ML19343A768
| ML19343A768 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 11/19/1980 |
| From: | Lundvall A BALTIMORE GAS & ELECTRIC CO. |
| To: | Clark R Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8011210291 | |
| Download: ML19343A768 (31) | |
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BALTIMORE G AS AND ELECTRIC COMPANY P.O. B O X 14 7 5 B ALTIMOR E. M ARYL AN D 21203 AITMun E. LumovALL.Jm. l
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, November 19, 1980
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3 A A Office of Nuclear Reactor Regulation T v3 i ll. S. Nuclear Regulatory Commission Q '35 Washington, D. C. 20555 ., ,e. y,"o ,
a 1,7; ATTENTION: Mr. R. A. Clark, Chief y, s
$3 l Operating Reactors Branch #3 '
ha l Division of Licensing % Q -
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SUBJECT:
Calvert Cliffs Nuclear Power Plant '" u ' , ,
Unit No.1, Docket No. 50-317 Amendment to Operating License DPR-53 Fif th Cycle License Application Responses to NRC Staff Questions Gentlemen:
Enclosed are our reponses to questions posed by NRC staff on the subject application. t Very truly yours, BALTIMORE S L TRIC C .iPANY C
A . f. u vall, Jr. V
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Vice Pr sident - Su y AEL/WJL/mit Copy To: 3. A. B'iddison, Esquire (w/out Encl.)
G. F. Trowbridge, Esquire (w/out Encl.) ,
Messrs. E. L. Conner, Jr., NRC i P. W. Kruse, CE ,
Enclosures:
A(StartupTestProgram,40 copies) ,
B (Asyninetric Steam Generator Trip Instrumentathn, 40 copies)
C (TM/LP Trio Setpoint Determination, 40 copies, D (Core Performance Branch, 40 copies) i E (Transient Analysis - 11/12/80 meeting, 40 copies) ;
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ENCLOSURE A QUESTION Al Provide review as well as acceptance criteria for startup test results.
RESPONSE
Paragraph 10.5 of Reference A1 !! rewritten as follows:
10.5 ACCEPTANCE CRITERIA Acceptance and review criteria for the above startup tests are ,
listed below:
Parameters Acceptance Criteria Review Criteria CEA Groups Worth 1,15% of predicted 1,15% of predicted Total Regulating CEA -10% of predicted 110% of predicted Group Worth .
Critical Baron Concen- 1100 ppm of predicted 1,50 ppm of predicted tration Isothermal Temperature Technical Specifica- 1,0.3 X 10-4ap/0F Coefficient tion limits for Moderator Tempera-ture L afficient Power Coefficient 10.3 X 10-4ap/% Power 1,0.2 X 10-4'ap/% Power CEA Symmetry Check None $10% tilt; a tilt of >10%
will be resolved prior to exceeding 20% of maximum allowable thermal power level for the existing RCP combination.
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QUESTION Al Page 2 Parameters Acceptance Criteria Review Criteria Power Distribution Technical Specifica- Measured radial box power '
tion limits on F T, r
distributions will be com-F xyT, and Tq. pared with predictions. If a measurement varies from prediction by more than
, 110% (115% for fuel assemblies on core periphery) then the -
difference will be resolved and the validity of the safety analysis confirmed prior to submittal of the summary report of startup test results.
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QUESTIm A2 Provide an Action and Review plan .for startup test results.
RESPONSE
A paragrcnh 10.6 is added to Reference Al as follows:
10.6 ACTION AND REVIEW PLAN The Supervisor, Nuclear Fuel Management, shall review the comparison -
of measurements with Review / Acceptance Criteria.
If any Review Criteria are exceeded, an evaluation shall be made to detennine first, the applicability of the prediction to the precise plant conditions under which the measurement was performed and, second, the accuracy of the measurement. As a result of this review the measomcant may be repeated.
If any Acceptance Criteria are exceeded, in addition to the above actions, the validity of the physics data input to the Safety Analysis for the entire cycle shall be determined. If it can be demonstrated that the measured value of the particular parameter in question when combined with the values of the other safety related parameters does not increase the severity or consequences of accidents or anticipated operational occurrences, the test results shall be deemed acceptable.
