ML20046B624
| ML20046B624 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 07/16/1993 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20046B622 | List: |
| References | |
| NUDOCS 9308050262 | |
| Download: ML20046B624 (14) | |
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I NUCLEAR REGULATORY COMMISSION
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WASHINGTON, D.C. 20555-0001 j
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION 1
RELATING TO THE TRANSIENT ANALYSIS METHODOLOGY TEXAS UTILITIES ELECTRIC COMPANY COMANCHE PEAK STEAM ELECTRIC STATION UNITS 1 AND g DOCKET NOS 50-445 AND 50-446
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1.0 INTRODUCTION
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In a submittal dated February 28, 1991, the Texas Utilitie's Electric Company (TU Electric or licensee) provided its topical report RXE-91-001, " Transient i
Analysis Methods for Comanche Peak Steam Electric Station.(CPSES) Licensing Applications." This report describes the TU Electric transient analysis d
methodology for application to CPSES Units I and 2.
The TV Electric transient j
analysis methodology is an adaptation of the RETRAN computer code which is used in approved transient methodologies of other utilities.
Information provided in RXE-91-001 was extensively revised and supplemented in j
TU Electric submittals of August 9,1991, and March 22, 1993, which are 1
integral to the description of the methodology.
' l l
2.0 STAFF EVALUATION The staff performed its review of the TU Electric transient analysis methodology with the technical assistance of International Technical Services, Incorporated (ITS).
The evaluation of the methodology and review findings are described in detail in the ITS technical evaluation report (TER), which is attached as part of this report. The review covered areas related to applicability'of the RETRAN models to the CPSES design, qualification of the RETRAN models as used by TV 4
Electric against CPSES plant data, and the adequacy of ada'pted models for TV Electric's intended use in CPSES licensing applications, including the core i
operating limits report (COLR).
The staff concurs with the review and recommended findings of the attached ITS TER with clarifications as discussed in Section 3.0.
j 3.0 REVIEW CLARIFICATIONS TER Section 3.0 identifies that a: comparison of input data describing plant physical geometry versus actual plant physical data was not within the scope of its' review..The licensee uses the same, processes to generate plant physical data input to its transient analysis methodology as it uses for loss of coolant accident (LOCA) analyses, including all data control provisions and 9308050262 930716 PDR ADOCK 05000445 P
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consideration of differences between the two CPSES units.
These processes were found acceptable in the staff safety evaluation report (SER) regarding the TU Electric LOCA evaluation model of April 26, 1993. We find the processes equally applicable and act.eptable for use with the TU Electric transient analysis model.
TER Section 3.1.1.2 finds the use of a three-node steam generator model adequate for use in the transient analysis methodology, and indicates that an alternative four-node steam generator model might be used for certain cases in the future as justified by sensitivity analyses. This statement is clarified in that the staff also finds the four-node model acceptable for use without further justification. The " justification" referred to in the TER statement is more correctly stated as " determination" of licensee preference as determined by sensitivity analyses. Therefore, it is acceptable for the licensee to use either the three-node or the four-node steam generator model in using the TU Electric transient analysis methodology.
TER Section 3.3 and Table IV-1 appended to the TER discuss analysis of locked rotor events. The staff clarifies that discussion as follows.
In using the TU Electric transient analysis methodology as approved herein for evaluation of locked rotor events, the licensee shall include analysis assumptions as performance criteria as specified in NUREG-0800, Review Plans 15.3.3 and 15.3.4, including assumption of loss of offsite power concurrent with event initiation, calculated maintenance of reactor coolant system pressure to less than 100 percent of design pressure, and calculated limitation of doses at the site boundary to a small fraction of 10 CFR Part 100 exposure guidelines.
Changes to these analysis assumptions and criteria were not within the scope i
of this review.
TER Sections 1.0 and 4.0 identify that application of the TU Electric I
transient analysis methodology to evaluation of inadvertent opening of a steam generator safety valve, steam system piping failure, steam generator tube rupture, and power distribution anomaly events are not within the scope of this review. Similarly, TER Sections 3.1.1.3 and 4.0 identify that use of the RETRAN boron transport model was not requested in the licensee's submittal and was not within the scope of the current review. The staff's acceptance of the TV Electric transient analysis methodology does not include these items at i
this time. The licensee will submit these for separate review, when needed.
