Text

VEPCO Reactor Sys Transient Analysis Using Retran Computer Code
	ML20129E988
Person / Time
Site:	Surry, North Anna, 05000000
Issue date:	05/30/1985
From:	Berryman R, Cross R, Smith N; VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	;
Shared Package
ML18142A562	List: ML20129E988;
References
	VEP-FRD-41A, NUDOCS 8507170163
	Download: ML20129E988 (120)
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~~i R. W. Cross Supervisor, Nuclear Fuel Engineering APPROVED:~~

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~~R. M. Beriyman Director, Nuclear Fuel Engineering~~

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~~UNITED STATES~~

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~~NUCLEAR REGULATORY COMMISSION E~~

~~W ASHINGTON. D C. 20555~~

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l April 11,1985 Mr. W. L. Stewart Vice President Nuclear Operations viroinia Electric and Power Company P. O. Pox 26666 Richnond, Virainia 23361
~~Dear Mr. Stewart:~~
SURJECT: ACCEPTANCE FOR PEFERENCING pr LICENSING TOPICAL REPORT VEP-FRD-41, "VEPC0 PEACTOR SYSTEM TRANSIENT ANALYSIS USING RETPAN COMPUTEP CODE" We have completed our review of tha sub.iect topical report submitted by Viroinia Electric and Power Company (VEPC01 by letters dated April la,19P.1, February ?7,1984, sluly 12,1984 and Au9ust 24,10F4 We find the report to be acceptable for referencino in license applications to the extent specified and under the limitations delineated in the report and the associated NRC evaluation, which is enclosed. The evaluatinn defines the basis for acceptance of the report.
We do not intend to rapeat nur review of the matters described in the report and found acceptable when the report appears as a re'erence in license applications, except to assure that the material presentad is applicable to
+he speci'ic plant invnived. Our acceptance applios only to the ma+ters described in the repnet.
In arcnedance with proceduras establishad in NUREG-0390, it is reouestad that VEPCO publish accepted versions of this raport, proprietary and non-proprietary, within three months of receipt of this letter.
The accepted versions shall incorporate this letter and the enclosed evaluation between the title pace ard the abstract. The accepted versions shall include an -A (designatina acceptedi followina the report identification symbol.
Should nur criteria or regulations chanaa such that nur conclusions as to the acceptability of the report are invalidated, VFPCO and/or the applicants referencing the topical report will be expected to revise and resubrit thair respective documentation, or submit.iustification for the continued effective app'icability of the topical report without revision of their respective documentation.
Sincerely, G/AO.06%w Cecil 0. Thomas, Chief Standardization and Special Prn.iects Branch Division of Licensino Enclosur e:
As stated
ENCLOSURE
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SAFETY EVALUATION REPORT ON THE VEPCO TOPICAL REPORT VEP-FRD-41, " REACTOR SYSTEMS TRANSIENT ANALYSIS USING THE RETRAN COMPUTER CODE" 1.
Introduction The VEPCO topical. report VEP-FRD-41, "Rea,ctor System Transient Analysis Using the R5RAN Computer Code" was submitted to demonstrate the capabil-ity which VEPCO has developed for performing transient analysis using the RETRAN 01/ MOD 03 Computer Code.
This submittal.is cansistent with nur.
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Generic Letter 83-11. This analysis capability is to be' utilized by VEPCO to support plant operation and ;irovide future reload safety analyses for both Surry and North Anna Nuclear Power Stations. The report provides some overview ~ of the RETRAfi Computer Code, tiut refers to EPRI documenta-tion for' further material on the RETRAN models and for qualification support of these models.
The staff evaluation of the RETRAN Computer
' Cod [, has been completed.
A staff safety evaluation report has been issued on the acceptability of that RETRAN computer code for analyzing reactor transients for licensing applications.
The acceptance was subject to restrictions as specified in the staff SER for the generic RETRAN Computer Code. The VEPCO. topical report VEP-FRD-41 was submitted by VEPCO l
o in a letter dated April 14, 1981.
In response to the staff requests for additional information, additional supporting materials were submitted in VEPCO letters dated February 27, 1984, July 12, 1984 and August 24, 1984.
The staff evaluation is addressed below.
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2.
VEPCO NSSS Models Discussion of the RETRAN plant models developed for the three-loop West-ingho'use designed Surry and North Anna Units is provided in the topical report VEP-FRD-41.
The transient analysis to be performed detemines the level of detail required by the model. A single-loop and a two-loop RETRAN nodalization were submitted for staff review.
The single-loop model has been formulated by representing the three reactor coolant loops as a single loop.
This model was developed for use on transients which produce symmetric p1_ ant response in all unaffecte.d reactor coolant loops.
Exam-ples of such transients would include a complete loss of a.c. power to all of the reactor coolant pumps (a loss, of flow transient), a core reactivity. insertion resulting from the uncontrolled.. withdrawal qf a Rod Cluster Control Assembly, or a loss of external electrical load transient.
The two-loop model was developed with one loop representing a single primary coolant loop and.the other representing the remaining two primary
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coolant loops.
The two-loop model was designed for use on transients which produce asymmetric thermal-hydraulic conditions among one of the three loops.
Examples of such transients would include a postulated main
' steam line break resulting in the rapid cooldown of one reactor cooling loop, or a loss of power supply to a single reactor coolant pump, which results in a rapid flow coastdown of one reactor cooling pump.
In response to the staff request for additional infomation, VEPCO in letters dated July 12,.1984 and August 24, 1984, provided detail descrip-tions in the following areas:
1)' Volume and flow path network including heat slabs, 2) Component models used and user modif,ications to default models, 3) Control system models, and 4) RETRAN input option selections.
The staff has reviewed the above VEPC0 model descriptions and finds them acceptable for demonstrating understanding of the RETRAN code.
3.
Analysis Methodologv VEPC0 intends to reference VEP-FRD-41 as their basic model for reload applications.
Following determination of the key reload parameters, the safety analyst will apply the appropriate boundary conditions required for
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the specific application.
The evaluation is to ensure that those key parameters which may influence the transient response are consistent with the bounds or limits established by the technical specifications and parameters used in the reference analysis.
For cases where a parameter falls outside these previously defined limits an evaluation of the impact of the change on th'e results for the appropriate transients must be made.
For casies where significant variations occur, or for parameters which have a strong influence on accident results, reanalysis of the' affect.ed transient ~is required.
The results of. a reanalysis are com;2ared to the appropriate analysis acceptance criteria.
If the results of a reanalysis.
meet the acceptance criteria, the reload evaluation process is complete.
If the analysis acceptance criteria are not met, more detailed analysis methods or Technical Specification changes may be required to meet the acceptance criteria.
The NRC will be informed of the results of the evaluations in accordance with the requirements of 10 CFR 50.59.
VEPCO will use analysis methodology and acceptance criteria identified in the O
following documents:
1) Surry Power Station Units 1 and 2, Final Safety Analysis Report, 2) North Anna Power Station Units 1 and 2, Final Safety Analysis Report, and 3) WCAP-9272, " Westinghouse Reload Safety Evaluation Methodology,' which has been reviewed and approved by NRC in 1980.
We 3
require that the licensee fully document all assumptions and boundary conditions used in each application..This review does not constitute a
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transient specific methodology approval.
o 4.
Qualification Comparisons The VEPCO has developed a system transient analysis capability using the RETRAN Computer Code for non-LOCA initiating events.
In order to demonstrate VEPCO's ability'to correctly use the RETRAN Computer Code, verification work has been performed by benchmarking both actual plant transient dda and independent safety analyses previously performed by the NSSS vendor and documented in the FSAR.
For plant transient data benchmarking, the VEPCO RETRAN Computer Code was developed to model both Surry and North Anna power stations in a best estimate mode.
This permits direct compar'isons to the actual measured plant data.
Comparisons were made with flow coastdown tests performed at both-the' Surry and North Anna plants and a plant cooldown transient which occurred at North Anna Unit 1.
In the comparison of RETRAN analyses to
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'the data obtained from the flow coastdown tests, both single-loop and two-loop 'RETRAN models were used to simulate pump coastdown tests of various configurations (i.e. one pump coastdown, three pump coastdown).
The results of the comparison as documented in the topical report indicate that the VEPCO RETRAN predictions are in close agreement with the data obtained from Surry and North Anna.
A RETRAN analysis was performed to simulate the plant cooldown transient which occurred at North Anna Unit 1 on September 25, 1979.
The transient was initiated by a turbine trip and 4
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subsequent reactor trip.
Safety injection was actuated on a low pressur-izer pressure during the transient due to RCS depressurization in response to a ' fully stuck open steam dump valve.
The VEPCO RETRAN model used to simulate the cooldown scenario was a single-loop representation of the North Anna Unit.
The calculated transient parameters including steam pressure, RCS. temperatures, pressurizer pressure, and pressurizer level, were compared to the actual data taken during the event.
The results of the comparison show agreement between the best estimate calculation and the getual transient data.
VEPCO provided comparisons of FSAR licensing safety analysis with analyses performed using the RETRAN Computer Code.
Ibe. basis..for the event. selec-4 tion were:
1) Cons'ideration of those events which have previously been determined limiting and have bee'n most frequently subjected to reanalyses during each reload (e.g. Rod Withdrawal from Power and Complete loss of flow); 2)' Selecting analyses in each of the major categories of initiating events which include changes in reactivity (e.g. rod withdrawal tran-sients), variations in primary coolant flow rate (e.g. loss of flow
' transient), and variations in primary to secondary system heat transfer rates (e.g. main steam line break); and 3) Transients which are both symmetric (e.g. loss of load transient) and asymmetric (e.g. single pump flow coastdown) with respect to the thermal hydraulic response of the reactor coolant loops.
The results of analyses performed by VEPC0 (using the RETRAN Computer Code) for the above stated events compared favorably to those obtained by 5
its NSSS vendor.
The similarities in system response hold for a broad variety of transients and result in identical conclusions regarding core and system conditions.
O e
In res y nse to the staff request, VEPCO, in a letter dated July 12, 1984, provided results of RETRAN sensitivity studies for the following tran-sients:
~~1) Rod withdrawal at power, 2) Rod withdrawal from sub-critical,~~
~~3) loss of load, 4) excessive load increase, and 5) Complete loss of flow.~~
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The staff has evaluated the results of the VEPCO's sensitivity studies and finds them consistent with the NSSS Vendor's analyses, as documented in the Surry and North Anna,FSARs.
To further verify the comparability of the VEPCO RETRAN model to the NSSS Vendor's analysis model, VEPCO, in a lette'r dated August 24.
, sub-mitted a supplement'to VEP-FDR-41 which compa' red parallel calculations of RETRAN and LOFTRAN performed by VEPCO.
The LOFTRAN code is an NRC approved analytical program developed and maintained by the Wes:tinghouse Eled.ric Corporation for use in performing general non-LOCA transient and accident' analyses.
VEPCO has obtainec access to LOFTRAN via a special licensing agreement with Westinghouse.
The comparisons were performed with a LOFTRAN model of the Surry plant assembled by VEPCO applying the same data base used for developing the VEPCO RETRAN models. Thus the basic plant geometric and thermal parameters are consistent for the two models.
The following transients' were calculated and compared using both computer models:
~~1) Reactor trip from hot full power followed by a turbine trip, 2) Turbine trip from hot full power.~~
No credit taken for 6
direct reactor trip on the turbine trip, and 3) Simulataneous trip of all three' reactor coolant pumps at hot full power.
No credit taken for react'or trip on pump under voltage or under frequency.
The results of these analyses confimed that the VEPCO RETRAN models could produce compatible analysis results with that from the LOFTRAN models.
5.
Conclusions Based on the VEPCO RETRAN model and the qualification comparisons discussed above, the staff concludes that VEPCO has demonstrated their capability to analyze non-LOCA initiated transients and accidents using the RETRAN Computer Code.
VEPCO intends to perform future reload analyses and support;ing plant operations for Surry and Nortit. Anna plants.
We f.ind VEPCO qualified to ' perform the non-LOCA initiated transients and accident analyses using the RETRAN modelsl and methodology.
This topic report does not include the Rod Ejection Accident analysis which has been a'ddressed in a separate VEPCO Topic Report VEP-NFE-2 and'a separate staff safety evaluati~on report.
VEPCO has not provide infomation to address the restrictions stated in the staff SER for the generic RETRAN Computer Code.
'The$icceptanceoftheVEPC0RETRANmodelsissubjecttotherestrictions to the general RETRAN computer code specified in the staff safety evalua-tion report issued in July 1984 on RETRAN.
VEPCO has not provided an input deck to the NRC staff as was required by the staff SER for the generic RETRAN code.
We continue to require that this input deck be provided to us as a condition of this approval.
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With respect to the quality assurance requirement of the VEPCO RETRAN Computer Code, the staff has performed an audit at VEPCO with satisfactory results.
The staff requires that all future modification of VEPCO RETRAN model and the error reporting and change control models should be placed l
under full quality assurance procedures.
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l CLASSIFICATION /DISCLAUTR l
l The data, information, analytical techniques, and conclusions in this rapart have been prepared solely for use by the Virginia' Electric and Power l
Cstptny (the Company), and they may not be appropriate for use in situations I
cth3r than those for which they were specifically prepared. The Company there-l foro makes no claim or warranty whatsoever, express or i= plied, as to their cccuracy, usefulness, or applicability.
In particular, TdE COMPANY MAKES NO WARRANTY OF lERCHANTABILITY OR FITNESS FOR A PARTICULAR PUR?OSE, NOR SHALL XiY WARRXiTY BE DEEMED TO ARISE FROM COURSE OF DEALING OR 1* SAGE OF TPRE, with l
rcspset to this report or any of the data, infor=ation, analytical techniques, l
,or etnc us ons in it.
By making this report available, the Cc=pany does not l i l
tuchsrize its use by others, and any such use is expressly forbidden exc'ept with the prior written approval of the Company.
Any such written approval ch-ll itself be dee=ed to incorporate the disclai=ars of liability and dis-i clcimers of warranties provided herein.
In no event shall the Co=pany be litble, under any legal theory whatsbever (whether contract, tort, warranty, or strict or absolute liability), for any property dacage, = ental or physical injury or death, loss of use of property, or other damage resulting from or crising out of the use, authorized or unauthorized, of this report or the data, i
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infor=ation, analytical techniques, or conclusions in it.
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ACKNOWLEDGEMENTS This report is the culmination of several years of development effort in which numerous individuals have participated. The analytical work was performed by Messrs. R. W. Cross, S. M. Mirsky, G. M. Suwal and the author. Valuable technical and i
l Gditorial comments were provided by Messrs. W. C. Beck, M. L. Bowling, E. R.
Smith, Jr. and T. L. Wheeler.
Ms. Miranda Cooper and Ms. Sharon Kulp provided patient typing support. The efforts of these and numerous other individuals are gratefully acknowledged.
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- TADLE OF CONTENT 3 SECTION PAGE
.1 - INTR O D U CTIO N.. '.......................
1.1 2 - OVERVIEW OF THE RETRAN COMPUTER CODE 2.1 3 - REPRESENTATIVE VEPCO NSSS MODELS.............
3.1 3.1 In trodu c tion.......................... 3.1 3.2 Single Loop Model..........'. >............
3.2 3.3 Multi-loop Model......................., 3.5 4 - SYSTEM THERMAL HYDRAULIC ANALYSIS METHODOLOGY....
4.1 4.1 Introduction.........................
4.1 4.2 Licensing Evaluation Process..................
4.1 4.3 System Transient Analysis...................
4.2 l
4.3.1 System Model Application................
4.3 l
4.3.2 Transient Specific Input.................
4.4 4.3.2.1 Initial Conditions....... '.........
4.4 4.3.2.2 Reactivity Parameters..............
4.5 1
4.3.2.3 System Performance Assumptions.........
4.5 l
4.4 Use of System Thermal Hydraulic Results 4.6 5 -- QUALIFICATION COMPARISONS.................
5.1 5.1 Introduction.........................
5.1 5.2 Verification Against Licensing Analyses.............. 5.1 5.2.1 Transients Resulting from Changes in Reactivity......
