ML20086J829
| ML20086J829 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 11/27/1991 |
| From: | Gates W OMAHA PUBLIC POWER DISTRICT |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| LIC-91-311R, NUDOCS 9112120095 | |
| Download: ML20086J829 (24) | |
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Omaha Public Power District 444 South 16th Street Mall November 27, 1991 Omaha, Nebraska 68102-2247 l
LIC-91-311R 402/636 2000 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Mail Station Pl-137 Washington, DC 20555
References:
1.
Docket 50-285 2.
Letter from OPPD (W. G. Gates) to NRC (Document Control Desk) dated April 19, 1991 (LIC-91-0126R) 3.
Letter from OPPD (W. G. Gates) to NRC (Document Control Desk) dated July 31, 1991 (LIC-91-198R)
Gentlemen:
SUBJECT:
Cycle 14 Core Reload Analysis Methodology Updates and Cycle 14 CEA Ejection Report In Reference 2, Omaha Public Power District (0 PPD) submitted proposed changes to the Core Reload Analysis Methodology Topical Reports OPPD-NA-8301-P and OPFD-NA'8303-P for Cycle 14.
During a discussion with Mr. Larry Kopp and Mr.
Steve Bloom of your staff on October 22, 1991, OPPD agreed to revise the description of the Westinghouse CEA Ejection Analysis methodology in accordance with Topical Report, WCAP-7588, Revisicn 1.
In the Westinghouse methodology topical report, one of the acceptance criteria is Departure from Nucleate Boiling Ratio (DNBR). Since Westinghouse assumes 10% fuel cladding failure rather than explicitly performing DNBR calculations, OPPD is clarifying 0 PPD-NA-8303-P, Rev. 03 to reflect the Westinghouse methodology and practices being used for Fort Calhoun Station (FCS) and has had Westinghouse revise the Cycle 14 Control Element Assembly (CEA) Ejection Analysis.
l Attachment A contains replacement pages for page number 10 of OPPD-NA-8301-P, Rev. 04 and page numbers 50 and 51 of OPPD-NA-8303-P, Rev. 03.
Page 10 of OPPD-NA-8301-P, Rev. 04 was revised to expand the list of parameters included in the Core Operating Limits Report (COLR) as discussed on October 22, 1991 with your staff.
Pages 50 and 51 of OPPD-NA-8303-P, Rev. 03 were revised in accordance with the CEA Ejection Analysis methodology described above.
Replacement pages for both the proprietary (P) and nonproprietary (NP) versions of OPPD-NA-8301-P, Rev. 04 and OPPD-NA-8303-P, Rev. 03 are included.
- However, OPPD is not submitting an application for withholding proprietary information from public disclosure since these particular pages do not contain proprietary information.
Please replace the corresponding pages of these reports with the attached pages.
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PDR ADOCK 03000205 p
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U. S. Nuclear Regulatory Commission LIC-91-311R Page 2 Attachment B is a replacement for the " Control Element Assembly Ejection Accident Methodology Summary Report" submitted with Reference 2.
This overview report has been revised to reflect the use of the DNBR acceptance criterion discussed above.
The corresponding report submitted with Reference 2 should be replaced in its entirety with this update.
Attachment C is the revised FCS Cycle 14 CEA Ejection Analysis, which assumes a 10% fuel cladding failure for the radiological consequences analysis consistent with the above methodology revisions. The report titled "CEA Ejection Analysis Report" submitted with Reference 3 should be replaced in its entirety with this update.
If you should have any questions, please contact me.
Sincerely, s
'1 Y
QY W. G. Gates i'q Division Manager Nuclear Operations WGG/sel Attachments c:
LeBoeuf, Lamb, Leiby & MacRae (w/o attachments B & C)
R. D. Martin, NRC Regional Administrator, Region IV (w/o attachments)
D. L. Wigginton, NRC Senior Project Manager (w/o attachments)
R. P. Mullikin, NRC Senior Resident Inspector (w/c attachments)
H. Borchert, Director - Division of Radiological Health State of Nebraska (w/o attachments)
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ATTACIBIENT A c
4.
1
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7.0 CORE OPERATING LIMITS REPORT (Continued) g The COLR will consist of the following items:
g Thermal Margin / Low Pressure for 4 Pump Operation Refueling Boron Concentration Limiting Conditions for Operation for Excore Monitoring of LHR Limiting Conditions for Operation for DNB Monitoring Power Dependent Insertion Limit (PDIL)
Unrodded Integrated Radiai Peaking Factor ( Fr)
Unrodded Planar Radial Peaking Factor (F.)
