ML20091C628
| ML20091C628 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 07/30/1991 |
| From: | Dipiazza R WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | Murley T NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML19302E976 | List: |
| References | |
| CAW-91-192, NUDOCS 9108070031 | |
| Download: ML20091C628 (29) | |
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Westinghouse Energy Systems Bon 355 Pinsbmgn PennsyNania 15230 0355 Electric Corporation i
July 30, 1991 ,
CAW-91-192 Document Control Desk US Nuclear Regulatory Commission
- Washington, DC 20555 Attention: . Dr. Thomas Murley, Director APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE
Subject:
Westinghous ECCS Model for Analysis of CE-NSSS
Dear Dr. Murley:
The proprietary information for which withholding is being requested in the enclosed letter by Omaha Public Power District is further identified in
' Affidavit CAW-91-192. signed by the owner of the proprietary information, Westinghouse Electric Corporation. The affidavit, which accompanies this letter, sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR Section 2.790 of the Commission's regulations.
-Accordingly, this letter authorizes the utilization of the accompanying Affidavit by Omaha Public Power District.
Correspondence with respect to the proprietary aspects of the application for withholding-or the Westinghouse affidavit should reference this letter, CAW-91-192, and should be addressed to the undersigned.
Very ruly yours,
, / / ( % s R. P: 'D azza, Man '
, Enclosures Operhting Plant Licensing Support cc: M. P. Siemien, Esq.
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,4 Copyright Notice
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The reports transmitted herewith each bear a Westinghouse copyright notice, j The NRC is permitted to make the number of copies of the information contained
{
in these reports which are necessary for its internal use in connection with l
generic and plant-specific reviews and approvals as well as the issuance, denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license, permit, order, or regulation subject to the requirements of 10 CFR 2.790 regarding restrictions on public disclosure to the extent such information has been identified as proprietary by Westinghouse, !
copyright protection not withstanding. With respect to the non-proprietary !
versions of these reports, the NRC is permitted to make the number of copies '
beyond those necessary for-its internal use which are necessary in order to have one copy available for public viewing in the appropriate docket files in the public document room in Washington, DC and in local public document rooms as may be required by NRC regulations if the number of copies submitted is insufficient for this purpose. The NRC is not authorized to make copies for the personal use of members of the public who make use of the NRC public document rooms. Copies made by the NRC must include the copyright notice in all instances and the proprietary notice if the original was identified as proprietary.
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Proprietary Information Notice Transmitted herewith are proprietary and/or non proprietary versions of documents furnished to the NRC in connection with requests for generic and/or plant-specific review and approval.
In order to conform to the requirements of 10 CFR 2.790 of the Commission's regulations concerning the protection of proprietary information so submitted to the NRC, the information which is proprietary in the proprietary versions is contained within brackets, and where the proprietary information has been deleted in the non proprietary versions, only the brackets remain (the information that was contained within the brackets in the proprietary versions having been deleted). The justification for claiming the information so designated as proprietary is indicated in both versions by means of lower case letters (a) through (g) contained within parentheses located as a superscript .
immediately following the brackets enclosing each item of information being identified as proprietary or in the margin opposite such information. These lower case letters refer to the types of information Westinghouse customarily holds in confidence identified in Sections (4)(ii)(a) through (4)(ii)(g) of the affidavit accompanying this transmittal pursuant to 10 CFR 2.790(b)(1).
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CAW-91-192 AfflDAVIT COMMONWEALTH OF PENNSYLVANIA:
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COUNTY OF ALLEGHENY: '
Before me, the undersigned authority, personally appeared Robert W. Beer, who, being b: me duly sworn according to law, deposes and says that he is autnorized to execute this Affidavit on behalf of Westinghouse Electric Corporation (" Westinghouse") and that the avermcnts of fact set forth in this Affidavit are true and correct to the best of his knowledge, information, and belief:
Robert W. Beer, Manager Operations Engineering Technology l Sworn to and subscribed before me this M iay of % _ , 1991.
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l Notary Public I
NOI AA;Ai O! AL LoRR AiNE M P!PLH",A, f40TAnY PUBUC MONROiVILLE 0040, ALLEGHENYCOUNIY MY CoMM:sstCid EXP:HES oEC 14.17)!
Member.Pern./ vane hwann r.f tP/et.s
CAW-91-192-()) I am Manager, Operations Engineering Technology, in the Nuclear and Advanced Technology Division, of the Westinghouse Electric Corporation and as such, I am authorized to perform, on the behalf of Ronald P. DiPiazza, i the function of reviewing the proprietary information sought to be withheld :
from public disclosure in connection with nuclear power plant licensing and rulemaking proceedings, and am authorized to apply for its withholding on behalf of the Westinghouse Energy systems Business Unit. ,
(2) I am making this Affidavit in conformance with the provisions of 10CFR Section 2.790 of the Commission's regulations and in conjunction with the Westinghouse application for withholding accompanying this Affidavit.
(3) I have personal knowledge of the criteria and procedures utilized by the Westinghouse Energy Systems Business Unit in designating information as a trade secret, privileged or as confidential commercial or financial information.