5.1 5.2.1.1 Uncontrolled Control Rod Assembly Withdrawal from a Suberitical Condition - FSAR Analysis..... 5.2 5.2.1.2 Uncontrolled Control Rod Assembly Withdrawal from a Suberitical Condition - Current Analysis.... 5.3 l~
5.2.1.3 Uncontrolled Control Rod Assembly Withdrawal at Power Transient - FSAR Analysis......... 5.4 5.2.1.4. Uncontrolled Control Rod Assembly Withdrawal at Power Transient - Current Analysis........ 5.5
5.2.2 Transients Resulting from Changes in Primary System Flowrate. 5.6 5.2.2.1 Complete Loss of Flow - FSAR Analysis......
5.7 5.2.2.2 Complete Loss of Flow Transient-Current Analysis.
5.7 5.2.2.3 Partial Loss of Flow Transient-FSAR Analysis 5.8 5.2.3 Change in Primary to Secondary Heat Transfer.......
5.9 5.2.3.1 Loss of External Electrical Lead Transient -
FS A R Analysis.................
5.9 5.2.3.2 Loss of External Electrical Load Transient -
Current Analysis................
5.11 5.2.3.3 Excessive Heat Removal Due to Feedwater System Malfunction Transient - FSAR Analysis
...... 5.11 5.2.3.4 Accidental Depressurization of the Secondary System / Main Steam Line Break Transient - FSAR Analysis....
5.12 5.2.3.5 Accidental Depressurization of the Secondary System / Main Steam Line Break Transient - Current Analysis...
5.1.4 5.2.4 Conclusions - Licensing Transient Analyses.........
5.14 5.3 Verification Against Operational Data..............
5.15 5.3.1 Introduction......................
5.15 5.3.2 Pump Coastdown Tests 5.15 5.3.2.1 Surry Pump Coastdowns.............
5.15 5.3.2.2 North Anna Pump Coastdown...........
5.16 5.3.3 North Anna Cooldown and Safety Injection Transient.
. 5.17 5.3.4 General Coneb: vans - Best Estimate Transient Analyses...
5.19 6 CONCLUSIONS........
6.1 7 REFERENCES..........................
7.1 APPENDIX -
~~SUMMARY~~
OF IMPORTANT ASSUMPTIONS USED IN TRANSIENT ANALYSES DISCUSSED IN SECTION 5 i
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LIST OF FIGURES
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Number Title
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5.1 Nuclear Power, Uncontrolled Rod Withdrawal from Suberitical Transient, FSAR Analysis
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5.2 Average Fuel Temperature, Uncontrolled Rod Withdrawal from Suberitical
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Transient, FSAR Analysis
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s a;; 11' 5.3 Average Clad Temperature, Uncontrolled Rod Withdrawal from Suberitical 1 c;.N Transient, FSAR Analysis
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l 5.4 Core Heat Flux, Uncontrolled Rod Withdrawal from Suberitical Transient,
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5.5 Nuclear Power, Uncontrolled Rod Withdrawal from Suberitical Transient, Surry 2 Cycle 4 Reanalysis
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5.6 Average Fuel Temperature, Uncontrolled Rod Withdrawal from Suberitical
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Transient, Surry 2 Cycle 4 Reanalysis
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5.7 Average Clad Temperature, Uncontrolled Rod Withdrawal from Suberitical f~y Transient, Surry 2 Cycle 4 Reanalysis
{%: :.3 5.8 Core Heat Flux, Uncontrolled Rod Withdrawal from Suberitical Transient, y p.k i Surry 2 Cycle 4 Reanalysis f.].
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5.9 Nuclear Power, Uncontrolled Rod Withdrawal from Power Transient,
).,7 i.J g FSAR Analysis
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5.10 Pressurizer Pressure, Uncontrolled Rod Withdrawal from Power Transient, cwa FSAR Analysis if}/.? ' ~.'
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5.11 Average Coolant Temperature, Uncontrolled Rod Withdrawal from
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Power Transient, FSAR Analysis
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5.12 DNB Ratio, Uncontrolled Rod Withdrawal from Power Transient, FSAR
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Analysis 5.13 Variation of Minimum DNBR with Reactivity Insertion Rate, Rod Withdrawal O4-from 102% Power, Steam Generator Tube Plugging Reanalysis G[j l.y Q~W' Jj 5.14 Variation of Minimum DNBR with Reactivity Insertion Rate, Rod Withdrawal from 62% Power, Steam Generator Tube Plugging Reanalysis 5.15 Flow Coastdown, Complete Loss of Flow Transient, FSAR Analysis 5.16 Nuclear Power, Complete Loss of Flow Transient, FSAR Analysis 5.17 Average Heat Flux, Complete Loss of Flow Transient, FSAR Analysis 5.18 DNB Ratio, Complete Loss of Flow Transient, FSAR Analysis
5.19 Flow Coastdown, Complete Loss of Flow Transient, Steam Generator Tube Plugging Reanalysis 5.20 Nuclear Power, Complete Loss of Flow Transient, Steam Generator Tube Plugging Reanalysis 5.21 Average Heat Flux, Complete Loss of Flow Transient, Steam Generator Tube Plugging Reanalysis 5.22 DNS Ratio, Complete Loss of Flow Transient, Steam Generator Tube Plugging Reanalysis 5.23 Flow Coastdown, Partial Loss of Flow Transient, FSAR Analysis 5.24 Nuclear Power, Partial Loss of Flow Transient, FSAR Analysis 5.25 Core Average Heat Flux, Partial Loss of Flow Transient, FSAR Analysis 5.26 DNB Ratio, Partial Loss of Flow Transient, FSAR Analysis 5.27 Pressurizer Pressure Change, Loss of Load Transient, BOL-FSAR Analysis 5.28 Nuclear Power, Loss of Load Transient, BOL - FSAR Analysis 5.29 Pressurizer Water Volume Change, Lor.s cf Load Transient, BOL - FSAR Analysis 5.30 Coolant Inlet Temperature Change, Loss of Load Transient, BOL - FSAR Analysis 5.31 DNB Ratio, Loss of Load Transient, BOL - FSAR Analysis 5.32 Pressurizer Pressure Change, Loss of Load Transient, EOL - FSAR Analysis 5.33 Nuclear Power, Loss of Load Transient. EOL - FSAR Analysis 5.34 Pressurizer Water Volume Change, Loss of Load Transient, EOL - FSAR Analysis 5.35 Coolant Inlet Temperature Change, Loss of Load Transient, EOL - FSAR Analysis 5.36 DNB Ratio, Loss of Load Transient, EOL - FSAR Analysis 5.37 Pressurizer Pressure Change, Loss of Load Transient, Positive Moderator Coefficient Reanalysis 5.38 Nuclear Power, Loss of Load Transient, Positive Modcrator Coefficient Reanalysis 5.39 Average Coolant Temperature, Loss of Load Transient, Positive Moderator Coefficient Reanalysis 5.40 DNB Ratio, Loss of Load Transient, Positive Mcderator Coefficient Reanalysis
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5.41 Feedwater Temperature Change, Excessive Heat Removal due to Feedwater System Malfunction Transient, FSAR Analysis 5.42 Nuclear Power, Excessive Heat Removal due to Feedwater System Malfunction Transient, FSAR Analysis 5.43 Change in Average Coolant Temperature, Excessive Heat Removal due to Feedwater System Malfunction Transient, FSAR Analysis 5.44 Pressurizer Pressure Change, Excessive Heat Removal due to Feedwater System Malfunction Transient, FSAR Analysis 5.45 DNB Ratio, Excessive Heat Removal due to Feedwater System Malfunction Transient, FSAR Analysis 5.46 Break Flow Rate, Main Steam Line Break, FSAR Analysis 5.47 Pressurizer Pressure, Main Steam Line Break, FSAR Analysis 5.48 Total Reactivity, Main Steam Line Break, FSAR Analysis 1
5.49 Core Heat Flux, Main Steam Line Break, FSAR Analysis 1
5.50 Pressurizer Pressure, Main Steam Line Break, Surry 1, Cycle 4 Reanalysis 5.51 Total Reactivity, Main Steam Line Break, Surry 1, Cycle 4 Reanalysis 5.52 Core Heat Flux, Main Steam Line Break, Surry 1, Cycle 4 Reanalysis 5.53 Flow Coastdown, Operational Test at Hot Zero Power, Surry Three-Pump Coastdown 5.54 Flow Coastdown, Operational Test at Hot Zero Power, Surry One Pump Coastdown 5.55 Core Flow Coastdown, Operational Test at Hot Zero Power, Surry One Pump Coastdown 5.56 Flow Coastdown, Operational Test at Hot Zero Power, North Anna Three-Pump Coastdown 5.57 Steam Pressure, North Anna Cooldown Event 5.58 Cold Leg Temperature, North Anna Cooldown Event 5.59 Hot Leg Temperature, North Anna Cooldown Event 5.60 Pressu.*izer Pressure, North Anna Cooldown Event 5.61 Pressurizer Level, North Anna Cooldown Event
I LIST OF TABLES Table Title 3.1 Thermal-Hydraulic Design Parameters - Surry Plant 4.1 Protection System Characteristics Assumed in Safety Analyses 5.1 Limiting Predicted Results - Main Steam Line Break Transient
- Surry 1, Cycle 4 Reanalysis 1
SECTION 1 - INTRODUCTION The Virginia Electric and Power Company (Vepco) has developed the capability to perform system transient analyses of the North Anna and Surry Nuclear Power Stations.
This capability, coupled with the core thermal / hydraulic analysis capability discussed in Reference 1, encompasses the conservative non-LOCA licensing 1
analyses required for tne Conditions I, II and III transients addressed in the Final Safety Analysis Report (limited application to Condition IV transients is also included). In addition, the capability for performing best estimate analyses for plant operational support applications has also been developed.
The purpose of this effort is to 1) develop expertise in the system transient analysis area, 2) support reactor operation and 3) provide a basis for the reload core safety analysis and licensing process.
The principal analysis tool is the RETRAN 2
computer code which determines the time ' dependent or transient thermal-hydraulic response of a Nuclear Steam Supply System (NSSS). The RETRAN computer code calculates 1) general system parameters as a function of time and 2) boundary conditions for input into more detailed calculations of Departure from Nucleate Boiling or other thermal and fue) performance margins. The theory and numerical algorithms, the programming details, and the user's input information for the RETRAN computer code have been documented by its developers, Energy Incorporated (EI) and the Electric Power Research Institute (EPRI), in Volumes I through IV of Reference 2. Volume IV of Reference 2 provides the results of the extensive verification and qualification of the code which was performed by a group consisting of EI, EPRI, and 15 utilities including Vepco. The verification activity consisted of qualification of the code by comparison of code results with separate effects experiments, with systems effects tests, and with I
integrated system responses based on actual plant data or FSAR results.
f Performance of system transient analysis requires both single and multiloop l
1.2 modeling of the NSSS in order to analyze the required range of FSAR and operational support transients. Those transients for which the system thermal-hydraulic response of all reactor coolant loops is essentially identical require only a sing!e loop represen-tation. However, some transients are expected to have different responses in one or more of the reactor coolant loops, and these transients require multiloop representation of the NSSS. The RETRAN computer code, which is a variable geometry code, has the high degree of flexibility necessary for various system representations. Consequently, several models, including both single and multiloop representations, have been developed for the Vepco nuclear power stations.
In conjunction with both an analysis tool and system models, the development of a non-LOCA licensing analysis capability requires conservative analysis assumptions and input data. For licensing calculations, the Vepco analysis assumptions are consistent with those documented in the units' FSAR's (References 3 and 4). However, the specific analysis input may change as a result of plant modifications such as core reloads.
Consequently, the appropriate licensing analysis input consists of the current limiting values for the important safety parameters. For best-estimate analyses, nominal input values and actual operating histories of the Vepco nuclear power stations are used.
l The remainder of the report is organized in the following manner. Section 2 provides an overview of the RETRAN computer code, and Section 3 describes the Vepco models appropriate for the Surry and North Anna Nuclear Steam Supply Systems, as illustrated by a discussion of models developed for the Surry units. Section 4 provides a discussion of the Vepco transient analysis techniques and their relationships to other aspects of the licensing analysis process. Section 5 provides the results of a range of comparative analyses using the RETRAN code and the models of the NSSS discussed in Section 3 with calculations performed for the 1) design and licensing of the Surry Nuclear Power Station and 2) actual Surry and North Anna transient data. The report I
conclusions and references are provided in Sections 6 and 7, respectively.
l
SECTION 2 - OVERVIEW OF THE RETRAN COMPUTER CODE The RETRAN computer code was developed by Energy Incorporated under the auspices of the Electric Power Research Institute 2. As such, the RETRAN package is based upon the computer code RELAP4/003 Update 85 which was released by the United States Nuclear Regulatory Commission (NRC) as part of the Water Reactor Evaluation Model (WREM) 5. A detailed description of the RETRAN computer code can be found in Volume I of Reference 2. The following paragraphs summarize the important features of the code.
RETRAN contains the same fluid differential and state equations as RELAP4 for describing homogeneous equilibrium flow in one dimension. The representations used in previous RELAP codes for control volumes and junctions are also used in RETRAN and allow the analyst to model a system in as much detail as desired.
The modeling flexibility of the code is important and will be discussed in more detail in Section 3.
The equation systems, which describe the flow conditions within the channels, arc obtained from the local fluid conservation equations of mass, momentum and energy by l
use of mathematical integral-averaging techniques. Forms of the momentum equation are available for both compressible and incompressible flow.
The heat conduction representation capabilities of RETRAN have been increased over previous RELAP versions.
The principal augmentation to RETRAN is the capability to more accurately calculate two-sided heat transfer. The appropriate heat transfer correlation is selected based on thermodynamic conditions in each of two flow streams, on either side of a heat conducting solid. Consequently, representations of the heat transfer processes occurring in the steam generator, for example, are more accurate than previously possible.
Reactor kinetics are represented in RETRAN using a point kinetics model with reactivity feedback.
The reactivity feedback can be represented by constant
2.2 i
coefficients or in tabular form and accounts for explicit control actions (e.g., rod scram) and changes in fuel temperature, moderator temperature and density, and soluble boron concentration.
The system component models utilized in RETRAN include a pump model that describes the interaction between the centrifugal pump and the primary system fluid, and valve models that represent either simple valves, check valves or inertial valves. The flexibility of the valve representation and their configuration is important in allowing a wide variety of options to the user for the modeling of system dynamics.
Several representations for heat exchangers can be modeled by the code. These include the previously discussed two-sided heat transfer and several representations of one-sided heat transfer in conjunction with user specified boundary conditions.
A non-equilibrium pressurizer can be modeled in which the thermodynamic state solutions of the liquid and vapor regions of the pressurizer are determined from a distinct mass and energy balance for each region.
As in RELAP, a variety of trip functions can be modeled in the RETRAN code to represent various reactor protection system actions. A refinement of the RETRAN code over the RELAP code is the additon of a reactor control system modeling capability.
Consequently, the dynamics of linear and non-linear control systems are represented with RETRAN models of the more common analog computer elements. This additional capability is necessary for both best-estimate and licensing analysis, since the responses of various control and protection systems may have a significant effect on the overall system response.
l SECTION 3 - REPRESENTATIVE VEPCO NSSS MODELS 3.1 Introduction The RETRAN computer code is a variable-geometry code which allows the analyst to model a system in as much detail as required for a particular analysis. To illustrate this concept, two models developed for the Surry Nuclear Power Station will be discussed in detail in this section. (The modeling methodology is also applicable to the North Anna Nuclear Power Station).
The Surry Nuclear Power Station consists of two units, Surry Units No.1 and 2, which are identical Westinghouse designed three coolant loop pressurized water reactors with core thermal ratings of 2441 Mwt. The three similar heat transfer loops are connected in parallel to the reactor vessel with each loop containing a centrifugal pump, loop stop valves and a steam generator. The system includes a presssurizer and the associated control system and instrumentation necessary for operational control and protection.
The reactor vessel encloses the reactor core consisting of 157 fuel assemblies with each assembly having 204 fuel rods and 21 thimble tubes arranged in a 15 x 15 array. The fuel used in the Surry cores consists of slightly enriched uranium dioxide fuel pellets contained within a Zircaloy-4 cladding.
General thermal and hydraulic design parameters for the react ~ system are listed in Table 3.1.
The RETRAN thermal hydraulic model is formulated by representing individual portions of the hydraulic system as nodes or control volumes.
Control volumes are specified by the thermodynamic state of the fluid within the volume and basic geometric data such as volume, flow area, equivalent diameter and elevation. The flow paths connecting volumes or boundary conditions associated with a volume are designated as junctions. Junctions are described by specifying the flow, flow area, elevation, effective geometric inertia, form loss coefficient and flow equation specifi-cation for that particular flow path. Thermal interactions with system metal in the I
3.2 NSSS are modeled with heat conductors. Heat conductors may represent heat transfer from passive sources such as the metal of the reactor coolant system piping or the steam generator tubes. In addition, the internal generation of heat in the core may be Heat represented by active heat conductors designated as powered conductors.