4.0 CONCLUSION
S Based on our review, as summarized in Sections 2.0 and 3.0, the staff concludes that the TV Electric transient analysis methodology, as described in the topical report RXE-91-001, as supplemented by letters of August 9, 1991 and March 22, 1993, has been acceptably adapted, has been shown to be suitably applicable to the CPSES design, and is properly supported with acceptable programs covering user qualification, methodology maintenance and control, and
' ~.
interfaces with plant operation. The staff finds the TU Electric transient analysis methodology described in RXE-91-001 is acceptable for performing CPSES transient analyses, for reference in CPSES licensing applications, and for incorporation into or reference by CPSES core operating limits report.
Attachment:
Technical Evaluation Report Principal Contributor:
Frank Orr, SRXB/NRR Date: July 16,1993 t
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l TECHNICAL EVALUATION:
THE SYSTEM TRANSIENT METHODOLOGY USING RETRAN-02 RXE-91-001 f_QE THE TEXAS UTILITIES ELECTRIC COMPANY COMANCHE PEAK STEAM ELECTRIC STATIONS w,
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1.0 INTRODUCTION
RXE-91-001, dated December 1988 (Ref.1), was submitted by Texas Utilities Electric Company (TV Electric) for NRC review and approval.
Additional information was submitted on August 9, 1991 (Ref. 2) and March 1993 (Ref. 3).
This topical report, submitted in response to Generic Letter 83-11 (Ref. 4),
documents the development of transient safety analysis methodology based upon use of the RETRAN-02 MOD 005 computer code (Ref. 5) for the Comanche Peak i
Steam Electric Stations (CPSES).
The objectives of the subject topical report are: (1) to qualify RETRAN plant i
models for use in analysis of Comanche Peak Steam Electric Stations in a generic sense; and (2) to provide a description of TU Electric's transient safety analysis methodology using RETRAN.
This review, therefore, focused upon evaluation of acceptability of the RETRAN plant models for transient safety analysis through determination of the adequacy of qualification of such models against CPSES plant data and demonstration of their ability to perform FSAR-type analyses.
- However, review of the analysis methodologies for the Inadvertent Opening of a Steam Generator Safety Valve, the Steam Piping System Failure, the Steam Generator Tube Rupture and the Reactivity and Power Distribution Anomalies are beyond the scope of this review since these are covered under separate topical reports.
For the purpose of this review, it was further assumed that the two units at the CPSES are identical and therefore, a generic model was reviewed for acceptability.
Any plant specific differences will be addressed, if necessary, during the actual application of this methodology.
However, TV Electric stated that these differences are of a minor nature and would not be expected to result in any significant impact on the transient analysis.
2.0
SUMMARY
OF TOPICAL REPORT The topical report RXE-91-001 documents descriptions of TU Electric's RETRAN-
~
02 models for transient safety analysis for Comanche Peak Nuclear Stations (all Westinghouse plants).
TU Electric's objectives in submitting the topical report were: (1) to respond to NRC Generic Letter 83-11 (Ref. 4) by demonstrating its technical competence through qualification of RETRAN models 1
l and benchmark analysis against plant data and current FSAR analyses (Ref. 6);
and (2) to qualify RETRAN CPSES models for use as part of the TV Electric l
methodology to analyze FSAR Chapter 15 non-LOCA events for Comanche Peak Steam Electric Station (CPSES) Units 1 and 2.
l The RETRAN-02 MOD 005 CPSES models developed by TU Electric were described.
Sensitivity studies were performed to determined and qualify an adequate steam generator (SG) nodalization.
The plant model was benchmarked against five sets of the plant data.
The RETRAN model s were also used in
. generally described its transient analysis philosophy and apl5M,,TU Electric demonstration analyses of eight FSAR transients.
In additi.on ches towards performance of these analyses.
TV Electric's transient analysis methodology using the RETRAN computer code for the class of transients described in the submittals was found to be acceptable for application to Comanche Peak analysis.
3.0 EVALUATION TV Electric's methodology to perform Chapter 15 type licensing analysis using the RETRAN-02 M00005 computer code is reviewed below.
In order to develop a generic CPSES evaluation model, TV Electric first developed best-estimate models for benchmark against four plant transient events.
Qualification of l
models with modifications was further conducted against the eight current FSAR transients.
TV Electric's approach to generation of input to the RETRAN-02 code was reviewed for acceptability.