(4) Pursuant to the provisions of paragraph (b)(4) of Se; tion 2.790 of the
. Commission's regulations, the following is furnished for consideration by
- - the Commission in determining whether the information sought to be withheld
-from public disclosure should be withheld.
(i) The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse.
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CAW-91-192 (ii) The information is of a type customarily held in confidence by Westinghouse and not customarily disclosed to the public.
Westinghouse has a rational basis for determining the types of information custoniarily held in confidence by it and, in that connection, ut111zes a system to determine when and whether tn hold certain types of information in confidence. The application of that system and the substance of.that system constitutes Westinghouse policy and provides the rational basis required.
Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows:
(a) The information reveals the distinguishing aspects of a process (or component, structure, tool, method,- etc.) where prevention of its use by any of Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.
(b) It consists of supporting data, including test. data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage, e.g., by optimization or improved marketability.
CAW-91-192
-(c) Its.use-by a competitor would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing a similar product.
(d) It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers.
L (e) It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse.
.(f) It contains patentable ideas, for which patent protection may be desirable.
(g) It is not the property of Westinghouse, but must be treated as proprietary by Westinghouse according to agreements with the owner.
There are sound policy reasons behind the Westinghouse system which
-include the following:
(a) The use of such information by Westinghouse gives Westinghouse a competitive advantage over its competitors. It-is, therefore,-
withheld from disclosure to protect the Westinghouse competitive position.
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CAW-91-192 (b) It is information which is marketable in many ways. The extent to which such information is available to competitors diminishes the Westinghouse ability to sell products and services involving the use of the information.
(c) Use by our competitor would put Westinghouse at a competitive disadvantage-by reducing his expenditure of resources at our expense.
(d) Each component of proprietary information portinent to a particular competitive advantage is potentially as valuable as the total competitive advantage. If competitors acquire components of proprietary information, any one component may be the key to the entire puzzle, thereby depriving Westinghouse of a competitive advantage.
! (e) Unrestricted disclosure would jeopardize the position of j prominence of Westinghouse in the world market, and thereby give a market advantage to the competition of those countries.
(f). The Westinghouse capacity to invest corporate assets in research .
l and development depends upon the success in obtaining and maintaining a competitive advantage.
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CAW-91-192 l 1 i
(iii)-- The information is being transmitted to the Commission in ;
confidence and, under the provisions of 10CFR Section 2,790, it )
is to be received in confidence by the Commission. .
(iv)- The information sought to be protected is not available in public l sources or available information has not been previously employed l l in the same original manner or method to the best of our )
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knowledge and belief.
(v) The proprietary information sought to be withheld in this )
j- suhiittal is that which is appropriately marked in " Westinghouse ECCS Evaluation Model for Analysis of CE-NSSS", WCAP-13027, (Proprietary), July 1991, for fort Calhoun Station Unit 1, being transmitted by the Omaha Public Power District Company (0 PPD) letter and Application for Withholding Proprietary Information from Public Disclosure, Mr. W. G. Gates, OPPD, to Document Control Desk. The proprietary information as submitted for use by Omala Public Power District for the fort Calhoun Station Unit 1- is expected to be applicable in other licensee submittals in resptnse to certain NRC requirements for justification of Westinghaus; evaluation models for Combustion Engineering NSSS.
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2 CAW-91-192 This information is part of that which will encble Westinghouse to:
(a) Justify tne applic4 tion of the Westinghouse evaluation model to CE NSSS.
(b) Provide analysis methodology to perform large and small break LOCA analyses or CE NSSS.
(c) Assist the customer in obtain a licensee.NRC approval, further this information har substantial commercial value as follows:
(a) Westinghouse plans to sell the use of similar information to its customers for purposes of satisfying NRC requirements for licensing documentation.
(b) Westinghouse can sell support and defense of this methodology to its customers in the licensing process, I
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c;-
CAW-91-192 Public disclosure of this proprietary infornation is likely to 1 cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar analytical methodologies and licensing defense services for commercial power reactors without commensurate expenses. - Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without-purchasing the right to use the information.
The development of the technology described in part by the information is-the result of applying the results of many years of experience in an intensive Westinghouse effort and the expenditure of a considerable sum of money.
In order for competitors of Westinghouse to duplicate this information, similar technical programs would have to be performed and a significant manpower effort, having the requisite talent and experience, would have to be expended for the development of analytical methods and testing,-
Further the deponent sayeth not,
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FORT CALHOUN UNIT 1 CONTROL ELEMENT ASSEMBLY EJECTION ANALYSIS REPORT -
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CONTROL ELEMENT ASSEMBLY EJECTION ACCIDENT
- Genera 1 The CEA ejection accident is defined as the mechanical failure in the form of a complete circumferential rupture of a CEDM housing or nozzle on the reactor vessel head resulting in the ejection of a control rod. The consequence of this mechanical failure is a rapid reactivity insertion which <
when combined with an adverse power distribution may result in localized fuel damage.
