ML20024G590

Forwards Technical Basis for Allowable Rod Worth Specified in Tech Spec, in Response to Verbal Request for Addl Info Re 720922 Application for Amend to License DPR-22 Concerning Control Rod Worth Based on Postulated Rod Drop Accident

ML20024G590
Person / Time
Site:	Monticello
Issue date:	04/11/1973
From:	Mayer L NORTHERN STATES POWER CO.
To:	Giambusso A US ATOMIC ENERGY COMMISSION (AEC)
References
NUDOCS 9102130437
Download: ML20024G590 (23)

Text

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Agalatory F CF NORTHERN STATES POWER COMPANY MIN N E A PO U S MIN N E sOTA 95401 5%

y:W a , 5 April 11, 1973 s'f

/p .d b> 4 0 1 h h\@$- , , ;///h' 6 Mr. A Giambusso Deputy Director for Reactor Projects

.y& N/

Directorate of Licensing A '$'/

United States Atomic Energy Commission d' Washington, D C 20545

Dear Mr. Giambusso:

MONTICELLO NUCLEAR GENERATING PLANT Docket No. 50-263 License No. DPR-22 Supplementary Information on the Rod Drop Accident On September 22, 1972 we requested a change to our Technical Specifications which determines limiting control rod worth based on a postulated rod drop 'i accident. On March 3, 1973 we submitted additional information in answer to your December 28, 1973 questions. Per your verbal request, we are pro-  !

viding you with the attached document prepared by General Electric entitled

" Technical Basis for Allowable Rod Worth Specified in Technical Specifica-tions." You will find this additional information particularly relevant to =aterini previously submitted on the subject.

Yours very truly,

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I L 0 Mayer, P. .

Director of Nuclear Support Services a h LOM/MHV/br /. -

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r-cc: B H Grier G Charnoff lh

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g 4.4 e TECHNICAL BASIS F0R ALLOWABLE R00 WORTH SPECIFIED IN TECHNICAL SPECIFICATI0N i

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_ o o TECHNICAL BASIS FOR ALLOWABLE R00 WORTH SPECIFIED IN TECHNICAL SPECIFICATION 1 INTRODUCTION A topical report and two supplements (1), (2), (3) have been issued in the 17 last year which document new techniques and models being used to analyze

^

the Rod Drop Accident (RDA). The infonnation in these documents have been used for the development of design approaches on new projects to make the consequences of the RDA acceptable to all concerned. In the case of the operating plants where safety analyses and resulting Technical Specifications were previously established with the old approaches, the new information in the topical reports was not easily applied. The purpose of this document is to bridge that gap and provide a technical basis and recommended Technical Specification with the current design basis safety philosophy applied to operating plants in the RDA area, o

II

SUMMARY

& RECOMMENDATIONS Recommendations have been provided to operating plants previously to establish a Technical Specification for a 1.54K maximum allowable worth of in-sequence control rods based on judgement application of recent RDA y work. This document provides supporting detail on how the 1.54 K value could be specifically derived from detailed calculations on a plant-by- l plant basis. However, in view of the fact that this would not be practical '

yr- to do on all plants, a " worst case" comprehensive value of 1.4kK is '

recommended for general and immediate application at all operating plants.

This reconnendation is obtained from a comparison of available specific '

plant calculations, based on operating data, to those used in deriving a l

• (1) NEDO-10527 " Rod Drop Accident Analysis for Large Boiling Water Reactor", C. J. Paone, et al, 3-72 l i

(2) Suppl.1 to Ref.1, 7-72 (3) Suppl. 2 to Ref.1,1-73 l l

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'd L) 280 cal /gm peak fuel enthalpy boundary for the RDA with the key parameters affecting the outcome of the RCA. The 1.4%AK value represents a combination of conservative inputs which are inherently-fixed (e.g. use of the Doppler coefficient corresponding to a Beginning-of-l.ife (BOL) condition, which will always be conservative and judgement inputs which could vary significantly in the future but are not expected to be " worse" than those picked (e.g.

use of a maximum local peaking factor L[P ] of 1.30 for hot startup conditions).