Additional measurements of safety related parameters including CEA Group worth may be perfonned in order to support this demonstration, i If the combination of safety parameters detennined above fall outside of the range of safety parameters used to support the proposed opera-tion of the plant, the plant operating limits shall be adjusted to prevent conditions which could result in exceeding the Specified Acceptable Fuel Design Limits.
QUESTION A2 Page 2 ,
If any measurement from the low power physics tests exceeds its Review Criteria, the Plant Operations and Review Committee shall review results of the lower power physics tests and ensule that Acceptance Criteria are met prior to recannending operation above 5% of Rated Thermal Power. If, as a result of this review, it is determined that a Technical Specification limit has been exceeded, then appropriate action as required by Technical Specifications shall he taken. A similar action plan for escalation to power test measurement results shall be followed prior to exceeding 50% of Ratad Thermal Power.
A-summary report of the results of this test shall be submitted to NRC within 90 days of the completion of the startup test program.
The report shall include a comparison of the measured with predicted value for each test. If the difference between a measured and pre-dicted value exceeds its Review and/or Acceptance Criteria, the report will discuss the actions taken and also substantiate the adequacy of those actions.
REFERENCE Al A. E. Lundvall to R. A. Clark letter, dated 9/22/80, Unit 1, Fifth Cycle License Application i
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QUESTION A3 Provide the results of an analysis which demonstrates that a partial'inser-tion of Shutdown Group CEA's is as sensitive as the full insertion of those same CEA's for the purpose of detecting an azimuthal reactivity tilt.
RESPONSE
An eight percent (8%) azimuthal tilt was simulated at Beginning of Cycle
- using ROCS. A peripheral Shutdown CEA located near the low point of the tilt was inserted fully into the core and then inserted partially into the core (40% insertion). This same procedure was repeated for : 3hutdown CEA in the same Group located 180 0 away from the above CEA and at the high point of the tilt. The reactivity worth of the CEA's associated with those four (4) cases is displayed below.
Tilt Location Fully Inserted CEA Partially Inserted CEA Low .065%ap .032%Ap High .08% ap .038%ap From our measurement experience, the repeatability of the reactivity worth of a partially inserted Shutdown CEA is less than +.003%ap, which is well within the difference in reactivity worth.s of the partially inserted CEA's.
Therefore, it is concluded that the partial insertion of Shutdown CEA's will generate reactivity worths which when combined to determine an ' azimuthal tilt will unambiguously detect a tilt of greater than 10%.
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1 l ENCLOSURE B l
QUESTION B1 What effect does the ASGTPTF modification have on the perfonnance of the existing RPS functions?
RESPONSE
The Asymmetric Steam Generator Transient Protective Trip Function is incorporated into the RPS as part of the Thermal Margin / Low Pressure Trip Function. When the conditions for actuation of ASGTPTF are satisfied
.the TM/LP Bistable Trip unit is forced into a trip condition through satura-tion of its trip setpoint. When the conditions for ASGTPTF actuation are not satisfied then the TM/LP calculator operates as originally designed.
The ASGTPTF in no way can prevent a valid TM/LP trip from occurring.
The ASGTPTF is a four channel system meeting all the design criteria of the present RPS. Incorporation of the ASGTPTF into the TM/LP calculator does not adversely affect the perfor'mance of the existing RPS.
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ENCLOSURE C QUESTION C1 In previous cycles the thermal margin / low pressure trip setpoints included an allowance to compensate for pressure measurement error, trip system processing error, and time delay associated with providing effective tennination of the occurrence that exhibits the most rapid decrease in margin to the safety limit. For Cycle 4, this 84 psia allowance was made up of a 22 psia pressure measurement allowance and a 62 psia time delay allowance. The CEA withdrawal transient initiated at rated power resulted in the maximum pressure bias factor of 62 psia. In the CEA withdrawal analysis for Cycle 5, a maximum pressure bias factor of 70 psia is obtained but it appears that no additional pressure measurement allowance term is included in the TM/LP Tech Spec bases (8 2-7). Also, power measurement error potential (5%) and temperature mreasurement uncertainty (_2 F) were -
dropped. Since the SCU and CEAW topicals will not be reviewed for this reload, shouldn't the TM/LP trip setpoints be rederived as in previous cycles?