conductors are primarily specified by providing the heat transfer area, volume, hydraulic diameter, heated equivalent diameter and channel length of the particular part of the system being modeled. Temperature - dependent materials properties (specific heat, thermal conductivity and linear thermal expansion coefficient) are also input. In general, the basic NSSS model is formulated with the code capabilities discussed above. An extensive research effort was conducted to determine the appropriate input required for the models of the Surry and North Anna units.
Information was obtained from plant drawings, the Final Safety Analysis Reports, 4 3
Vepco internal operating documents, equipment technical manuals and specific information requested from the NSSS vendor.
Specific control capabilities and constitutive models of system components will be discussed in the following paragraphs.
3.2 Single Loop Model The analysis to be performed and level of detail required dictates the general form of the models which are required.
Many transients are expected to produce similar responses simultaneously in all reactor coolant loops. Examples of such transients would include a complete loss of power simultaneously to all reactor coolant pumps resulting in a pump coastdown, a core reactivity insertion resulting from the uncontrolled withdrawal of a Rod Cluster Control Assembly (RCCA), or a loss of external electrical load resulting in a large, rapid steam load reduction.
To perform these transients, a single loop model of a Surry unit has been formulated by representing the three actual reactor coolant loops as one loop. This 0
approach is consistent with currently used safety analysis methodology. The resulting representation is provided in Figure 3.1 and consists of 19 volumes, 28 junctions and 7 heat conductors. While the specific model input for the Surry and North Anna plants is
3.3 different, the basic model description is the same for the single loop models of both plants. The reactor vessel includes representation of the downcomer, upper and lower plenums, core bypass, and reactor core. The steam generator is represented by four volumes on the primary side, one volume on the secondary side and four heat conductors representing the tubes. Single volumes represent the hot leg piping, steam generator inlet plenum, pump suction piping, reactor coolant pump, cold leg piping, pressurizer, and pressurizer surge line.
Primary system boundary conditions are specified with junctions representing the pressurizer relief and safety valves. Junctions representing the feedwater inlet, steam outlet, atmospheric steam relief and steam line safety valves provide secondary system boundary conditions. Specific aspects of the basic model will be discussed below.
The RETRAN code contains several system component models which are used in the Surry Single Loop Model. These include pump models which describe the interaction between the centrifugal pump and the primary system fluid. These models calculate pump behavior through the use of empirically developed pump characteristic curves which uniquely define the head and torque response of the pump as functions of vclumetric flow and pump speed. RETRAN includes " built-in" pump characteristics which are representative of pumps supplied by the major reactor coolant pump manufacturers. These curves may be modified, as appropriate, by the user to more realistically represent a specific pump design. Although the built-in data are not appreciably different from Vepco's plant-specific curves, Vepco's Single Loop Models incorporate the specific head vs. flow response for first quadrant operation found in the Units' FSAR's, 4 3
The Single Loop Model incorporates the RETRAN pressurizer model which defines two sepa' ate thermodynamic regions that are not required to be in thermal equilibrium. A non-equilibrium capability is particularly necessary when the transient involves a surge of subcooled liquid into the pressurizer. In addition, the Single Loop
3.4 Model represents the effects of subcooled spray, electrical immersion heaters, liquid droplet rainout and vapor rise in the pressurizer.
The reactor systems trip logic is modeled to the detail required for a specific analysis. RETRAN trip functions are used to model 1) protective functions, such as the overtemperature AT trip, which result in reactor scram, 2) control system bistable element logic, such as coincidence trips which model " majority" logic and 3) general problem control (e.g., problem termination, etc.).
The protective function trips necessary for the analyses documented in Section 5 and modeled in the Single Loop Model include:
1.
High flux 2.
Overtemperature AT 3.
Overpower aT 4.
Low /high pressurizer pressure 5.
liigh pressurizer level 6.
Low coolant flow 7.
Loss of power to reactor coolant pumps.
l The Single Loop Model also incorporates the RETRAN control system capability to model the following NSSS control and protection features:
1.
Overtemperature aT setpoint 2.
Overpower AT setpoint 3.
Pressure controller l
4.
Lead / lag compensation of the low pressure trip signal.
The core power response is determined by the point kinetics model in con-junction with explicit reactivity forcing functions and thermal feedback effects from moderator and fuel in the three core regions. The point kinetics model specified for the Single Loop Model incorporates one prompt neutron group and six delayed neutron groups with decay heat represented by 11 delayed gamma emitters and the important radioactive actinides, U-239 and Np-239.
Explicit reactivity forcing functions l
L-_ __ _ ____.
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3.5 represent reactor scram arm reactivity insertion due to control rod withdrawal in the Single Loop Model as the particular analysis requires.
Constant temperature coefficients or reactivity tables as a function of temperature (fuel), density (moderator)-
cr power represent feedback' effects. Core power is distributed axially among the three core conductors approximating a symmetric cosine shape. Three core materials regions fuel pellets, the helium filled gap and the Zirealoy cre used to represent the UO2 cladding. Several radial nodes are specified in the pellet region, in the gap and in the cladding. Direct moderator heating is appropriately accounted for in the model. The transient fuel and clad temperatures are calculated based on temperature-dependent thermal properties, which are input in tabular form.
The preceding paragraphs have discussed the Surry Single Loop Model in some detail. Some of the input is transient specific and the important assumptions and parameter values will be discussed for each analysis presented in Chapter 5.
3.3 Multi-loop Model Some transients are expected to have different responses in one or more of the reactor coolant loops. These transients require multi-loop representation of the NSSS. Several examples include the rupture of a main steam line resulting in the rapid cooldown of only one reactor coolant loop or the loss of power to a single reactor coolant pump resulting in a flow coastdown in only one coolant loop.
Consequently, a two loop model has been developed which represents the Surry units. One loop of the model represents a single primary coolant loop while the other loop is structured to represent two primary coolant loops. This approach is 6
consistent with current system transient analysis methodology. The model is designed with a geometrical noding which is detailed enough to analyze transients where flow and tcmperature asymmetries within the reactor vessel are significant.
The Surry Two Loop RETRAN Model, with a reactor vessel configuration appropriate for analyzing a Main Steam Line Break (MSLU) transient is shown in Figure
3.6 3.2. (The input structure of RETRAN allows rapid alterations in noding and flow path r: presentations, as may be appropriate for analyzing multiloop transients requiring less rtactor vessel detail.)
This particular configuration consists of 42 volumes, 56 junctions and 16 hrat conductor nodes. Single volumes in each loop represent the hot leg piping, steam g;nerator inlet plenum, pump suction piping, reactor coolant pump and cold leg piping.
Etch -steam generator is represented by four primary side volumes and four heat conductor nodes for the tube region.
The reactor vessel representation includes a two volume," split" downcomer, and similarly divided inlet and outlet plena. Junctions representing interloop flow mixing in the inlet and outlet plena allow for a range of mixing assumptions to be specified, such as " perfect" or. complete mixing or an incomplete mixing assumption based on actual test data (see, for example, Reference 7). The latter assumption, c:mbined with appropriate azimuthal weighting factors applied to the temperature coefficients, may be used to conservatively model the core kinetics response to a MSLB transient.- This is facilitated by a split core model in which the reactor core is ripresented by two azimuthal sectors, with each sector being divided axially into four coolant volumes. Thus, for an analysis in which an imperfect interloop flow mixing essumption is conservative, each azimuthal core sector receives more of its flow from
- the nearest loop than would be dictated by complete mixing.
Eight powered heat conductors represent the core and four passive heat conductors represent the tube region in each steam generator. Junctions representing the feedwater inlet and steam outlet in each steam generator provide secondary side boundary conditions. A junction representing safety injection of borated water via the cold leg injection path models a primary side boundary condition. Specific model
- espects will be discussed in more detail below.
As in the Single Loop Model, the Two Loop Model incorporates a Surry specific first-quadrant pump head curve and the non-equilibrium pressurizer option.
3.7 The Two Loop Model also makes use of the RETRAN valve system component model.
The simple valve option models the main steam valves and the break opening simulation associated with the severance of a main steam line.
. Trip functions are modeled in a manner similar to that discussed for the Single Loop Model. Specific protective function trips currently in the Two Loop Model include:
A.
Steam Break Protection 1.
Safety injection initiated by any of the following:
a.
Low Pressurizer pressure b.
High header / steam line pressure differential c.
High steam flow coincident with either 1) low steam pressure or
~~2) low primary system average temperature 2.~~
Main steam line isolation B.
Other-Reactor trip on low coolant loop flow.
The core power response is calculated via point kinetics in the Two Loop Model as previously discussed for the One Loop Model. A specific reactivity forcing function represents the effects of increased soluble boron levels in the core following safety injection for transients, such as the Main Steam Line Break, where safety injection is important. The time-varying core boron concentration is generated by a submodel using the RETRAN control system capability which performs a detailed calculation of the dilution and transport of safety injection fluid.
Moderator and Doppler feedback effects are represented using reactivity functions in a manner consistent with that reported in References 3, 4 and 7.
The feedback effects are weighted axially based on certurbation theory approximations; azimuthal weighting may be by volume, or for situations where skewed inlet temperature distributions are important, a conservative non-uniform weighting scheme such as discussed in Reference 7 is used. Noding in the fuel, gap and cladding regions is the same as that discussed for the One Loop Model.
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Table 3.1 Thermal - Hydraulic Design Parameters - Surry Plant Total core heat output, Mwt 2441 Heat generated in fuel, %
97.4 System operating pressure, psi 2250 Total coolant flow rate, Ib./hr.(gpm) 100.7 x 106 (265,500)
Coolant Temperatures, F (@l00% power)
Nominal inlet 543 Average rise in the core 65.5 Average rise in vessel 62.6 Average in the core 577.0 Average in vessel 574.
Nominal core outlet 608.5 Nominal vessel outlet 605.6 Average linear power deasity, Kw/ft.
6.2 e
'SECTION 4 - SYSTEM TRANSIENT ANALYSIS METHODOLOGY '
4.1 Introduction As discussed in the introduction, Vepco system transient analysis is intended for both best estimate and licensing applications. Since core reloads are the most common and expected reason for accident reanalysis, Vepco's system transient methodology will be discussed in that context.
In general, Vepco intends to continue the reference analysis approach which has been employed by our nuclear fuel vendor in support of our nuclear plants.
This approach is fully explained in Reference 8 and requires reanalysis of an accident, which is part of the licensing basis for our plants, only under certain conditions. These conditions and the licensing evaluation process are summarized in Section 4.2. Section 4.3 discusses the system transient analysis methodology and its relation to the licensing process.
4.2 - Licensing Evaluation Process The actual execution of transient analyses forms part of an integrated system of a
cvaluations performed to verify the acceptability of a reload core design from the standpoints of safety, economics and operational flexibility. The purpose of this section is, therefore, to provide a brief overview of the relationship of transient analyses to the Integrated reload design and licensing process.
The reload design process will be d; scribed in detail in a future Vepco topical report. However, the process has been g:nerally described in Reference 8 and consists of a design initialization, design of the core loading pattern, and detailed characterization of the core loading pattern by the nuclear designer. The latter process determines the values of key reload parameters.
l These key reload parameters are provided to the safety analyst who uses them in f
l conjunction with current plant operating configurations and limits to evaluate the impact of the core reload on plant safety, i
i i
4.2 In performing this evaluation, it is necessary to ensure that those key parameters which influence accident respoase are maintained within the bounds or " limits" established by the parameter values used i's the reference analysis (i.e. the currently applicable licensing calculation). The reference analysis (and the associated parameter limits) may be updated from time to time in support of a core reload or to evaluate the impact of some other plant parameter change.
For cases where a parameter falls outside these previously defined limits, an evaluation of the impact of the change on the results for the appropriate transients must be made. This evaluation may be based on known sensitivities to changes in the vrrious parameters in cases where a parameter change is small or the influence on the cecident results is weak. For cases where larger parameter variations occur, or for pirameters which have a strong influence on accident results, explicit reanalysis of the affected transients is required and performed as discussed in Section 4.3.
Past enalytical experience has allowed the correlation of the various accidents with those p:rameters which have a significant impact on them.
The results of such a correlation are summarized in References 3, 4 and 8.
If required, a reanalysis is performed and the results are compared to the appropriate enalysis acceptance criteria identified in References 3,4 and 8. The reload evaluation process is complete if the acceptance criteria are met, and internal documentation of the reload evaluation is provided for the appropriate Vepco safety review. If the analysis acceptance criteria are not met, more detailed analysis methods and/or Technical Specifications changes may be required to meet the acceptance criteria. The NRC will be informed of the results of the evaluation process in accordance with the requirements of 10CFR 50.59.
4.3 System Transient Analysis The production of a conservative, reliable safety analysis of a given anticipated or postulated transient is accomplished by combining a' system transient model with
_ _ ~
4.3 l-cppropriate transient specific input. A system transient model, such as those discussed l
in Section III, is designed to provide an accurate representation of the reactor plant and those associated systems and components which significantly affect the course of the tr:nsient. Transient specific input ensures that the dynamic response of the system to the postulated abnormality is predicted in a conservative manner, and includes a) initial conditions, b) core reactivity parameters such as Doppler and moderator temperature coefficients, and control rod insertion and reactivity characteristics, and c) assumptions concerning overall systems performance. Important systems performance assumptions include the availabili,ty of certain system components (such as pressurizer spray or relief valves) and control and protective characteristics (setpoints, instrument errors, d;1ay times).
A summary of key analysis assumptions for those transients discussed in Chapter 5 is included in the Appendix. A general discussion of this transient specific input is provided in the paragraphs which follow.
4.3.1 System Model Application While RETRAN affords the modeling flexibility to develop an infinite number of r: presentations for a given nuclear plant, practical considerations dictate that a small number of standard plant models be assembled and maintained for performance of the entire spectrum of system transient analyses. Section 3 provides examples of the types of models that are required for system transient analysis. RETRAN makes use of an input structure which allows modification of the base deck input for specific cases by use of override cards. Thus, specific transient cases may be executed without altering the base plant models.
The base models are designed to provide a basic system description comprised of those parameters which would not ordinarily change from cycle to cycle. Thus such parameters as system volumes and flow areas, characteristics of various relief and j
s:fety valves, primary coolant pump characteristics, etc. form part of the base models.
l
c 4,4 i