III , DISCUSSION A. Design Basis The design basis for evaluating the consequences of the RDA are described I in the topical reports (pgs. 3/4 of ref. 3). The difference in the application of these bases between the new projects and the operating plants is in the definition of the worst single inadvertent operator error or equipment malfunction to cause the RDA. Previously for new projects and currently for the opereting plants, the Rod Worth Minimizer (RWM) and operator were the redundant controls on rod selection so that a single failure could not cause the drop of an out-of-sequence rod; if the RWM were out of service, a second independent operator was acceptable as a substitute. This has not been accepted on new projects e and a third system, the Rod Sequence Control System (RSCS), has been applied. Since this new system is not operative beyond the 50% rod density point, the design basis for new projects has shifted so that the drop of an out-of-sequence rod at that point is analyzed. If it cannot be assumed that the RWM or operator will prevent the selection of an out-of-sequence rod, then the worst case accident for new projects becomes the drop of an out-of-sequence rod at the point where the RSCS is no longer operative.

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I Since the contents of the topical report supplements were developed ll in conjunction with the new design basis on new projects, it became ,j necessary to review and provide other means for applying the new RDA results to the current Technical Specification application on operating )

plants. i.e.. The current Technical Specifications on operating plants are applied on the basis that the maximum reactivity value of any in-sequence rod must be limited in order to maintain the consequences of a RDA within those analyzed and accepted. The topical reports also covered only particular plants at particular reactivity / exposure conditions, and since this added more variable parameters to an analysis that already contained many variables, it became necessary to develop worst case values that would assuredly cover a wide range of conditions.

In this case, available data from calculations performed for particular operating plants and conditions was compared with the same parameters used in calculating RDA consequences for the topical reports. It was found that the TVA Beginning-of-Life (BOL) data described in ref. (2) was suitable as a worst case encompassing operating plant data and as a means of comparison. These parameters and comparisons are described in detail below.

B. Parameters Considered & Design Assumptions Used e

Although there are many input parameters to the rod drop accident f

analysis, theresultant peak fuel enthalpy is most sensitive to the 4

i following input parameters:

7

1. Steady state accident reactivity shape function k 2. Tctal control rod reactivity worth

3. Maximum inter-assembly local power peaking factor (PL -n rmalized 9

5 over four bundles)

'h 4. Delayed neutron fraction Q

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5. Scram reactivity shape function

6. Doppler reactivity feedback

7. Moderator temperature For a fixed control rod drop velocity and scram insertion rate, these parameters can be varied and combined to yield a peak fuel enthalpy of 280 cal /gm. This was done using the data developed for the TVA BOLcasesinref.(2).

Rod drop velocity was assumed to be that justified by the statistical evaluation in the appendix of Ref. (1) 1.e., the average measured value plus three standard deviations was used. Also, the current standard Technical Specification scram times tabulated below were used in developing the scram reactivity curves for the 280 cal /gm design limit boundary corresponding to the third basi c condition specified below:

% of Rod Insertion Time from De-Energization of Scram Solenoid Valve (sec.)

1 l

5 0.475 20 1.10

~

50 2.0 90 5.0 In order to meet the RDA design limit of 280 cal /gm the above parameters I b'

+ are combined to meet three basic conditions. These are (A) the accident

) reactivity characteristics, (B) the Doppler reactivity feedback, and

[ (C) the scram reactivity feedback. If any one of these conditions are

• not satisfied, then a more detailed analysis would have to be perfonned

j. to establish compliance with the 280 cal /gm design limit.

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O O I Y

C. Three Basic Conditions

1. Accident Reactivity Characteristics - Accident reactivity shape function total control rod reactivity worth, inter-assembiv local power peaking factor, and the delayed neutron fraction The sensitivity of the rod drop accident to the first three parameters at cold startup and hot startup are shwon by Figures 1 and 2 and the effect of the delayed neutron fraction (beta) can be seen by comparing Figures 1 and 2 with Figures 3 and 4 respectively. To determine whether or not a specific condition will meet the 280 cal /gm design limit at cold startup or hot startup, the accident reactivity characteristics (i.e.,

accident shape function, local peaking, etc.) for the plant being analyzed should be matched to those presented in Figures 1 4 through 5. If the accident reactivity characteristic curves are equal to or les:1 than those shown as solid lines in Figures 1 through 4 then one of the three conditions needed to conserva-tively ensure RDA peak fuel enthalpy equal to or less than 280 cal /gm is satisfied. If the actual plant accident reactivity characteristics are greater, a more detailed analysis would have to be performed.

