing. To complete the second phase analysis, we need the geometrical parameters used in calculating G, , G, and G, . First, to determine a,, & b, for Cases A & B (ring fractures into n segments), we use relations applicable to th!n rings as an opproximation. For the nth part of a complete ring, the C.G. Is Iccated at a radius (from the ring center) of: D-14) 7-yk4{ where k,is the square of the radius of gyratien of the complete ring Then, for Case A, impact on the end of a ring segment, we obtain:'
Nt75L211 4n 28 D-15) f as = 7 4 (h t s be . A - F c=(%) , where A is the radius of the impact point Similarly, for Case B, impact at the center of a ring segment, wie obtain: (a=0s D-16) q b1 A-7 For both cases, the square of the radius of spation of the ring fragmenr I, , is related to that of the entire ring, k,by the axis transformation: D-17) E,,=4 1 -7* This results from the general form <** 4' .c d.' where d is the distanc i1.etween the axis and the CG. Case C, in which the ring is considered as o group of particles, require: more detailed examination. in place of eg. D-2) we can write for the I.th.partfelt h e m;{U,; - % ) : K.
- i. u,3,... t 0-1a) m, (.7,; - 4) a, q *t E c (thi - Ws.) = Ost d i
- hi 7l If these equctions are now summed over all L pcrticles, and it is assumed (os an approximetion) that all the peticles impact at th: some point on the wheel fragment, and the particles are all idential with:
#s, u,; . J.
a ( me- 4 .
$ I Yes '$ $
a,e , a, (*A Then we obtain: L
~
LAr,($n - u u ) s h, C' ' 1 ' L D-20) L 4, (fr, . c,, ) . . 42.. N,. #
< ^ . I.l,,, ic, (w s -wu ) , d, Al > l, I" If the porticles are compact, with a gene. '., pherical shape with rce , we may use: ,
i
TR675L211 Page 29
$'O 3
D-21) k*-f 5, $;-F* Take now the limit p.o , but L (no. of porticles) increases, so Ldi, = d a s, where dm s is a specified fraction of the total ring mass. Then: 4L
~
D-22) [, ,,,, y zc; - a
- -- o a, &,
With these, the relations following eq. D-6) become: g , m', ,d m, &,
,E x%
6 * -N~~ * , G , *A i . , 4. The last three initici velocities of eq. D-3) cro similcrly replaced by: L-Y A' au - s k, g D-24) k 4,, = i% a Wu
- b And, the instantonecus velocities of eq. D-7) become: .
a, = u ,
- gg T s
D-25) 11. ' % e k#
' 7" m , , rou- 4f 4M s Note that the Ic..' n;9atien contains p in the denomincter, indicating the angular velocity becomes infinite os the particle radius becemes vanishingly small. This, however, is the only place the perticle redius appects exolicitly. Furthermore, the total kinetic encroies in t enslation end rotatien (all the perticles) are finite and
- independent of f . In replace eq. D-10) and D-11), then
l l I
TR675L211 Page 30 O.26) 6.,= f d A {ud ru f N * ["s* D 'sa 0 ' u s V T& 10 T' . D-27) W,* d4g Note that the latter equation has no " initial energy" term; it disappears as /~o. Further, D-23) which we see that determine thethe size Nof&the impulses particle T, and from eq.(as measured D-26 byp))has and D-27) which disap determine the energies. The ".s hape" of the particle (assumed spherical) is contained in the numerical fectors 3.5 in the equation for G and 1.25 in the expression for KE4 . The value of < , the fraction of the total ring mass which is acted on bp each wheel fragment, was chosen somewhat arbitrarily (and censervatively). The ratio was formed by dividing the minimum projected radial distance on the wheel (view "A" of Fig.14) by the circumference of the ring at the striking radius. This gives the formula:
<-r,<eff)h D-28) c(
- gg _
where 4 is outer radius of the wheel fragment 4 si bore radius of the whcel fragment p is the central angle of the wheel fragment A is the striking radius of the ring ,
.s
TR675L211 Page 31
~'
E. Energy Relattens In Terms of irr. pulse Valu6s and Relative Velocitiest Deformatien Energy Limit !
\
i It was noted in 6q. C-19) cnd D-13) flict the leu In system (wheel fragment plus ring) kinetic energy could be writtens E-1) K16 - 4 T + C.N - I 6.7 - s6 NT - y Y The differential of this can be put in a revealing form by using eq. C-4) or eq. D-6): 1 E-2) Mioss ) ' he I ~ 6: b
# ~b 3
- = ssr , csN \
If this is Integrated, en alternato ferm for eq. E-1) is seen to her l T M E-3) KF :r Sdr + Cdil , K{,, r KE, e e This is perticularly convenient, sines now the loss In kinetic energy can be
'pertitioned Ints fricticn and deformation components. The first Integral of eq. E-3) defines the energy loss due to friction sinca It is develcped by sliding metten and tangentici forces. Tha second intocrol, involving con pression volocit!.4 cnd normal forces defines the energy len which results from d ferr.iction cf the parts. The frictional cficet !: visualized as being vury lacci la chcracter sin % the Impact takes place very rapidly. It may involva hecting and/or malt!ng of suface layers, shearing cf bonds which have " weld.da durins tho impcct, src. The d=iormatten les, on the other hand, can result from creu deformation of either the whewl or ring cnd need not be confined to the impact point, in eq. E-3), since S & di chveys havo the semo sign, the frictional energy less must incr ease monotonically (irreversibly). The dN cf the deformation Integral is always positive, since N must increasa menetenically, but C may change sign from plus to minus, in fact, this la precisely the effect determined by the restitution coefficlent. Through-out the approcch ported, C is pcsitive, and the deferaition Inte;;rul must increase. Its maximum Is obtained for C = 0, the end cf the cppresch. If there is a rebound, C must be negative, decreasing the defermation Integral from its maximum. For non-zero restitution, then, samo (rv ull) cf the defor;natlon en:.rgy may be recovered. (It must be pcssible to store the required energy, as elast!c dsformation, semswhere In the system for thl to be valld).