Since occasional changes to such " fixed" parameters do occur as a result of equipment modifications or replacement or upgrades to various safety-related systems, the base models are reviewed periodically to ensure that the latest system-related changes have been adequately reflected. Generally this review is performed duchig the initial core design stages of each reload cycle.
4.3.2 Transient Specific Input As discussed earlier, input parameters which may be varied for a specific analysis to ensure a conservative representation of the system response include initial conditions, core reactivity parameters and assumptions concerning systems perfor-mance. For a given type of accident, not all parameters have a significant influence on the accident response. Those parameters which are significant, and their limiting directing (i.e., maximum or minimum) are determined from:
a) the unit's FSAR b) sensitivity studies such as those summarized in Reference 8.
The most important of these safety-related parameters are examined in more d; tail in the following discussions.
4,3.2.1 Initial Conditions Most accidents exhibit some sensitivity to the initial conditions assumed. For eccident evaluation, the initial conditions are obtained by adding or subtracting, as gppropriate, maximum steady-state errors to or from rated values. Steady-state errors which are applied are:
e)
Core Power
+ 2 percent allowance for calorimetric error b)
Average reactor coolant 14 F (Surry) system temperature allowance for deadband and measurement error.
c)
Pressurizer pressure 130 psi allowance for operational fluctuations and measurement error.
In general, errors are chosen in the directions which minimize core thermal margin or margin to other plant design criteria and are therefore dictated by the type of analysis being performed.
1
i 4.5
)
4.3.2.2 Reactivity Parameters s
Reactivity parameters, which may have a significant impact on the transient response to an abnormal condition, include the Doppler and moderator temperature coefficients of reactivity, delayed neutron fractions, the trip reactivity and insertion characteries, and the differential control bank worth. The reactivity parameters are normally chosen in a manner which tends to maximize the nuclear power during the transient. The limiting value of a given parameter is dictated by the type of transient involved as indicated by the examples in Chapter 5. For example, for transients where large decreases in moderator temperature are a concer.) (such as a steamline break),
large negative moderator temperature coefficients tend to be limiting. On the other hand, for transients where increases in moderator temperature are the major concern (for example, a loss of external electric load or turbine trip) the most positive value of moderator temperature coefficient tends to' produce a more severe transient. The i
choice of the limiting reactivity parameter value, as discussed earlier, is made to i
ensure that the accident analyses are bounding with respect to the range of parameter values realized over the life of the reload core.
4.3.2.3 System Performance Assumptions The predicted transient performance is influenced by assumptions concerning the 1
cvailability of various system components and the characteristics of the reactor protection and control system.
l In many instances the mitigating effect of various system design features on l
postulated trcnsients are ignored. This provides additional conservatism and confidence l
that the calculation conservatively " bounds" the actual expected system performance.
4 For example, the analysis of the Uncontrolled Rod Withdrawal from Suberitical transient conservatively takes no credit for the source range or intermediate range flux level trips or for the intermediate range control rod stop function. For certain control system components (e.g., relief and spray valves), it is conservative to assume
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4.6 availability for some transients and unavailability for others. The choice of whether or not to include the effect of a particular system component is based on prior experience end sensitivity studies. These assumptions normally remain constant from analysis to analysis of a given transient.
In order to adequately account for the impact of instrumentation errors and signal delays, conservative protection system characteristics are assumed when performing tecident analyses. Thus, expected instrument errors and system response times are conservatively bounded by the analysis assumptions, thereby adding to the previously discussed conservatisms employed in a transient analysis.
Examples of protection system setpoints and delays used in performing Surry safety analyses are shown in Table 4.1. Periodic review of protection system setpoints as defined in the plant Precautions, Limitations and Setpoints is performed to ensure that the safety analysis models continue to conservatively reflect current safety system settings.
4.4 Use of System Transient Results The results of a system thermal hydraulles analysis are used either for direct comparison to accident analysis acceptance criteria (e.g. system pressure limits) or as a boundary condition for more detailed core thermal hydraulic analyses, using the Vepco capability documented in Reference 1, or for more detailed fuel rod analyses, as required for some condition IV transients.
TABLE 4.1 - PROTECTION SYSTEM CHARACTERISTICS ASSUMED IN SAFETY ANALYSIS Mode of Protection Surry Setpoint (Delay time, sec.)
High neutron flux, Fraction of Rated Low Power Range 0.35(0.5)
High Power Range 1.18(0.5)
Overtemperature AT Variable (6.0*)
Loss of Pump Power
"(1.2)
L:w Reactor Coolant Loop Flow, Fraction of Full Flow 0.87(0.6)
High Pressurizer Pressure, psia 2425(1.0)
Initiation of Safety Injection flow on high steamline A P, psi 150.0(Variable) on low pressurizer pressure, Psia 1715(Variable)
This value includes loop and RTD bypass line transport delays, the RTD thermal o
time constant and electronic signal processing delays.
00 Undervoltage trip setpoint not used in analysis.
SECTION 5.0 - QUALIFICATION COMPARISONS 5.1 - Introduction As discussed in earlier sections, the primary Vepco objectives in developing I
a system transient analysis capability are to provide a basis for the reload core safety i
enalysis and licensing process and to support reactor operations. As verification of this cIpability, appropriate results and comparisons are provided for a representative series
- of analyses of licensing and best estimate plant transients. The selection of licensing analyses for presentation was based on 1) consideration of those transients which are i
thermally limiting and have been most frequently subject to reanalysis during the reload 5'
licensing process (e.g. Rod Withdrawal ftom Power and Complete Loss of Flow); 2) i providing a selection of analyses for each of the major categories of initiating events f
which include changes in reactivity (such as rod withdrawal transients), variations in primary coolant flow rate (such as loss of flow transients) and variations in primary to s:condary system heat transfer rates (e.g. Main Steam Line Break); and 3) examination of transients which are both symmetric (such as a Loss of Load) and asymmetric (such as a single pump flow coastdown) with respect to the thermal hydraulic response of the reactor coolant loops.
Comparisons to plant startup flow coastdown test data and the data taken during a reactor cooldown transient experienced at North Anna in 1979 are also j
provided to illustrate typical best estimate modeling applications.
Comparisons for small and large break Loss of Coolant Accidents (LOCA) f cnd Rod Ejection are beyond the current intended scope of application of Vepco's models and are not presented.
5.2 Verification Against Licensing Analyses I
5.2.1 Transients Resulting from Changes in Reactivity l
ll Several transients result primarily from a postulated reactivity change.
l These transients include an Uncontrolled Control Rod Assembly Withdrawal From a s
5.2 i
Suberitical Condition (UCRW from Suberitical), an Uncontrolled Control Rod Assembly Withdrawal at Power'(UCRN at Power), Control Rod Assembly Drop, Chemical and Vclume Control System Malfunction, Startup of an Inactive Loop, Single Control Rod Assembly Withdrawal at Power and Control Rod Assembly Ejection. The first two cecidents were chosen for analysis because they are subject to reanalysis for reload 4
cores based on past Vepco experience. In addition, these two accidents represent a i
limiting condition for reactivity change rate (UCRW from Suberitical) and DNBR (UCRW from Power) with respect to the other Condition II accidents.
5.2.1.1 Uncontrolled Control Rod Assembly Withdrawal from a Suberitical Condi-
\\
tion Transient - FSAR Analysis A control rod assembly withdrawal incident is defined as an uncontrolled cddition of reactivity to the reactor core by withdcawal of control rod assemblies resulting in a power excursion. While the probability of a transient of this type is extremely low, such a transient could be caused by a malfunction of the Reactor.
i Control or Control Rod Drive Systems. Section 14.2.1 of the Surry FSAR (Reference 3) 1 discusses the mitigeting automatic safety systems appropriate for this transient in more 4
d: tall.
The nuclear power response to a continuous reactivity insertion from a suberitical condition is characterized by a very fast rise terminated by the reactivity fzedback effect of the negative fuel temperature coefficient. This self-limitation of the initial power excursion is of prime importance during a startup incident, since it limits the power to a tolerable level prior to external control action. After the initial power excursion, the nuclear power is momentarily reduced, and then, if the incident is not terminated by a reactor trip, the nuclear power increases again but at a much I
slower rate.
This is a Condition II event, and the analysis is performed to demonstrate that the DNB criterion for Condition II events is met.
5.3 In order to give comparable results, the analysis assumptions used in this investigation are the same as those indicated in Reference 3. The limiting input values cnd analysis assumptions assumed for this investigation are provided in the Appendix (Item la). The Single Loop Model, discussed in Section 3, was used for the analysis.
Figures 5.1 through 5.4 present the results of the analysis using the RETRAN computer code as compared to the FSAR results for nuclear power, average fuel and clad temperature and core heat flux, respectively.
The RETRAN results are based on a single integrated kinetics and thermal-hydraulic calculation. The FSAR results, in contrast, reflect separate core kinetics (power) and heat transfer calculations, performed with two computer codes, with distinct sets of input assumptions designed to conservatively maximize core heat flux.
This distinction in analytical approach most likely accounts for the differences in results for the average fuel and clad temperatures.
Note that both calculations result in pre'dicted heat fluxes, and fuel and clad temperatures which are well below steady-state full power values. Therefore large margins to the Condition II DNB limits are maintained throughout the transient.
5.2.1.2 Uncontrolled Control Rod Assembly Withdrawal from a Suberitical Condition Transient - Current Analysis Due to changes in the calculated limit for the reactivity insertion rate parameters, this transient was reanalyzed for several reload cores.
The latest reanalysis was for Cycle 4 of Surry Unit 2.9 The assumptions used for this analysis are the same as those discussed in Section 5.2.1.1, with the exception of the limiting reactivity insertion rate which was increased to a value of 75 pcm/sec*, and a modification in the trip reactivity (see the Appendix, Item Ib).
The comparison of the vendor reload analysis and RETRAN results is indicated by the excellent agreement for the core heat flux, the limiting analysis result,
[
e3 reported in the licensing submittal. The RETRAN and vendor reload analyses both
~~1 pcm = 1.0 x 10-0 A K/K i~~
5.4 yielded peak values of 69% of nominal full power core heat flux. Figures 5.5 through 5.8 provide the complete RETRAN transient response for the appropriate parameters.
The vendor transient results are proprietary and are omitted. The transient response is similar to and consistent with the comparisons indicated in Figures 5.1 through 5.4.
5.2.1.3 Uncontrolled Control Rod Assembly Withdrawal at Power Transient - FSAR Analysis This postulated transient, which is a Condition II event, was analyzed because it is a limiting reactivity perturbation transient with respect to the minimum DNBR criterion and because it is subject to reload reanalysis. This transient is defined as an uncontrolled addition of reactivity to the reactor core while in an at-power condition resulting in a power excursion and an increase in core heat flux. Since the
. heat extraction from the steam generator remains relatively constant until the steam generator pressure reaches the relief or safety valve setpoint, there is a net increase in reactor coolant temperature. Unless terminated by manual or automatic action, the power mismatch and resultant coolant temperature rise would eventually result in DNB.
Therefore, to prevent the possibility of damage to the cladding, the Reactor Protection System is designed to terminate any such transient before the DNBR falls below its limit. The automatic features of the Reactor Protection System, which would prevent core damage in a control rod assembly withdrawal incident at power, are discussed in detail in Reference 3.
In order to obtain conservative results (i.e., minimtim DNBRs) for this transient and to provide a consistent comparison, the analysis assumptions are the same as those indicated in the FSAR.
These assumptions, and the limiting values assumed for this analysis are provided in the Appendix (Item 2a). The Single Loop Model, discussed in Section 3, was used for this analysis.
It should be noted that the Overtemperature Delta T Trip setpoint equation, which is important for this transient, is explicitly modeled in the Single Loop Model using the control system capability in RETRAN.
5.5 The FSAR presents the results of this transient for several initial power levels and for various reactivity insertion rates. However, a full range of system parameter transient results is presented only for two analyses from en initial power level of 100% The two 100% analyses are for differing reactivity insertions rates to d2monstrate the protective action of both the liigh Flux and the Overtemperature Delta T Trip functions. Of the two transients, the more limiting results are for the slow tractivity insertion (2 pcm/sec) which is terminated by the Overtemperature Delte T Trip. Consequently, the analysis used for compur'ison of the RETRAN and FSAR results tssumed a slow reactivity insertion rate of 2 pcm/see starting from 102% of nominal full power.
Analysis results for a range of reactivity insertion rates are discussed in the next section.
Figures 5.9 through 5.12 present the RETRAN results, compared to the FSAR for nuclear power, pressurizer pressure, average coolant temperature and I
transient DNBR, respectively. The DNBR's were calculated with COBRA IIIC/MIT using input forcing functions of core heat flux, coolant inlet temperature, coolant inlet mass velocity and RCS pressure, all from the RETRAN analysis. Note the similarities in time of trip (Figure 5.9). The decay heat level shown in the FSAR result apparently reflects the conservatism used by the vendor prior to the development of the ANS standard decay heat curves. Note also the similarity in predicted pressure responses in Figure 5.10, including the effects of automatic spray and Power Operated Relief Valve (PORV) actuation. The RETRAN analysis shows, as does the FSAR, that the Condition 11 DNB criterion is met for this transient.
5.2.1.4 Uncontrolled Control Rod Assembly Withdrawal at Power Transient -
Current Analysis The most recent reanalysis of this accident was required as a consequence of the plugging of steam generator tubes at the Surry Nuclear Power Station.10 It was determined that steam generator tube plugging would result in lower initial flows with consequently less initial margin to DNB and the need for revision of the constants
5.6 associated with the Overpower and Overtemperature Delta Temperature setpoint equation. Consequently, the UCRW at Power transient was reanalyzed to verify that the new setpoint equation constants did in fact result in minimum DNBRs above the cppropriate criterion of 1.3.
The only information available for comparison purposes fr:m the licensing reanalysis was the minimum DNBR as a function of reactivity insertion rate. An enalysis of the transient was performed using the Single Loop Surry RETRAN Model with those assumptions specified in the Appendix (Item 2b), including several modeling changes to reflect the impact of the low flow assumption (i.e. lower fl:ws, lower steam generator heat transfer areas, etc.). Key input parameter values cssumed for this analysis are also provided in the Appendix (Item 2b).
The RETRAN results were then used as boundary conditions in the Vepco I
v:rsl6n of the COBRA IIIC/MIT code. The results of this transient reanalysis are presented in Figure 5.13.
Another analysis of the transient was performed at an initial power level of 62% of nominal full power. The results of this analysis and a comparison to licensing r: analysis results are provided in Figure 5.14. RETRAN results were generated with a
cnd without the assumption of operable steam generator relief valves, as shown. These results show that the RETRAN/ COBRA analysis supports the conclusion provided by the licensing reanalysis, i.e., that the updated setpoint equation constants are sufficient to provide margin to the Condition II DNBR limit for reactor operation with 90% or greater of thermal design flow.
5.2.2 Transients Resulting from Changes in Primary System Flowrate Several FSAR transients result primarily from the loss of Reactor Coolant System (RCS) flow and the corresponding decreased transfer of heat from the reactor core. Transients in this category include the Loss of Reactor Coolant Flow (partial and complete) and the Locked Rotor transients. The Complete Lose of Reactor Coolant Flow Transient was chosen for comparative analysis because it has been subject to rcanalysis for reload cores based on past Vepco experience. In addition, it is the most
Figure 5.1 NUCLTAR POWER UNCONTROLLED ROD WITHDRAWAL FROM SUBCRITICAL TRANSIENT FSAR ANALYSIS 1
3.5 PITE ~~

