ML20153H144

Topical Rept Evaluation of XN-NF-825(P), BWR/6 Generic Rod Withdrawal Analysis,Mcprp. Rept Acceptable for Ref in License Applications

ML20153H144
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Issue date:	12/11/1985
From:	Office of Nuclear Reactor Regulation
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ML20153H131	List: ML20153H127 ML20153H144
References
TAC-57460, NUDOCS 8602280658
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pEc 111985 EVALUATION OF EXXON TOPICAL REPORT XN-NF-825(P)

(TACS 57460)

By letter dated April 8, 1985, Exxon Nuclear Company (ENC) submitted a proprietary Licensing Topical Report (LTR) XN-NF-825, "8WR/6 Generic Rod  ;

Withdrawal Analysis, MCPRp", for staff review. The report describes the ENC analysis of the rod withdrawal event (RWE) for reactors (such as the BWR/6) which contain a Rod Withdrawal Limiter to limit the consequences of the "at power" rod withdrawal. We have reviewed the document and prepared the following evaluation.

1. Description of Report After an introduction (Section 1) and summary (Section 2) the report presents a brief overview (Section 3) of the protection systems for the rod withdrawal transient. The Rod Block Monitor (BWR/2-5) and the Rod Withdrawal Limiter (BWR/6) are described. It is noted that the analysis of core behavior during a rod withdrawal event is the same for both protection systems. The Technical Specifications for use with the Rod Withdrawal Limiter are described and typical curves are presented.

Section 4 of the report describes the methodology used to perform the statistical analysis required to demonstrate at a 95% percent confidence level that MCPR safety limits are not violated for at least 95 percent of the rod withdrawal events. The choice of initial conditions for the event simulations is described and the methods and procedures used to perform the simulations are discussed.

Section 5 contains the results of the analyses. Section 6 presents the statistical methodology employed. The application of the generic analyses to reload support is discussed in Section 7. _,.-

8602280650 DR 860214 TOPRP ENVEXXN PDR ENCLOSURE

2. Description of Evaluation The Exxon generic analysis of the rod withdrawal event assumes the presence of the Technical Specification on rod withdrawal limits as a function of power which have been established by the reactor designer (General Electric). These limits are one foot for core powers greater than 70 percent of rated power and .

two feet for core powers between 20 and 70 percent of rated. The analyses are then performed to establish the values of operating limit MCPR as a function of core power which are required to assure that the Safety Limit MCPR will not be violated for the rod withdrawal event. This is an acceptable approach.

The calculational methods and procedures used to simulate the rod withdrawal event are essentially the same as those described in Reference 1 which has been approved by the staff (Reference 2). The only difference from the methods of Reference 1 occurs because of the different means of effecting the rod block. This block is triggered by local power rise in the analyses of Reference 1 and by a fixed rod withdrawal distance in the BWR/6 analysis. This difference does not affect the calculation of the change in MCPR as a function of rod withdrawal. We thus conclude that the methods used to simulate the rod withdrawal event are acceptable.

A data base of initial conditions was constructed by choosing actual operating states obtained for a BWR/6 reactor. These states were chosen so that the entire power-flow map of the reactor was represented and the range of the

_ control rod patterns throughout the cycle was included. Emphasis was placed on states that are more of ten represented (e.g. , those new full flow-full power) and 37 statepoints were chosen.

The next step in the process consists of simulating the withdrawal of each gang of control rods in the data base in two one-foot incret.ents and obtaining the resulting Critical Power Ratio for each fuel assembly in the core at each increment. Calculations were performed to show that ganged withdrawals are always more limiting than single rod withdrawals. This result is consistent with previous analyses and is acceptable. .

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The core AMCPR values for each foot of control rod withdrawal are obtained by taking the difference between the initial and final state core MCPR values. A .

statistical analysis is then performed to determine the 95/95 generic rod withdrawal error transient limits. These calculations were performed with the Exxon XTGBWR code which has been approved (Reference 2) for BWR. core simulation. . ,

The statistical evaluation of the 95/95 value of the expected initial MCPR values is straightforward. The values of the mean expected MCPR as a function of power are well above the typical Technical Specificaticn limit for this quantity. Adherence to this limit is not expected to result in operational inconvenience.

The statistical analysis of the AMCPR data is complicated by the presence of zero values among the data base which invalidate the assumption of normality of the random variable. The pr'oblem was resolved by regarding AMCPR=0 as a censoring point with positive values of AMCPR representing that portion of the random distribution which is observable. This is an acceptable treatment of the statistics.

The operating limit MCPR required by the Rod Withdrawai Error transient may be obtained by adding the AMCPR value as a function of power to the safety limit value. This results in a value of 1.13 at 100 percent power, for example.

The typical Technical Specification value at 100 percent power is 1.18-1.20.

Similar margin exists at other powers.

The values of MCPR operating limit (OLMCPR) as a function of power are typically determined by other events (e.g., LOCA and flow-run out events).

Section 7 of the present topical report presents a curve of OLMCPR as a function of power which is required for the RWE. This represents a lower f limit for this quantity. As noted above considerable margin exists between this curve and the current Technical Specification curve.

The results described above are valid for operation within the power flow, domain illustrated in Figure 4.1 of the present report and for theifuel management scheme used for determining the operating states of the data base.

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Use of other power flow domains (e.g. , the Maximum Extended Operating Domain-ME00) or other fuel management schemes (e.g., the single sequence loading pattern) will require verification by analysis that the conclusions of this report are valid.

3. Regulatory Position -

Based on our review which is described above we conclude that Exxon Topical Report XN-NF-825 may be used as a reference to the analysis of the Rod Withdrawal Error transient for BWR/6 reloads. We further conclude that a (

cycle specific analysis of this event is not required provided that the operating power-flow region is bounded by that used in the analysis reported in the present topical report and the core loading and control rod patterns are consistent with those used in the data base. Reloads which do not meet these provisions will need analyses to confirm the conclusions reached in the present topical report.

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6 REFERENCES l

1. XN-NF-20-19(P), Including Supplements 1 and 2, " Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Design and Analysis, May, 1981.

2. Letter to G. F. Owsley, Exxon Nuclear Company, from J. R. Miller, NRC dated April 7, 1982.

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