When applying these functions a linear interpolation can be employed to determine intermediate points with regards to the local L peaking factor and beta variables.

Some example curves resulting from calculations with operating 1 plant data is also plotted as dotted lines on Figures 3 and 4 c

to demonstrate compliance with the condition, including the one 47 with the highest K,ff, Other data (not plotted to avoid confusion) is shown in Table 1. Comparisons have been made on Figures 3 and

{ y 4 because the betas most closely coincide. The beta for Figures q

ff

- ( ,l eo l n- ,

F 1 and 2 correspond to Beginning-of-Life (BOL) conditions which no longer exist for operating plants. Although the betas associated with the operating plant curves are not precisely the same as the value used for the 280 cal /gm boundary curves, the differences are in the conservative direction, i.e., as shown in Table I, betas for operating plant conditions are generally higher than those used in Figures 3 and 4 for the 280 cal /gm boundary curves, thus allowing higher PL 's or rod worths within the boundary.

A typical plant local peaking factor map is shown in Figure 8. As can be seen the maximum value on this map is 1.217. While this is not the maximum that could be expected for a hot startup condition, values above 1.30 would not be expected to occur at any plant.

Actual maximum local peaking factors (PL ) would be expected to g be slightly higher in the cold startup condition than in the hot w startup condition; however, as can be seen by comparison of Figures 3 and 4, a substantially higher Pt can be tolerated for cold startup conditions at the 280 cal /gm boundary, other conditions e being equal. Thus, in reviewing the compensating factors involved, it is apparent that the " worst case", or lowest rod K,ff allowable at the 280 cal /gm boundary would be represented by the solid curves (r

in Figure 4, which are for the hot startup condition with the minimum G beta.

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2. Doppler Reactivity Feedback g;

V E, p The Doppler reactivity coefficients used for these analyses to identify a 28 cal /gm boundary were held fixed at the beginning g of life (BOL) condition. The Doppler reactivity coefficients M for the cold and hot startup conditions are presented in Figure S.

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If the Doppler reactivity coefficeints are equal to or more ,

negative than those given as solid lines in Figure 5, then another I one of the three conditions needed to conservatively ensure RDA i peak fuel enthalpy 280 cal /gm is satisfied.

i Using the BOL Doppler reactivity coefficient will be conservative since the Doppler coefficient always becomes more negative with increasing exposure. This effect is typically demonstrated by '

the exposed core data shown as dotted lines on Figure 5, and is due primarily to the Pu-240 buildup and contribution as a function of exposure.

3. Scram Reactivity Feedback The scram reactivity feedback function is unique in that the total scram feedback is not required to tenninate the accident and limit peak fuel enthalpy in the time scale of interest. The combined Doppler and .01 Ak scram will be more than sufficient to terminate the accident and bring the reactor core suberitical for control rod worths of interest. This is not meant to imply that total scram is not required for complete shutdown but rather to emphasize the fact that partial scram bank insertion would be sufficient to limit the resultant RDA peak fuel enthalpy to 280 cal /gm in the time scale ,

of interest. Therefore, up to .01 Ak, the actual plant scram reactivity feedback function must be equal to or greater than the

[ data presented in Figures 6 and 7 for the cold and hot startup operating states respectively in order to satisfy the third of

• the three conditions needed to conservatively ensure RDA peak ,

fuel enthalpy 7 280 cal /gm, ,

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- A typical example derived from operating plant data is also plotted on these figures as dotted lines to demonstrate that the Londition is met in actual scram perfomance. Additional available data was not plotted to avoid graphic confusion, but is summarized with total scram worths in Table I.

O. Application of the 280 cal /gm Boundary In summary, all three conditions 1, 2, and 3, as stated above, must be satisfied in order to conservatively stay within the 280 cal /gm design limit boundary. If any of the conditions are not met then a more detailed calculation would have to be performed to demonstrate compliance with the design limit.

Likewise, given a particular set of conditions, a maximum rod worth could be determined which could show compliance with a Technical CN Specification based on keeping RDA consequences below the peak fuel (1/ enthalpy design limit of 280 cal /gm.

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i As an example, assume the following conditions:

/ w; . Hot startup

k. > . beta = .0055 isRP .P L = 1.20 ce

. Doppler coefficient = Figure 5 solid curve for hot startup

. Scram reactivity = Figure 7 solid curve

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. Accident reactivity shape = Figure 2 and 4 solid curves  ;

7; For the above conditions linear interpolation between Figures 2 and a  !