In design and fabrication, the CEDM is considered to be an extension of the reactor coolant system boundary; hence the probability of such a failure is equivaient to any other rupture of the reactor coolant system and is considered highly unlikely. Further, even if the CEA nozzle should separate from the reactor vessel head, its potential vertical upward travel is limited by the missile shield blocks placed over the reactor head and drive mechanisms. The missile shield block placement will allow an upward movement of only 18 inches; therefore, an additional failure in the drive train must be postulated for the continued CEA ejection. In addition, if the ejection continues, it will do so at a substantially lower rate.
In the following analysis, it is assumed that a CEA is ejected instantaneously from the core, although no mechanism for such an event has been identified. The analytical results presented in this section deal with the nuclear' portion of the transient, which is terminated within several seconds.
The analysis was performed for hot zero power and hot full power initial conditions assuming the most adverse initial CEA configurations which are determined from the Technical Specification on power dependent insertion limits (PDIL). 'Oual CEAs are not considered, because the PDil prohibits their insertion when critical. At zero power Groups 1 and 2 must be totally withdrawn and Group 3 at least 20% withdrawn. At full power all Groups except Group 4 must be withdrawn, and the Group 4 insertion is limited to 75%
withdrawn (see Figure 2-4 of-Technical Specifications).
If.the reactor is subcritical, Technical Specifications require all shutdown CEA's to-be withdrawn before any regulating CEA's are withdrawn and all regulating CEA's to be inserted before any shutdown CEA's can be inserted.
These specifications require that during shutdown dissolved boron concentration must be maintained such that all shutdown CEA's and Groups.1 and
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2 regulating CEA's must be fully withdrawn and Group 3 r gulating CEA's must be at;1 east 20% withdrawn in order to achieve criticality. Ejection of any
- one-dual:CEA when the reactor--is subcritical under the above conditions cannot result in criticality, because the worth of any one dual CEA is less than the combined worth of all shutdown and regulating-CEA's.
Following _the rapid ejection of a CEA, either from full power or zero power (critical) initial conditions, the core power rises rapidly for a brief-period until the increasing reactivity loss due to the widening absorption-resonances (Doppler effect) in U-238 terminates and reversesthe increasing power transient. Inc easing power will initiate a variable high power trip at 19% for the zero power case and a high power trip for the full-power case, causing the CEA banks to insert which reduces the neutron power to negligible levels.
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The loss of coolant resulting from the circumferential rupture of a CEDM housing or nozzle, and its consequences are bounded by.the scope of the small-break. loss of coolant accident which is discussed in USAR Section 14.15.
Method of Analysis The analysis of the CEA ejection accident s performed in two '
stages:-(a):an average core nuclear power transient calculation and (b) a hot spot heat transfer calculation. The average core calculation is performed using spatial neutron kinetics methods to determine the average pnw generation with time including the various total core feedback ef fects, i.e., Doppler reactivity and moderator reactivity. Enthalpy and temperature transients in the hot spot are then determined by multiplying the average core energy generation by_ the hot channel factor and performing a fuel rod transient heet transfer calculation.
The power distribution calculated without feedback is conservatively assumed to exist throughout the transient. A detailed discussion of the method of analysis can be found in Reference 1.
The spatial kinetics computer code, TWINKLE (Reference 2), is used for the average core transient analysis. This code solves the two group neutron diffusion theory kinetic equations in one, two, or three spatial dimensions (rectangularcoordinates)forsixdelayedneutrongroupsandupto2000 spatial points. The computer code includes a detailed multiregion, transient fuel-clad-coolant heat transfer model for calculating pointwise Doppler, and moderator feedback effects.
In this analysis, the code is used as a one-dimensional axial kinetics code since it allows a more realistic representation of the. spatial effects of axial moderator feedback and CEA movement. However, since the radial dimension is missing, it is still necessary to employ very conservative methods (described below) of calculating the ejected rod worth and hot channel factor.
The average core energy addition, calculated as described above.
- is multiplied by'the appropriate hot channel factors, and the hot spot analysis is performed using the detailed fuel and cladding transient heat transfer computer code, FACTRAN (Reference 3). This computer code calculates the
- transient. temperature distribution in a cross section of a metal clad UO2 fuel rod,- and the heat flux at the surface of the rod, using as input the nuclear
-power versus time and the local coolant conditions. The zirconium-water reaction is explicitly represented, and - all material properties are represented as functions of temperature. A conservative parabolic radial pellet power generation is used within the fuel rod.
FACTRAN uses the Dittus-Boelter (Reference 4) or Jens-Lottes (Reference 5) correlation to determine the film heat transfer before DNB, and the Dishop-Sandberg-Tong correlation -(Reference 6) to determine the film boiling coefficient af ter DNB. The DNB heat flux is not calculated; instead the code is forced into DN8 by specifying a conservative 'DNB heat flux. The gap heat transfer coef_ficient can be-calculated by the code; however, it is adjusted in order to force the full power steady state pellet temperature distribution to agree with that predicted by design fuel heat transfer codes.
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L for full power cases, the design initial hot channel peaking factor is input to the code. The hot channel factor during the transient is assumed to increase linearly from the initial steady state _ design value to the maximum transient value in 0.05 seconds,-and remain at the maximum for-the duration of the transient. The values for ejected rod worths and peaking factors are calculated using multi-dimensional calculations. No credit'is taken for the flux-flattening effects of reactivity feedback. This is conservative, since ,
detailed spatial kinetics models show that the hot channel factor decreases shortly after the nuclear power peak due to power flattening caused by
- preferential feedback in the hot channel. Appropriate margins are added to the results to allow for calculational uncertainties.
