ML20046C781
| ML20046C781 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 08/05/1993 |
| From: | Richard Anderson NORTHERN STATES POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| GL-93-04, GL-93-4, NUDOCS 9308120103 | |
| Download: ML20046C781 (9) | |
Text
{{#Wiki_filter:_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ Northem States Power Company 414 Nicollet Mall Minneapolis, Minnesota 55401 1927 Telephone (612) 330-5500 August 5, 1993 Generic Letter 93-04 U S Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 -{ 'l PRAIRIE IS1AND NUCLEAR GENERATING PLANT l Docket Nos. 50-282 License Nos. DPR-42 ) 50-306 DPR-60 Response to NRC Generic Letter 93-04 Rod Control System Failure and Withdrawal of Rod Control Cluster Assemblies ) Pursuant to the requirements of 10 CFR 50.54(f),_the NRC issued Generic Letter 93-04, " Rod Control System Failure and Withdrawal of Rod Control Cluster I Assemblies," on Monday, June 21, 1993 and was addressed to.all licensees with the Westinghouse Rod Control System (except Haddam Neck) for action and to all other licensees for information. i The generic letter requires that, within 45 days from the date of the generic letter, each addressee provide an assessment of whether or not the licensing basis for each facility is still satisfied with regard to the requirements.for system response to a single failure in the Rod Control System (GDC 25 or equivalent). If the assessment (Required Response 1.(a)) indicates the licensing basis is not satisfied, then the licensee must describe compensatory l short-term actions consistent with the guidelines contained in the generic letter,.and within 90 days, provide a plan and schedule for long-term resolution. Subsequent correspondence between the Westinghouse. Owners Group-and the NRC resulted in schedular relief for Required Response l.(a)(NRC Letter to Mr. Roger Newton dated July 26, 1993). This portion of the required actions will now be included with our 90-day response, i. We hereby submit our response to the Generic _ Letter as it applies to Prairie Island. This response summarizes the compensatory actions we have taken in l. response to the Salem rod control system failure event. It also providesa li summary of the results of the_ generic safety analysis program conducted by'the l-Westinghouse Owners Group and its applicability to Prairie Island. We consider this action to be complete with respect to the 45 day required'. response to GL 93-04 (as amended by July 26 NRC letter to Mr Roger Newton). 110027 p 9308120103 930805-U PDR ADOCK 05000282 d 1 L1 - = _N_
1 ) USNRC NORTHERN STATES POWER COMPANY August 5, 1993 Page 2 In response to this generic letter, we have done or will do (as noted) the following: The biweekly rod exercise test will be changed to add steps to document receipt of the rod deviation alarm. The rod deviation alarm response procedure contains a step to place the rod control system in Manual as one of the initial actions. Following completion of the transient analysis and the evaluation of the. rod control system response, training via.the licensed operator requalification program will provide updated information on this subject. In our 90-day response, we will identify any long-term NRC commitments that we determine prudent at that time. Please contact Jack Leveille (612-388-1121, Ext 4662) if you have any questions related to our response. / (((1 ( ficd 2'Af)^L-Roger O Anderson Director Licensing and Management Issues c: Regional Administrator - Region Ill, NRC Senior Resident Inspector, NRC NRR Project Manager, NRC J E Silberg -Attachments
- 1. Affidavit
- 2. Generic Letter 93-04 Response I
e 4
i i UNITED STATES NUCLEAR REGULATORY COMMISSION NORTHERN STATES POWER COMPANY PRAIRIE ISLAND NUCLEAR GENERATING PLANT DOCKET NO. 50-282 50-306 R0D CONTROL SYSTEM FAILURE AND WITHDRAWAL OF ROD CONTROL CLUSTER ASSEMBLIES Northern States Power Company, a Minnesota corporation, with this letter is submitting information requested by NRC Generic Letter 93-04. This letter contains no restricted or other defense information. NORTHERN STATES POWER COMPANY /> 1 / i rfuh Y? $ 4 8 L-By ,pfloger 0 Anderson Director Licensing and Management Issues On this ay of /Nbefore me a notary public in and for said County,personallyap;(jaredRoger0 Anderson, Director,Licensingand Hanagement' Issues; and being first. duly sworn acknowledged that he is authorized to execute this document on behalf of Northern States Power Company, that he knows the contents thereof, and that to the best of his knowledge, information, and belief the statements made in it are true and that it is not interposed r lay. s m :::::::: :::::::::::::::::::::::: MARCIA K. LaCORE NOTARY PUBUC--lMNNESOTA HENNEPIN COUNTY fay commenon Expires Sept 24,1933 =wwvowwwovwwwwwwwwWn
