ML20046C335
| ML20046C335 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood, Zion File:ZionSolutions icon.png |
| Issue date: | 08/04/1993 |
| From: | Chrzanowski D COMMONWEALTH EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| GL-93-04, GL-93-4, NUDOCS 9308100156 | |
| Download: ML20046C335 (9) | |
Text
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N Ccmmonwaalth Edison g
I / 1400 Opus Place Downers Grove, Illinois 60515 i
August 4,1993 U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Washington, DC 20555 Attention: Document Control Desk
Subject:
Generic Letter 93-04, " Rod Control System Failure and Withdrawal of Rod Control Cluster Assemblies."
1 Byron Station Units 1 and 2, (NRC Docket Numbers 50-454 and 50-455)
Braidwood Station Units 1 and 2, (NRC Docket Numbers 50-456 and 50-457)
Zion Station Units 1 and 2, (NRC Dockets 50-295 and 50-304) 1
Reference:
Letter from A. Thadani (NRR) to R. Newton (Westinghouse Owners Group), dated July 26,1993.
The purpose of this letter is to provide the 45 day response to the subject Generic Letter. Based on guidance contained in the referenced letter, Commonwealth Edison (CECO)is limiting this response to address only the second part of the " Required Response 1.(b)" section of the Generic Letter. This response discusses the compensatory actions taken at Byron, Braidwood and Zion Stations and is contained in Attachment 1 to this letter. Attachment 2 contains a discussion of the results of the generic safety analysis program conducted by the Westinghouse Owners Group and its applicability to Byron, Braidwood and Zion Stations.
Per the conditions of the schedular relief granted in the referenced letter, CECO will be providing additional information regarding the effect of the Salem Rod Control event on the licensing basis of Byron, Braidwood and Zion Stations in an additional response due in 90 days from the date of the Generic Letter.
To the best of my knowledge and belief, the statements contained in this document are true and correct. In some respects these statements are not based on my personal knowledge, but on information furnished by other CECO employees, contractor employees, and/or consultants. Such information has been reviewed in accordance with company practice, and I believe it to be reliable.
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If there are any questions or comments, please contact me at (708) 663-7292 f
Sincerely, David. Chrzanowski Generic Issues Administrator Nuclear Regulatory Services Attachments:
(1) Commonwealth Edison Response to Generic Letter 93-04 for Byron, Braidwood, and Zion Stations.
i (2) Summary of the Generic Safety Analysis Program cc:
J. Martin, Regional Administrator-RIII J. Hickman, Byron Project Manager-NRIUPDIII-2 R. Assa, Braidwood Project Manager-NRIVPDIII-2 C. Shiraki, Zion Project Manager-NRR/PDIII-2 H. Peterson, Senior Resident Inspector (Byron)
. S. DuPont, Senior Resident Inspector (Braidwood)
J. D. Smith, Senior Resident Inspector (Zion)
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ATTACHMENT 1 COMMONWEALTH EDISON RESPONSE (BYRON, BRAIDWOOD & ZION)
GENERIC LETTER 93-04 The CECO response to the second paragraph in 1.(b) of the " Required Response" section of the Generic Letter is as follows.
Describe any compensatory short-term actions taken or that will be taken to address any actual or potential degraded or nonconforming conditions such as
- additional cautions or modifications to surveillance and preventive maintenance procedures CECO reviewed the response provided in OG-93-42, and agrees there is no need to increase the frequency of rod testing on a permanent or generic basis.
Byron, Braidwood, and Zion Station Management have considered the Salem event in light of our Licensed Operator Training Program, General Operating Procedures, Abnormal Operating Procedures, Administrative Procedures, Surveillances, and Preventative Maintenance Procedures. Though we recognize the severe reactivity ramifications of such an event, we firmly believe our current procedural and administrative controls are more than adequate. Thus, we did not identify the need for any additional administrative controls, cautions, or modifications for plant startup and power operation.
t CECO has confirmed the functionality of the rod deviation alarms at Byron, Braidwood and Zion. This action was taken in response to Westinghouse Letter NSAI 93-007, dated June 11,1993. This functionality test verified that any control rod in Control Banks C or D which deviates from its bank demand position by more than 12 steps will generate an alarm. Simulation of the process computer inputs for each Control Bank C and D rod Digital Rod Position Indication (Byron and Braidwood) or Rod Position Indication (Zion) input adequately demonstrated alarm functionality. -
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- additional administrative controls for plant startup and power operation As previously stated, PSE&G committed the Salem units to startup by dilution. CECO, Westinghouse, nor the WOG have endorsed this requirement. In actual operation, the operators would be aware of abnormal rod movement and terminate rod demand prior to ever reaching criticality. The operator would be manually controlling the rod withdrawal such that the detection of rod mis-stepping in under 1 minute would be reasonable. In fact, as demonstrated during the R.E. Ginna event, abnormal rod motion was terminated after only one step both in automatic and manual rod control. It is entirely too unrealistic to believe that the operators would permit an unchecked rod withdrawal during startup such that criticality would be reached. Thus, CECO, Westinghouse, and the WOG have concluded that startup by dilution is not required in response to the Salem rod control system failure event.