Results The magnitude of fuel failure can be determined by the following limits:
(1) The average fuel pellet deposited energy at the hot spot is no greater than 200 cal / gram (clad damage threshold). ,
-(2)= The centerline enthalpy threshold for incipient melting is no greater than 250 cal / gram.
_(3) The centerline enthalpy threshold for the fully molten condition is no greater than 310 cal / gram.
The criterion for determining the fraction of fuel rods that will release their radioactive fission products during the CEA ejection is the same asitem(1)abov'efordeterminingcladdamace. Thus, it is assumed that any fuel rod that exceeds a total average enthalpy of 200 cal / gram releases all of its gap activity. The gap activity corresponding to the most-limiting fuel rod during the cycle is conservati'vely assumed for each rod that suffers clad damage.
Table 1 lists the significant input variables for the limiting analyses at full power and zero power. All of the ejected CEA worths and radial i peaking factors include appropriate allowances for calculation uncertainties.
In all cases. analyzed, a conservative value of 0.05 seconds was assumed-for the total ejection time. For the full-power and zero power ct.ses, a Variable Overpower trip is conservatively assumed to initiate at 112% and 29.1% (19.1%
+10% uncertainty)offullpower,.respectively. The initial conditions assume thecorewasoperatipgat102%offullpowerforthefull_powercaseswhilean initial power of 10' of nominal was ' assumed for. the zero power case.
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1 TARLE 1
.QffA EJECTION ACCIDENT ASSUMPTION,$,
Analysis Parameter Units value Full Power Moderator Temperature 10dap/ UF 40.5 Coefficient Doppler Defect %4p -1.25 Ejected CEA worth %6p 0.36 ,
Delayed Neutron 0.0061 Fraction, b-Pre-ejected Rod 2.52 Hot Spot Peaking Factor Post-ejected Rod 6.93 Hot Spot Peaking factor
~CEA Worth at Trip %6p 4.2 Zero Power Nominal Core Power 4 10 Fraction l
Ejected CEA worth
- %4p- 0.69 l
l Delayed Neutron 0.0061 Fraction, b.
Post-ejectedRod 10.51 Hot Spot Power CEA Worth at Trip %4p 1. 8i The results of the full and zero power CEA ejection events may be:found in Table 2. This analysis was assessed against the Regulatory Guide 1.77 criteria
- (Reference 7) which . limits the average hot pellet enthalpy to -less than 280 l cal / gram. The previous acceptance critaria of 200 cal / gram is more conservative with respect to the Regulatory Guide .mit. The centerline melt criterion was not assessed in this analysis since the Regulatory Guide does not require it.
TABLE 2 CEA EJECTION ACCIDENT RESULTS Analysis Parameter value fullLPower Total Average Enthalpy of Hottcst Fuel Pellet (cal / gram) 182.1 Total Centerline Enthalpy of Hottest Fuel Pellet (cal / gram) 286.6 Fraction of Rods That Suffer Clad Damage (AverageEnthalpy1200 cal / gram) 0.0 Fraction of Pellet at Hot Spot Having at least (Centerline Incipient Centerline Enthalpy 1 250~ cal Melting
/ gram ) 0.09 Fraction of Fuel Having a Fully Molten Centerline Condition'(Centerline Enthalpy 1 310 cal / gram) 0.0 Zero Power Total Average Enthalpy of Hottest Fuel Pellet (cal / gram) 60.6 Total Centerline Enthalpy of Hottest Fuel -
Pellet (cal / gram) 71.8 1 Fraction of Rods That Suffer Clad Damage (Average Enthalpy 1 200 cal / gram) 0.0 Fraction of Pellet at Hot Spot Having at.
Least Incipient Centerline Melting
-(Centerline Enthalpy-1 250 cal / gram) 0.0 Fraction of Fuel Having a Fully Holten CenterlineCondition(Centerline'Enthalpy
-2 310 cal / gram) 0.0 Radiolooical Consecuences~
The analysis of radiological consequences of a CEA- ejection accident '
considers the_ release of secondary -coolant activity as well as the reactor coolant activity released through the ruptured CEDM housing. The major assur.ptions used in the analysis are:
- 1. CEA ejection occurs while the reactor is operating at 102% of 1500 MWt with 1% failed fuel and a 1.0 gpm primary-to-secondary leak.
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' 2. The steam generator equilibrium activity for both steam generators is assumed to be 0.1 iCi/gm DEC 1-131,
- 3. Offsite power is lost; the main condenser is not available for steam relief via the turbine bypass system.
- 4. The activity available for leakage from containment is based on the equilibrium reactor coolant activity. The activity instantaneously available for release from the containment is 100% of the noble gases and 25% of the halogens.
- 5. The containment leakage rate is assumed to be 0.2 volume percent per day for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and 0.1 volume percent per day for the uuration of the accident (1-30 days).
- 6. A post-accident decontamination factor of 10 was used in the steam generator between the water and steam phases.
