ML20056D339

Responds to GL 93-04, Rod Control Sys Failure & Withdrawal of Rod Cluster Assemblies

ML20056D339
Person / Time
Site:	Catawba, McGuire, Mcguire
Issue date:	08/05/1993
From:	Tuckman M DUKE POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
GL-93-04, GL-93-4, NUDOCS 9308120190
Download: ML20056D339 (9)

Text

l DuieIbwerCompany M S Dxutav'

. P.0 Box 1006 Senior Vice President

. Owrlotte, NC2820M006 Nuclear Generation (704)382-22000thce (704)3824360 Fax DUKEPOWER August 5, 1993 U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Attention: Document Control Desk

Subject:

McGuire Nuclear Station Docket Numbers 50-369 and -370 Catawba Nuclear Station Docket Numbers 50-413 and -414 Duke Power Company's Response to Generic Letter 93-04 In accordance with 10 CFR 50.54 (f) , the NRC on June 21, 1993 issued Generic Letter 93-04, " Rod Control System Failure and Withdrawal of Rod Cluster Assemblies" to all licensees with the Westinghouse Rod Control System (except Haddam Neck).

The Generic Letter (GL) required that within 45 days various actions be taken, including:

• Assess whether the licensing basis of the plant is satisfied with regard to system response assuming a single failure in the rod control system-(RCS) (Item 1(a)); and e If the assessment in 1(a) indicates the licensing basis is not satisfied, assess the impact of potential single failures in the RCS on the licensing basis of the facility and indicate what short-term compensatory measures have been taken (Item 1(b)).

Subsequent correspondence (Reference letter, A. C. Thadani (NRC) to R. Newton (Westinghouse) , July 26, 1993) provided schedular relief for Item 1(a), delaying its required submittal an additional 4r days.

Attached is Duke Power Company's response to GL 93-04 as it applies to McGuire and Catawba Nuclear Stations. This response provides a summary of the generic safety analysis provided by Westinghouse, and its applicability to the two stations (Attachment-I), and the short-term compensatory actions taken at McGuire and Catawba (Attachment II).  !

If you have any questions regarding this response, please call

-.. e 9308120190 9 -i PDR ADOCK 0S000369 PDR E O l P L mm m eur- ), g6

U. S. Nuclear Regulatory Commission August 5, 1993 Page 2 Scott Gewehr at (704) 382-7581.

I declare that the statements and matters set forth herein are true and correct to the best of my knowledge.

Very truly yours, N.5. S M. S. Tuckman cc: Mr. V. Nerses, Project Manager Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Mail Stop 14H25, OWFN Washington, D. C. 20555 Mr. L. A. Wiens, Project Manager.

Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Mail Stop 14H25, OWFN Washington, D. C. 20555 Mr. R. E. Martin, Project Manager Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Mail Stop 14H25, OWFN Washington, D. C. 20555 i

Mr. S. D. Ebneter, Regional Administrator U.S. Nuclear Regulatory Commission - Region II i 101 Marietta Street, NW - Suite 2900 Atlanta, Georgia 30323 Mr. Mark Proviano Westinghouse Electric Corporation  :

P. O. Box 344 ECE 4-08 i Pittsburgh, PA 15230-0355 i

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1 Attachment I Summary of Generic Safety Analysis Provided by Westinghouse i

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Introduction As part of the Westinghouse Owner's Group initiative, the WOG Analysis subcommittee 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 subcritical and power conditions to demonstrate that DNB does not occur. .

The current Westinghouse analysis methodology for the bank withdrawal at power and from subcritical 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. 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 temperature and pressure. If the reactivity worth of the withdrawn rod 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 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 i 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

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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 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 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 the LOFT 4 code (Reference 1), '

which has been used by Westinghouse for many years in the analysis ,

of the RCS behavior to plant transients and accidents, and the advanced code SPNOVA (Reference 2).

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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 >

rod powers to simulate the effects of plants with different initial FAH values. A " hot" rod represents the fuel rod with the highest FAH in the assembly, and is 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 calculation 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 i plants, maximizes DNBR margins, is approved by the NRC, and is '

licensed for a number of westinghouse plants. The LOFTS-calculated i DNBRs are conservatively low when compared to the THINC-IV results.

Assumptions The initial power levels chosen for the performance of bank and l i

multiple RCCA withdrawal cases are 100%, 60%, 10%, and hot Zero I power (HZP). These power levels are the same powers considered in j 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 analyses. For the at-power cases, all reactor coolant pumps are assumed to be in operation. For the hot zero power case

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I (subcritical event), only 2/4 reactor coolant pumps are assumed to '

be in operation. A " poor mixing" assumption is used for the reactor l vessel inlet and outlet mixing model.

Results ,

A review of the results presented in Reference 4 indicates that for I

the asymmetric rod withdrawal cases analyzed with the LOFT 5 code, l

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 l pattern, and core design. The results of the A-factor approach also l demonstrate that the cases analyzed with the LOFTS 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 j 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. I This result is applicable to all Westinghouse plants.