3 Allowablo Peak Linear Heat Rate vs. Bumup F
T, p,T and Core Power Limitations y
Core inlet Temperature in accordance with Reference 7-1 requirements, updates to the COLR during the operating cycle will be issued to the NRC (NRR, Region IV and Senior Resident inspector) concurrent with internal OPPD distribution.
OPPD-NA-8301-P Rev. 04 Page 10 of 25
t.
9 7.0 CORE OPERATING LIMITS REPORT (Continued) g The COLR will consist of the following items:
g Thermal Margin / Low Pressure for 4 Pump Operation Refueling Boron Concentration Limiting Conditions for Operation for Excore Monitoring of LHR Limiting Conditions for Operation for DNB Monitoring Power Dependent insertion Umit (PDIL)
Unrodded Integrated Radial Peaking Factor ( Fr)
Unrodded Planar Radial Peaking Factor (Fxy)
Allowable Peak Linear Heat Rate vs. Bumup Fxy, F,T and Core Power Limitations T
Core inlet Temperature in accordance with Reference 7-1 requirements, updates to the COLR during the operating cycle will be issued tc the NRC (NRR, Region IV and Senior Resident inspector) concurrent with internal OPPD dictribution.
l l
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OPPD-NA-8301-NP Rev. 04 Page 10 of 25 I
l 5.0 TRANSIENT AND ACCIDENT ANALYSIS METHODS (Continued) 5.10 CEA Eiection Accident (Continued) 5.10.2 Analysjs Criteria The CEA ejection event is classified as a postulated accident. The design and limiting criteria are:
A.
Fuel cladding and enthalpy threshold (Reference 5-5) which is characterized as the Prompt Fuel Clad Rupture Threshold. The value used is the radial Average Pellet Enthalpy (at hot spot) 1280 cal / gram.
B.
The peck reactor pressure during a portion of the transient will be less than tne value that will cause stress to exceed the emergency conditions stress limits as defined in Section 111 of the ASME Boiler g
and Pressure Vessel Code. This objective is achieved if the peak RCS pressure does not exceed 2750 psia (110% of design).
C.
Fuel melting will be limited to keep the offsite dose consequences g
within the guidelines of 10 CFR 100.
D.
Offsite dose consequences will be based on the number of fuel rods that fail due to DNBR values less than the analysis CHF Correlation limit. An explicit number of rods will not be calculated, however, a conservatively bounding value of 10% will be used besed on the approved vendor methodology.
5.10.3 Obie.ctives of the Analysis The objective of the cnalysis is to demonstrate that fuel failures are less than those reportad in Section 14.13.4.1 of the Fort Calhoun Station Unit No.1 USAR in that site boundary doses are within the 10 C~R 100 limits. A further objective is to ensure that the "well within" review criteria are met, however, the license basis for Fort Calhoun is maintained as the 10 CFR 100 limits since the plant is not part of the SRP program.
5.10.4 Arlalyti9 Metho_d OPP." utilizes the CEA Ejection Accident Analysis of our current fuel vendor, Westinghouse. This analysis methodology is documented in Reference 5-5 and is performed by Westinghouse. This methodology utilizes physics parameters, calculated by OPPD in accordance with the methods out!ined in Reference 5-6.
OPPD-NA-8303-R Rev. 03 l
Page 50 of 125
5.0 TRANSIENT AND ACCIDENT ANALYSIS METHODS (Continued) 5.10 CEA Ejection Accident (Continued) 5.10.2 Analvsis Criteria The CEA ejection event is classified as a postulated accident. The design and limiting criteria are:
A, Fuel cladding and enthalpy threshold (Reference 5-5) which is characterized as the Prompt Fuel Clad Rupture Threshold. The value used is the radial Average Pellet Enthalpy (at hot spot) 1280 cal / gram.
B.
The peak reactor pressure during a portion of the transient will be less than the value inat will cause stress to exceed the emergency conditions stress limits as defined in Section lli of the ASME Boiler g
and Pressure Vea el Code. Yhis objection is achieved if the peak RCS pressure does not exceed 2750 psia (110% of design).
C.
Fuel melting will be limited to keep the offsite dose consequences g
within the guidelines of 10 CFR 100.
D.