However, no review was conducted of the input data in comparison to the actual physical geometry.
i 3.1 RETRAN Model Descriotion 3.1.1 Plant Nodalization For most of the analyses, TV Electric used a generic asymmetric 2-loop base i
model where one loop represents the single affected loop and the second loop represents the other three unaffected loops.
In addition, TM Electric
)
maintains a single loop model for analysis of symmetric transiFnts and a l
four-loop model for detail analysis.
3.1.1.1 Reactor Coolant System Modelina Pressurizer The pressurizer (PZR) model is a single node non-equilibrium volume including a separate surge line, power-operated relief valves (PORVs), relief valves, PZR sprays and heaters.
Simulation of the actuation of the PZR pressure control system is accomplished through use of the RETRAN control blocks.
Pressure Vessel Eight major regions within the pressure vessel are modeled including a three-2
volume active core and a core bypass region.
3.1.1.2 Steam Generator Model and Nodina Sensitivity Although TV Electric assumed, for the purpose of model qualification, that the steam generators at Unit 1 and 2 are identical, TU Electric will, at the time of actual application of the model, verify the acceptability of a particular model selection on a plant /model specific basis.
l l
The steam generator model includes an eight volume primary-side and the steam l
line modeling.
Through an extensive secondary-side nodal.ization sensitivity study, TU Electric developed four steam generator secondary side models:
(1) two-node, (2) three-node, (3) four-node and (4) detailed multi-node model.
The two-node model consists of separate nodes for the steam dome region and the boiling region. The three-node SG model refined the two-node model by adding i
a separate preheater region. The four-node model breaks the preheater region of the three-node model into two nodes.
The multi-node model was developed to assure nodalization convergence of the model determined to be acceptable for use in licensing applications.
TV Electric demonstrated, through benchmark and sensitivity studies, that the three-node model resulted in acceptably conservative predictions when l
compared with plant data and the current FSAR analysis (Ref. 6). Therefore, l
TU Electric intends to use the three-node model generally in the future. The licensee also stated that any use of an alternative SG nodalization, such as the four-node model, in the future would be justified by sensitivity analysis.
3.1 1.3 Reactor Trio Loaic and Control System Models RETRAN trip functions are utilized to simulate trip logic required for the transient analysis such as the following: reactor protection functions, control system bistable element logic, and general problem control.
Four main control systems (SG level, PZR pressure, PZR level and rod movement) are simulated in the RETRAN base model.
N-16 Transoort 4
Unique to the CPSES plants is the Nitrogen-16 (N-16) power monitoring system.
The short-lived N-16, which is generated in the core fluid regions by activation of the Oxygen-?.6 in the coolant, emits high energy gammas.
The measurement of the N-16 activity provides an indication of the core power.
Overtemperature and overpower protection signals are also generated by this system.
Simulation of the N-16 power monitoring system is performed by the use of the RETRAN-02 control blocks.
Boron Transoort Since boron injection is not required for mitigation of any event, TU 3
Electric did not qualify its boron transport model as required in the RETRAN SER.
TV Electric, however, stated that in the future when it becomes necessary, this model will t'e qualified to assure its conservatism.
SG Low-Low level Setooint Calculation For two transients (loss of normal feedwater and feedwater line break), TU Electric relies upon actuation of the SG low-low level setpoint for actuation of reactor scram and/or initiation of the auxiliary feedwater.
Because RETRAN is unable to compute the mixture level accurattTf,' TU Electric
~
simulated the Westinghouse algorithm for estimating the water level based upon the SG secondary side liquid mass.
The Westinghouse algorithm was developed using the LOFTRAN computer code.
In order to allow a margin of uncertainty in adapting this algorithm to the RETRAN model, TV Electric increased the initial steam generator mass to delay actuation on the low-low steam generator water level.
This was demonstrated to be the case through benchmark analysis against the current FSAR data.
Reactor Kinetics Model A point kinetics model was used to determine the core power response together with specific reactivity forcing functions and thermal feedback effects from the moderator and fuel temperature coefficients in the three core regions.
RETRAN hot Soot Model TV Electric developed a hot spot model for computation of the cladding temperature response during the locked rotor event. Comparison of the RETRAN computed clad surface tev erature and the FSAR data shows that this model is computing roughly 200 'c higher peak temperature than the FSAR data.