~. RESPONSE TO NRC CENERIC LETTER 93-04 Compensatory Actions taken or that will be taken: 1. " additional cautions or modifications to surveillance and preventive maintenance procedures" - Westinghouse did not make any initial recommendations regarding surveillance or preventive maintenance procedures. Based on the response provided in Westinghouse Owners Group Letter OG-93-42, there was no perceived need to increase the frequency of testing on a permanent or i generic basis. PSE&G had committed to a temporary increase in testing, but only until it was demonstrated that the rod control system was operating properly and with confidence. A recommendation was made for utilities to ensure that their surveillance testing will demonstrate rod control system operability and address maintenance trouble-shooting. Increased surveillance testing is contrary to the general trend and philosophy of surveillance testing in that increased testing can, in and of itself, result in higher rates of system and component failures. Therefore, the WOG and Westinghouse have concluded that increased frequency in surveillance testing is not appropriate. 'l Assessment of functionality testing of the rod deviation alarm at Prairie Island has been done. The rod deviation alarm is generated by the plant process computer. The alarm is functionally tested each refueling, and a verification is done monthly to ensure that.the plant process computer is properly monitoring rod position. Verification of alarm functionality was also performed by reviewing the alarm printout l from a recent rod exercise test. The biweekly rod exercise test will be changed to add steps to document receipt of the rod deviation alarm. The rod deviation alarm response procedure has been changed; a step has been added to place the rod control system in Manual as one of the initial actions. 2. " additional administrative controls for plant startup and power operation" - l PSE&G committed the Salem units to start up by dilution. Neither Westinghouse'nor the WOG has endorsed this requirement. In actual operation, the operators would be aware of abnormal rod movement and terminate rod demand prior to reaching criticality. The operator would be manually controlling the rod withdrawal such-that the detection of rod mis-stepping in less than 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 unrealistic to believe that the operators would permit an unchecked rod withdrawal during startup such that criticality would be reached. Thus, the WOG and Westinghouse have concluded that.startup by ) dilution is not warranted in response to the Salem rod control system failure event. i I l l I
.. _. ~. ? a August 5, 1993 Page 2 of 6 3. " 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" - Both Westinghouse and the WOG have, at various times, recommended that licensees provide additional discussion, training, standing orders, etc., to ensure that their operators are aware of what occurred at Salem. The recommendations of the Westinghouse Nuclear Safety Advisory Letter (NSAL) 93-007, which was subsequently endorsed by the WOG via Letter OG-93-42, recognize the benefits of ensuring that plant operators are knowledgeable of the Salem rod control system failure event. Operations personnel were informed of the event in writing and by routing a' copy of the Westinghouse NSAL, Following completion of the transient analysis and the evaluation of the rod control system response, training via the licensed operator requalification program will provide updated information on this subject. P A f t i s i m
l August 5, 1993 Page 3 of 6 Summary of the Generic Safety Analysis Prozram Introduction As part of the Westinghouse Owners Group initiative, the WOG Analysis subconnittee is working on a generic approach to demonstrate that for all 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 suberitical and power conditions to demonstrate that DNB does not occur. The current Westinghouse analysis methodology for the bank withdrawal at power and 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 (LOFTS /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. I Differences in plant designs are addressed by using conservative adjustment factors to make a plant-specific DNB assessment. Descrintion 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 temperature and pressure. If the reactivity worth of the withdrawn rods is sufficient, 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 decrease due to the negative feedback effect caused bysthe moderator temperature rise. The accidental withdrawal of a bank or banks of RCCAs in the normal overlap mode is a transient which is specifically considered in I plant safety analysis reports. The consequences of a bank withdrawal accident meet Condition II 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 l symmetrically located around the core, this can cause.a " tilt" in the core j 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 i
4-August 5, 1993 Page 4 of 6 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 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 core where the 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 asymmetric rod withdrawal. The LOFT 5 code is a combination of an advanced l version of the LOFT 4 code (Reference 1), which has been'used for many years by 1 Westinghouse in the analysis of the RCS behavior to plant transients and accidents, and the advanced nodal code SPNOVA (Reference 2). LOFT 5 uses a full-core model,~ consisting of 193 fuel assemblies with one node i per assembly radially and 20 axial nodes. Several " hot" rods are specified with different input multipliers on the hod rod powers to simulate the effect l 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 i LOFT 5. DNBRs are calculated for each hot rod within LOFT 5'with a simplified ] DNB-evaluation model using.the VRB-1 correlation. The DNBRs resulting from the LOFT 5 calculations are used for comparison purposes.