Offsite dose consequences will be based on the number of fuel rods that fait due to DNBR values less than the analysis CHF Correlation limit. An explicit number of rods will not be calculated, however, a conservatively bounding value of 10% will be used based on the approved vendor methodology and previous license submittals.
5.10.3 ObitC1!YR1D11beAlatysis The objective of the analysis is to demonstrate tnat fuel failures are less than those reported in Section 14.13.4.1 of the Fort Calhoun Station Unit No.1 USAR in that site boundary doses are within the 10 CFR 100 limits. A further objective is to ensure that the "well within" criteria are met, however, the l
license basis for Fort Calhoun is maintained as tne 10 CFR 100 ';mits since the plant is not part of the SRP program.
5.10.4 Analysj1 Method l
OPPD utilizes the CEA Ejection Accident Analysis of our current fuel vendor, l
Westinghouse. This analysis methodology is documented in Reference 5-5 and is perforrned by Westinghouse. This methodclogy utilizes physics parameters, calculated by OPPD in accordance with the methods outlined in Reference 5-6.
OPPD-NA-8303-NR Rev. 03 j
Page 50 of 125 l
5.0 TRANSIENT AND ACCIDENT ANALYSIS METHODS (Continued) 5.10 CEA Election Accident (Continued) 5.10.5 Ana_!ysis Results and 10 CEB3059_Clitena_
The results of the CEA Ejection Analysis are reported in Section 14.13 of the Fort Calhoun Station Unit No.1 USAR. The criteria of 10 CFR 50.59 are satisfied if clad damage and the amount of fuel melting are less than or g
equal to the values assumed for input to the Radiological Consequences portion of the analysis.
5.10.6 Conservatism of Results The major area of conservatism is the calculation method used to obtain the ejected CEA worth and the ejected radial peak. The ejected worth and the ejected radial peak are calculated without any credit for Doppler or Xenon feedback. In addition, the hot full power ejected worth and ejected peak are calculated assuming the no-load temperature of 532*F. The lower temperature is more adverse since this causes a power roll to the core periphery which also happens to be the location of the ejected CEA. Also, the ejected worth is calculated assuming the CEAs are fully inserted for hot full power case regardless of PDIL. Thus, the ejected worth is conservative.
5.11 Loss of Coolant Accident OPPD does not perform the Loss of Coolant Accident Analysis. The large and small break loss of coolant analyses were performed by Westinghouse. The large break and small break topicals are mentioned in Reference 5-7. OPPD verifies that the physics input assumptions and the maximum red burnup are within the bounds assumed in the % ! rge break analysis.
g 3
5.12 Loss _of Load to Both StearD_G9D01alGrs_Eyg 5.12.1 Definition of the Event A total loss of load to both steam generators usually results from a turbine trip due to a loss of external electrical load or to abnormal variations in electrical network frequencies. Other possible causes include the simultaneous closure of all turbine stop valves or main steam isolation valves. Allinitiating mechanisms result in a corresponding reduction in heat removal from the reactor coolant system due to the loss of secondary steam flow. Although a Reactor Protective System trip signal would normally result from a turbine trip, no credit is taken in the analysis of this event for the turbine trip signal.
OPPD-NA-8303-P, Rev. 03 Page 51 of 125
5.0 TRANSIENT AND ACCIDENT ANALYSIS METHODS (Continued) 5.10 CEA Ejection Accident (Continued) 5.10.6 AnalysiaBeaulisltrid1CLCER5039_Cutena.
The results of the CEA EJoction Analysis are reported in Scction 14.13 of the Fort Cethoun Station Unit No.1 USAR. The cr'tvria of 10 CFR 50.59 are satisfied it clad damage and the amount of fuel melting are less than or equal to the values assumed for input to the Radiological Consequences portion of the ana'.ysis.