RESPONSE
If SCU was implemented,'then all of the uncertainties cited in the ques-tion (except for the 70 psia bias) would-be included in the combined ,
uncertainty. Therefore, these uncertainties would not have to be detailed separately in the Tech Spec Bases, This is why this paragraph was initially -
changed. When we went back to the old method of combining uncertainties, t the Calvert Cliffs Unit 1 Cycle 5 Bases should have been changed back to I the same content as Unit 1, Cycle 4 except for changing the bias input from 62 to 70 psia. The paragraph, as written for the case using SC.
is in error for the revised submittal as it relates to TM/LP uncertainties actually used in the revised submittal. However, the calculation has been ~'
properly performed and does indeed explicitly account for all the uncer-tainties cited in the question in the same manner as they were applied in Cycle 4.
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l ENCLOSURE D-Question D3:
What are the target burnups of the PROTOTYPE assemblies (in terms of both assembly average and lead rod)?
Response -
The target burnups for both the assembly average and -lead rod for the prototype assemblies and 'the average cycle lengths for Cycles 5 through 8 are:
' Cycle Assembly Average Lead Pin (MWD /T)- .(MWD /T) 5 12,600 17,700 6 29,100 32,600
> 7 42',300 47,800 8 50,000 54,900 b
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QUESTION D6 i Section 15.4.7 of the Standard Review Plan for the Review of Safety Analysis
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Reports for Nuclear Power Plants (NUREG-75/087) requires an analysis of possible fuel loading errors such as the loading of one or more fuel assemblies into improper locations. Discuss the analyses for each misloading case f (including the worst case) considered and show that either the error is l detectable .(and thus remedial) or that the error is inconsequential and i within the nuclear uncertainty or that the offsite consequences of any core damage due to undetected errors are a small fraction of 10 CFR 100 guidelines.
' RESPONSE See A. E. Lundvall to R. A. Clark letter dated 11/10/80, Responses to NRC Staff Questions.
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QUESTION D7 A partial list of physics characteristics for Cycles 4 and 5 was presented in the refueling license amendment. Provide a list of final Cycle 5 physics characteristics if different from the original submittal (Tables 5-1 thru 5-6) including the maximum radial power peaks expected to occur (Frand Fxy with uncertainties and biases).
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RESPONSE
See A. E. Lundvall To R. A. Clark letter dated 11/10/80, Responses to NRC Staff Questions.
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Question 08: .
i Please provide the following burnup values:
a) The actual core avarage exposure achieved during Cycle 4.
b) Each batch average exposure for E0C 4.
c) Each batch average exposure for E0C 5.
Respong a) Cycle 4 core average end of cycle exoosure is 21,275 MWD /T.
b) Batch average exposure for EOC 4. .
Batch Exposure (MWD /T)
B 43,200 l D 30,400 E 21,900 -
t l F 11,400 c) Batch average exposure for E0C 5 based on a Cycle 5 length of 13,0C9 MWD /T:
Batch. Exposure (MWD /T) i i D 41,800 E 33,200 F 24,600 G 13,200 l 1
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Question 09:
Cycle 5 will employ several different guide tube and sleeve des.igns that are specifically chosen to mitigate or investigate guide tube wear. Please -
provide a table for the Cycle 5 core inventory that identifies these different, fuel assembly designs. The table should indicate the number of fuel assemblies in each category--sleeved, sleeve design,. guide tube design--and the number of fuel assemblies in each category to reside under CEA's.
Response
Fuel Assembly Category (I) Total Number in Cycle 5 Number Under CEA's Standard w/o Sleeves' 57 1 Standard w/ Sleeves (2) 128 44 Revised Design No. 1 ,
20 20 Revised Design No. 2 12 12 Total 217 77 (I)The fuel assembly categories are as defined in CEN-138(B)-P (Appendix E of
- Calvert Cliffs Unit 1 Cycle 5 Reload Submittal).
(2)The guide tube sleeves are the same design as previously described in CEN-83(B)-P.