- j A more detailed DNBR analysis is done at the limiting transient statepoints from LOFT 5 using THINC-IV (Reference 3) and-the Revised Thermal Desiga l
Procedure (RTDP). RTDP applies to all Westinghouse plents, maximizes'DNBR margins, is approved by the NRC, and is licensed for a number of Westinghouse 'i plants. The LOFTS-calculated DNBRs are conservatively low when compared to the THINC-IV results. 3 Assumptiong The initial power levels chosen for the performance of bank and multiple RCCA withdrawal cases are 100%, 60%, 10% and hot zero power (HZP). The-power levels-are the same powers considered in the RCCA-Bank Withdrawal at Power and-Bank Withdrawal from'Subcritical events presented in the plant Safety Analysis = Reports. The plant., in'accordance-with RTDP, is assumed to be operating at' nominal _ conditions for each_ power level examined. Therefore, uncertainties. 1 will not _ affect the results of the IJDFIS transient analyses. For the at-power: Leases, all coolant pumps are assumed to be in operation. For-hot'zero power case (suberitical' event),_only 2/4 reactor coolant pumps are assumed to be in operation. 'A " poor mixing". assumption is used for the reactor. vessel inlet 'I and outlet mixing model.
t -E August 5, 1993 Page 5 of 6 a Results i A review of 'he results presented in Reference 4 indicates that for the 't asymmetric 4 withdrawal cases analyzed with the LOFT 5 code, the DNB design l basis is met. As demonstrated by the A-Factor approach (described below) for r addressing various combinations of asymmetric rod withdrawals, the single s most-limiting case is plant-specific and is a function of rod insertion limits, rod control pattern, and 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. j At HZP, a worst-case scenario (3-rods withdrawn from three different banks which is not possible) shows a non-limiting DNBR. This' result is applicable to all other Westinghouse plants. 1 ? Plant Applicability l The 3-D transient analysis approach uses a representative standard 4 Loop l 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 credit is assumed for the f(AI) penalty function. ~ The f(A1) penalty function reduces the OTDT setpoint for highly skewed l positive or negative axial power shapes. Compared to the plant-specific OTDT j setpoints including credit for the f(AI) penalty function, the setpoint used' in the LOFT 5 analyses is conservative, i.e., for those cases that tripped on l OTDT, a plant-specific OTDT setpoint with the f(A1) penalty function will. j result in an earlier reactor trip than the IDFT5 setpoint. This ensures that the rtatepoints 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-factor is defined as the ratio between the design FAH,from'the symmetric and asymmetric RCCA withdrawal 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. i 1 J/ M +
~- .v.- ~. t .[ i t August 5, 1993 Page'6 of 6= 1 Conclusion i Using this approach, the generic analyses and their plant-specific npplication i demonstrate that, for Prairie Island, DNB does not occur for the worst-case asymmetric rod withdrawal. References 1.
- Burnett, T.W.T., et al., "LOFTRAN Code Description,"
1 WCAP-7907-A, April 1984. 2.
- Chao, Y.A.,
et al., "SPNOVA - A Multi-Dimensional Static and Transient Computer Program for PWR Core Analysis," WCAP-12394, September 1989. 3. Friedland, A.J. and S. Ray, " Improved THlNC-IV Modeling for PWR Core Design," WCAP-12330-P, August 1989. 4.
- Huegel, D., et cl., " Generic Assessment of Asymmetric Rod Cluster Control Assembly Withdrawal," WCAP-13803, August 1993.
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