~~FSAR g~~

3.0 l
3 w
e 2.5 l
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z
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q 0
5 10 15 20 Time, Seconds
t i
d Figure 5.2 AVERAGE FUEL TEMPERATURE UNCONTROLLED ROD WITHDRAWAL FROM SUBCRITICAL TRANSIENT FSAR ANALYSIS
' ~ - ~ ~
l' RETRAN ---
i A
-..- -. _.. _ \\ - - -.. _. -

~~FSAR g~~

I.
800
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600 500 0
5 10 15 20 Time, Seconds
Figure 5.3 AVERAGE CLAD TEMPERATURE UNCONTROLLED ROD WITHDRAWAL FROM SUBCRITICAL TRANSIENT FSAR ANALYSIS' RETRN!

~~FS AR l~~

610
,u.
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U 600 i
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I Figure 3.i CORE HEAT FLUX UNCONTROLLED R0D WITHDRAWAL FROM SUBCRITICAL TRANSIENT FSAR ANALYSIS
=
._.e..
..-_e sp,.m 0.7 RETPE
_-..._.._.____.___..-------FSAR
. _ _.. _ \\
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Figure 5.5 NUCLEAR POWER UNCONTROLLED R0D WITHDRAWAL FROM SUBCRITICAL TRANSIENT SURRY 2 CYCLE 4 REANALYSIS RETRAN -----
5.0 l
q
.5 Eg 4.0 w
,o t
c h
3.0 g
9 cy 8
2.0 Z
1 l
l l.0 0
5 10 15 20 Time, Seconds
Figure 5.6 AVERAGE FUEL TEMPERATURE UNCONTROLLED ROD WITHDRAWAL FROM SUBCRITICAL TRANSIENT SURRY 2 CYCLE 4 REANALYSIS
..__.._J 800 y
a g
g u
a 700 g
w g
y
~ ~ ~
_\\.
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600 500 e
l l
0 5
10 15 20 Time, Seconds l
i Figure 5.7 AVERAGE CLAD TEMPERATURE UNCONTROLLED ROD WITHDRAWAL FROM SUBCRITICAL TRANSIENT SURRY 2 CYCLE 4 REANALYSIS 620 l
610 i
g 600 j
C v
e 590
-_\\..---._-.--....-.-.--
3 1_
l z;
5 580 g
570 560 f
i l
l 550 l
0 5
10 15 20 l
Time, Seconds
)
i 1
Figure 5.8 CORE HEAT-FLUX UNCONTROLLED R0D WITHDRAWAL FROM SUBCRITICAL TRANSIENT l
SURRY 2 CYCLE 4 REANALYSIS l
t l
l z
0.6 7
ay o
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0.5
-. _. - =. - _ - -.
g i"
~~ ---
0.4 3
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u 0.2 0.1 1
0.0 0
5 10 15 20 Time, Seconds
i Figure 5.9
{
hTCLEAR POWER l
UNCOUROLLED R0D WITHDP.AWAL FROM POWER TRANSIENT FSAR ANALYSIS l
l EETEM m
1.2 1.0 l
,c S
0.8 I
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10 20 30 40 50 60 Time, Seconds l
l
Figure 5.10 PRESSURIZER PRESSURE UNCONTROLLED ROD WITHDRAW L FROM POWER TRANSIENT FSAR ANALYSIS REH'RNi g.gp
.e.e z
L
~ ~ - -
5 2400 l
g 3
t i
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0 10 20 30 40 50 60 Time, Seconds I
t
i i
i i
i i
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Figure 5.11 4
AVERAGE COOLANT 4EMPERATURE UNCONTROLLED ROD WITHDRAWAL.FROM POWER TRANSIENT i
FSAR ANALYSIS j
i gg7g... _
l FSAR I
590 g
=
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D 10 20 30 40 50 60 Tic 2e, Seconds I.
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Figure 5.12 1
DNB RATIO UNCONTROLLED ROD WITHDRAWAL FROM POWER TRANSIENT FRAR ANAT.YST S l
vepco FSAR._._.
2.3
~
~
.,l l
g i.
5 2.1
- I--
___ j l_._ _ _ _
l 1.9 I
I.
l l.
1.7
's 8
I.
I
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0 10 20 30 40 50 60 Time, Seconds
]
Figure 5.13 VARIATION OF MINIMUM DNBR WITH REACTIVITY INSERTION RATE ROD WITHDRAWAL FROM 102* POWER STEAM GENERATOR TUBE PLUGGING REANALYSIS 1.50 1.40 p-i j
/
l
=
N
/
I 2
//
%</
1.30
- VEPCO
/endor 1.20
-6 10 10' 10 10' Reactivity insertion Rate. AK/Second
Figure 5.14 VARIATION OF MINIMUM DNER WITH REACTIVITY INSERTION RATE ROD WITHDRAWAL FROM 627. POWER STEAM GENERATOR TUBE PLUGGING REANALYSIS l
l
/
vac0 1.60 Venc or
/
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-6 10 10' Reactivity Insertion Race A K/Second
5.7 stvere credible loss of flow condition.
The Partir.1 (one-punip) Loss of Flow was analyzed to provide qualification of the Two Loop Mcdel.
5.2.2.1 Complete Loss of Flow Transient - FSAR Analysis This postulated transient, which is a Condition III event, is defined as the simultaneous loss of electrical power to all reactor coolant pumps at full power resulting in a rapid RCS flow reduction and consequent coolant temperature increase with the possibility of Departure from Nucleate Boiling (DNB) if the reactor is nct
-I tripped promptly. The necessary protection action to preclude DNB is discussed in more s
dstall in Reference 3.
The conservative assumptions used in the RETRAN analysis, which are delineated in the Appendix, (Item 3a) are the same as those presented in Reference 3.
Specific limiting parameter values assumed are also provided in the Appendix. The
' RETRAN analysis was performed with the Single Loop Model discussed in Section 3.
Figures 5.15 through 5.18 present the results of the comparisons for this transient for flow coastdown, nuclear power, core heat flux and DNBR, respectively.
As discussed previously, the DNBRs were calculated with the Vepco version of the COBRA IIIC/MIT computer code using boundary conditions obtained from the RETRAN analysis. The minimum DNBR predicted by the Vepco analysis was 1.50 and compares very favorably with the value of 1.46 reported in the FSAR analysis. Time of occurrence of minimum DNBR also compared well and was approximately 2.3 seconds for both analyses.
Thus the RETRAN/ COBRA results support the FSAR conclusion
. that, while Complete Loss of Flow is a Condition III transient, the Condition II DNB criterion is met for this event.
5.2.2.2 Complete Loss of Flow Transient - Current Analysis The Complete Loss of Flow transient has had to be reanalyzed in the past t
- for the Surry plants. The most recent analysis was required as a consequence of the j
plugging of steam generator tubes.10 The tube plugging resulted in reduced primary coolant flow and less initial margin in DNB. Since the Loss of Flow transient was i
5.8 potentially affected, the transient was reanalyzed to verify the continued acceptability of the results.
An analysis of the transient was performed with RETRAN using the assumptions specified in the Appendix (Item 3b).
The specific parameter values essumed for this analysis are also provided in the Appendix. The Single Loop Model, as modified to reflect the effects of steam generator tube plugging (lower flows, steam g:nerator heat transfer areas, etc.), was used for the analysis. A conservatively low value of initial flow was assumed in the analysis.
The comparative results of this reanalysis are provided in Figures 5.19 through 5.22. Figure 5.19 shows the comparison of pump coastdown for the respective analyses, and Figure 5.20 compares the nuclear power response. Figure 5.21 presents the results for core average heat flux, and the DNBR response using the RETRAN/ COBRA methodology is compared in Figure 5.22 to the prediction reported in the licensing reanalysis. The Vepco predicted minimum DNBR again agrees wellin both magnitude and time of occurrence to the licensing reanalysis results and confirms that the Condition II DNB criterion is met for tilis event.
5.2.2.3 Partial Loss of Flow Transient - FSAR Analysis In addition to the Complete Loss of Flow transient, discussed in the two previous sections, various Partial Loss of Flow Accidents may be postulated, in which power is lost to one or more reactor coolant pumps, with the remaining pumps continuing to operate at full speed. Such a transient would result from failure of a single pump bus. Since this does not constitute loss of line voltage or frequency, no credit is taken for the direct reactor trip on low voltage. Instead, protection of the core is provided by a reactor trip actuated by low measured reactor coolant flow in any primary coolant loop.
Since this transient involves unbalanced reactor coolant loop flow rates, the Surry Two Loop Model is used for the RETRAN analysis. The case analyzed assumes initial operation of all reactor coolant loops, with a subsequent loss of pump power in a
5.9 single loop. Specific parameter values and initial conditions assumed for this analysis see shown in the Appendix (Item 4). The low coolant flow trip setpoint and delay time assumed are consistent with Table 4.1.
The results of the RETRAN analysis are compared to the corresponding 3
FSAR results in Figures 5.23 to 5.26 for core flow, nuclear power, core average heat flux and DNBR, respectively.
Asirs previous DNB analyses presented in this section, the Vepco curve was gtnerated with the Vepco' version of COBRA IIIC/MIT, using input forcing functions from the two loop RETRAN analysis. Again, the Vepco results confirm the conclusion that the Condition II DNB criterion is met for this transient.
5.2.3 Change in Primary to Secondary Heat Transfer The remaining types of non-LOCA perturbations analyzed for a nuclear plant in a FSAR are characterized by changes in primary system pressure and temperature resulting from changes in primary to secondary heat transfer.
Accidents in this category would include Excessive Heat Removal Due to Feedwater System Malfunction, Loss of External Electrical Load, Excessive Load Increase Incident, Loss of Normal Feedwater, Loss of all AC Power to the Station Auxiliaries, Turbine Generator Unit Overspeed and Main Steam Line Break. The majority of these transients are nonlimiting and have not been reanalyzed since the FSAR. However, the Main Steam Line Break and Loss of Load transients have required reanalysis as a result of core reloads and for that reason were chosen for comparative analysis. In addition, the Main Steam Line Break transient reanalysis required a multiloop capability and served to qualify the Two i
Loop Model discussed in Section III.
Finally, the Feedwater System Malfunction transient was analyzed to further demonstrate the capability of the Single Loop Model
(
to represent a secondary side initiated transient.
5.2.3.1 Loss of External Electrical Load Transient - FSAR Analysis The Loss of Load transient is defined as the loss of external electricalload wisich may result from an abnormal variation in network frequency, or other adverse i
f L
l I
Figure 5.15 i
i FLOW COASTDOWN 1
COMPLETE LOSS OF FLOW TRANSIENT FSAR ANALYSIS 1
RETRAN FSAP.
u I
3e 100 1
i b
y M
.5 80 s
i s N c
... N __.
s O
"y 60 m
y
. _.. _s%
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= - +
1 20
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0 0
2 4
6 8
10 Time, Seconds l
1
l l
l l
Figure 5.16 4
1 NUCLEAR-POWER COMPLETE LOSS OF FLOW TRANSIENT FSAR ANALYSIS i
gpg e
g
^
c 0
100 w
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80
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4 5
Time, Seconds w
Figure 5.17 AVERACE HEAT FLUX COMPLETE LOSS OF FLOW TRANSIENT FSAR ANAT.V9TR i
FSAR
=
C>
100
~
~ ~ ~ ~
c w
c 80
^
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xy N
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u, 0
60 g
l o
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=
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5 i
Time, Seconds l
Figure 5.18 DNB R1TIO COMPLETE LOSS OF FLOW TRANSIENT FSAR ANALYSIS u_
Vepco y
2.4 2.3 7
~
g 2.2
-~
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.-_._:..___.a_-_.-.__.-
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1 2
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5 Time, Seconds
[
l I
i l
l Figure 5.19 i
FLOW COASTDOWN COMPLETE LOSS OF FLOW TRANSIENT STEAM GENERATOR TUBE PLUGGING REANALYSTS l
n l
i RETRAN ---
l Vendor _.--
l i
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~
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1 2
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Time, Seconds l
i l
l Figure 5.20 NUCLEAR POWER COMPLETE LOSS OF ILOW TRANSIENT STEAM GENERATOR TUBE PLUGGING REANALYSIS RETRAN Vendor.._
(
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0 1
2 3
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Time, Seconds l
l l
e Figure 5.21 AVERAGE HEAT FLUX COMPLETE LOSS OF FLOW TRANSIENT STEAM GENERATOR TUBE PLUGGING REANALYSIS 4
a e
-ee
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=
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Figure 5.22 DNB RATIO COMPLETE LOSS OF FLOW TRANSIENT STEAM GENERATOR TUBE PLUGGING REANALYSIS Vepco
-~

~~Vendor 1.9 s~~

/. _ _.... _
z s
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Figure 5.23 COPI FLOW COASTDOWN PARTIAL LOSS OF FLOW TRANSIENT FSAR ANALYSIS i
t 1
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RETFAN I