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Question D12: ,
Differential growth between fuel rods and the fuel assembly structure is burnup dependent and progresses in such a manner that mechanical inter-ference will o: cur when the shoulder gap is consumed. What assurance '
exists that the shoulder gap in the high burnup fuel assemblies comprising the Cycle 5 core is adequate to preclude this interference.
Response
Fuel rod shoulder gap clearances are reviewed as part of the preparation for each reload core. Clearance must be demonstrated for each fuel type included in the core.
The Batch D assembly, scheduled for loading in the core center position for
- Cycle 5, was inspected at-the end of Cycle 3 (EOC3) and will be inspected at the end of Cycle 4 (E0C4). Shoulder gap measurements are included as part of the inspections. .Using the actual EOC3 and E0C4 values (representing 2 and 3 cycles of irradiation on the D assembly), an extrapolation will be made to EOC5 using conservative Cycle 5 exposure values. This extrapolation must show no interference (correction will be made for guide tube compression under the holddown force, and for differential thermal expansion between the fuel rods and guide tubes).
- For Batch E assemblies the. dimensional change correlations presented in CENPD-198 have been utilized. An EOC5 analysis of the lead exposure case has shown that no interference is predicted at a confidence level exceeding 95%. This represents the worst case for shoulder gap, since other Batch E cases have lower exposures. -
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QUESTION 013 -
Section 7.3.1 of the Calvert Cliffs Unit 1 Cycle 5 submittal (D13.1) shows CEA ejection accident results based on an enthalpy threshold for cladding failure. The enthalpy values provided in the submittal have never been approved and have been shown (D13.2) to be non-conservative with respect to ragulatory guides assumptions for this event at similar C-E NSSS-designed p'.sn ts . Provide a reanalysis of the CEA ejection accident based on a minimum DNBR for cladding failure as described in Regulatory Guide 1.77.
ANSWER
-The criteria for fuel damage during an ejected CEA event are the same as those in the FSAR and approved by NRC in all subsequent reload license submittals. These criteria were formulated prior to issuance of Regulatory Guide 1.77 and corresponding SRP's. The analyses of the CEA ejection event for Unit 1, Cycle 5 assumed a conservative rod average enthalpy criterion of 200 cal /gm rather than 280 cal /gm coolable geometry criterion and DNBR criterion for clad rupture, as suggested by the Regulatory Guide. No DNBR calculations were performed for Cycle 5, since C-E does not equate exceed-ing the DNBR limit with onset of cladding damage. Furthennore, the peak (radially averaged) fuel enthalpy calculated for Unit 1, Cycle 5 are sig-nificantly less than the maximum enthalpy permitted by Regulatory Guide 1.77. It is expected, therefore, that if a reanalysis was performed with DNBR as a criterion, acceptable results will also be attained. Such an analysis will be performed and results transmitted to NRC within 4 to 5 months after Calvert Cliffs Unit 1, Cycle 5 startup. date.
REFERENCES D13.1 Calvert Cliffs Unit 1, Cycle 5 Refueling License Amendment, BG&E, September 22, 1980.
D13.2 Arkansas Nuclear One, Unit 2, FSAR, Amendment No. 34, Pages 15.120-3a/7.
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QUESTION D15_
l The NRC staff has been generically evaluating three materials models that j
are used.in ECCS evaluation models. Those models are cladding rupture We temperature, cladding burst strain, and fuel assembly flow blockage.
have (a) met and discussed our review with Combustion Engineering and other industry representatives (1), (b) published NUREG-0630, " Cladding Swelling and Rupture Models for LOCA Analysis" (2), and (c) required fuel vendors and licensees of LWRs with Zircaloy cladding to confirm that their plants would continue to be in confomance with the ECCS criteria of 10 CFR 50.46 if the materials models of NUREG-0630 were substituted for those models of their ECCS evaluation models (3 and 4).
BG&E responded (5) to the NRC request of Reference 3. Please either amend ,
as necessary or confirm that the BG&E response (5) continues to be applicable for Cycie 5 operation of Calvert Cliffs Unit 1 in light of changes in plant variables for Cycle 5 (e.g., the increase in peak LHGR to 15.5 kw/ft).