During the Irrpet, S & C era glysn la terms cf N & T by eq. C'-6) er I)-6), N & T cre censtralned to follow n c? rte.!n trcisce-ry (one of the !!nc-; cf F!c. 9) es d<. fined by l eq. C-9). Lhing this, tha fric:Isa and dai.rmatica lea Intecrois con be evalueted cs follows: 9
~, e- --- .,.
TR675L211 osN* M s (until 5 ' 0) ge, = C, M - } G, N'- { G, N T E-4) (gg , y, T - { G, T *- { G, N T N># 3 (oHer 5 -o) gE, = C,N - { G, N * - { G, MT - { G, u G (T, - T) E-5) {f gg, $,T - { G, T'- Q G,NT - { G, (TH, - N T ) 3 where G6 , N3, T 3cre given following eq. C-9) wi+h (N3 , T 3) being the condition fcr which S = 0. , i Observe as it must. that thethe Similarly sumsum of KEDEF of eq. E-5)nd KEives g.FRICthe eg. l.O n correct E-4)is result, so but notyields it .SS c KE obvious, it can be shown, however, that: E-6) Gu (T - T) + TNs ~ A Ts"0 3 i so the proper result is obtained. I From the dafinition of the friction energy loss, it is seen thct no contributien to the integral is modo if the state-line follows along the line S = 0 where there is no sliding. Thus, for trajectories typified by case @, Fig. 9, the friction loss increases along the dotted part of the path. When the solid (no sliding) part of the path is reached, however, the KE IC of eq. E-5) remains constant at the value reached at the end of the dotted leg. Depenhng on the initial slope of the dotted path, the frictional energy at a point on the solid path may have several values. There is a similar lack of uniqueness in the deformation energy component for case @ state-lines. This obtains from the condition that KELOSS f eq. E-1) is unique, giving elliptical contours shown on Fig.10. Fortunately, for all points in the N-T space for which S > 0, both energy components are unique functions of N and T. l A typical set of contours for friction and defermation energy losses is shown on Fig.13. This case is the same.one for which the total energy losses are given on Fig.10. i Because of the lack of uniqueness along the line S = 0, contours cre shown only for 5 > 0. l This includes almost all of the imoact state-lines of interest for this study. i
~ .s
TR6'75L211 3- Page 33 t Now consider how the preceeding information con be used to terminate the first phase impoet if there is a limited deformation absorbing capability in the inner cosing parts. For this purpose, we must evaluate the deformation energy loss, KEDEF I '9' E-4) and E-5) along the impact state-line defined by eq. C-9). From the discussion following eq. E-3), it is clear that the caly part of the state-line of interest is that for which C > 0, since KE DE3 increcses monotonically to C = 0 and then decreases. Therefore, if the energy which con be cbsorbed is limited, the first phase impact must end while C > 0. This is also shewn on Fig.13, where the solid lines are tangent to rays from the origin on the line C =,0. The region of concern is fer N values below this line. In the onelyses of section IV-C and IV-D, this N is denoted
- Nj , the end of the approcch. N denotes 3 the crossing of the line S = 0 where.the sliding stops. .
There are actually three cases to consider.
- l. O < N <j N . The 3 approach ends before sliding stops. For this, we use the first of eq. C-9) to eliminate T in eq. E-4). Then the deformction energy varies according to:
E-7) ME,,7 = C. N - i (6 s+ )N' l
- 11. N; > N 3, but 0 < N < N . 3The approach ends offer sliding steps, but consider the pcrt befere sliding stops. We oscin use the first of eg. C-9) l to eliminate T from eq. E-4), giving the same result es cbove:
E-8) k'G ou*C.N-I(6* t N 111. N y> N ,3and N > N . The3 approach ends eftu sliding steps end we consider the part ofter sliding steps. For this, the second of eg. C-9) is used to eliminate T, with the result:
) n - -f- {G, + {';) N' - + Ns E-9)' k'E ,y = (C. +
, if now there is en energy absorbing copcbility of KEgg, it is a simple matter to check whether this value is reached along the impact state-line. If it is not, the first l phase impact goes to con pletion and there is no second phase. If KEDEF c n become higher than KE pp along the state-line, the first phase of the impact is terminated , when equality is reached. The values of N & T for this condition cre denoted N3 cnd l l TS. These values cre used to determine the veiocities of the ring and wheel fragments at the end of the first phcse and the beginning of the second phase (if there is a second phase). N3& T; jare also used to cciculate KEDEF *"d KE FRICby eq. E-4) er E-5) at the end of the first phcse. If there is a second phase, it is treated as a new impact state-line and eq. E-4) and E-5) then provide increments in energy losses to be cdded to the first phase values.
l ' TR675L211 Page 34 F. Energy Lesses During inner Casing impact Severci mechanisms were investigcted for absorbing the kinetic energy lost during the impact. The first three, below, deal with plastic defermation of the various parts. The fourth deals with heating and melting, which might be obtained from the frictional effects. Other mechanisms, such as elastic energy storege are much smaller in magnitude and were therefore neglected.