~~FSAR i~~

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Figure 5.24 NL'CIIAR POWER PART.IAL LOSS OF FLOW TRANSIENT FSAR ANALYSIS
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1 Figure 5.25 i
CORE AVERACE HEAT FLUX PARTIAL LOSS OF FLOW TRANSIENT l
FSAR ANALYSIS l
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Figure 5.26 DNB RATIO PARTIAL LOSS OF FLOW TRANSIFNT FSAR ANALYSIS m.
i i
Vepco f

~~FSAP.~~

,t i
2.2 4
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c
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5.10 network operating conditions and is considered a Condition II event. The interaction of the mitigating systems for the various credible initiating actions for this transient are discussed in further detail in References 3 and 11. For analysis purposes, the limiting condition of a complete loss of load from 102% of nominal full power without a direct reactor trip is assumed to demonstrate 1) the adequacy of the pressure relieving devices to maintain the RCS within the Condition II pressure boundary criterion (i.e.110% of design pressure) and 2) that the Condition II DNB limits are not violated for both beginning of life (BOL) and end of life (EOL) core conditions.
The conservative assumptions used in the Reference 3 analysis were assumed for the RETRAN comparative analysis (note that the limiting FSAR analysis condition for the reactor in manual control was assumed). These asumptions and the specific analysis parameter values are indicated in the Appendix (Item Sa). Note, in particular, that many of the system pressure relieving devices are assumed to be inoperative in order to produce conservative results. The RETRAN analysis was performed with the Single Loop Model.
The comparative results for this analysis are provided in Figures 5.27 through 5.31 for the BOL parameters and Figures 5.02 through 5.36 for the EOL case.
The constraining result for this analysis is the pressurizer pressure and the change in this parameter is provided in Figure 5.27. Note that the rate of pressure change, the time of peak pressure and the magnitude of the peak pressure calculated for the respective analyses are in close agreement for the pressurization period of the transient. However, some deviation exists during the depressurization phase of the transient. This deviation most likely results from steam generator secondary side modeling differences used in RETRAN and the FSAR analyses.
Both analyses demonstrate that the RCS pressure criterion for Condition II events is met.
Figures 5.28-5.31 provide the RETRAN and FSAR responses for nuclear power, pressurizer water volume, coolant inlet temperature and DNBR, respectively.
5.11 u
The DNBR's were generated with COBRA. DNB is not limiting for this event, as can be s:en from Figure 5.31..
Figures 5.32 through 5.36 present the results for the Loss of Load EOL cnalyses and again confirm that the Condition II pressure and DNB criteria are not violated.
5.2.3.2 Loss of External Electrical Load Transient - Current Analysis The Loss of Load has been reanalyzed since the FSAR to support a Technical Specification change allowing core operation with a slightly positive moderator temperature coefficient at powers less than hot full power at BOL.12 The licensing reanalysis, to be used for comparison purposes, was only p;rformed for the BOL case, since the moderator temperature coefficient would be highly negative at EOL and, therefore, not impacted by the proposed Technical Specifications change. The RETRAN analysis assumptions and parameter values are provided in the Appendix (Item Sb); note that for the moderator temperature coefficient, a value of +3.0 pcm/ F was assumed. The Single Loop Model was used for the RETRAN analysis.
A comparison of the RETRAN and licensing reanalysis results is shown in Figures 5.37 through 5.40. Comparisons are provided, for nuclear power, pressurizer pressure, coolant average temperature, and DNBR.
As discussed previously, the secondary side heat transfer modeling differences resulted in some differences in the predictions during the depressurization phase. The RETRAN analysis results confirm the conclusion drawn in the licensing reanalysis, i.e., that the pressure relieving devices are adequate to limit the peak pressure to a value below the Condition II Criterion and that the Condition II DNBR Criterion is also met.
5.2.3.3 Excessive Heat Removal Due to Feedwater System Malfunction Transient -
FSAR Analysis Excessive heat removal incidents resulting from feedwater system malfune-tions result from either 1) excessive feedwater flow, such as might result from a failure of the feedwater flow control valve or 2) reductions in feedwater
~~temperature. An~~
5.12
- example of the second type of transient, which consists of the accidental opening of the feedwater bypass valve resulting in diversion of flow around the low pressure feedwater heaters, was chosen for analysis.
The case examined, which was analyzed in Reference 3, assumed no reactor c:ntrol and a zero moderator temperature coefficient. The resulting transient is a very gradual increase in core power in response to the primary coolant and fuel temperature rzduction resulting from the decreased temperature of the feedwater to the steam g:nerators. After the core power increases to a level which essentially matches the
-s:condary side heat removal rate, the temperature begins to stabilize and the system pressure increases in response to the pressurizer heaters.
The Appendix (Item 6) summarizes the important analysis assumptions made 3
' for both the FSAP and RETRAN analpes, including specific analytical parameter values assumed. The RETRAN analysis was performed with the Single Loop Model discussed in Section 3, and conservatively assumes constant steam flow throughout the transient.
The RETRAN analytical results are compared to the results reported in the FSAR, in Figures 5.41 through 5.45. It should be noted that this transient is calculated over a long time period, approximately 900 seconds.
Figure 5.41, which represents the variation in feedwater temperature with
~ time, depicts the forchig function assumed in the two anulyses.
Figures 5.42 - 5.45 present the results for core power, average coolant
('
l tamperature, pressurizer pressure and DNBR.
The primary FSAR conclusion, that DNBR increases monotonically as the i-l transient proceeds, is supported by both analyses.
5.2.3.4 Accidental Depressurization of the Secondary System / Main Steam Line Break Transient - FSAR Analysis l
This class of accidents includes any uncontrolled steam release from a steam generator, such as might be caused by failure of a safety or relief valve or rupture of a main steam pipe. A Main Steam Line Break (MSLB) Transient, which is a Class IV event
5.13 and the limiting transient in this category, was chosen for analysis.
The increased steam flow resulting from this accident causes a reduction in primary coolant system temperatures and pressures. The reduced temperature causes a positive reactivity insertion (assuming a negative moderator temperature coefficient).
This insertion, coupled with the assumption that the most reactive rod cluster control assembly (RCCA) sticks in its fully withdrawn position, increases the possibility that the reactor will return to a critical condition and resume power generation following i
reactor trip. This is a potential problem because of the high power peaking factors associated with the stuck RCCA assumption. The core power is limited by the negative Doppler and moderator reactivity effects for which conservative values are assumed in the analysis. The core is ultimately returned to a suberitical condition by boric acid dalivered by.the safety injection system. A more detailed discussion of the transient i
and the various mitigating systems is provided in the units' FSAR's.
3,4 The Several different MSLB transients are discussed in the FSAR limiting MSLB case, which was analyzed with RETRAN for comparison to the Surry FSAR analysis, consisted of a break adjacent to a steam generator outlet nozzle with
{
continued availability of offsite power. The MSLB was analyzed with the Two Loop Model (See Section 3) which calculates both the primary and secondary system responses, the reactivity effects of safety injection and the core power response I
following return to criticality.
A summary of important analysis assumptions, which correspond to the assumptions made in the FSAR, is given in the Appendix (Item 7a). Specific analytical values used for the analysis are also shown in the Appendix. Representative results from the FSAR analysis are presented and compared to vendor results in Figures 5.46 to 5.49, for steam flow, pressurizer pressure, core reactivity and core average heat flux, respectively. The slight differences in the shapes of the core heat flux response are i
believed' to be related to differences in the treatment of core boron concentration buildup following safety injection.
5.14
- The comparisons indicate that Vepco's RETRAN Models provide an appro-
[
priate basis for calculating the system transient portion of the Main Steam Line Break l
' analysis.
5.2.3.5 Accidental Depressurization of the Secondary System / Main Steam Line Break Transient - Current Analysis The Main Steam Line Break Transient has been reanalyzed for several Vepco reload e' ores.
The reanalyses have been necessary to confirm the continued i
seceptability of the MSLB transient results for variations in the reload core designs.
[
For example, a recent licensing update reanalysis of the system response was performed for the Surry Unit 1, Cycle 4 reload core (see Reference 13). The basic analytical
[
essumptions and parameter values for this reanalysis are shown in the Appendix, (Item 7b.) The comparative results of the two analyses are summarized in Table 5.1. As can j
be seen the results for temperature, pressure and core heat flux for the two analyses are quite similar.
The dynamic response to the MSLB reload reanalysis is shown in Figures t
5.50 - 5.52.
Comparison to the FSAR results (Figures 5.46 - 5.49) shows that the i-general characteristics of the transient responses are the same for the two cases. The vendor results for the analysis are considered proprietary and are omitted.
5.2.4 General Conclusions - Licensing Transient Anclyses The analysis results shown in Figures 5.1 - 5.52 show that Vepco's analysis approach yields results which are comparable to those obtained by our NSSS vendor for I
previous licensing submittals. The similarities hold for a broad variety of transients of i
varying levels of severity and result in identical ecnclusions regarding core and system safety. These comparisons illustrate Vepco's general capability to perform analyses of Condition I-III transients, and the system transient aspects of certain Condition IV transients.
i
m TABLE 5.1 LIMITING PREDICTED RESULTS MAIN STEAM LINE BREAK TRANSIENT SURRY 1, CYCLE 4, REANALYSIS Parameter Peak Value Licensing Analysis RETRAN Core heat flur, % of rated.
28.6 25.8 Reactor inlet temperature (failed loop), F 373 373 Reactor inlet temperat'ure (intact loop), F 502 504 Pressurizer Pressure, Psia 1167 1255
Figure 5.27 PRESSURIZER PRESSURE CHANGE LOSS OF LOAD TRANSIENT BOL - FSAR ANALYSIS RET E FSAR
}
400 3
u b_
320 2
~
I
\\
5 240
~
/
. _ _ _\\
!m g
\\
m
/ _ _ _ _..
.=
80
\\ _.. ___ _.
i
\\
\\,
g 0
5 10 15 20 25 Time, Seconds
1 Figure 5.28 NUCLEAR POWER LOSS OF LOAD TRANSIENT BOL - FSAR ANALYSIS RETRAN 1.2
=. _ -
1.0
-- \\ -- --
y
._.._ _ \\.
g
\\
{
0.8
-- {
p.
S 0.6 M
J.
l j
i 0.4
}.
g Q
._\\
N 0.2
%'%C 1 -.
0.0 0
5 10 15 20 25 Time, Seconds
Figure 5.29 PRESSURIZER WATER VOLUME CHANGE LOSS OF LOAD TRANSIENT
)
BOL - FSAR ANALYSIS l
RETRAN I

~~FS AR 120 r,~~

g E
100 d
o l
\\.
=
g o
\\
3
...__ N +.
60 u
\\__
=m 40 20 i
0 0
5 10 15 20 25 Time, Seconds
i I
l l
l Figure 5.30 COOLANTINLETTEMhERATURECHANGE LOSS OF LOAD TRANSIENT BOL - FSAR ANALYSIS i
35 RETRX;

~~FSAR 30~~

^
p 3
g
/ _..._s g
5 N
25 l
0
/.
N --._
\\
u
/
o e
/~
n g
/
i c.
20 j
e "c.
15 l
5
..__/
.]. _
1
/
_/
l
_ _ ____ _ _7 j
.j _.-
/
0 i
0 5
10 15 20 25 t
Time, Seconds l
Figure 5.31 DNB RATIO LOSS OF LOAD TRANSIENT BOL - FSAR ANALYSIS 4.0
/
3 -
-~~_~.~.
/
.-.-------f_...
/
yepeo
{
j...-. - -..

~~FSAR 3.0~~

__f
./
/
J.
/ _. _ _ _. _. _ _ _.. _. _ _. _
2.0
,?
p..___
._._.g-...
1.0 0
5 10 15 20 25 i
i Time, Seconds
i l
l l
Figure 5.32 l
PRESSURIZER PRESSURE CHANGE LOSS OF LOAD TRANSIENT i
EOL-FSAR ANALYSIS
~
350 PITPE - ---