RESPONSE
The ECCS perfonnance analysis for Calvert Cliffs Unit 1, Cycle 5 (. Reference
- 6) used models for cladding rupture temperature, cladding burst strain and fuel assembly flow blockage which are part of the NRC approved C-E evalua-tionmodel(Reference 7). The analysis verified acceptable ECCS performance using those models.
The proposed NRC model for new swelling and rupture curves (Reference 2) has been previously reviewed by C-E and our position to the models remains the same as expressed earlier in Reference 8. However, C-E has previously demonstrated acceptable ECCS perfomance (Reference 9) using the most adverse rupture conditions from the proposed NRC model (Reference 2) in conjunction with the heat transfer portion of the C-E alternate flow blockage / heat transfer model (Reference 10). The Reference 9 study applied to Calvert Cliffs Unit 1, Cycle 4 and Unit 2, Cycle 2. Because of the r -
QUESTION D15 Page 2 similarity between Unit 2, Cycle 2 and Unit 1, Cycle 5 (both cores consist of only high density stable fuel and both have the same peak linear heat generation rate, 15.5_kw/ft), the ECCS performance analysis for Unit 2, Cycle 2 was selected as the reference analysis for Unit 1, Cycle 5. Based on.these facts, C-E has no reason to expect that Unit 1, Cycle 5 would not show acceptable ECCS performance if analyzed using the methods described in Reference 9.
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References for C15
- 1. Memorandum -from R.P. Denise, NRC, to R. J. Mattson, " Summary Minutes of Meeting on Cladding Rupture Temperature, Cladding Strain, and Assembly Flow Blockage," November 20, 1979.
- 2. .D. A. Powers and R. O. Meyer, " Cladding Swelling and Rupture Models for LCCA Analysis, " NRC Report NUREG-0630, April 1980.
- 3. Letter fron D. G. Eisenhut, NRC, to all Operating Light Water Reactors, dated November 9,1979.
- 4. Memorandum from H. R. Denton, NRC, to Commissioners, " Potential Deficiencies in ECCS Evaluation Models," November 26, 1979.
- 5. Letter from A. E. Lundvall, CG&E, to C. G. Eisenhut, NRC, dated Jaruary 31, 1980.
- 6. Calvert Cliffs Unit 1 Cycle 5 Refueling License Amendment, BG&E, September 22, 1980.
5 7. CENPD-132, "r ;1culative Methods for the C-E Large' Break LOCA -
Evaluation whdel', August,1974 (Proprietary).
Supplement 1, " Calculational Methods for the C-E Large Break CENPD-132, LOCA Evalur.cion Model", February, 1975 (Proprietaryj.
CENPD k?2, Supplement 2, " Calculational Methods for the C-E Large Break LOCA Evaluation Model", July,1975 (Proprietary).
- 8. Letter from A. E. Scherer, Combustion Engineering, Inc., to R. P. Denise, NRC,
Subject:
Review of Draft Report NUREG-0630, " Cladding Swelling and Rupture Models for LOCA Analysis," dated December. 11, 1979.
- 9. Letter from A. E. Lundvall, BG&E, to D. G. Eisenhut, NRC, dated January 31, 1980.
- 10. LD-78-069, Enclosure 1-P, "C-E ECCS Evaluation Model Flow Blockage Analysis", September,1978 (Proprietary). -
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Question 016:
For Cycle 4 operation, BG&E imposed (13) restrictions on CEA movement whi'.a at primary system temperatures below 4000F. This restriction was in response
. . _ , to a C-E recommended operational guideline to reduce CEA drag force on relaxed crimp joints (i.e., the bond which holds sleeve inserts to guide tubes). Is the continuation of this restriction prudent for Cycle 5 operation? If rot, please describe the basis for discontinuation.
Response
Sampling of ,leeve crimp sizes in bundles going under CEA's in Cycle 4 indicated that there were undersized crimps (reference CEN-116(B)-P for crimp size criterion and basis) in some of those bundles. The operational restriction on CEA movement below 4000F for Cycle 4 was, therefore, imposed to preclude CEA drag forces causing movement of the sleeves with undersized crimps.