5.10.0 C_oDscryatistitotBesujts The major area of conservatism is the calcitlation method used to obtain the ejected CEA worth and the ejected radial peak. The ejected worth and the ejected radial peak are calculated without ar.y credit for Doppler or Xanon icedback. In addition, the hot full power ejected wotth and ejected peak are calculated assuming the no-load temperaturo of 532* F. The lower temperature is more adverse since this causes a power roll to the core periphery which also happens to be the location of the ejected CEA. Also, the ejected worth is calculated assuming the CEAs oro fully inserted for hot full power case regardlest of PDil. Thus, the ejected worth is conservative, b.C Lostof Coolant Accident OPPD does not perfor:7 the Loss of Coolant Ar.cident Analysis. The large and small break loss of coolant analyses were perfozmed by Westinghouse The large break and small break topicals are mentioned in R,sference 5-7. OPP! verifies that the pi sics input r.ssumptions cnd the maximum rod burnun are within tne bounds
/
assumed in the Westinghouse largo broak analysis, g
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5.12 Lostofl oarttoEnt1Steaanenerator1EYeDI
- 5. 2.1 Definition of the Evant A total loss of load to both steam generators usually results from a turbine trip due to a loss of external electrical load or to abnormal variations in electrical network frequencies. Other possible causes ulclude the simultaneous closure of all turbine stop valves or main steam isolation valves. Allinitiating mechanisms result in a corresponding rsducticn it' icat removal from the reactor coolant system due to the loss of secondary steam flow. Although a Reactor Protective System trip signal would normally result from a turbino trip, no credit is taken in the analysis of this event for the turbine trip signal.
OPPD-NA-8303-NP Rev. 03 Page 51 of 125
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ATTACHMENT B
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l1 OMAHA PUBLIC POWER DISTRICT FORT CALHOUN UNIT 1 CONTROL ELEMLNT ASSEMBLY EJECTION ACCIDENT METHODOLOGY
SUMMARY
REPORT t -
November, 27, 1991 h
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1.0 HETHOD OF ANALYSIS A complete description of the Westinghouse analysis methodology for the CEA ejection event is described in Reference 1.
The methodology described within this report has been approved by the NRC for numerous applications on Westinghouse plants as well as for a Combustion Engineering plant loading Westinghouse fuel. Also found in Reference 1 are numerous sensitivity studies performed which provide the basis for the conservative choice of core physics characteristics used in this analysis. A brief discussion of the methodology found in Reference 1 follows.
The calculation of the CEA ejection event is performed in two stages.
First, an average core channel calculation is done using TWINKLE: and then, a hot spot analysis is done using FACTRAN.
The average core calculation is performed using spatial neutron kinetics to determine the average power generation with time, including the various core reactivity feedback effects, i.e., Doppler and moderator reactivity. The nuclear power increase during this transient will lead to elevated fuel pellet and fuel cladding temperatures.
The TWINKLE code is utiliad, in conjunction with Fort Calhoun Unit 1 plant-specific abysicsdata,.operformaone-dimensional (axial)averagecoreneutron cinetic analysis allowing for a more realistic representation of the spatial effects of axial moderator feedback and CEA movement. Hoq ver, since the radial dimension is missing, it is still necessary to employ very conservative methods of calculating the CEA worth and hot channel peaking factor as discussed below.
The resulting average core nuclear power transient is input to FACTRAN along with the appropriate parameters such as fuel geometry, initial
)ower, nominal average heat flux and core flow rate, initial and final lot spot total peaking factors, pellet power distribution, and gap heat transfer coefficients vs. time. Enthalpy and temperature transients in the hot spot are determined by multiplying the average core energy generation by the hot channel peaking factor and performing a fuel rod transient heat transfer calculation. During the transient, the steady-state heat flux hot channel factor is linearly increased to the transient value in 0.05 second, the assumed time for full ejection of the CEA. Prior to ejection, the power in this region will be depressed.
However, the assumption is made that the hot spots before and after ejection are at the same axial location. This is conservative since the peak power af ter ejection will occur in or adjacent to the assembly with the ejected CEA.
In the hot spot analysis, the transient temperature distribution in a cross section of a metal clad uranium-dioxide fuel rod, and the heat flue. at the surf ace of the rod, is calculated, 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.
The FACTRAN computer code uses the Dittus-Boelter or Jens-Lottes correlation to determine the film heat transfer before DNB, and the Bishop-Sandberg-Tong correlation af ter DNB. Prior to DNB, the code l
automatically selects between the forced convection (Dittus-Boelter) and local boiling (Jens-Lottes) correlations based on the clad temperatures calculated by each.
The Bishop-Sandberg-Tong correlation is conservatively used, assuming zero bulk fluid quality. The DNBR is not calculated; instead, for the full power cases, the code is forced into DNP 0.05 seconds af ter the start of the transient while in the zero p er cases, the code is forced into DNB by specifying a conservative Dhd 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 temperature distribution to agree with the fuel heat transfer design codes.