- 1. , Ring deformation This is the largest compenent of energy loss, but con only be estimated. The total weight of the diaphragm outer ring, the inner casing end ring, the inner casing wrapper, and certain ribs were calculated. This Includes only material within cpproximately two feet (extolly) from the wheel center-line.* lt was assumed that each fragment was capable of deforming 25% cf this total, so that four fregments deform all the material, whereas two fragments d.ferm only half of It.
It was assumed that the deformed material was stressed to the speelfication minimum tensile strength (50,000 ps!) and the specificatien m!ntgum alongation
; (21%) is cbtained. This requires epproximately 10,500 in //in. , or 3090 ft #/#.
- 2. Diophragm pcrtitlen deformatien The airfoil section partitions cf the diaphragm cre in direct line of the wheel hub cnd must be deformed before the wheel con centact the inner casing. It was assur-:d that each wheel fragment deforms 10% cf the total partition weight by eendi q; and/or crushing. With specification minimum tensile strength of 100,000 ps!
and e!cngation of 20%, this absorbs on energy of 5890 fr #//.
- 3. Wheel fragment deformation Tbn fragment motion In the cases studied results In impact at or near the cuter corners cpposite the numbered ones of Fig. 4. The impact then tends to squash the corner region, as 15 commcnly observed following disk bursting tests.
The energy required for such do,fermation, neglecting elastic effects, con be estimated from Hill's (3) analysts (pp. 221) of the compres;!on of a wedge by a flat dia. For a right-angled corner, this results In the expresslom
- Refer to71gure 15. Generally all parts to right of vane tip and parts within two feet to left of vane tip are included.
i I
TR675L211 Page 35 W = 2. 89 wc 2 y F-1) where W is the deformation energy, inch # w is the wheel rim thickness (axial), inches e is the depth of flattening, inches Y is the tensile yield strength, psi Converted to it #, this energy is small (about 5%) of the two components above, but was included in the total.
- 4. Heating and melting of parts From the data of Camp and Francis (4) pp. 842, the heat required to heat from room temperature and mcit steel and ecst iron are 419,000 ft #/# and 319,000 ft #/#,
'respectively. This is an attractive " energy sink" as melting only 23.9# of steel or 31.3 #of cost iron would cbsorb 10 million ft # of energy.
This energy dissipation is Irroortant in explaining the frictional energy losses. By friction, relatively thin layers con be heated, rubbed-off, " welded" and broken. It is probably not feasible to transfer heat cway from the impact zone by conduction alone, since the impcet is too rcpid. !f heated layers con be physically
" scrubbed-off", however, it's not difficult to account for fairly large energies.
*4 .
i
TR675L211 p Page 36 G. Outer Casing Penetration When the wheel fragment passes through the Inner casing, the remaining barrierI is the outer casing. Most designs have relatively large crees of 1-1/4" thick plate. It is assumed that the fregment tears through such a region. An Individual design mcy have Inner or outer ribs of various kinds; th .se would provide greater resistance to ' penetration but were ignored. As will be seen, the present outer ecsing designs provide a small, but not negligible reductlon In fragment energy. There are at least two analyses which could be used for the penetrotton calculation. Recht and Ipson (5) give a semi-analytical method for calculating exit velocities of miss!!es perforating flat plates. This includes a relation for determining the " minimum perforation velocity", which they Interpret In terms of the energy required for perforation. A second expression was developed by Stenferd Resecrch Institute and reported in a series of reports. Moore (6) summcrizes their results, giving their empirical formula for minimum energy for perforation. Comparison calculottens using the two cpproaches were mode. In both ecses, there cre some arbittery factors which must be assumed in order to relate the wheel frogment geometry to the Idealized cases for the formulae. Since the formulae could be made to give the some or similar resuits for reasonable values of the fccters, there is little reason to use one fermula over the other. The Srcnferd ferraula wcs used for all the data repcrted. Recht and Ipsen (5) cover the obilque perferation case (fragment Initial motion not necmal to th s plate), with the general result that moro energy Is required. Although ; I; h likely thct tho wheel fragment strikes the outer casing obliquely, the effect was Ignored. This, as is the ecse with most previous assumptions, tends to overestimate the ! wheel frccment energy outside the turbine casing. The Sinnicrd formula is given by Mcore (6) tn the form G-1) .b = U D
. 344 T 2
*.00606 %
where E is the criticci perferation energy, it # D is the missile diameter, Inches U !s the ultimate tensile ttrength of the plate, psi ( T is the plcte thickness, Inchas W is the window width, Inches l l
1 i TR6751.211 Page 37
. 3 The perforation energy was used as the energy absorbed when the missile " punched" a hole in the outer casing. Fermulo G-1) was bc>cd on tests in which cylindrical projectiles were fired end-on at plates whose edges were rigidly supported. The diameter D, and the window width, W, (width of plate between supports) are not well defined for the wheel fragment impacting on the outer casing.