This model was approved as part of the NRC review of RXE-91-002, " Reactivity Anomaly Events Methodnlogy," and was not further reviewed here.
Core Power For the demonstration analyses, TU Electric used the same powers as were used in the FSAR analyses.
For the licensing analyses, for the "non-ESF" analyses, a core thermal power of 3479.2 MW was used instead of 3479.5 MW.
This is a result of adding a 3 power uncertainty only, and not specifically the result of considering the uncertainty associated with the net RCP heat, since TV Electric stated that uncertainty is already reflected in the measurement of the core thermal power through the use of a plant calorimeter.
In the "ESF" analyses, such as the feedwater line break event, TU Electric used a core thermal power of 3635.8 MW instead of 3630.6 MW used in the FSAR.
In the TU Electric analysis, heat addition due to the RCPs was calculated by RETRAN based on the input homologous curves.
3.2 RETRAN Model Oualification Five plant transient events and tests were analyzed to globally qualify CPSES 4
i RETRAN Plant models.
In addition, selectively choosing transients to be performed, TU Electric intended to qualify adequacy of RETRAN control block modeling and reactivity feedback models.
Transients selected were (1) full load rejection test, (2) 50% load rejection test, (3) complete loss of feedwater event, (4) single main feedwater pump trip event, and (5) partial loss of RCS flow event.
~
A four-loop version of the CPSES model was used for each of the plant data J
comparisons. The three-node steam generator model was used. A best-estimate ~
~
representation of the performance characteristics was usedWthe control system modeling.
Similarly the best-estimate reactivity and decay heat
[
feedback were simulated.
For each event detailed comparison discussion of analysis was provided.
In addition to the three-node SG model, TU 'Ilectric performed the analysis using l
the two-node SG model.
In all cases, the three-node secundary side SG model predicted plant parameters adequately - and resulted in good agreement with plant data and i
better agreement than was achieved using the two-node SG model.
Use of the detailed SG model did not improve the agreement, indicating that.the three-node SG model is adequate.
3.3 Demonstration Analysis Comoarison In order to demonstrate adequacy of the TU Electric RETRAN models in licensing type applications, TU Electric reanalyzed eight analyses from the current FSAR Chapter 15.
The RETRAN predictions were compared to the FSAR data.
Comparison of initial conditions used in the reanalyses to the current FSAR is shown on Table 13-B through 13-1 of Reference 4.
Where there are some differences, they are primarily due either to the manner in which the uncertainties are accounted for or to selection of more conservative values.
Three of eight transients were reanalyzed in depth: loss of normal feedwater, loss of nonemergency AC power and feedwater system pipe break.
These are discussed in detail below.
Four of the remaining transients were reanalyzed as part of sensitivity studies to demonstrate adequacy of the use of the three-node SG model with the preheater explicitly modeled, and include excessive increase in secondary steam flow,. turbine trip and partial and complete loss of RCS flow.
i In all four cases, the use of the three-node SG model improved the agreement with the current FSAR data.
It also resulted in faster initial PZR l
pressurization or deeper depressurization.
The Locked Rotor event was re-calculated to show that the peak pressure and i
the peak cladding temperatures were within the threshold with the three-node SG model.
5 j
3.3.1 Loss of Nonemeroency AC Power / Loss of Normal Feedwater These two transients are discussed together since for Comanche Peak, as stated by TV Electric, the method of analysis is the same and the transients follow similar sequences of events.
The transient analysis objective was to demonstrate the adequacy of the long-term, decay heat removal capability of the RCS and the auxiliary feedwater system.
In addition, the event acceptance criterion is that the PZR does not become water solid prior to the time that the heat resovaMapability of
~ these systems more than compensates for the decay heat generation rate.
l l
For this transient TU Electric selected a low-low SG trip setpoint of 15%
span.
This allowed for uncertainties in the initial SG water level, yet permitted the inventory to be roughly 7000 lbm lower that the value used by l
the vendor for the 0% span trip setpoint.
Having less SG mass resulted in a significant amount of tube uncovery following the reactor trip (which was predicted to be later than was obtained l
by the vendor) causing significantly higher post-trip RCS temperatures and therefore higher PZR pressure and PZR liquid volume than predicted by the I
vendor.
l As a result of the use of the three-node SG model and its associated low-low SG mass setpoint in the analysis, in order to demonstrate compliance with the event acceptance criteria, it became necessary to decrease the auxiliary feedwater purge volume to 75 ft.