- 7. The total activity released from the' secondary system is presented in Table 3.
TADLE 3 ACTIVITIES RELEASED FROM THE SECONDARY SYSTEM Nuclide Activity (Ci)
Kr-83m 5.0 E-02 Kr-85m 2.6 E Kr-85 4.4 E+01 Kr-87 1.4 E-01 Kr-68 4.8 E-01.
Xe-131m 3.6 E-01 Xe-133m 5.5 E-01 Xe-133' 5.0 E+01 Xe-135m 3.1 E-02 Xe-135 8.5 E-01 Xe-138 1.1 E-01 1-131 2.9 E+00 I-132 2.2 E-01 I-133 1.3 E+00-I-134- 3.7 E-02 1-135 4.9 E-01 A
- 8. The total activity released from the containment, 0-2 hours and for 0-00 days, is presented in Table 4.
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l TABLE 4 ACTIVITIES RELEASED FROM THE CONTAINMENT Nuclide~ Activity (Ci) 0-2 hrs 0-30 days Kr-83m 2.23 E-03 2.23 E-03 Kr-85m 1.40 E-02 1.46 E-02 4 Kr-85 2.66 E+00 4.57 E+02 Kr-87 5.37 E-03 5.37 E-03 Kr-88 2.36 E-02 2.37 E-02 Xe-131m' 2.16 E-02 1.72 E+00 Xe-133m 3.46 E-02 5.34 E-01 Xe-133 3.15 E+00 1.26 E+02 Xe-135m 3.75 E-04 3.75 E-04 Xe-135 4.95 E-02 7.80 E-02 Xe-138 1.11 E-03 1.11 E-03 1-131 1.99 E-02 1.18 E+00 1-132 3.75 E-03 3.76 E-03 1-133 1.94 E-02 8.75 E-02' I-134 1.25 E-03 1.25 E-03 I-135 9.93 E-03 1.21 E-02 9.. The dispersion factors for the EA8 gnd the LPZ outer boundary are 2.55 E-04 sec/m and 4.53 E-06 sec/m , respectively (Reference 8).
- 10. The ad it breathing rate for the EAB and LPZ is assumed to be 3.47 E-04 m sec.
Based on these assumptions, the results doses are as follows:
Thyroid Whole Bod (Rems) (Rems)
EAB 4.4 E-01 1.8 E-03 LPZ 9.7 E-03 1.7 E-04 Conclusions The analysis of the CEA ejertion accident shows that the energy increase
.at the hottspot is limited and that no fuel rods suffer any significant damage following a.CEA ejection from full or zero power at beginning or end of cycle.
The results of radiological consequences of a-CEA ejection accident are
-presented above. The calculated values for thyroid dose and whole body dose show-
' that the doses based on conservative assumptions are well within the . limits specified in 10CFR, Part 100.
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References-
- 1. D. H. Risher, Jr., ,An Evaluation of the Rod Eiection Accident in Westinohouse Pressurized Water Reactors Usina SD6tial Kinetics Methods, WCAP-7588, Revision I-A, January 1975.
- 2. D. H. Risher, Jr., and R. F. Barry, TWINKLE - A Multi-Dimensional Neutron Kinetics Computer Code, WCAP-7979-P-A (Proprietary),
WCAP-8028-A (Non-Proprietary), January 1975.
- 3. H. G. Hargrove, FACTRAN - A FORTRAN IV Code._for Thermal Transients in a 002 Fuel Rod, WCAP-7908-A, December 1989.
- 4. F. W. Dittus and L. M. K. Boelter, University of California (Berkeley), Pubis.Eng., 2, 433, 1930,
- 5. W. H. Jens and P. A. Lottes, Analysis of Heat Transfer, Burnout, Pressure Droo. and Density Data for Hiah Pressure Water, USAEC Report ANL-4627, 1951.
- 6. A. a. Bishop, et al., " Forced Convection Heat Transfer at High Pressure After the Critical Heat Flux," ASME 65-HT-31, August 1965.
- 7. Regulatory Guide 1.77, " Assumptions Used for Evaluating a Control Rod Ejection Accident for Pressurized Water Reactors", U. S.
Nuclear Regulatory Commission, May, 1974.
- 8. Gebers, S., " Radiological Services, Atmospheric Dispersion: USAR Calculations", October 31, 1990.
- 9. OPPD Engineering Analysis EA-FC-91-001, "1% Failed Fuel" Rev. O.
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FORT CALHOUN UNIT 1 CONTROL ELEMENT ASSEMBLY EJECTION ANALYSIS REPORT
CONTROL ELEMENT ASSEMBLY EJECTION ACCIDENT
. General The CEA ejection accident is defined as the mechanical failure in the form of a completo circumferential rupture of a CEDM housing or nozzle on the reactor vessel aead resulting in the ejection of a control rod. The ;
consequence of this mechanical failure is a rapid reactivity insertion which !
when combined with an adverse power distribution may result in localized fuel 1 damage. !