-Four cases are considered for this event to cover the spectrum of power levels and reactivity conditions which can occur throughout the fuel cycle. Full-power and zero-power cases are analyzed with reactivity coefficients consistent with end-of-life and beginning-of-life core physics conditions.
2.0 Computer Codes 2.1 TWINKLE The TWINKLE code is a neutron kinetics code which solves the multidimensional, two-group transient diffusion equations using a finite-difference technique. The code contains a detailed six-region fuel-clad-coolant transient heat transfer model at each spatial point for calculating Doppler and moderator feedback effects. The method used to calculate feedback is similar to that used in Westinghouse nuclear design codes. TWINKLE handles up to 2000 spatial points in one, two-or three-dimensional rectangular geometry and performs its own steady-state initialization. Aside from basic cross-section data and thermal-hydraulic parameters, the code accepts as input basic driving functions such as inlet temperature,
- pressure, flow, boron concentration, CEA motion and others to produce output of nuclear power as a function of time.
The TWINKLE code is used to predict the neutron kinetic behavior of a reactor core for transients, such as CEA ejection, which cause a major perturbation in the spatial neutron flux distribution.
TWINKLE is further described in Reference 2.
-2.2 FACTRAN The FACTRAN computer code calculates the transient temperature distribution in a cross section of a metal clad, uranium dioxide fuel rod and the transient heat flux at the surface of the clad, using as input the nuclear power and the time-dependent coolant parameters (pressure, flow, temperature and density). The code uses a fuel model containing a sufficiently large number of radial space increments to model even fast transients. FACTRAN also uses material properties
j which are functions of temperature and has the capability to perform a l
sophisticated fuel-to-clad gap heat transfer calculation. Additionally, I
the code can perform the necessary calculations to deal with post-DNB transient phenomena such as:
film boiling heat transfer correlations, the zirconium-water reaction, and partial melting of the materials.
j For the CEA ejection analyses, the nuclear power input is a time-dependent parameter from TWINKLE; however, the input parameters for I
pressure, flow, temperature and density are maintained at a constant conservative value so as to produce a limiting calculation in FACTRAN.
j FACTRAN is further discuSN i 'n Reference 3.
3.0 Safet_y Limits The real physical limits et u,o accident are that any consequential damage to either the wre of the reactor coolant system mot not prevent long-term core cooling and 6at any offsite dose consequences must be within the guidelines of 10 CFR Part 100.
However, the magnitude of fuel failure will be determined by the folloiting limits:
1.
The radial average pellet deposited energy at the hot spot is no greater than 280 cal / gram (prompt fuel clad rupture).
2.
Fuel rods that have a DNBR less than the critical heat flux correlation limit are assumed to fail.
An explicit number of rods will not be calculated, however, a conservatively bounding value will be used based on the approved vendor methodology and previous license submittals.
3.
Fuel melting will be limited to ensure the offsite dose consequences are'within the guidelines of 10CFR100.
The criterion for determining the. fraction of fuel rods that will release their radioactive fission products during the CEA ejection is conservatively assumed to be a radial average pellet deposited energy no greater than 200 cal / gram.
4.0 References 1.
Risher, D. H., Jr., "An Evaluation of the Rod Ejection Accident in Westinghouse Pressurized Water Reactors Using Spatial Kinetics Methods,"WCAP-7588, Revision 1-A, January 1975(Proprietary).
2.
Risher, D. H., Jr., and Barry, R. F.,
" TWINKLE - A Multi-Dimensional Neutron Kinetics Computer code," WCAP-7979-P-A, January 1975 (Proprietary) and WCAP-8028-A, January 1975 (Non-proprietary).
3.
Hunin, C.,
"FACTRAN Code Description," WCAP-7337, Rev 1-P, December 1989(Proprietary).
ATTACHMENT C 3
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FORT CALHOUN UNIT 1 CONTROL ELEMENT ASSEMBLY EJECTION ANALYSIS REPORT l
b l
CONTR0t ELEMENT ASSEMBLY EJECTION ACCIDENT General 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 incalized fuel damage.
In design and fabrication, the CEDM is considered to be an extension of l
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 i
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.
1 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 i
limits (PDIL).
Dual CEAs are not considered, because the pDil prohibits their t
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 I
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 of zero power (critical) initial conditions, the cora nower rises rapidly for a brief period until the-increasing reactivity loss due to the widening absorption resonances (Doppler effect) in U-238 terminates and reverses the increasing power transient.