An orbitrary choice was made for W, setting it equal to the inner radius of the inner casing. ' This provides a number which increases with casing size, and is representative of the unsupported distances in the exhaust hood. It makes the second term of eq. G-1) comparable to the first term. Further, it is within the dimensionless limits given by Moore so us to be covered by the test data. For missile diameter, Moore recommends using on equivalent diameter like the hydraulic radius concept used in fluid friction. We choose a circular missile which has tho some ratio of perimeter to cross-sectional area as the actual missile. For a
. cylinder 1r G-2) Area ,
TDs , 14 Perimeter Tr D So, we use for D: D=4 *"
- E'. I'"' 'O *h' 9**"I G-3) t'erimeter of the minimum arca X-section in all calculations, the tensile strength, U, was taken as 55000 psi, the minimum specification value for the outer casing plate traterial. The thickness, T, is a constant i-1/4 inches for all machines.
E, in eg. G-1) represents only one part of the energy transfer in perforation. The second port, as described by Recht and Ipson (5) involves the inelcstic collision between the mass punched from the outer casing and the wheel fragment mass. Taking the plug mass, M , to be the product of the outer casing thickness with the minimum projected crea of t. e wheel fragment, we then obtain: G-4) KE ouissa e g m, r ms ( %sior l wiiere m, is the wheel fragment mass l rnS si the plug mass KE & KE are the energies before and INSIDE OUTSIDE offer perforation. ( l
TR675L211 Page 38
~ .' H. Energy Losses Due to Air Drag .
Two classes of wheel fragments are potentially dengerous to reactor contcInment. Low-ongle fregments may strike the side of the containment after a short, almost straight f!!ght although there may be intervening obstacles. Necrly vertically flying missiles may strike the top of the reactor containment after a high, arching flight. These missiles must be " precisely" aimed, else they will miss the target. Further, they con lose significant energy due to air drog over the relatively long pcth. The following analysis is similar to one made by J. E. Corr (unpublished memo of 11/20/66). We assume a drog force on the fragment In the form: H- F, = C,(?f) FD = drog force, # CD = drog coefficient, dimensionless p = air density, f-sec2/p,4
' A = projected creo, ft V = velocity, ft/sec < We are concerned with nearly verticci missiles and therefore consider only the vertical motion. The equatiens of motion cre: i dk d' AN M g, 5 M g/7 = - Mg C 3 P-H-2) Upward flight:
d4 all
-M9+CD
$2 Cownward flight: My:M g where M = missile mass, #-sec277, g = ceceleration of gravity, it/see The equations may be integrated by eliminating time in favor of x. We note the relation:
H-3)
- - 3. g/
Substituting this in the up-eque:lon, we obtain: b=V + 6V H-4) dx
\
where B = gAC, g l W= missile weight, # w = cir density, #/ft 3 .. n.-- - .., ,,,y a- . . . , .
TR675L211 Page 39 A t The variables are separable in eq. H-4), so that the solution con be obtained in the form:
' U,
- g, 7 H-5) An x= 2B ya,. 9/g ,
where V g is the initial velocity of the missile at x . o. The maximum height attained by the missile (V = 0) is:
~
H-6) x, = 2 8 ' " I* 9 ] For the dreg, hence B, opproaching zero, this hes the well-known expression as a limit: y.' H-7) Nu (no am4) = gg Similarly integrcting the downward equation: H-8) X=Xw+ lg- L (1- ) We assume the final striking velocity, Vp, is reached at x = 0, although the l impcet may be at a different elevation than the turbine. Then fro a .g. H-0) and H-6): fg M fl-H-9) g )=-Xmr=-fs'M(I* g l 1 The ratio of final to initial ~ velocity can be found from: 1 I H-10) 4** ya gv=c
- 9 This may also be written in terms of finct and initial kinetic energies:
l
TR675L211 Page 40 T, I
- s' H-11) y * ; jfs, - 6 WL Tp,T~t = final and Initial missile kinetic energies, ft #
N' L = 78 = cf S For the calculations made In this report, w, the otr density was taken es .074 f /ft3 , The drog coefficient CD w s taken as 1.0, which represents a recsonable estimate for blunt bodies In the velocity range of Interest. The crea was teken as the average of the minimum and maximum projected crecs of the wheel fragment In a plane perpendicular to the. shaft exis. This overeging takes account of the probable tumbling of the fragment. e 9 f l l l
TR675L211 Page , 41 (3 1. Penetration of Cencrete Slob The evaluation of missile effects in a particular plant is a matter for the plant designer. In this report, a comparison between various missiles was desired. For this, the work of Amiriklan (7) appears most applicable. Moore (6) also summarizes Am!rik!an's approach. In these papers, the missile penetration is predicted by the empiricci relation: 1-1) D = k Ap V' where D is the penatration depth in feet k is the penetration coefficient (experimentally determinod, see below)
- A is the sectional pressure, #/ft 2, obtained by dividing the missile P weight by its cross-sectional creo ys -,
V'Is a velocity factor = leg 10 b
- 215000 a
~ V is the missile velocity, ft/sec.