In design and fabrication, the CEDM is considered to be an extension of the reactor coolant system boundary; hence the probability of such a failure is equivalent to any other rupture of the reactor coolant system and is considered highly unlikely. Further, even if the CEA nozzle should separate from the reactor vessel head, its potential vertical upward travel is limited by-the missile shield blocks placed over the reactor head and drive mechanisms. The missile shield block placement will allow an upward movement of only 18. inches; therefore, an additional failure in the drive train must be postulated for the continued CEA ejection. In addition, if the ejection continues, it will do so at a substantially lower rate.
In the following analysis, it is assumed that a CEA is ejected instantaneously from the core, although no mechanism for such an event has been identified. .The analytical results presented in this section deal with the nuclear portion of the transient, which is terminated within several seconds.
The analysis was performed for hot zero power and hot full power initial conditions assuming the most adverse initial CEA configurations which are determined from the Technical Specification on power dependent insertion limits (PDIL). Dual CEAs_are not considered, because the PDIL prohibits their insertion when critical. At zero power Groups 1 and 2 must be totally withdrawn and Group 3 at least 20% withdrawn. At full power all Groups except Group 4 must be withdrawn, and the Group 4 insertion is limited to 75%
withdrawn (see Figure 2-4 of Technical Specifications).
If the reactor is subcritical, Technical Specifications require all shutdown CEA's to be withdrawn before any regulating CEA's are withdrawn and all regulating CEA's to be inserted before any shutdown CEA's can be inserted.
These specifications require that during shutdown dissolved boron concentration must be maintained such that all shutdown CEA's and Groups 1 and 2 regulating CEA's must be fully withdrawn and Group 3 regulating CEA's must be at least 20% withdrawn in order to achieve criticality. Ejection of any one dual CEA'when the reactor is subcritical under the above conditions cannot result-in criticality, because the worth of any one dual CEA is less than the combined worth of all shutdown and regulating CEA's.
Following the rapid ejection of a CEA, either from full power or zero ,
power (critical) initial conditions, the core power rises rapidly for a brief period until the increasing reactivity loss due to the widening absorption resonances (Doppler:effect) in U-238 terminates and reversesthe increasing power transient. Increasing power will initiate a variable high power trip at 19% for the zero power case and a high power trip for the full power case, causing the CEA banks to insert which reduces the neutron power to negligible levels.
The loss of coolant resulting from the circumferential rupture of a CEDM housing or nozzle, and its consequences are bounded by the scope of the small break loss of coolant accident which is discussed in USAR Section 14.15.
Method of Analysis The analysis of the CEA ejection accident is performed in two stages: (a)-an average core nuclear power transient calculation and (b) a hot spot heat transfer calculation. The average core calculation is performed using spatial neutron kinetics methods to determine the average power generation with time including the various total core feedback effects, i.e., Doppler reactivity and moderator reactivity. Enthalpy and temperature transients in the hot spot are then determined by multiplying the average core energy generation by the hot channel factor and performing a fuel rod transient heat transfer calculation.
The power distribution calculated without feedback is conservatively assumed to exist throughout the transient. A detailed discussion of the method of analysis can be found in Reference 1.
The spatial kinetics computer code, TWINKLE (Reference 2), is used for the average core transient analysis. This code solves the two group neutron diffusion theory kinetic equations in one, two, or three spatial dimensions (rectangularcoordinates)forsixdelayedneutrongroupsandupto2000 spatial points. The computer code includes a detailed multiregion, transient fuel-clad-coolant heat transfer model for calculating pointwise Doppler, and moderator feedback effects.
In this analysis, the code is used as a one-dimensional axial kinetics code since it allows a more realistic representation of the spatial effects of axial moderator feedback and CEA movement. However, since the radial dimension is missing, it is still necessary to employ very conservative methods (described below) of calculating the ejected rod worth and hot channel factor.
The average core energy addition, calculated as described above, is multiplied by the appropriate hot channel factors, and the hot spot analysis is performed using the detailed fuel and cladding transient heat transfer computer code, FACTRAN (Reference 3). This computer code calculates the transient tenperature distribution in a cross section of a metal clad U02 fuel rod, and the heat flux at the surface of the rod, using as input the nuclear power versus time and the local coolant conditions. The zirconium-water reaction is explicitly represented, and all material properties are represented as functions of temperature. A conservative parabolic radial pellet power generation is used within the fuel rod.
FACTRAN uses the Dittus-Boelter ~ (Reference 4) or Jens-Lottes (Reference 5) correlation to determine the film heat transfer before DN8, and the Bishop-Sandberg-Tong correlation (Reference 6) to determine the film boiling coefficient after DNB. The DNB heat flux is not calculated; instead the code is forced into DNB by specifying a conservative DNB heat flux. The gap heat transfer coefficient can be calculated by the code; however, it is adjusted in order to force the full power steady state pellet temperature distribution to agree with that predicted by design fuel heat transfer codes.
C
For-f ull power cases, the design initial hot channel peaking f actor is input to the code. The hot channel factor during the transient is assumed to increase linearly from the initial steady state design value to the maximum k nsient value in 0.05 seconds, and remain at the maximum for the duration of u:e transient. The values for ejected-rod worths and peakfr,1 factors are calculated using multi-dimensional calculations. No credit is taken for the flux-flattening effects of reactivity feedback. This is conservative, since detailed spatial kinetics models show that the hot channel factor decreases shortly after-the nuclear power peak due to power flattening caused by preferential-feedback in the hot channel. Appropriate margins are added to the results to allow for calculational uncertainties.