Increasing power will initiate a variable high power trip at 19.1% 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.
~ -
l 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 TheanalysisoftheCEAejectionaccidentisperformedintwostages;(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 multip1ying 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 (rectangular coordinates) for six delayed neutron groups and'up to 2000 spatial points. The computer code includes a detailed multi-region, transient fuel-clad-coolant heat transfer model for calculating pointwise Doppler, and moderator feedback ef fects.
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 t.nd 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 clidding transient heat transfer computer code, FACTRAN Reference 3). This computer code calculates the transient temperature 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 s
T 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 Bishop-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 DNB by specifying a conservative DNB heat flux. The gap heat transfer coefficient can be caiculated 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.
4
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 f actors are calculated using multi-dim.nsional calculations.
No credit is taken for the flux-flattening effects of reactivity feedback.
This is conservative, since detailed spatial kir.etics 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 radial average fuel pellet deposited energy at the hot spot is no greater than 280 cal / gram (prompt fuel clad rupture).
(2) Fuel rods that have a DNBR less than the critical heat flux correlation limit are assumed to fail.
An explicit number of rods is not calculated, however, a conservatively bounding value is used based on the approved vendor methodology and previous license submittals.
(3) fuel melting is limited to ensure the offsite dose consequences are within the guidelines of 10CFR100.
The criterion for determining the fraction of fuel rods that will release their radioactive fission products during the CEA ejection is similar to item (1) above for determining clad damage.
It is conservatively assumed that any fuel rod that exceeds a radial 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.
The peak reactor coolant system pressure which occurs during the transient will be less than the value that will cause stress to exceed the emergency condition stress limits as defined in Section 111 of the ASME Boiler and Pressure Vessel Code.
This objective is achieved if the peak RCS pressure is less than or equal to 2750 psia.
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 11?% and 29.1% (19.1% + 10's uncertainly) of full power, respectively. The initial conditions assume the core wasoperatiggat102%offullpowerforthefullpowercaseswhileaninitial power of 10 of nominal was assumed for the zero power case.
9 TABLE 1 CEA EJECTION ACCIDENT ASSUMPTIONS Analysis Parameter Units value Full Power Moderator Temperature 10"Ap/ F
+0.5 Coefficient Doppler Defect
%Ap
-1.25 Ejected CEA 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 5 sap 4.2 Zero Power 4
Nominal Core Power 10 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
%Ap 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 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.
TABLE 2 CEA EJECTION ACCIDENT RESULTS Analysis Parameter Value Full Power Radial Average Enthalpy of Hottest Fuel Pellet (cal / gram) 182.1 Total Centerline Enthalpy of Hottest Fuel Pellet (cal / gram) 286.6 Fraction of Rods that Suffer Clad Damage (AverageEnthalpy>200 cal / gram) 0.0 i
(promptfuel_cladruptureT Zern Power Radial Average Enthalpy of Hottest Fuel Pellet (cal / gram) 60.0 Total Centerline Enthalpy of Hottest Fuel Pellet (cal / gram) 71.8 Fraction of Rods that Suffer Clad Damage (AverageEnthalpy>200 cal / gram) 0.0 (promptfuelcladruptureT
Radioloaical Conseauences A conservative analysis of the potential radiological consequences of a CEA Ejection event has been performed in accordance with the guidelines presentedinRegulatoryGuide1.77(Reference 7).
Two radioactivity release paths to the environment are assumed to contribute to the radiological consequences of a CEA ejection accident.
The first is through containment leakage of fission products contained in the primary system. The second is tnrough leakage from the primary system to the steam generators (primary-to-secondary leakage) and release to the environment via the secondary side relief valves.
The salient assumptions used to calculate the activity releases and offsite doses follow.
1.
Prior to the accident, the primary coolant iodine and noble gas concentrations are assumed to equal the 1% fuel defect level, based on plant operations at 1500 MWt (Reference 9).
2.
Prior to the accident, the secondary coolant iodine concentration is assumed to equal the Technical Specification limit for full power operation - 0.1 uCi/ gram of dose equivalent 1-131.
3.
TenpercentofthecoreisassumedtofailasaresultofDNB(Reference 1).
This is assumed to result in the instantaneous release of 10% of the core gap activity to the primary coolant. The fraction of core activity contained in the gap (gap fraction) is assumed to be 10% for all nuclides.
Thus, a total of 1% of the core activity is released.