The following types of sleeved fuel assemblies will be located under CEA's during Cycle 5:
- 1) 4 Batch F assemblies sleeved unirradiated in 1979
- 2) 16 Batch F assemblies sleeved irradiated in 1980
- 3) 24 Batch G assemblies sleeved unirradiated in 1980 i
The four (4) bundles s~.eeved in 1979 were sleeved in accordance with the same procedure as was employed at Calvert Cliffs Unit II for the bundles whose crimp sizes were sampled in October of 1979. The results of that sampling (Reference CEN-118(B)-P) indicate that the procedure yields acceptable crimp sizes in unirradiated guide tubes. The remaining sleeved fuel assemblies under CEA's (16 Batch F and 24 Batch G) were sleeved in accordance with a revised procedure which included an Eddy Current Test of the lower portion of the installed sleeves to confirm the presence of an adequate crimp. Therefore, since none of the sleeved fuel assemblies- going under CEA's in Cycle 5 contain
' undersized crimps, the operational restriction on CEA movement below 4000F.will be discontinued.
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ENCLOSURE E QUESTION El.
What code is used to calculate peaking factors for asymmetric core conditions? Is the code approved by NRC7
RESPONSE
As reported in an earlier response to a question on the Cycle 5 reload submittal, the ROCS computer code was used to calculate changes in the 3-D core power distributions that result frora inlet temperature maldis-tributions (asymmetric steam generator transient). The changes in the peaking factors (distortion factors) calculated by 3-D ROCS are then applied to the maximum peaking values used in the safety analysis, i.e.,
Tech Spec values. Best estimate calculations have Indicated that the expected radial peaking factors will be below the safety analysis values and, thus, use of the calculated distortion factors as applied to the Tech Spec values is considered to be conservative. ,
The use of ROCS for specific applications, including the one mentioned above, was approved by NRC in the SER for Cycle 4.
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QUESTION E2' How do you specify the boundary conditions for the core inlet temperature profile during a s mmetric core conditions?
ANSWER A nominal core inlet temperature profile representative of-temperature tilting associated with asymmetric core conditions was used as input to ROCS. It was derived from inlet plenum mixing factors obtained from flow tests performed by Battelle Memorial Institute (Reference E2.1). ,
The referenced flow tests measured the fraction of flow to each of 133 bundles from a single traced cold leg. The fractional contribution from each of the other cold legs was found from symmetry.
The fraction of flow to each bundle from one steam generator was found by summing the contribution from two adjacent cold legs.
The resulting flow distribution was scaled to a 217 bundle core geometrically by overlaying a 217 bundle grid on a 133 bundfe grid with the same area. The fraction of flow to each of the 217 bundles was taken to be the area-weighted sum of the mixing factors from the 133 bundle grid. Finally, these mixing factors were applied to the cold leg coolant temperatures to determine the bundle-wise temperature profile due to a given core inlet temperature .: symmetry.
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QUESTION E2 Page two-
- REFERENCE E2.I._
H'. ;L. Crawford and L. J. Flanigan,. " Final Report on Studies of
- Flow in a 0.248 Scale Model of the Omaha PWR," Battelle Memorial Institute, Columbus, Ohio, April,1970.
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- QUESTION E3 Clarify the statement:
"However, the use of cycle specific data woulu i t ~.e extensive reanalysis and possibly restrict operational flexibility for future reload cycles if the data was determined to be more adve rse."
- a. Does this mean Asymmetric Core Condition analysis is one time analysis and is to be applied to the future cycles?
- b. How does using cycle specific data place restrictions on operation?
ANSWER
- a. The use of enveloped data for the increases in the integrated radial peak as a function of core inlet temperature tilt in combination with the Asymmetric Steam Generator Protective Trip Function (ASGPTF) is expected to preclude the need to reanalyze the Loss of Load to one steam generator (LL/ISG) event for future cycles on a cycle specific basis.
- b. The use of Cvele 5 specific data for the LL/ ISO event will not impose any operational restrictions for Cycle 5 operation.
However, the use of cycle specific data would require reanalyses for future reload cycles. If the cycle speci fic data becomes more adverse, this event could become more limiting than events which now establish margin requirements. This could restrict i 1
c.norational flexibility by forcing the use of more restrictivo 1 limiting conditions for operation.
QUESTION E4 If the TORC code is used to ca!culate the DNBR, how do you specify the core inlet mass flux and inlet temperature for the asymmetric core condition?