Results The magnitude of fuel failure can be determined by the following limits:
(1) The average fuel pellet deposited energy at the hot spot is no greater than 200 cal / gram (clad damage threshold).
(2) The centerline enthalpy threshold for incipient melting is no granter than 250 cal / gram.
(3) The centerline enthalpy threshold for the fully molten condition is no greater than 310 cal / gram.
The criterion for determining the fraction of fuel rods that will release their radioactive fission products during the CEA ejection is the same as item (1) above for determining clad damage. Thus, it is assumed that any fuel rod that exceeds a total average enthalpy of 200 cal / gram releases all of its gap activity. The gap activity corresponding to the most limiting fuel rod during the cycle is conservatively assumed for each rod that suffers clad damage.
Table 1 lists the significant input variables for the limiting analyses at full power and zero power. All of the ejected CEA worths and radial peaking factors include appropriate allowances.for calculation uncertainties, in all cases analyzed, a conservative value of 0.05 seconds was assumed for the total ejection. time. . For the full power and zero power cases, a-Variable Overpower trip is conservatively assumed to initiate at 112% and 29.1% (19.1%
+ 10% uncertainty) of_ full power, respectively. The_ initial conditions assume
-the core was operatipg at 102's of full _ power for the- full power cases while an initial power of 10' of nominal-was assumed for the zero power case.
e
..w,- , - . . - . 3y-
- 1. -
TABLE 1 CEA EJECTION ACCIDENT ASSUMPTIONS Analysis Parameter Units Value Full Power Moderator Temperature 1044p/ F +0.5 Coefficient Doppler Defect %Ap -1.25 Ejected UA worth- %Ap 0.36 Delayed Neutron 0.0061 Fraction, b Pre-ejected Rod _ 2.52 Hot Spot Peaking Factor Post-ejected Rod- 6,93 Hot Spot Peaking Factor CEA Worth at Trip fsAp 4.2 Zero Powel Nominal Core Power 10 4
Fraction Ejected CEA worth %Ap 0.69 Delayed Neutron 0.0061 Fraction, b Post-ejected Rod 10.51 Hot Spot Power CEA Worth at Trip
%6p 1.5
-The results of the full and zero power CEA ejection events may be'found in Table 2. This analysis was assessed against the Regulatory Guide 1,77 criteria
-(Reference 7) which limits the average hot pellet enthalpy to less than 280-cal / gram. The previous acceptance criteria of 200 cal / gram is more conservative with respect to the Regulatory Guide limit. The centerline melt criterion was not assessed in this analysis since the Regulatory Guide-does not require'it.
l l
l i
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t TABLE 2 -
! CEA EJECTION ACCIDENT RESULTS Analysis Parameter Mgg s
Full Power Total Average Enthalpy of Hottest Fuel i Pellet (cal / grain) 182.1 Totti Centeritne Enthalpy of Hottest Fuel Pellet (cal / gram) 286.6 ,
Fraction of Rods That Suffer Clad Damage (AverageEnthalpy1200 cal / gram) 0.0 Fractica of Pellet at Hot Spot Having at leastincipientCenterlineMelting)
(Centerline Enthalpy 1 250 cal / gram 0.09 I Fraction of fuel Having a Felly Holten ConterlineCondition(CenterlineEnthalpy 1 310 cal / gram) 0.0 1ero Power ,
Total Average Enthalpy of Hottest Fuel '
Pellet (cal / gram) 60.6 Total Centerline Enthalpy of Hottest Fuel !
Pollet(cal / gram) 71.8 Fraction of Rods That Suffer Clad Damage lverageEnthalpy1200 cal / gram) 0.0 Fraction of Pellet at Hot Spot Having at t leastIncipientCenterlineMelting)
(CenterlineEnthalpy1250 cal / gram 0.0 Fraction of fuel Having a fully Holten Centerline Condition (Centerline Enthalpy 1 310 cal / gram) 0.0 Radiolooical Conseauences -
The 1alysis of radiological consequences of a CEA ejection accident considers the release- of secondary coolant activity as well as the reactor coolant activity released through the ruptured - CEDM housing. The major assumptlons used in the analysis are:
- 1. CEA ejection occurs while the reactor is operating at 102% of 1500 MWt with 1% failed fuel and a 1.0 gpm primary-to-secondary leak.
- - - - . - , %- . , .,-~-.w-c..c.,v--c-r,-c .+.nww,,.n- .,,,,.m-,--,-.-- --ww-...- , , ,,...r,,-,.%.,v._,.
+,e 7m,, . - , , ,- , ., 4-,c,- v-*
I i'
TABLE 4 E 11VITIES RELEASED FROM TM CONTAINMENT, Nuclide .
Activity (Ci) __.