This formula cpplies to penetration into an infinite slab. Amiriklen reports Navy experiments which resulted in a correction factor for finite sicbs
. 4 (<'- 2 ) . ,
1-2) D, = D [1 + e ; E where D is the penetration depth in an Infinite sich, ft. D' is penetration depth in a finite thickness slab, ft. el = T/D T is the sich thickness, ft. For complete paneiration of a slab, we must have
. .y(T/p -2) ~
I-3) D' = T~ = D [I + 2- .. Rearrangement cf this equation shows that D = T/2 gives complete penetrat!on. Therefere, from eq. (1-1), the thickest slab which will be perforuted by a missile is l-4) T = Zk gA I This formula was used fer the present calculations. A, was determined using the weight divided by the overage of the minimum and maximum pro [ected areas of the wheel. It was ; felt that the fragment rotation would tend to reduce penetration somewhat so that using the i I minimum wheel area was too conservative. Amiriklen gives several values for the penetration coefficient, k. From his table 1 (also given by Moore) and his Fig.10, we have the values:
.,)
- . -c
TR675L211 Page 42
,o
. k Meterial
.00799 ,
2200 psi concrete
.00476 3200 psi concrete,1.4% reinforcement
- -? .00282 5700 psi concrete,1.4% reinforcement
.00348 3000 psi concrete Specially reinforced 00277 4000 psi concrete according to
.00224 5000 psi concrete Amirikien Fig.10 in the present calculations, k = .00476 was used as probcbly representative of current construction. Obviously, from eq.1-4) and the table, this may overestimate the penetrction by a factor of 2 if special construction is used. .
One further remark should be made. By use of eq.1-1) and I-2), it een be seen that the penetration depth in a sicb at locst twice as thick as the perforction thicknesses given on pgs. 60 and 62 (i.e., for T > 2D') would be one-half the
~
perforation thicknesses shown. s 1 l l l l l p
=
i l
. TR675L211 Page 43
,O J. Wheel F egment Properties The preceding calculations make use of several prcperties of pie-section fragments of the turbine wheel. All of these qucntities may be found from the distribution of axial thickness vs. redius for the wheel cross-section. A typical fragment is shown on Fig.14 The shaded part of view "A" is the cross-section shown on normal turbine layouts. From this, the lower right graph con be derived. Although the wheel contour may have curves, it is es2umed that it can be represented by straight-line segments es shown.
Most of the properties of interest are functions of the following integrals: J-1) Q= J h# 3 de , j o,s, 2, 3
- where (, is the inner radius and q is the outer radius The creo of the shaded cross-section of view "A" is:
J-2) Area = Q , The volume of the pie-shcped segment with centrol angle, f , is: J-3) Vol = f Q3 From this, the weight end mess follow directly: Weight = f[d i J-4) g Mass = - p 0, 3 where / si the wt/ unit volume of the wheel. 490f/ft was used g is acceleration of gravity = 32.2 ft/sec. The distance from the shaft center to the fragment C.G. is: i Qn \ { 4 Plt J-5) Rec , ' i- 5--)( p,, j The polar moment of inertio cbeut the C.G. is: a S
1 i l TR675L211 l Page 44 j The crea of view "A" is: J-7) 43 = Q, + h, r, (1- 2 k2) l The crea of view "B" is: ! l J-8) A:g 2(Q+h,c,)if/z l To evaluate the Q; , we assume that the wheel geometry can be described by straight-line segments as shown on the lower right in Fig. IV-J-1 We take from the graph the pairs of numbers, (4,k, ), (<a , A,) ' -(<, fi ). Note that certain pairs, j
. 2 & 3 on the figure for example, may have the some radius but different thickness; other pairs, like 4 & S, may have the same thickness but different rodii. The contribution to Q; for one of the line segments is:
'rCes i
J-9) 4 Q. = h <I d,'" - J s 9 For each segment, h is a linear function of r, so the above integral can be evolucted. To facilitate this calculation, we define the following function: (i,i,,) 3.,1 * ' _4" 1
' 9. , - fi l where the superscript on D indicates the segment for l which it is used. l The D's are calci'!ated recursively, to enwre accurate evaluation and avoid division by zero (if (;%, ):
U i* d , , J-11) v v.c,.) 1 Opa.t.0 * # ,, ii *$ , ]*E3.4,S i i
- I, 2, 3 o L l
Then using eq. J-9) and J-11), we obtain:
~
L-1 J-l 2) 0; =, z[ f, O <.~.- Ad [ dp' Xh s - ^<. 9 )D ,, e j Finally, it is necessary to obtain the perimeter of the rninimum projected crea, or
- view "A". Assuming the wheel is symmetric about its mid plane, and described by
'c ' straight-line segments, this is:
L1 ~ J-13) Perimeter = hi e uh + 2fi (1- em M4)..+ .Z, j 2&c.,- A,)* + 4(4,-f) *,' .q
TR675L211 Page 45
-s,
. K. Wheel Bunting Speed The calculation of wheel bursting, speed is based on two criterla. The first, for
" sound" wheels, is bcsed on the overage bursting stress. The second considers the effects of bore defects and is therefore dependent en the wheel bore stress. Average bursting stress (ABS) is the limiting criteria for bore defects smaller than .34 Inches. For the 38" '
turbine, the wheel bursting speed based en ABS, and the vone bursting speed based on tensile stress are: Minimum spec. wheel TS, ksi 110 ABS at 1800 RPM, ksi 32.63
% overspeed (for ABS = .85TS) 69 '
%overspeed for vene 68 It is noted that the vene falls first, provided failure is governed by the minimum tensils strength, it has been assumed throughout the analysis that the wheel fragment consisted of both the wheel proper and the bucket dovetail. '
There is an inconsistency in these numbers, however, in that the wheel ABS has been calculated with the venes present. If the vanes fall, it could be crgued that the wheel is unlikely to fcII, or et any rete would fall at a higher speed then shown in the tcble. If the wheel falls befere the vene, the potential fragment demcoe could be reduced because of two effects. The frcgment crea Is, of course, increased somewhat, but mere Important, the defermable material to cbsorb energy is increased. l I e
. 1 1
' TR675L211 Page 46 IV-L. Numerical Dato Used for CcIculations For illustratien, the following dimensions and data were used in the calculations for the 38" lost stege. Similar information would be used for other machines.