TV Electric stated that this modified d
amount of volume is still greater than the actual plant geometry but less than assumed by the vendor.
With this assumption, the event acceptance criterion was met.
In both instances, the TV Electric analysis was conservative relative to the vendor's analysis.
3.3.2 Feedwater System Pioe Break Differences in the computed results were attributed by the licensee to a significant difference in modeling assumptions between the vendor analysis and the TU Electric analysis.
The vendor forced break flow quality to be at certain conservative values, while the TU Electric used the RETRAN code's built-in choked flow correlation to calculate the break flow through the l
This modeling difference resulted in less break flow being predicted in the RETRAN analysis between the time of reactor trip and the time for the affected steam generator to dryout and later steamline isolation. TU Electric provided adequate explanations of these results.
3.4 Licensina Analysis Acoroach The TU Electric licensing analysis approach is summarized in Table IV-1 of Reference 3 and also attached to this evaluation report.
This table describes how the input selection in system thermal / hydraulics, reactivity 6
1
coefficients and transient assumptions is made on a transient-by-transient basis according to the transient acceptance criterion.
The table also indicates parameters whose input selection criterion is affected according to the method of DNBR calculation (if the statistical core uncertainty methodology is used the nominal values are selected for those parameters identified).
In this section, TU Electric stated that it will use event-specific sensitivity studies to determine appropriate steam generator models by selecting from among those already studied. This approach. is.g. ceptable.
4.0 Conclusions This review was performed to evaluate acceptability of the RETRAN plant models for transient safety analysis through determination of the adequacy of qualification of such models against CPSES plant data and demonstration of their ability to perform FSAR-type analyses. However, review of the analysis methodologies for the Inadvertent Opening of a Steam Generator Safety Valve, the Steam Piping System Failure, the Steam Generator Tube Rupture and the Reactivity and Power Distribution Anomalies are beyond the scope of this review since these are covered under separate topical reports.
We find that the subject topical report, together with TU Electric responses, contains sufficient information to satisfy the RETRAN-02 SER requirement that each RETRAN-02 user qualifies the models for its intended applications.
We further find that the use of the TU Electric developed RETRAN models for Comanche Peak analysis results in predictions with adequate assurances of conservatism, and is therefore acceptable, subject to the conditions set forth above regarding selection of input and models for specific application.
Since boron injection is not required for mitigation of any event, TV Electric did not qualify its boron transport model as required in the RETRAN SER.
TV Electric, however, stated that in the future when it becomes necessary, that model will be qualified to assure its conservatism.
5.0 REFERENCES
1.
" Transient Analysis Methods for Comanche Peak Steam Electric Station Licensing Applications," RXE-91-001, February 1991.
2.
Letter from D.R. Woodlan (TUEC) to USNRC, " Comanche Peak Steam Electric Station Reload Analysis Topical Review Transient Analysis Methods for CPSES Licensing Applications, August 9, 1991.
3.
Letter from D.R. Woodlan (TVEC) to USNRC, " Comanche Peak Steam Electric Station Reload Analysis Topical Review Transient Analysis Methods for CPSES Licensing Applications, March 22, 1993.
4.
Licensee Qualification for Performing Safety Analyses in Support of Licensing Actions (Generic Letter No. 83-11)," USNRC, February 8, 1983.
7 l
t i
5.
" Acceptance for Referencing of Licensing Topical Report -RETRAN-02: A Thermal-Hydraulic Code-for-Reactor Cores, EPRI NP-2511-CCM Vols.1-4,"
May 1, 1986.
i 6.
" Chapter 15," CPSES/FSAR, January 15, 1990.
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e lable IV-1 Page 2 of 3 FSNt SEC110N 15.2.7 15.2.8 15.3.1 15.3.2 15.3.3 15.4.1 15.4.2 Event Acronym 10FW NFt8 Pt OF Ct OF tR/55 RV5 RWP Acceptance Criterlon Dm' Om' ONOR DNOR RCS Pres /DNBR DNBR OMER Power' High High High High High lero Range Prrr Pressure' High High low low High/ tow low low Prrr levet High High High High High High High RCS 1-avg' High High High High High High High RCS toop Flow' 10w low tw low tow low tow
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S Decay Heat Removal with Aunit isry Feedwater System 154
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