9-2 hrs 0-30 dan Kr-83m 2.23 E-03 2.23 E-03 Kr-85m 1.40 E-02 1.46 E-02 Kr-85 2.66 E400 4.57 E402 Kr-87 5.37 E-03 5.37 E-03 Kr-88 2.36 E-02 2.37 E-02 Xe-131m 2.16 E-02 1.72 E+00 '
Xe-133m 3.46 E-02 5.34 E-01 Xe-13*- 3.15 E+00 1.26 E+02 Xe-l', n. 3.75 E-04 3.75 E-04 :
Xe-135 4.95 E-02 7.80 E 02 Xe-138 1.11 E-03 1.11 E-03 1-131 1.99 E-02 1.18 E400 1-132 3.75 E-03 3.76 E 03 '
1-133 1.94 E-02 8.75 E-02 1-134- 1.25 E-03 1.25 E-03 1-135 9.93 E-03 1.21 E 02
- 9. ThedispersionfactorsfortheEADgndtheLPZouterboJndaryare 2.55 E-04 sec/m and 4.53 E-06 sec/m , respectively (Reference 8).
10.
The E-04 m ady/sec.lt breathing rate for the EAB and LP7 is assumed to be 3.47 Based on theso-assumptions, the results doses are as follows:
Thyroid Whole Bod (Rems) (Rems)
EAB 4.4 E 01 1.8 E 'D LPZ 9.7 E-03 1.7 E-04
_ Conclusions lhe analysis of the CEA ejection accident shows that the energy increase at the hot spot is-limited and that no fuel rods suffer any significant damage :
following a CEA ejection from full or zero power at beginning or end of cycle.
The results of radiological consequences of a CEA ejection accident are presented above. The calculated values for thyroid dose and whole body dose show that the doses based on conservative assumptions are well-within the limits specified_in 10CfR, Part 100.
l
- _ , - . . . - . - , _ . - , _ . - - . , - . . _ , . . _ . _ . . . _ . . , _ , _ _ _ . . -.__..-.m..-___ m..,--,- - --
h
- 2. The steam generator equilibrium activity for both steam generators is assumed to be 0.1 iCi/gm DEC 1-131. ,
- 3. Offsite power is lost; the main condenser is not available for steam relief via the turbine bypass system.
- 4. The activity available for leakage from containment is based on the equilibrium reactor coolant activity. The activity instantaneously available for release from the containment is 100% of the noble gases and 25% of the halogens.
- 5. The containment leakage rate is assumed to be 0.2 volume percent per day for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and 0.1 volume percent per day for the duration of the accident (1-30 days). '
- 6. A post-accident decontamination factor of 10 was used in the steam
- generator between the water and steam phases, t
- 7. The total activity released from the secondary system is presented in Table 3.
IABLE 3 ACTIVITIES RELEASED FROM THE SECONDARY SYSTEM Nuclide Activity (C1)
Kr-83m 5.0 E-02 Kr-85m 2.6 E-01 Kr 85 4.4 E+01 Kr-87 1.4 E-01 Kr-88 4.8 E-01 Xe-131m 3.6 E-01 '
Xe-133m- 5.5 E-01 Xe-133 5.0 E+01 Xe-135m 3.1 E-02 Xe-135 d.5 t-01 Xe-138 ' 1 E-01 1-131 U.9 E+00 1-132 U.2 E-01 1-133 . 3 E+00 1-134 3.7 E-02 ,
- I-135 4.9 E-01
- 8. The total activity released from the containment. 0-2 hours and for 0-30 days, is presented in Table 4.
!Leferences
- 1. D. H. Risher, Jr., An Evaluatior of the Rod Fiection Accident in Westin0 house Pressurized Water Reactors UsirLg Spatial Kinetics Methods. WCAP-7588, Revision 1-A, January 1975.
- 2. D. H. Risher, Jr., and R. f. Barry, LWLNKlE - A Multi-Dimensiq0M fleutron Kinetics Computer Code, WCAP-7979-P-A (Proprietary),
WCAP-8028-A (Non-Proprietary), January 1975.
- 3. H. G. Hargrove, [ACTRAN - A FORTRAN IV Code for Thenyl Transienti in a U02 Fuel Rod, WCAP-7908-A, December 1989, j l
F. W. Dittus and L. M. K. Boelter, University of California 4
(Berkeley), Pubis. Eng., 2, 433, 1930.
- 5. W. H. Jens and P. A. Lottes, Analysis of Heat Trg g er, Burnout.
Pressure Drqa d _n_d Density Data for Hiah Pressure Water, USALC ,
Report ANL-4627, 1951. l
- 6. A. A. Bishop, et al., "forceo Convection Heat Transfer at High Pressure Af ter the Critical Heat Flux," ASME 65,,1]L-R, August 1965.
- 7. Regulatory Guide 1.77, " Assumptions Used for Evaluating a Control Rod Ejection Accident for Pressurized Water Reactors", U. S.
Nuclear Regulatory Commission, May, 1974.
- 8. Gebers, S., " Radiological Services, Atmospheric Dispersion: USAR Calculations", October 31, 1990.
- 9. OPPD Engineering Analysis EA-TC-91-001, "1%- Failed fuel" Rev. O.
- 10. OPPDcalculationPED-fC-91-lj57,"AtmosphericDispersion: USAR Calculations". g.
f