For the containment leakage release,-100% of this activity is assumed to be instantaneously released to the containment atmosphere.
For the secondary system release, 100% of this activity is assumed to be 1
contained in the reactor coolant.
The core activity is summarized in Reference 11.
4.
One fourth of one percent (0.25%) of the core is assumed to melt. For the containment leakage release, 100% of the noble gases and 25'S of the iodines are assumed to be instantaneously released to the containment atmosphere. For releases through the secondary system, 100% of the noble gases and 50% of the iodines are assumed to be released to the primary coolant.
The melted fuel fraction was determined as follows:
a, A conservative upper limit of-50% of the rods experiencing clad dama e are assumed to experience centerline melting (5% of the Core.
b.
For rods experiencing centerline melting, 10% percent of the rod-volume is assumed to actually melt (equivalent to 0.5% of the core).
c.
A conservative maximum of 50'5 of the axial length of the rod is assumed to experience melting due to the power distribution (0.5 of 0.5% of the core is equal to 0.25'5 of the core).
_ -. - - -. ~. -..
4 i
5.
The total primary-to-secondary leak rate is assumed to be at the Technical Specification limit of 1 gpm.
6.-
Activity released to the environment via the primary to secondary leakage j
pathway is assumed to be released directly to the environment without mixing with the secondary coolant. An iodine decontamination factor of 10 is applied to this activity release.
7.
Offsite power is lost.
8.
Steam re:1 ease to the environment 0 to 50 seconds - 9354 lbm This steam release is used with an iodine partition coefficient of 0.1 to t
determine the release of the initial secondary coolant iodine activity 7
(Item 2).
9.
Containmentleakagerate(volumepercent/ day): 0 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> - 0.1 1 to 30 days - 0.05
- 10. Atmosphericdispersionfactors(sec./cu. meter)(Reference 8):
siteboundary(0to2 hour) 2.55 E-4 low population zone (0 to 30 days) 4.53 E-6 i
11.
Breathing rates (cu. meter /sec.): 0 - 8 hr, 3.47 E-4 8 - 24 hr, 1.75 E-4
> 24 hr, 2.32 E-4 12.
Thyroiddoseconversionfactors(rem / curie): ICRP Publication 2 Results The activity released to the environment from the secondary system is presented in Table 3 TABLE 3
. ACTIVITY RELEASE 0 FROM THE SECONDARY SYSTEM Nuclide Activity (Curies)
(0 - 50 sec.)
Kr-85m 1.5 E 0 Kr-85 7.5 E-2 Kr-87 2.8 E O Kr-88 3.9 E 0 Xe-131m 6.6 E-2 Xe-133 1.1 E 1 Xe-135m 2.2 E O Xe-135 2.8 E 0 Xe-138 9.2 E O I-131 8.9 E-1 1-132 6.8 E-1 1-133 9.6 E-1 1-134 1.1 E 0 1-135 9.0 E-1
The activity releasea to the environment from the containment is presented in Table 4.
TABLE 4 ACTIVITY RELEASED FROM THE CONTAINMENT Nuclide Activity (Curies) 0-2 hours 0-30 days Kr-85m 2.3 E 1 8.3 E 1 Kr-85 1.4 E O 2.5 E 2 Kr-87 3.1 E 1 4.6 E 1 Kr-88 5.6 E 1 1.4 E 2 Xe-131m 1.2 E O 1.1 E 2 Xe-133 2.0 E 2 1.0 E 4 Xe-135m 7.6 E 0 7.6 E O Xe-135 4.8 E 1 3.1 E 2 Xe-138 3.4 E 1 3.5 E 1 1-131 8.5 E 1 6.0 E 3 1-132 9.3 E 1 2.1 E 2 1-133 1.7 E 2 2.1 E 3 1-134 9.6 E 1 1.2 E 2 1-135 1.5 E 2 7.6 E 2 The resulting doses at the exclusion area boundary (EAB) and at the outer boundary of the low population zone (LPZ) are presented in Table 5 below.
TABLE 5 RADIOLOGICAL CONSE0VENCES
SUMMARY
Dose (rem)
Thyroid Whole body aamma Total offsite dose 0-2 hour dose at EAB 19.7 1.6 E-2 0-30 day dose at LPZ 10.0
~.3 E-4 Conclusions The 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 erd 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 dosed based on conservative assumptions are well within the limits specified in 10CFR, Part 100.