~~FSAR 2~~

300
\\. _.__
q 250 a
5; Y
0 h
\\
200 u
I~
-~
\\
4 l
z
,8 150
/
k
~
I I
k
\\_
100
\\
~ ~ ~ ~ ~
\\
/
so _ __ /.__ _
~\\ i 7
g 0
5 10 15 20 za Time, Seconds
i l
1 Figure 5.33 NUCLEAR POWER LOSS OF LOAD TRANSIENT EOL - FSAR ANALYSIS RETPE
-FSAR 4
1.2 j
?
y x
1.0 C
g e"
.i.. _
g g
0.8 w
g 3
e c
0.6 u
g z
0.4 0.2
\\g l
i l
W
=-
o,o l
0 5
10 15 20 25 1
Time, Seconds l
Figure 5.34 PRESSURIZER WATER VOLUME CHANGE LOSS OF LOAD TRANSIENT EOL-FSAR ANALYSIS l
m U
RETRA" - - -
._u
-FSAR l
E c
g q
100 p
3 f
u N
80 N
s N. _ _._ _
i a
\\
60 i
N g
40 g
20 8
f i
i 0
l 0
5 10 15 20 25 Time, Seconds
t d
I I
I Figure 5.35 l
COOLANT INLET TEMPERATURE CHANGE LOSS OF LOAD TRANSIENT EOL-FSAR ANALYSIS i
l 35 RETPAN
~ - - -
FSAR
~~~~~
j ci 30 g
g 3
25
/
2
..._N. g i
__. /
ey g
20
-~
g
~
I c
C q
15 j
j l
10
/
-7 5
- - - - - -- -(
.. - _ _,/..
_. ~. -.
/
e s
i 0
5 10 15 20 25 Time, Seconds I
Figure 5.36 DNB' RATIO LOSS OF LOAD TRANSIENT EOL FSAR ANALYSIS 4.0 1
p-~~'."""
1
,A,
,/
/
3.0
/
. /.
VEPCO t
./

~~FSAR g~~

5
..___..__7 y-
_j l
__ /.
/
2.0 j _ _ _
/..
- ~ ~ ~~~ 'y 1.0 0
5 10 15 20 25 Time, Seconds i
l
\\
i Figure 5.37 PRESSURIZER PRESSURE LOSS OF LOAD TRANSIENT POSITIVE MODERATOR COEFFICIENT REANALYSIS 5
. ~...
_ _.. _ _ _ _. _.._ z_
RETRAN 1

~~VENPOP j~~

.5 2600 r.
C.
g
'n mz 2500
____._..J
_s 8
w p.
.\\.
\\
i u
/
.y g
,/
. g _.., _ _. _ _.
w g 2400 m
i j
/
.1
_ _ _ _ -.... _..._ _ _. A - -. _..-.
s
'\\. -.. _.... - -
/
2300
\\-
g l
^
/
l 2200 0
5 10 15 20 25 Time, Seconds j
l l
...~
Figure 5.38 NUCLEAR POWER LOSS OF LOAD TRANSIENT POSITIVE MODERATOR COEFFICIENT REANALYSIS
_.._____._...______.._._.__________J_
l EETPR:
1.2

~~VENDOR j~~

1.0
.._____._ _ _.. T..____..__.
e i
n y
~~i... _. _.~~
\\
cc t 0.8
-s- - -
g n
s e
b C
I g 0.6 c
g g
a 0.4 i
t s
g _ _ __._
0.2 m...__
4
~~..
_*+
e,,
0.0
~
"-- ~~' ~
0 5
10 15 20 25 Time, Seconds
j l
8 1
Figure 5.39 AVERAGE COOLANT TEMPERATURE LOSS OF LOAD TRANSIENT POSITIVE MODERATOR COEFFICIENT REANALYSIS RETRAN ___. -
595

~~VENDOR 590 w~~

e g
u 3
g E.585
-- - - --r E
H a.4 N
s
/
\\
N D
/
s g
.f.._._,
g e-*
N C
g O
580 y
's C
o g
s 57
~-
-~
570 0
5 10 l'5 20 25 Time, Seconds I
i f
i
Figure 5.40 DNB RATIO LOSS OF LOAD TRANSIENT POSITIVE MODERATOR COEFFICIENT ASSUMPTION REANALYSIS
...~.----_
m.e a
s-w
-w--
e
+.w.
en-M
-mm
-e
-e-em VEPCO
.-e.w,.e--.
.ew..
h i
3.0 j
g E
4 l
- -- - - - +
e-=w-W
-g-
-+4se+-+e-em e
-w 4
e
-m 2.0 f
.- +
+.*g mm m e-Mhw.e_--
am h
.em am-
- eur ease me*-
M 4E4'28 1.0 0.0 2.5 5.0 7.5 10.0 12.5 Tirae, Seconds l
l s
Fi,gure 5.41 FEEDGTER TEMPEPATURE GR4CE EXCESSIT/E HEAT RDOVAL DUE TO FEEDWATER SYS'IEM ELEWCTICN TPRISID71 FSAR ANALYSIS l_
, d..
I RE1WN I
l l
I i
E--
FSAR i
T~~
i I.
i!... j.
._{ _..a..
i s
.}..
i I
i w
]
r i
c' O
m.
I e.c e
i-i 3
u
. _.!. _j.
3
__.p.______.._.__'
6.e 3
-5 o
g e
o h
l Q
%g
-10 o
-15 L
\\\\_.-
V
-20 0
200 400 600 800 1000 i
Time, Seconds 1
-~r-p,
-e..--
Figure 5.42 NUCLEAR POWER EXCESSIVE HEAT REMOVAL DUE TO FEELWATER SYSTEM MALFUNCTION TRANSIENT FCAR ANALYSIS 1
RETRAN

~~FSAR g~~

y j
1.1 n -. -. _. - - - -
h t
aa---
/
g 1.0 1
g Z
h 0.9 0
200 400 600 800 1000 Time, Seconds I
(
i Figure 5.43 CHANGE IN AVERACE COOLM;T TDtPERATURE EXCESSIVE HEAT RDIOVAL DUE TO FEEDWATER SYSTEM MALFUNCTION TRANSIENT FSAR ANALYSTS l
RETPA'i ~

~~FS AR g-o y~~

aw B
i--
0 o
8 g
5 c
-10 g
N s
j g
-20 I
-30 l
-40 0
200 400 600 800 1000 Time, Seconds
Figure 3.44 PRESSURIZER PRESSURE CHANGE EXCESSIVE HEAT REMOVAL DUE TO FEEDWATER SYSTEM MALFUNCTION TRANSIENT FSAR ANALYSIS t
RETPE

~~FSAR~~

~ - ' ~
g 6
g g
g 0
u g
g
-20
/- --
g
/
uw
/
-40
/
\\
_/
^3
-60
\\__
s
/__ __ _
g _ _ __j.
-80 0
200 400 600 800 1000 Time, Seconds i
Figure 5.45 D E RATIO EXCESSIVE HEAT RE20 VAL DJE TO N.TER SYSTEM bMELUCTICN TFANSIENT FSAR RFJf1 SIS FSAR l
_7 2.5 l
~.
2.3 g
2.1 1
f 1.9
/
/ s, -.. -
.7_.--_--
- - / ---
/
f
)
~~
1.7 r
1.5 0
200 400 600 000 1000 Time, Seconds
Figure 5.46 BREAK FLOW RATE MAIN STEAM LINE BREAK FSAR ANALYSIS pg7g.
12000-

~~FSAR g~~

j 10000-3
{
3
[
8000-
.___ _ _._ _ l E
\\..._.._..
e j.
m 6000-
)
-\\_
4000-
- _ \\. _...
A
\\..
2000-
\\
~~
~ -.
, _Q_
0 10 20 30 40 50 60 Time, Seconds
g Figure 5.47 PRESSUNIZER PRESSURE MAIN STEAM LINE BREAK FSAR ANALYSIS i
3000
'- - - - - - ~ "
PJ2TPE
=FSAR l
2500-
- ~ - - - - -
e
~~c e~~
N
- - ~ ~ ~ ~ - ' -
E 2000 -
N
_g_
i c.
g l
t
\\
g g
5 1500-
\\
e
.. - _ =. _. - _..... _...
L 1000.
..\\
~m
. - + -.
e
-.e.-.ew..W.
500 1
0 10 20 30 40 50 60 l
l
~~ime. Seconds~~
4 i
l l
Figure 5.48 l
TOTAL REACTIVITY MAIN STEAM LINE BREAK FSAR ANALYSIS FITRAN I
_____rs An
.02 -
. 01.
e
/
&~
l-T~
0
/
D
/
U e
I
. 01 -
cc
/
e l
u H
/
.02 i
O 10 20 30 40 50 60 Time, Seconds
Figure 5.49 CORE HEAT FLUX MAIN STEAM LINE BREAK FSAR ANALYSIS
.2 PETPE
-FSAP 12-A 10
\\.
5x
./
's
?
l
--N -
8-r w
i o
i I.
5 6
--.).
y l
2 f_
e
/
0 4-l
/
/
/
2-
/-
/ _.._.
/
/
~
0
/
0 10 20 30 40 50 60 Time, Seconds
--~
Figure 5.50 PRESSURIZE. PRESSURE MAIN STEAM LINE BREAK TRANSIENT SURRY 1, CYCLE 4 REANALYSIS
.___...2 RETRAN - ---
_g_
u.
mee.
7.*-
2500.
o' b
u t
2000 -
I; E,
s 1500 1000 -
- - ~
~
- ~ - -
~
500 0
10 20 30 40 50 Time, Seconds
Figure 5.51 TOTAL REACTIVITY MAIN STEAM LINE BREAK TRANSIENT SURRY 1. CYCLE 4 REANALYSIS
.03 RETRAN
.02 g
4
.01
,g y
C~~~~~~~~~~
~~
0 g
.01
-0.3 0
10 20 30 40 50 60 Time, Seconds
Figure 5.52 CORE HEAT FLUX MAIN STEAM LINE BREAK TRANSIENT SURRY 1, CYCLE 4 REANALYSIS 30 -
yy. -..
e++..e,.-
mc g
20.
u g
u g
=
y U
10-i 0
0 10 20 30 40 50 60 Time, Seconds
5.15 5.3 Verification Against Operational Data 5.3.1 Introduction The purpose of comparing RETRAN predictions to plant operational data is to demonstrate that the code, coupled with appropriate plant models and best estimate input values, provides physically realistic predictions of integrated system response to various perturbations. Vepco RETRAN comparisons are for the pump coastdown tests performed at both the Surry and North Anna plants and a plant cooldown event which occurred at North Anna Unit 1.
5.3.2 Pump Coastdown Tests Pump coastdown tests of various configurations (i.e., coastdown of a single pump, two pumps, three pumps, etc.) are performed as part of the initial startup test sequence for new nuclear units. The sections below discuss RETRAN comparisons for a single pump and a simultaneous three pump coastdown for Surry Unit No. I and for a simultaneous three pump coastdown performed on North Anna Unit No.1. Both single loop and two loop RETR AN models were used for the comparisons, as discussed below.
5.3.2.1 Surry Pump Coastdowns Pump coastdown tests were performed at the Surry Power Station Unit No.
1 in January 1975.
The tests were performed with the reactor at hot shutdown conditions with all Rod Cluster Control Assemblies (RCCA) fully inserted. The test results of reactor system flow versus time have been compared with the flow coastdown associated with the Loss of Flow transients reported in the Surry FSAR and with RETRAN analytical predictions using both the Single Loop and Two Loop Surry Models described in Section 3.
The comparison for the simtItaneous three pump coastdown is shown in Figure 5.53. The RETRAN code predicts a flow coastdown curve which lies between the 3
FSAR prediction and the test dats. Results for this case (3 pump coastdown) were generated with both the Single Loop and Two Loop Surry RETRAN Models. The coastdown curves generated by the two models were essentially identical.
5.16 Figure 5.54 compares analytical predictions made with the Surry Two Loop.
Model with test data for a single pump (two pumps remaining at full speed) coastdown.
The data are presented in terms of loop flows.
As may be seen, the RETRAN predictions are in close agreement with the data. The same data are presented in terms of core flow fraction in Figure 5.55 to allow an additional comparison to be made, i.e.,
to the FSAR accident analysis results.
As with the three-pump coastdown, the RETRAN curve lies between the FSAR and the data in the region of interest (minimum DNBR for the single pump loss of flow accident occurs at #3 seconds - see Figure 5.26).
It should be noted that although the data indicate a slightly more rapid flow coastdown than either the FSAR or the RETRAN predictions, use of either analytical curve in combination with the conservative FSAR assumptions concerning trip delay times has been shown to provide conservative results for the postulated loss of flow accident.
5.3.2.2 North Anna Pump Coastdown The three pump coastdown test was performed on North Anna Unit No.1 in April, 1978.
As with the Surry Unit No.1 test, hot shutdown conditions were maintained. The reactor coolant flow versus time was measured out to 10 seconds 4
following the loss of pump power. The comparison to the FSAR flow coastdown predictions and to the RETRAN analytical predictions is shown in Figure 5.56.
The RETRAN results agree quite well with the FSAR, particulary over the first four seconds of the transient, which in a complete loss of flow accident is the most limiting period from the standpoint of DNB. Note that both the FSAR and RETRAN predict a slightly slower coastdown than the data indicates over this same period. As discussed above, slight deviations are evaluated at the time of the test to ensure the overall conservatism of the FSAR analyses, in summary, the RETRAN pump coastdown calculations performed with the Surry One and Two Loop Models and the North Anna One Loop model have been shown to give results which agree well with the measured data.
5.17 5.3.3 North Anna Cooldown and Safety Injection Transient An analysis was performed to simulate the unplanned cooldown event which occurred at North Anna Unit 1 on September 25,1979.
The following sections provide
1) a brief description of the event; 2) a description of the RETRAN model used for the analytical simulation; 3) comparisons of RETRAN results with plant data taken at the time of the event; and 4) conclusions regarding the analysis and data comparisons.
The North Anna cooldown event resulted from a turbine trip and subsequent reactor trip on high feedwater heater condensate level. The high level signal was the result of tube leakage inside the heater drain cooler. Following the trip the eight condenser dump valves tripped fully open to supplement the reactor trip in providing load rejection capability. As the plant began to cool down seven of the eight dump valves modulated closed as designed. The remaining valve stuck in its fully open position. This resulted in additional cooldown beyond the no-load temperature, causing a depressurization of thc reactor coolant system and initiation of Safety Injection on low pressurizer pressure. Following Safety Injection, the operator tripped the reactor cootant pumps in accordance with procedures and the system rapidly repressurized to the normal pressure range. One of the two high head safety injection pumps was tripped; *.he RCS coatinued to repressurize at a slower rate until one of two pressurizer Power Operated Relief Valves (PORV's) opened on a high pressure signal. This valve then cycled to maintain RCS pressure at the relief setpoint.
Normal pressure was restored by a combination of operator actions, including initiation of auxiliary spray, realignment of the charging pumps to the normal charging path, throttling the charging flow and reestablishment of letdown flow.
The RETRAN model used to simulate the cooldown event is a 20-volume, single Icop representation of the North Anna Reactor Coolant System, steam generators and associated control systems. The general description of Vepco's Single Loop Models, given in Section 3, is also applicable to this model. Additional features included in this
F 5.18 model to provide a best estimate analysis capability include the following:
1)
Representation of the automatic steam dump control system.
2)
Simplified representation of the feedwater control (steam generator level) system.
3)
Representation of the High Head Safety Injection system 4)
Automatic charging flow (pressurizer level) control in combination with RCS letdown.
5)
Representation of the following operator actions as boundary conditions:
-Manual tripping of the primary coolant pumps shortly after Safety Injection
-Manual tripping of one charging pump after Safety Injection had restored pressurizer pressure and level to their normal values
-Manual tripping of the Main Steam Isolation Valves to terminate the steam release shortly after Safety Injection initiation
-Manual termination of auxiliary feedwater flow.
The following discussion provides a comparison of analytical results to plant data obtained at the time of the cooldown. Plant data sources include alarm typewriter printout and control room strip chart recordings. The resolution of the alarm printout, which is the source of most of the data, is plus or minus thirty seconds.
Figure 5.57 shows the depressurization of the main steam system. The alarm typewriter data are representative of all three loops. Examination of the data indicated that the depressurization took place in a symmetric manner. Note from the figure the pronounced impact of operator intervention on the pressure response.
Figures 5.58 and 5.59 compare calculated and observed cold and hot leg temperatures, respectively. The cold leg temperature data in Figure 5.58 from 0 to 300 seconds are based on alarm typewriter printout of narrow range Teold. The data points
5.19 represented by triangles are Teold values inferred from clarm typewriter steam pressure data. These points were derived by table lookup of the saturation temperature of the steam system and correction by the calculated primary to secondary temperature difference.
The dashed line represents control room strip chart data. As can be seen, the general agreement of the model with the data is good. The predicted reactor vessel A T under natural circulation conditions is slightly lower than the measured value.
Figure 5.60 shows the pressurizer pressure response. The calculated initial depressurization and repressurization following Safety injection initiation at 300 seconds show excellent agreement. This good agreement provides further qualification for the RETRAN nonequilibrium pressurizer model.
Figure 5.61 shows the pressurizer level response. Both the observed data and the model indicate that pressurizer level indication was lost for a brief portion of the transient. The model predicted a slightly lower drain rate during cooldown than was observed. This may reflect a difference in the assumed initial pressurizer mixture quality and the actual plant condition. The general agreement is still quite good over the first 10 minutes of the transient. The underprediction at 1400 seconds is possibly related to the integral effects of RETRAN's underprediction of the safety injection flow rate at elevated system pressures.
5.3.4 General Conclusions-Best Estimate Transient Analyses The comparisons of best estimate RETRAN predictions to plant data presented in sections 5.3.1-5.3.3 (Figures 5.53-5.61) are indicative of Vepco's best estimate analytical capabilities; the favorable results shown here provide a sound basis for applying this capability to general plant operational support.
1
Figure 5.53 FLOW COASTDOWN OPERATIONAL TEST AT HOT ZERO POWER SURRY THREE - PUMP COASTDOW'N RETRAN Vender -
p
_. _ _ _. - _ ___ o o o Data 3
o 4
~
3 100 w
3 80 N..
\\
,,E, g
a N _ _ _ __
3
.N
,c g%
60
_ _. _*\\
40 20
~~
i 2
4 6
8 10 Time, Seconds
Figure 5.54 FLOW COASTDOWN OPERATIONAL TEST AT HOT ZERO POWER SURRY ONE PDiP COASTDOWN i
i i
RETRNi-
..__.L_-
.I i
j-
}
-i l'4 iLoop6 B +:C-
_. ~
-- i e
.2 100
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^
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4 t
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i Fi
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1
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20 i
i t
j----~r---------
i t
i o
1 k
.L f
. _. _.. I i
i i
0 2
4 6
8 10 Tirne, Seconds
Figure 5.55 CORE FLOW COASTDOWN OPERATIONAL TEST AT HOT ZERO P0bTR SURRY ONE-PUMP COASTDOWN n
g g
RETRAN O
e 3
FSAR g
O
.f
==_-
g 90 s.
s G
g_
e g_%g 80
..--em..
.........-...._J-70
.,.ee-e.
60 0
2 4
6 8
10 Time, Seconds i
Figure 5.56 FLOW COASTDOWN OPERATIONAL TEST AT HOT ZERO POWER NORTH ANNA THREE - PUMP COASTDOWN RETRAN o o o DATA

~~FS AR C~~

~
g c
I
~~1. 0 -~~
c s
o 3
h s
t 0.8 Ew
\\
0.6 w
-o 0
- -e 0.4 0.2 0.0 0
2 4
6 8
10 Time. Seconds
Figure 5.57 STEAM PRESSURE NORTH ANNA COOLDOWN EVENT o
PODEL PLN T PATA 1100 -
-- - - - - - - ~ - ~ ~ - ~ ~ - - ~
~ '-~-
. -..- 3. -. -._--
1000-
~---
I
..g
- -y
}
900 4 - ------ - - - - - - ~ ~ - - TURBINE TRIP /REACTOP TRIP 2 _ MAT *: 9 TEA" T90LATION VALVE TRIPPED CLOSED 800. _.
- - - AUyILIARY FEEPUATER TEEPINATION c
g E
am 700
' - - - ~ - - - - - -
~~~
___ y y
~ 00 --
6 500 0
400 800 1200 1600 2000 Tirne, Seconds
l l
Figure 5.58 COLD LEG TEMPERATURE r
NORTH ANNA C00LDOWN EVENT MODEL PLANT DATA (N.R.TCOLD) 570-
~ ~ - -. _ _
A INFERRED FROM STEAM PRESSURE DATA