- 1. Wheel dimensions Radius & thickness, Inches r h
- 16.5 20.5 20.5 d 23 23 13.4 38.25 8. 7 44.65 8. 7
- 2. Diaphragm and Inner casing dato l The following data were cbtained using a 2 ft length of Inner casing wrapper and I ribs.
1 Partition weight, Ib. 10500 Diaphragm outer ring w. sight, Ib. 15483 . I Inner casing ports weight, Ib. 16924, Total ring weicht, ib. 42907 . Radius of gyration, inches 85.1 l 1 4 ll
~
TR675L211
] Page 47
- 3. Energy losses The energy obsorpticn, per fragment, are given in million ft #:
25% of casing parts 13.7 10% of partitions 6.2 Wheel corner deformation 1.0 Total energy obscrption 20.9
- 4. Other Data As described in earlier sections, the outer casing was assumed to be 1-1/4" picte having 55000 p:1 tensile strength. The penetration coefficient for concrete was
.00478.
A range of values for friction coefficient (0.1 to 1.0) and restitution coefficient (0 to 1.0) were reviewed et verlous steges of the study. The results section lists the ones actually used for the complete analysis. These cover the significent vorlations.
.V O
*..wr e
TR6751.211
- Page 48 IV-M. Computer Program Description ',
A computer pregram has been written to perform the calculations required by the previous sections. It operates en the GE-265, being written in TIME-SHARING FORTRAN. A listing (5 pages) of the sourc~e language program follows this description. The program consists of two linked parts, celled EEZP and EEZPl . These mcy be renamed, but if changes cre made, line no.1250 of EEZP should be citered cecordingly. The program, as it is written could not be written in the BASIC time-shcring langucge because of its length. I The program listing includes a key to formulo numbers in this report. The notation ) used in the program is similar, but not identical to that of the report. A glosscry of symbols should not be necessary however. l Data for individual calculations is entered in EEZP beginning after line no. 90CO. j
- The input data consist of the following numbers. They may be separated by commas or blanks, and must be in order. The icllowing line numbers are suggested:
9100 Number of pairs of rodii and thicknesses for wheel dimensiens (20 max.) Radius, thickness, radius, thickness, . . .etc., in inches ; 9110 Additienal radii cnd thicknesses, if needed l 9200 Percent speed Total ring weight, Ib. Fraction of ring weight fct'redial fc.ce, use .25 Impcet redius on ring, inches Radius of cyretien of ring, inches Defermation energy ecpability per wheel frcgment, ft-lb. 9300 Tensile strength cf outer casing meterial, psi Thickness of outer casing, inches Penetration coefficient (Amirikian formula) 9400 30 chcracters of identification
TR675L211 Page 49 l The program for Case C, second phase impact is not shown on the listing. It is - obtained by adding /or replacing the following lines in EEZP1: 470 PRINT, " STRIKING COMMINUTED RING" 480 A2=0.; B2=.001; Z3=(R(L)-R(1)*COS(P2/2.))/(6.2832*A) 490 K2=.0004 500 TRIG 1=2 560 G1=1./Ml+(Bl +2)/(M1"K1)+3.5/(Z3*M2) 570 G2=1./M1-KAl **2)/(Mi *Kl)+1./(Z3*M2) 580 G3=(Al *B1)/(M1 *K1) 750 W1=W10-(31 *T2+Al *N2)/(M1 *K1); U2=U20&T2/(Z3*M2)
. 760 V2=V20+N2/(Z3*M2) 780 TRTF=.5*Z3*M2*(U2"V2"2)+((Z2-Z3)/Z2) *TRTF 790 TRRF=(1.25*(T2**2))/Z3*M2)+((1./M3-23)*M3*TRRF); TK=TRTF+TRRF
' The Case A and Ccse B calculation will not be run if these chenges are made, although certain statements used by them will still be necuted.
