ML20046C211

Submits Response to Generic Ltr 93-03, Rod Control Sys Failure & Withdrawal of Control Cluster Assemblies. Attachment 2 Summarizes Compensatory Actions Taken in Response to Rod Control Sys Failure Event

ML20046C211
Person / Time
Site:	Comanche Peak
Issue date:	08/05/1993
From:	William Cahill TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
GL-93-03, GL-93-3, TXX-93287, NUDOCS 9308090346
Download: ML20046C211 (8)

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Log # TXX-93287 Fi1e # 10035 TUELECTRIC William J. Cahllt. Jr.

Gr <mp Yk e hesialent U. S. Nuclear Regulatory Commission Attn: Document Control Room Washington, DC 20555

SUBJECT:

COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)

DOCKET NOS. 50-445 AND 50-446 TRANSMITTAL 0F RESPONSE TO GENERIC LETTER 93-04 Gentlemen:

Pursuant to the requirements of 10CFR50.54(f), the NRC issued Generic Letter 93-04, " Rod Control System Failure and Withdrawal of Rod Control Cluster Assemblies," on Monday, June 21, 1993.

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 that the licensing basis is not satisfied, the licensee must provide an assesment of the impact of a potential single failure and ,.

describe compensatory short-term actions consistent with the guidelines contained in the generic letter (Required Response 1.(b)); and within 90 days, provide a plan and schedule f or long-term resolution (Required Response 2). Subsequent correspondence between the Westinghouse Owners Group and the NRC resulted in schedular relief for Required Response 1.(a)

(NRC Letter to Mr. Roger Newton dated July 26, 1993). This portion of the required actions will now be included with the 90-day licensee response.

TV Electric hereby submits a response to the Generic Letter as it applies to CPSES Units 1 and 2. Attachment 2 to this letter summarizes the compensatory actions taken by TV Electric in response to the Salem rod control system f ailure event. Attachment 3 to this letter addresses the safety impact of this issue. TV Electric considers 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).

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TXX-93287 Page 2 of 2 If you have any questions, please contact Mr. J. D. Rodriguez at

-( 214 ) 812-8674.

Sincerely,

. f gran William J. Cahill, Jr.

JDR/grp Attachments c- Mr. J. L. Milhoan, Region IV Resident Inspectors, CPSES (2)

Mr. T. A. Bergman, NRR r

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Attachment 1 to TXX-93287 Page 1 of 1 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of )

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Texas Utilities Electric Company ) Docket Nos. 50-445

) and 50-446 (Comanche Peak Steam Electric )

Station, Unit 1 & 2) )

AFFIDAVIT William J. Cahill, Jr. being duly sworn, hereby deposes and says that he is Group Vice President, Nuclear Production of TV Electric, the licensee herein; that he is duly authorized to sign and file with the Nuclear Regulatory -

Commission this response to Generic Letter 93-04

• Rod Control System Failure and Withdrawal of Rod Control Cluster Assemblies,10CFR50.54(f)" for the Comanche Peak Steam Electric Station, Units 1 & 2; that he is f amiliar with the content thereof; and that the matters set forth therein are true and correct to the best of his knowledge, information and belief.

William J. Cahvil, Jr. /

Group Vice President, Nuclear STATE OF TEXAS ) '

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COUNTY OF DALLAS )

Subscribed and sworn to before me, on this 5th day of Augurl , 1993,

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Nota ry/ Public

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Attachment 2 to TXX-93287 Page 1 of 1 _;

Comoensatory Actions i Taken in Response _12 l Salem Rod withdrawal Eveit As requested by required action 1(b) of GL-93 04 TV Electric has taken the f ollowing actions:

1' Additional cautions or modifications to surveillance and preventive maintenance procedures:  !

Station procedures for Unit I have been revised to require  ;

verification of the functionality of the rod deviation alarm after every refueling. The Unit 2 rod deviation alarm was functionally verified as part of the startup test program. Changes to Unit 2 .

procedures, to require verification of the functionality of the ,

rod deviation alarm, will be incorporated prior to restart from ,

2RF01.

t Additional administrative controls for plant startup' and power

• operation:

An internal review of event response procedures has been  !

completed. The procedures were determined to be adequate.  !