Based on these considerations and the applicable Byron, Braidwood, and Zion procedures, CECO has not initiated any additional administrative controls for plant startup and power operation.
- additional instructions and training to heighten operator awareness of potential rod control system failures and to guide operator response in the event of a rod control system malfunction CECO in consideration of Westinghouse and WOG recommendations has provided additional discussion, training, standing orders, etc. to ensure that their operators are aware of what transpired at Salem.
Specifically, Byron, Braidwood,and Zion operations personnel were made aware of the Salem event and reminded to continue to verify expected rod motion is indicated on their Rod Position Indication system.
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l ATI'ACIIMENT 2 i
Summarv of the Generic Safety Analysis Procram Introduction As part of the Westinghouse Owners Group initiative, the WOG Analysis aubcommittee is working on a generic approach to demonstrate that for all i
Westinghouse plants there is no safety significance for an asymmetric RCCA withdrawal. The purpose of the program is to analyze a series of asymmetric rod withdrawal cases from both subcritical and power conditions to demonstrate that DNB does not occur.
The current Westinghouse analysis methodology for the bank withdrawal at power from suberitical uses point-kinetics and one dimensional kinetics transient models, respectively. These models use conservative constant reactivity feedback assumptions which result in an overly conservative prediction of the core response for these events.
A three-dimensional spatial kinetics / systems transient code (LOFT 5/SPNOVA) is being used to show that the localized power peaking is not as severe as current codes predict The 3-D transient analysis approach uses a representative standard 4-Loop Westinghouse plant with conservative reactivity assumptions. Limiting asymmetric rod withdrawal statepoints (i.e., conditions associated with the limiting time in the transient) are established for the representative plant which can be applied to all Westinghouse plants. Differences in plant designs are addressed by using conservative adjustment factors to make a plant-specific DNB assessment.
Description of Asymmetric Rod Withdrawal The accidental withdrawal of one or more RCCAs from the core is assumed to occur which results in an increase in the core power level and the reactor coolant i
temperature and pressure. If the reactivity worth of the withdrawn rods is sufficiert, the reactor power and/or temperature may increase to the point that the transient is automatically terminated by a reactor trip on a High Nuclear Flux or Over-Temperature Delta-T (OTDT) protection signal. If the reactivity rise is small, the reactor power will reach a peak value and then decreased due to the negative feedback effect caused by the moderator temperature rise.
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The accidental withdrawal of a bank or banks of RCCAs in the normal overlap mode is a transient which is specifically considered in plant safety analysis i
reports. The ecmsequences of a bank withdrawal accident meet Condition H criteria (no DNB). If, however, it is assumed that less than a full group or bank of control rods is withdrawn, and these rods are not synunetrically located around the core, this can cause a " tilt"in the core radial power distribution. The " tilt" could result in a radial power distribution peaking factor which is more severe than is normally considered in the plant safety analysis report, and therefore cause a loss of DNB margin. Due to the imperfect mixing of the fluid exiting the core before it enters the hot legs of the reactor coolant loops, there can be an imbalance in the loop temperatures, and therefore in'the measured values of T-avg i
and delta-T, which are used in the Over-Temperature Delta-T protection system for the core. The radial power " tilt" may also affect the ex-core detector signals used for the High Nuclear Flux trip. The axial offset (AO)in the region of the i
core where rods are withdrawn may become more positive than the remainder of the core, which can result in an additional DNB penalty.
Methods The LOFT 5 computer code is used to calculate the plant transient response to an I
asymmetric rod withdrawal. The LOFT 5 code is a combination of an advanced version of the LOFT 4 code (Reference 1), which has been used for many years by Westinghouse in the analysis of the RCA behavior to plant transients and I
accidents, and the advanced nodal code SPNOVA (Reference 2).
LOFT 5 uses a full-core model, consisting of 193 fuel assemblies with one node per assembly radially and 20 axial nodes. Several " hot" rods are specified with different input multipliers on the hot rod powers to simulate the effect of plants with different initial FAH values. A " hot" rod represents the fuel rod with the highest FAH in the assembly, and is calculated by SPNOVA within LOFT 5.
DNBRs are calculated for each hot rod within LOFT 5 with a simplified DNB-evaluation model using the WRB-1 correlation. The DNBRs resulting from the LOFT 5 calculations are used for comparison purposes.