. ANSWER The inlet temperature used in the calculation corresponds to the average of the inlet temperature f rom the two loops, which at the time of minimum DNBR is approximately equal to 550.3 F. This assumption is conservative, since it has been shown that the maxi-mum increases in the integrated radial peak (aF r) and the hot channel location both occur on.the cold side of the core. Thus, the maximum AFp in conbination with the average inlet temperature are used to conservatively calculate the minimum DNBR.
The mass flow rate assumed in the analyses. corresponds to the initial value of 2.5x106 lbm/hr-ft2 at the start of the event.
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QUESTION ES If the CESEC code is used to calculate inlet temperature difference
-between coolant loops- for asymmetric core conditions, provide the following:
- a. The version of CESEC used?
- b. Why was CESEC-SLB not used?
- c. What .is the moderator temperature coefficient used to calculate ~
moderator reactivity feedback?
ANSWER
- a. The version of CESEC used to simulate the NSSS response is called _ Cycle 7 (Reference E5.1). This code was used to obtain the core inlet tem; 9< ;.ture dif ference between the loops and the steam generator pressure variations during this event.
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- b. The CESEC-SLB code was not used in this event since the code is being used to calculate core inlet temperature dif ferences
. between loops. These differences are calculated adequately .
by the old version of CESEC (i.e., CESEC, Cycle 7).
- c. The Moderator Temperature Coefficient used inthe analysis is equal to -2.5x10~4 Ap/ F.
- REFERENCE E5.1 CENPD 107, "CESEC Topical," April,1974.
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T QUESTION E6 Why has the F[ limit increased relative to Cycle 47 (i.e., why the change from 1.585~ to 1.62?)
ANSWER Cycle 4hadanF[limitof1.585. Cycle 5's limit'of 1.62.is approximately 2% higher. The approximately 2% higher F[ limit is due almost entirely to the different axial power. distributions used for the Cycle 5 analyses. Cycle 4 was based on an existing conservative shape analysis. For Cycle 5, the shape analyses, when compared to that of Cycle 3, shows an increase in overpower in the shape index range of Interest of approximately 25.
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- - ,s-QUESTION E7 .
Present a graph of CNBR versus time for the 1.oss of Flow' event at i
the most limiting shape index value.
i l ANSWER
, The limiting axial power distribution for the Loss of Flow event is one characterized by a 0.0 axial shape Index. Figure E7-1 presents the DNBR versus time for this axial shape Index.
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-QUEST 10N E8 .
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-Provide a graph of DNBR versus tine for-the Loss of Load to one steam generator event at the most -IimitIng shape index value.
ANSWER i
! The~ limiting axial power shape for the Loss of Load to one~ steam i
9enerator event' Is one characterized by a .16 axial shape index.
Figure E8-1 presents the DNBR versus time for this axial shape index.
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QUESTION E9
- a. For the Setzed Rotor event (Section 7.3.4) why was an ASI of .16 used?
- b. Why is negative axial shape index more conservative than positive . shape Indices for both the Loss of Flow event and
. the Seized Rotor event?
ANSWER
- a. The ASI of .16 was used in the Seized Rotor analyses because
-It results in the maximum margin degradation during the event and thus the greatest amount of predicted fuel failurc,
- b. The negative axial shape index is more limiting because the initial (i.e., time, t=0) DNBR value is much lower (i.e.,
closer to the DNBR limit) than for positive axial shape Indices. Thus, the transient minimum DNBR also reaches a lower value for this negative axial shape index.
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WESTION E10
- a. For the Seized Rotor. event, what is the value of core flow
' assumed in the analyses?
- b. What is the impact of_ Natural Circulation flow on this eventi
. ANSWER d
- a. The initial core mass flow assumed in the analysis is equal to 370000 GPM (or 2.5x10 6 lbm/hr-ft2 ). The final mass flow is equal to 77.2% of the initial flow.
- b. The Seized Rotor event assumes an instantaneous drop in core flow to 3 pump flow. The criteria of interest (i.e., DNBR) for this event are approached in seconds. Thus, natural circulation flow regimes are not encountered in the analysis time f rame for this event.
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