Additional instructions and training to heighten operator awa'reness of potential rod control system failures and to guide operator response in

• the event of a rod control system malfunction:

An operation shift order which cautions operators, summarizes the ,

Salem rod withdrawal event, and reinforces the need to continue to be attentive to possible malfunctions of the rod control system  ;

was issued.

Operator training in the form of Lessons Learned was issued. The Lessons Learned training required operators to read and understand the Nuclear Safety Advisory Letter (NSAL-93-007).

A classroom training lecture on current industry events as well as a simulator scenario will be developed and is to be completed by i Licensed Operator Re-qualification Training (LORT) cycle 94-01.  :

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i Attachment 3 to TXX-93287 Page 1 of 4  ;

Summary of the Generic Safety Analysis Proaram +

Introduction As part of the Westinghouse Owners Group initiative, the WOG Analysis subcommittee has developed a generic approach to demonstrate that for all Westinghouse plants there is no safety significance for an asymmetric Rod ControlClusterAssembly(RCCA) withdrawal. The purpose of the program is to analyze a series of asymmetric rod withdrawal cases from both subtritical and power conditions to demonstrate that Departure from Nucleate Boiling (DNB) does not occur.

The current Westinghouse analysis methodology for the bank withdrawal at power and from subtritical 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. When the current Westinghouse analysis methodology is used for a single RCCA withdrawal, it is .

shown that the minimum DNB limit is exceeded.

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 plant-specific DNB assessment.

Description of Asymetric 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 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 by the 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 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 symmetrically 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 r

Attachment 3 to TXX-93287 Page 2 of 4 -

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-cold, which is used in the Over-Temperature reactor protection system. The radial power " tilt" may also affect the excore 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 version of che LOFT 4 code (Reference 1), which has been used for many years by Westinghouse in the analysis of the RCS behavior to plant transients and accidents, an:1 the advanced nodal code SPNOVA (Reference 2). t 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 ',nput 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 h ghest FAH in the assembly, and is calculated by SPHOVA 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 LOFT 5 using THINC-IV (Reference 3) and the Revised Thermal Design .

Procedure (RTDP). RTDP applies to all Westinghouse plants, maximizes 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.

Assumptions The initial power levels chosen for the performance of bank and multiple RCCA 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 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 will not affect the results of the LOFT 5 transient analysis. 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.

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i Attachment 3 to TXX-93287 Page 3 of 4 Results A review of the results presented in Reference 4 indicates that for the asymmetric rod withdrawal cases analyzed with the LOFT 5 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 control pattern, and core design. 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 three different banks which is not possible) shows a non-limiting DNBR. This result is applicable to all other Westinghouse plants. -

Plant Applicability 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 Over-Temperature reactor trip, no credit is assumed for the f(AI) penalty function. The i f(AI) penalty function reduces the OT setpoint for highly skewed positive or negative axial power shapes. Compared to the plant-specific OT 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 OT, a ,

plant-specific OT setpoint with the f(AI) 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 OT 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 and the maximum transient 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. With the differences in  ;

plant design accounted for by the adjustment approach, plant-specific DNBR calculations can be generated for all Westinghouse plants.  :

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Attachment 3 to TXX-93287 Page 4 of 4 ,

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Conclusion The generic analysis combined with the plant-specific application demonstrates i that for CPSES Units 1 and 2, DNB does not occur for the worst-case asymmetric ~

rod withdrawal. ,

References ,

1) Burnett, T.W.T. et al., "LOFTRAN Code Description," 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 THINC-IV Modeling for PWR Core  ;

Design," WCAP-12330-P, August 1989. -

Huegel, D., et al., " Generic Assessment of Asymmetric Rod Cluster 4)

Control Assembly Withdrawal," WCAP-13803, August 1993.

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