A more detailed DNBR analysis is done at the limiting transient statepoints from l
LOFT 5 using THINC-IV (Reference 3) and the Revised Thermal Design Procedure (RTDP). RTDP applies to all Westinghouse plants, maximized DNBR margins, is approved by the NRC, and is licensed for a number of Westinghouse plants. The LOFT 5-calculated DNBRs are conservatively low when compared to the THINC-IV results.
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Assumptions The-initial power levels chosen for the performance of bank and multiple RCCA j
' withdrawal cases are 100%,60%,10% and hot zero power (HZP). These power levels are the same powers considered in the RCCA Bank Withdrawal at Power and Bank Withdrawal from Suberitical events presented in the plant Safety j
Analysis Reports. The plant,in accordance with RTDP,is assumed to be operating at nominal conditions for each power level examined. Therefore, uncertainties will not affect the results of the LOFT 5 transient analyses. For the at-power cases, all reactor coolant pumps are assumed to be in operation. For the hot zero power case (subcritical event), only 2/4 reactor coolant pumps are assumed to be in operation. A " poor mixing" assumption is used for the reactor vessel inlet and outlet mixing model.
Results A review of the results presented in Reference 4 indicates that for the asymmetric
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rod withdrawal cases analyzed with the LOPf5 code, the DNB design basis is met.
As demonstrated by the A-Factor approach (described below) for addressing various combinations of asymmetric rod withdrawals, the single most-limiting case is plant-specific and is a function of rod insertion limits, rod centrol pattern, and r
core design. The results of the A-Factor approach also demonstrates that the cases analyzed with the LOFT 5 computer code are sufficiently conservative for a wide range of plant configurations for various asymmetric rod withdrawals. In addition, when the design FAH is taken into account on the representative plant, the DNBR criterion is met for the at-power cases. At HZP, a worst-case scenario (3-rods withdrawn from the three different banks which is not possible) shows a non-limiting DNBR. This result is applicable to all other Westinghouse plants.
Plant Applicability
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The 3-D transient analysis approach uses a representative standard 4-Loop Westinghouse plant with bounding reactivity assumptions with respect to the core design. This results in conservative asymmetric rod (s) withdrawal statepoints for the various asymmetric rod withdrawals analyzed. The majority of the cases analyzed either did not generate a reactor trip or were terminated by a High Neutron Flux reactor trip. For the Overtemperature Delta-T reactor trip, no i
credit is assumed for the RAI) penalty function. The RAI) penalty function reduces the OTDT setpoint for highly skewed positive or negative axial power shapes.
Compared to the plant-specific OTDT setpoints including credit for the RAI) i penalty function, the setpoint used in the LOFT 5 analyses is conservative, i.e., for those cases that tripped on OTDT, a plant-specific OTDT with the RAI) penalty function will result in an earlier reactor trip than the LOFT 5 setpoint. This ensures that the statepoints generated for those cases that trip on OTDT are conservative for all Westinghouse plants.,
With respect to the neutronic analyses, an adjustment factor ("A-factor") was calculated for a wide range of plant types and rod control configurations. The A-i factor is defined as the ratio between the design FAH and the change in the
- maximum transient FAH from the symmetric and asymmetric RCCA withdrawal
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cases. An appropriate and conservative plant-specific A-factor was calculated and used to determine the corresponding DNBR penalty or benefit. With respect to the thermal-hydraulic analyses, differences in plant conditions (including power level, RCS temperature, pressure, and flow) are addressed by sensitivities performed using THINC-IV. These sensitivities are used to determine additional DNBR penalties or benefits. Uncertainties in the initial conditions are accounted for in the DNB design limit. Once the differences in plant design were accounted for by the adjustment approach, plant-specific DNBR calculations can be generated for all Westinghouse plants.
Conclusion In summary, the 3-D transient analysis of the representative standard 4 loop Westinghouse plant with bounding core design assumptions demonstrated 4
acceptable DNBR results for the worst case asymmetric rod withdrawal event. In addition, DNBR sensitivity evaluations were performed to address plant specific variations in core design (e.g., loading pattern, fuel type, and peaking factors),
plant design (RCS flow, temperature, pressure, etc.), tmd reactor protection setpoints OTDT. Thrse results demonstrate that the Byron, Braidwood, and Zion stations maintain acceptable DNBR margin with respect to the conservative Westinghouse representative analysis methodology for the worst case asymmetric rod withdrawal event.
References 1)
Burnett, T.W.T., et al., "LOPTRAN Code Description," WCAP-7907-A, April j
1984.
t 2)
Chao, Y.A., et al., "SPNOVA-A Multi-Dimensional Static and Transient Computer Program for PWR Core Analysis," WCAP-12394, September 1989.
i 3)
Friedland, A.J. and S. Ray, " Improved THWG IV Modeling for PWR Core
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Design," WCAP-12330-P, August 1989.
4)
Huegel, D., et al., " Generic Assessment of Asymmetric Rod Cluster Control Assembly Withdrawal," WCAP-13803, August 1993.
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