~~W.R.TCOLD w~~

C 550-8
=
\\
530-
\\
e ec 3
\\
g 510 l
_g N \\
490.
__g g.
.. j T >..--..
~A
/
g g, p. _ _
A.
470
-A-450 0
400 800 1200 1600 2000 Time, Seconds
Figure 5.59 HOT LEG TDIPERATURE NORTH ANNA C00LDOWN EVENT MODEL 600-
-WIDE RANGE TH0T l
~ ~~~~
CALCULATED FROM NARROW
- ~ ~
RANGE TAVG AND TCOLD 580.
l l
l I
c J
560-
,I r
\\
{
(
g
[
540.
F f
a f
520-fw f
y j_.
500-w.
480 0
400 800 1200 1600 2000 Time, Seconds
Figure 5.60 PRESSURIZER PRESSURE NORTH ANNA COOLDOWN EVENT
. ~. -
2400
.a 2300 2200-N-
H 8
4 eI 2100-
- ~ ~ ~ ~ - - ~
y E
2000- - - - -
c.
/
1900-
^
- - ~ - ~
1800.
1700 0
200 400 600 800 1000 1200 Time After Turbine Trip. Seconds
Figure 5.61 PRESSURIZER LEVEL NORTH ANNA C00LDOWN EVENT FDDEL 70.
PIR7f DATA
.M 60 -
]
w w
50 -
-~~
~ - -
e et l
40 -
a G
E 30 -
t C-e--
20 -
10 ~
~ ~ ~ ~ ~ - ~ ~
m 0
0 400 800 1200 1600 2000 Time After Turbine Trip, Seconds
SECTION 6 - CONCLUSIONS The Virginia Electric and Power Company (Vepco) has developed the capability to perform system transient analyses using the RETRAN computer code. The general code features and the types of models developed for analysis of the Surry and North Anna I
Units 1 and 2 have been discussed. The adequacy of these models and the associated i
accident analysis methodolcgy has been demonstrated by comparison of selected analytical results to vendor calculations and to plant data. The overall goed agreement realized in these comparisons demonstrates that these models and methods can be used for operational and licensing support of Vepco's nuclear plants.
SECTION 7 - REFERENCES 1.
VEP-FRD-33, "Vepco Reactor Core Thermal-Hydraulic Analysis Using the COBR A IIC/MIT Compute Code," Virginia Electric and Power Company, August 1979.
2.
Electric Power Research Institute Report EPRI CCM-5, "RETRAN - A Program for One-Dimensional Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems," Energy Incorporated, December 1978.
3.
Surry Power Station Units 1 and 2, Final Safety Analysis Report, Virginia Electric and Power Company, December 1,1969.
4.
North Anna Power Station Units 1 and 2, Final Safety Analysis Report, Virginia Electric and Power Company, January 3,1973 as amended.
5.
NUREG-75/056, "WREM:
Water Reactor Evaluation Model," Revision 1, May 1975.
6.
WCAP-7978, "LOFTRAN Code Description," Rev.
1, Westinghouse Electric Corporation, January 1977.
7.
WCAP-7909, " MARVEL - A Digital Computer Code for Transient Analysis of a Multiloop PWR System," Westinghouse Electric Corporation, October 1972.
8.
WCAP-9272, " Westinghouse Reload Safety Evaluation Methodology," Westinghouse Electric Corporation, March 1978.
9.
Letter from C. M. Stallings (Vepco) to E. G. Case (NRC), Serial No. 403, September 9,1977.
10.
Letter from C. M. Stallings (Vepco) to E. G. Case (NRC), Serial No. 344, August 9, 1977.
11.
WCAP-7769, " Overpressure Protection for Westinghouse Pressurized Water Reactors," Westinghouse Electric Corporation, June,1972.
12.
Letter from C. M. Stallings (Vepco) to K. R. Goller (NRC), Serial No. 553, June 5, 1975.
13.
Letter from C. M. Stallings (Vepco) to B. C. Rusche (NRC), Serial No. 256, September 27, 1976.
14.
Letter from C. M. Stallings (Vepco) to J. P. O'Reilly (NRC), Serial No. 829, transmitting LER 79-128/01T-0, October 9,1979.
I A PPEN DIX
~~SUMMARY~~
OF IMPORTANT ASSUMPTIONS USED IN TH ANSIENT AN ALYSES DISCUSSED IN SECTION 5 Transient Specifie input System Model Key System Perffmance Type of Analysis Description initial Conditions Reactivity Parameters Assumpti1ns
~~1. Uncontrolled Rod Withdrawal from Suberatical a) FSAR Ana!ysis Surry one Loop Core power = 10'I3xrated aMOD=+10 pcm/ F Pressure = 22'0 psia cDOP:-l.75pem'/ F d~~
T-intet = 550 F (c/550 F)
Delayed neutron No credit taken for fraction = 0.0072 I) Source range high flux trip Rent isty insertion 2)Intermediat'e range high flux Rate =60 pcm/see trip
~~3) Intermediate range control rod stop Trip Reactivity: Fig. A.I Source of protection:~~
curve (s), total = 2.8% 8 K/K Low power range high neutron flux trip b) Current Analysis Surry One Loop Same as case (a)
Same as case (a), except Other assumptions same as Hractivity insertion Rate case (a) 75 pcm/see Trip reactivity: Fig. A.t curve (b), total:4.0%8 K/K (1) Trip setpoints and delay times assumed are consistent with Table 4.8
-5
'l pcm=1.0 m 10 A K/K I
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APPEN DIX (Continued)
SUMM ARY OF IMPOhTANT ASSUMPTIONNED IN TR AN$ifRT AN AL.YSES DISCUSSED IN SECTION 5 Trarmient Specific Input System Model Key System Performance Type of Analysis Description initial Condit sons Hemetivity Parameters Assumptions
~~2. Uncontrolled Hod With-drawal from Power l~~
~~b) Current Analysss Surry One Loop Core power = I.92 x rated aMODa*l.0 pem/ F No creoit t=4en fon (Modafaed to reflect~~
~~1) liigh neutron flus rod stop steam generator Pressure = 2220 psis Doppler power coefficient~~
~~2) High Overtemperature~~
~~1) From 102%~~
~~tube plugging)~~
-6.0 pem/% at 100% power aT rod stop power T-inlet = 543.4 F 3)lingh overpower AT rod stop or trap HCS Flow s 90% of full Hemetivity insertson power thermal rate versed design Trip Remetivity: Figure A.!
Source of protection:
curve (m), total = 2.8%
liigh power range high AK/K neutron flux trip or liigh overtemperature A T trip
~~2) From 62%~~
Core power = 0.62xrmted Trip Hemetivitys power Fig. 4.1 curve (m), total:2.8% A K/K Pressure = 2220 psis aMOD=el.0pem/ F Assumptions same as T-inlet = 550.3 F uDOP =-7.3 pcm/%
mt 62% power HCS Flow = 90% of full power thermal design Hemetivity insertson rate versed A T trip equation used (includes errors):
A T(Setpoant) = (1.166.0095 (l
30s ) (T" 574.4) +.0005 (P-2250)) X AT Hated 1*4s
A PPEN DIX (Continued)
SUMM ARY OF IMPORTANT ASSUMPTIONS USED IN TR ANS!E"NT AN ALYSES DISCUSSED IN SECTION 5 Treenient Specific Input System Model Key System Performance Type of Analysis Description initial Conditions Hemetivity Parameters Ahsumptions
~~3. Complete Loss of Forced Remetor Coolant Flow a) ESAR Analysis Surry One Loop Core power = 1.02x rated 410D=0 Source of protection:~~
Pressure = 2220 psim aDOPPLEH =-l.6 pcm/"F T-inlet = $47 F Trip remetivity:
Fig. A.I. Curve (m)
Total = 2.8% AK/g b) Current Analysis Surry One Loop Core Power = 1.02x rated u MOD =+3.0 pcm/~F Source of protection:
(Modified to Low HC Pump voltage reflect steam Pressure a 2220 psia uDOPPLEH=-1.6 pcm/ F generator tube Conservative (low)anatial plugging)
T-intet = 547.l*F flor was assumed Trap remetivity:
Fig. A.I Curve (b)
Total = 4.0% AK/K
~~4. Partial Loss of Surry Two Loop Core power = 1.02xrmted uMOD = 0.0 Source of protection:~~
Forced Hemetor Pressure = 222gpsia uDOPPLER=-1.6 pcm/ F Low RC loop flow rate Coolant Flow T-inlet = 547.1 F Trip Hemetsvity:
Fig. A. l.curvela)
Total =2.8% oK/K l
e l
A PPENillX (Contmiaw1)
SUMM AR Y OF IMPOHTA NT ASSUMPTIONS USElklN TH A NSIENT AN AI.YSES DISPUSSED IN set' TION 5 Tranwent Specific Input Key Systein Performance System Model Type of Analyus Oncription Imtial Corulitice.
Hemetivity Parameters Assumptiorm
~~5. Im of Enternal Electrical Load a) FSAR Analysis Surry One Loup Core power = 1.02mrated lieginning of Life No credit taken form~~
~~1) Pressurizer spray Premure = 222g p>ia T-talet = 547.2 P aMOD - 0.0~~
~~2) Pressurizer power operated relief valves uDOPPLER a-l.6 pcm/"F~~
~~3) Atmospheric steam dump valves~~
~~4) Atmospheric steam Delayed neutron rehef valves fractions.0072~~
~~5) Direct reactor trip resulting from a End of Life:~~
tubinnenerator trip uMODx-35pcm/ F Source of protection:
liigh pressurizer pressure nDOPPLER=-l.6 pcm/*F trip Delayed neutron fractions.0048 l
Trap reactivity:
Fig. A.1, curve (s)
Total = 2.5% A K/K b) Current Analysts Surry One Loup Core power = 1.02xrmted oMODs+3.0pem! F Key assumptions are the Pressure a 2220Imia eDOPPLERa-l.6 pcm/ F T-sniet = 547.2 F Delayed neutron fraetson a.0072 Trip reactivaty Fig.A.I, curve (s)
Total = 2.8't. A E/K
A PP EN IH X (Cont mued)
SUMM ARY OF IMPOHTANT ASSUMPTIONS USED INTliANSIENT AN AR.YSES DISCUSSED IN SECTION 5 Transient Specahe Input System Model Key System Performance Type of Analyus Description Initsal Conditions Reactivity Parameters Assumptiorni
~~6. Excessa ve if est Surry One Loop Core power = 1.02arated uMOD = 0.0 Reactor masumed to Removal Due to Pressure = 2220 psia uDupplers-1.0pem/ F be in manual control (Tave l~~
Feedwater System T-inlet = 547.2 F control inactive)
Malfunction Trip Remetivity:
Source of protection:
Fig. A.1, curve (a) none required Total = 2.8% A K/K
~~7. Maan Steam Line Dreak~~
-8 m) FSAR Analysis Surry Two Loop Core power = 4al0 x aMOD=-25.4 pcm/ F Technical Specsfacetions rated C)S50 F,-13.8 pcm/ F value for initial shutdown Pressure = 225gpsis g)300*F remetsvity margin assumen T-sniet = 549.7 F the lughest worth control rod assembly stuck in aDOPPLER(Zero power) its fully withdrawn position
=-l.6pem/*F Safety en)=ction espability based on failure of one Total power defect high-head safety injection at 30% power =.0135 A K pump Differentist boron No credit is taken for the worth m-10pem/ ppm effect of the maan steam lane check valves in precluding dis-charge of secondary fluid from the intact steam generators prior to main steam isolation valve closure
l A PPEN DIX H'ont muest)
SUMM AHY OF iMPOHTANT ASSUMPTIONS USED IN TH ANS!ENT AN Al.YSES DISCUSSED IN SECTION 5 Transient Specific input System Model Key System Performance Type of Analysis Description Initial Conditions Heactivaty Parameters Assumptions b) Current Analybas Surry Two Loop Core power-4x10'8xrated aMOD:-25.4pem/ F Key performance assumptions o )S50*F, -13.8pem/*F are the same as for the Pressure = 2251 pssa c)300*F FSAR Analysis, above 0
T-intet a 549.7 F oDoppler(Zero power) a-l.Spctn/"F Total power defect at 30'it, power a.0148 A K Differential boron worth s-10pem/ ppm l
l i
l
t I
FIGUPI A.1 TRIP REACTIVITY INSERTION CHARACTERIETICS
. - -. - _ - Surry Analyses Prior to April 1977 __.__._..
g)
Surry Analyses After April 1977 (b)
=--
c North Anna
{f)' __.
~
c c
w f-0.8 t
_ I-h l
(a) l(b)
(c) i 0.6 o
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L..
D 0.4 l!
j j-J_ :_.. f-
/~
0.2
/
Au 0
1 2
3 TD'E, SECONDS
_ _ _ _ _

ML20129E988

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