7 4 show that a rod worth of .01514 Ak will satisfy the 280 cal /gm design y limit. This example is conservative since the BOL Doppler feedback has

< been coupled with a typical end of cycle delcyed neutron fraction.

.c.m Therefore, for an operating reactor with scram and accident reactivity 5

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, p characteristics equal to or better than those described above, a '

'[ .015 Ak Technical Specification on allowable rod worth is justifiable.

t It is important to recognize that there is no practical way to calculate all possible conditions or parametric values as they may occur during the cycle at a particular plant or plants. However, J

some calculations have been performed to obtain typical values as g.

shown in this document and judgement can be exercised to obtain n worst cases or perceive the effects of variatiuns. On this basis, it would be reasonable to pick some worst case values of the key parameters in the RDA based on the approaches used in this document  ;

and derive a rod worth for Technical Specification application that could be widely used without recourse to lengthy repetitive calculations for each reactor and each fuel cycle, i Such a process was conducted in the course of preparing this document,

1. with the following results:

1. Scram reactivity condition: While there could be significant variation in the shape and total worth of the scram reactivity g curve, actual operation in the future is not likely to degrade y down to the point where the net effect on a RDA calculation would 4 be any less than that represented by the solid curves of Figures

6 and 7.

& 2. Doppler reactivity condition: The lease effective (BOL) Doppler E feedback has been assumed in the 280 cal /gm boundary cases Q' calculated for this document and it would be simplest to maintain

@~ this assumption in deriving a comprehensive Technical Specification e- applicetion. This conservatism would also serve to compensate D for any concern in other areas where variations beyond the 280

[

4 cal /gm boundary might be postulated in extreme situations.

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3. Accident reactivity charactLristic condition: If it is assumed that the 280 cal /gm boundary conditions established in 1. & 2.

above represent worst case values that no operating plants are likely to exceed, then selection cf a recomended comprehensive

- Technical Specification on maximum allowable rod worth reduces to a consideration of the parameters associated with the accident reactivity characteristics discussed in C.1. above. There are four parameters considered for this 280 cal /gm boundary condition and it was established in C.1 that the closest approach of actual plant operating paramaters to this 280 cal /gm boundary was represented by Figure 4. It was also established that two of the parameters, the accident reactivity shape function and beta, derived from any actual plant operating data, generally could not reach those used in calculating the 280 cal /gm boundary shown in Figure 4. Thus, the maximum allowable rod worth can be derived by determining the maximumg P in the hot startup condition and using the corresponding solid curve. As stated in C.1, a P L above 1.30 would not be expected at any plant and a maximum allowable rod w6rth would therefore, be 1.44 K. This value is ,

0 recomended for comprehensive Technical Specification application i 4 on a " worst case" basis in the absence of specific detailed calculations on each operating plant.

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TA BLE I

' i TYPICAL RELOAD OPERATING CORES NUCLEAR DATA k A. In-Sequence Control Rod Worth PLANT CONDITION POINT IN 51AX. AK egg CYCLE _

1 A Cold SU BOC 0.007 B Cold SU BOC 0.011 B Cold SU EOC 0.003 C Cold SU DOC 0.005 B Hot SU BOC 0.003 C Hot SU BOC 0.005 B. Senun Bank Worth * ,

PLANT CONDITION POINT IN TOTAL NEG.

CYCLE A Koff.

A Cold SU BOC 0.071 B Cold SU BOC 0.049  !

B Cold SU EOC 0.051 A Hot SU BOC 0.131 B Hot SU BOC 0.125 B Hot SU EOC 0.121 D Hot SU BOC 0.147 D llot SU MOC 0.143 D Hot SU EOC 0.141

• Minus the dropping rod in the RDA

W' l$;&

l (Continued)

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. ., TABLEI TYPICAL RELOAD OPERATING CORES NUCLEAR DATA Li

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A C. Delayed Neutron Fraction (a)

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CONDITION POINT IN BETA G .' PLANT CYCLE g-Ilot SU BOC 0.0059

. A A Ilot SU EOC 0.0054 B Hot SU BOC 0.0059 Hot SU EOC 0.0054 B

,$ llot SU BOC 0.0060 C

llot SU EOC 0.0056 C

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