Plant Apolicability 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 credit is assumed for the f(AI) penalty function. The f(AI) 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 f(AI) 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 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 OTDT are conservative for all Westinghouse plants.

With respect to the neutronic analysis, 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 change in the maximum transient FAH i from the symmetric and asymmetric RCCA withdrawal cases. An l 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, l

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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 were accounted i for in the adjustment approach, plant-specific DNBR calculations can be generated for all Westinghouse plants.

The evaluation used by Westinghouse to calculate the amount of DNBR margin available for each unit compared to the reference plants used the WRB-1 Critical Heat Flux (CHF) correlations. The analysis method relies on the use of the CHF correlation's DNBR sensitivity to reactor power level, system flow, core inlet temperature, and core pressure to convert parameter differences into common units of percent DNB. This DNBR adjustment, which includes any applicable '

DNBR penalties, is used to determine if the specific unit being evaluated will violate the licensed DNBR design limit. Duke Power Company uses the BWCMV CHF correlation for DNB analysis of both the Mark-BW and OFA 17x17 mixing vane fuel assembly designs. The same analysis method described by Westinghouse was used with the BWCMV CHF correlation sensitivities substituted for all the evaluated parameters. The results show positive DNBR margin is available for both the Mark-BW and OFA designs at McGuire and Catawba.

Conclusion Using this approach, the generic analyses and their plant-specific.

application demonstrate that for Catawba and McGuire DNB does not

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occur for their worst-case asymmetric rod withdrawal.

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Referenceg

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 Ray, S. , Improved THINC-IV Modeling for DWR Core Design, WCAP-12330-P, August 1989.

4) Huegel, D., et al., " Generic Assessment of Asymmetric Rod Cluster Control Assembly Withdrawal," WCAP-13803, August 1993. l i

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Attachment II -

Short-term Compensatory Actions  !

1 The. purpose of this discussion is to provide a response to the three areas of short-term compensatory actions identified by the  !

-NRC (GL 93-04 Required Response 1(b)),- and any additional compensatory actions judged to be appropriate. i

1. " additional cautions or modifications to surveillance and a preventive maintenance procedures"' 5 i

There is no perceived need to increase the frequency of testing.

Increased surveillance testing is contrary to the general trend and  ;

philosophy of surveillance testing relaxation in that it has been -;

recognized that increased testing can, in and of itself, result in higher rates of system and component failures. Westinghouse, the Westinghouse Owner's Group, and Duke Power Company have concluded

• that increased' frequencies in surveillance testing are not  :

appropriate.

Testing has been performed at Catawba, and McGuire, to verify that i

the rod deviation alarms on the operator' Aid Computer -(OAC)  !

function as intended. A procedure was developed to verify'that the  !

rod position versus bank position deviation- alarms function  !

properly. This procedure was performed for all control rods in-  !

each unit at Catawba and McGuire.

2) " additional administrative controls for plant startup and I power operation" ,

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 l and terminate rod demand prior to reaching criticality. The i operator would be manually controlling the rod withdrawal such that q the detection of rod mis-stepping in under' 1 minute would~ be reasonable.In fact,.as demonstrated in the R. E. Ginna event,- i abnormal rod motion was terminated after only one step in both'  !

automatic and manual rod control modes. It is unrealistic to  !

believe' that the operators would permit an unchecked rod withdrawal 1 during startup, . such that criticality would _ be reached. Thus,  ;

Westinghouse, WOG, and Duke have concluded that startup by dilution '

is not required. l

3) " additional instructions and training to heighten operator i awareness of potential rod control system failures and to guide I operator response in the event of a rod control system  ;

malfunction."

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~ . l The operators at McGuire and Catawba were provided, in writing, a -

summary of the Salem event as described in - the Nuclear Safety Advisory Letter provided by Westinghouse, and Information Notice  ;

93-46. The summary noted that the potential exists for a simil2r  !

event to occur at any plant with a Westinghouse Solid State Rod Control System, and recommended the following actions:

1. Licensed. Operators should continue normal' process of verifying that rod motion is proper for the requested movement while l either in Automatic, Mr.nual, or Individual . Mode of operation j by: 1

- comparing Digital Rod Position Indication, Step Demand l Counters and General 76 Program on the OAC.

- monitoring the Rod Deviation Alarms on the OAC

- evaluating all other available indications of reactor operation, such as reactivity, Tavg, etc.

2. If correct rod motion cannot be verified or no rod movement has f taken place within at least six steps, the operators should stop, place the Rod Control System Bank Selector Switch in l Manual, and contact appropriate technicians to troubleshoot. ,

3. Licensed Operators should continue to perform the Monthly Rod Movement PT. ,

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4. Operators should review Information Notice 93-46 and the t Westinghouse Nuclear Safety Advisory Letter regarding the Salem Rod Control System Failure. These documents will more fully describe the event that occurred at the Salem facility.  !

Based on the above compensatory actions, Duke Power Company feels ,

that the McGuire and Catawba Nuclear Stations are adequately  !

safeguarded against an event similar to the one which occurred at Salem.

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