m O e
' ' "- = -, . - - - . _ . . _ . , _ . _ , , _ _ _ _ _ _ _ _ _ _ _ _ _
TR675L211 PROGRAM LISTING ~ N PAGE 1 t .- 99 'EZIP ,
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- p:gresrirs 300 Q3=HCI-1)onc!)-flCI)sRCI-1); COeHCI)-HCI-1) ,
310 00 INT 1, Jets 4 , 320 C3(J+ 13 a(RCI)*03(J))+(RCI-1)tJ) se. J iO l 330 IflT13 GCJ)=GCJ)+(Gic Cf4))/J+CO2*G3(J+1))/(J+g) to s it) , t 335 IUPUT C:~01CE2 # 3 40 PRIi!To "AN C.ZC DEC) s i R* CTICifs RCSTI TIONu",
' (*[" ,,77*;
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*= 8 t al 7 60 A ln O! $ 33(8 A ALn AL I-DI:. .".) J U1=D(1C05(AL) ~.< "8*I 7 '.*0 SOM V5 R 3 o t!C) *C'r( F d 2 . S 1079 63-C'43.TA) t B 1* WO ) ta, c s)
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- J
TR675L211 s PaGe 51 PROGRAM LISTING PAGE 2 9 30 FI RST PHA1E IMPt.CTt Typ= 1 9 40 IMPACT PROCESS 2 61=( 1./Fil)+(( 519 23 / (H 1cM 1 ) ) + ( CM3*( A 7 23 3 / ~ " ~ 950 *(Me*K2))# GEu( 1./M 1)+( 1./(M2*E2) ) +( C A l ? 23/ C M 1*K1)) 84 C48 . 9 60 G3a( A t=91)/CM1cK1) 9 65 ENTiCH3 saruv araras 9 70 Gas t./FI; I F( nn) 3n, sp g,7,pg; sg, g4..g4 950 Spit I FC FI-63/G1)FM3 03=-G1/G3J GO TO FN33 FNs 65m-l./FI o rcd'.* 185** 990 FN38 T3= 30/ ( G3= G4+ G 1) J tisaG4* T3 M 84 54 1000 IFC FI-03/ G1) C4; G&s50/00s 60 70 FN t. . 1010. GMt G4=0.J IFf S0)FfJ1J G6m2*N 3 1020.FHis Tl =( CO-62* G 6) / ( 02* G5+ G3) J N1*G"#Tt+04 et c.ic; 1030 IF(H1-N3)LN3* M OsMM3A Ltl3: T!= C0/ C 04* Ge+ G3) J Mle84*71 l 1040 NN3t N2=N1*(1.+Et) 10$0 IFCN2-N3)MSNsHSPsM2PJ H2Ns 72=N2/94J GO TO IMPACT END 1060 MOPS 72a(N2-G5J/G5 # l 1070 7*1 PACT EttO 03 70 (FIRSTS SECOND)sTRIq 1060 FIRSTt pamt 1200 PRINT TITLE wmas 1810 FIRST GUTPUTs FoliiMTC/" FAILURE AT "sI3s" PERCENT SPEID"//
\ 1120 +"MtEE1. FRAG Fi.0PCHTI EG"/I 7s 3X,"MG.t, DEQ"/
1130 +17,3Xs "WT, LU"/ F7 3s 3Xs "ECOs FT"/I 7.> 3X,"Jos LS-PT-SCC.30"/ 1,140 +77 34 3Xs"Ifj!!:s F.MT"/ F7 3, OXs ".**!i'"s 00-FT"// 1150 +" TOTAL RING PiOPCP.TI 21"/F7.Os 3Xs" hts LS"/
! ! 60 +77 3,3X,"R-I;:NZRs VT"/F7 3 3Xs"II-GYHs FT"/
1170 += 6PF7. !s 3Xs"X:F E3F.30Ys MILL'FT-LD"// t 100 +"137 PHASE IMPACT"/0PF7 1,3Xs"VZLs FT/ SEC"/ . 1190 +F7.As3X,"E3Ts TU 9/ SIC"/ F7 3,3Xs"A14 F7"/ 1200 +F7 3s3Xs"B!s FT**/;f 7. I e OXs "30s FT/ GEC"/ 1210 +F7.Is3Xs"COs FY, J4C"/ F 1. is 3Xa "FRI CTIOrJ"/ 1990 +F7. Is 3Xs "RESTITU M0fJ" //) 1930 PRIfJT FI RST CUTFU T, tDs P!s G4s R! e 054 G4s cts Mas As RG/12. emes 1240 + V0s W0s Als C1s EL CO, F1s E1 - 12 ;0 :mSE Egget i.ws To aamwen. er cocaar 1251 $US AROTAS*** - tw= ra sac,.w r ruawr. .c 9000 SCATA -
- N' ca ra 9100 5s 16 5s 20 5s 23s 20 5 23413. As 30 25sG.7s 44,65sG.T 9 200 169. 42707 25 G3 G5 1 20 9E6 9 300 55000s 1 25s .C0474 ,
9 409 39 INCN SUCXETs CAS7 ITM RING ~ : ' s N n 1
TR6751.211 m Page 52 PROGRAM LISTING FAGE3 i ICO'EC:FI 110'CUMT1t1UATION OF C2ZPs l:/ 3 / 4 7 120'TEiP13t:ATZ FIft37 M C2 ~W 2:12P.W o , ,,, , ,7, , ,, 130 G11aC CO.10 N/(c:nteng: :%.qe,33 c m,,c g.e0y;;g)f ga g .au ,,mr 140 613=C CC9453+G38G 6)/C L '.3cG0+429115) l 150 614'.-'(0::4Go*N3+2.v G5cMX)/ CA1 140 IFCGI-H3) ONE EMMcM !
. on a s .c. ., us, z ,
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- 9 220 OME D.PRCH3 l ircg:4.. a.s C@N 1) ' FI RST BRf.*iO 3 C3 TO K0ft1AL EMD
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, TR675L21T ' Page 53 PROGRAM LISTING PAGE 4 510 LA0Y UJTENs U2i.;wUOG-ii ate 03 VEGe V20-mio t 20 ... a.3 ,
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. ~l TR675L211
' Page 54 PROGRAM USTING PAGE 5
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- TR675L211 (m Page 55 V. Acknowledgements J. E. Corr, APED, was most helpful at various stages during the development of the analysis described in this report. Of special importance were discussions and comments which led to the "two-phasa" impact analysis. J. F. Proctor, U.S. Naval Ordnance Laboratory, originally posed some of the questions which the analysis is designed to answer. He raderived many of the impact equations from a brief outline (as a check on the analysis) and reviewed many cf the assumptions made.
l l l 1 i l l l I .
. _ _ _ A _ _ _ . _ _ __ .
l l l l t i TR675L211 Page 56 VI. References
- 1. Goldsmith, " Impact", Edward Arnold, Ltd.,1960.
- 2. Quad Cities Station, Units No.1 & 2, Safety Analysis Report, '
Amendment 3, Table 3-1.
- 3. R. Hill, "The Mathematical Theory of Plasticity", Oxford,1950.
- 4. J. M. Camp & C. B. Francis, "The Making, Shaping and Treating of Steel",
Cer;.agle Illinois Steel Cerpcration, 5th Ed.,1940.
- 5. R. F. Recht & T. W. Ipson, " Ballistic Perforation Dynamics", Trans. of ASME, Journal of Applied Mechanics, Sept.1963, pp. 334-390.
- 6. C. V. Mccre, "The Design of Barricades fer Hazardcus Pressure Systems",
Nuclear Eng. & Design, Vol. 5, No.1.,1967, pp. 81-ff.
- 7. A. Amir!klan, " Design of Protective Structures", NAVDOCKS P-51,
(] . August 1950 (Presented at annual meeting of the ASCE, Chicoso, Illinois, Octc' ecr 11-14, 1950) i -
.e 9 _ __ __ .-. - . _ _ . , ..,,_. ,
1 l TR675L211 1
- Pcge 57 l
l l l l 4 TABLE 1: WHEEL. FRAGMENT PROPERTIES l i (38" WHEEL) 1 Frogment Angle, Deg. 90 120 180 - l _ Fragment Weight, Lb. 4458 5944 8916 Radius of CG, Ft. 2.279 2.093 1.611 Polar Inertia, Lb-Ft-Sec2
^
234 462 1187 Min. Prol. Area, Ft2 3.170 3.65 7 4.831 Max. P.oj. Area, ft 2 6.832 8.368 9.662 O Failure Speed, % of 1800 RPM 169 169 169 initial Velocity, Ft/Sec 72 6. 0 666.8 513.3 Energies, Million Ft-Lb, per fragment Initial, Transiction 36.5 41. 0 36.5 Rotation 11.9 23.5 60.3 Total 48.4 64.5 96.8 1 4 h
w ,
- TP,675L211 Page 58 TABLE II: FRAGMENT TRANSLATIONAL XINETic ENERGY OUTSIDE THE CASING, 38" WHEEL Fregment Angle, Deg. 90 120 180 Trenslation K.E., Mill Ft 1 15.0 20.5 17.2
% of Total Initial K.E. 30.9 31.8 17.8
% of Inittel Translottenal K.E. 41 .1 50.0 47.1 e
f S 8 9 e
TR675L211 Page 59 TABLE 111:
SUMMARY
OF DATA TO BE FL'INISHED TO A CUSTOMER (TYPICAL, FOR 38" WHEEL) Fregment Angle, Deg. 120 Fragment Weight, Lb. 5944 Radius of CG, ft. 2.093 Polar inertia, Ib-ft-sec 2 462 Min. Prol. Area, ft.2 3.65 7 Max. Pro [. Area, ft.2 8.368
- Fallure Speed, % of 1800 RPM 169 Initial Velocity, ft/sec. 666.8 Energies, Militen Ft-Lb.
Initial, TransIction 41. 0 Initial, Rotation 23.5 Outside Turbine Costnu, Translation 20.5 After Air Dreg, (Vertical Tra[ectory) 16.3 a ? i
TR675L211
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