ANS 19.11-1997 Calculation and Measurement of the Moderator Temperature Coefficient of Reactivity for Water Moderated Power Reactors《水慢化动力反应堆反应性的慢化剂温度系数的计算和测量》.pdf

《ANS 19.11-1997 Calculation and Measurement of the Moderator Temperature Coefficient of Reactivity for Water Moderated Power Reactors《水慢化动力反应堆反应性的慢化剂温度系数的计算和测量》.pdf》由会员分享，可在线阅读，更多相关《ANS 19.11-1997 Calculation and Measurement of the Moderator Temperature Coefficient of Reactivity for Water Moderated Power Reactors《水慢化动力反应堆反应性的慢化剂温度系数的计算和测量》.pdf（28页珍藏版）》请在麦多课文档分享上搜索。

1、ii en tI) z tI) z c( ANSI/ANS-19.11-1997 calculation and measurement of the erator temperature coefficient of reactivity for water moderated power reactors This standard has been reviewed and reaffirmed with the recognition that it may reference other standards and documents that may have been super

2、ceded or withdrawn. The requirements of this document will be met by using the version of the standards and documents referenced herein. It is the responsibility of the user to review each of the references and to determine whether the use of the original references or more recent versions is approp

3、riate for the facility. Variations from the standards and documents referenced in this standard should be evaluated and documented. This standard does not necessarily reflect recent industry initiatives for risk informed decision-making or a graded approach to quality assurance. Users should conside

4、r the use of these industry initiatives in the application of this standard . Secretariat American Nuclear Society Prepared by the American Nuclear Society Standards Committee Working Group ANS-19.11 Published by the American Nuclear Society 555 North Kensington Avenue ANSIIANS-19.11-1997 American N

5、ational Standard Calculation and Measurement of the Modular Temperature Coefficient of Reactivity for Water Moderated Power Reactors La Grange Park, D1inois 60526 USA Approved September 25, 1997 by the American National Standards Institute, Inc. American National Standard Designation of this documen

6、t as an American National Standard attests that the principles of openness and due process have been followed in the approval procedure and that a consensus of those directly and materially affected by the standard has been achieved. This standard was developed under procedures of the Standards Comm

7、ittee of the American Nuclear Society; these procedures are accredited by the Amer ican National Standards Institute, Inc., as meeting the criteria for American National Standards. The consensus committee that approved the standard was balanced to ensure that competent, concerned, and varied interes

8、ts have had an opportunity to participate. An American National Standard is intended to aid industry, consumers, governmental agencies, and general interest groups. Its use is entirely volun tary. The existence of an American National Standard, in and of itself, does not preclude anyone from manufac

9、turing, marketing, purchasing, or using products, processes, or procedures not conforming to the standard. By publication of this standard, the American Nuclear Society does not insure anyone utilizing the standard against liability allegedly arising from or after its use. The content of this standa

10、rd reflects acceptable practice at the time of its approval and publication. Changes, if any, occurring through develop ments in the state of the art, may be considered at the time that the standard is subjected to periodic review. It may be reaffirmed, revised, or withdrawn at any time in accordanc

11、e with established procedures. Users of this standard are cautioned to determine the validity of copies in their possession and to establish that they are of the latest issue. The American Nuclear Society accepts no responsibility for interpretations of this standard made by any individual or by any

12、 ad hoc group of individuals. Requests for interpretation should be sent to the Standards Department at Society Headquarters. Action will be taken to provide appropriate response in accordance with established procedures that ensure consensus on the inter pretation. Comments on this standard are enc

13、ouraged and should be sent to Society Headquarters. Published by American Nuclear Society 555 North Kensington Avenue La Grange Park, Dlinois 60526 USA Copyright 1997 by American Nuclear Society. All rights reserved. Any part of this standard may be quoted. Credit lines should read “Extracted from A

14、merican National Standard ANSIIANS-19.11-1997 with permission of the publisher, the American Nuclear Society: Reproduction prohibited under copyright convention unless written permission is granted by the American Nuclear Society. Printed in the United States of America r Foreword (This Foreword is

15、not a part of American National Standard for the Calculation and Measurement of the Moderator Temperature Coefficient of Reactivity for Water Moderated Power Reactors, ANSIIANS-19.11-1997.) It is the intent of this American National Standard to provide guidance and specify criteria for the calculati

16、on and measurement of the moderator temperature coefficient of reactivity (MTC) in water moderated power reactors. At present, this standard addresses the calculation and measurement of the MTC only in pressurized water reactors (PWRs), because PWRs are the only type of power reactor currently sited

17、 in the United States for which measurement of the MTC is required. This standard was developed by Working Group ANS-19.11 of the American Nuclear Society. During the period when the standard was prepared, the working group had the active participation of the following members: R. D. Mosteller, Chai

18、r, Los Alamos NatUJTI4l Laboratory C. E. Apperson, Jr., Westinglwuse Savannah River Company S. P. Baker, Public Service Electric e IIJp otlom Pfoll (Eq.2) The AFD sometimes is referred to as power im balance. control rod (rod). One or more control mem bers mechanically attached to a single fixture;

19、for the purposes of this standard, the length of the neutron absorber material in the control member is nearly equal to the length of the fuel in the core; i.e., “full length.“ This definition excludes “partial length“ control rods. control rod group (bank). One or more control rods that are inserte

20、d or withdrawn simultane ously during normal operation. differential boron worth (DBW). The change in reactivity per unit change in soluble boron concentration. Doppler temperature coefficient of reac tivity (DTC). The change in reactivity per unit change in the fuel temperature. fuel assembly. A gr

21、ouping of fuel rods or pins which are mechanically or metallurgically joined together in a fixed geometrical arrangement. 2 fuel power. Power deposited directly in the fuel, as opposed to being deposited in the clad ding, moderator, or structural materials. full power. The licensed core thermal powe

22、r level. hot zero power (HZP). A reactor operating state where the core is not producing measurable heat from nuclear fission and the primary coolant system is at the no-load design temperature and pressure. isothermal temperature coefficient ofreac tivity (ITC). The change in reactivity per unit ch

23、ange in the fuel and moderator temperature when the fuel and moderator are at the same temperature. moderator temperature coefficient ofreac tivity (MTC). The change in reactivity per unit change in the average moderator temperature. For the purposes of this standard, the MTC includes all reactivity

24、 effects associated with a change in the average moderator temperature, whether direct or indirect. (An alternative defi nition used in portions of the commercial nuclear industry limits the reactivity effects to those resulting directly from the change in the aver age moderator temperature and the

25、associated change in the moderator density.) NI power. Core power, as indicated by excore nuclear instrumentation (0). partial length control rod. One or more con trol members mechanically attached to a single fixture; for the purposes of this standard, the length of the neutron absorber material in

26、 the control member is a fraction of the length of the fuel in the core; i.e., “partial length.“ power coefficient (PC). The change in reactiv ity per unit change in reactor power, with a con stant average moderator temperature. power Doppler coefficient (PDC). The change in reactivity per unit chan

27、ge in reactor power, with a constant moderator temperature distribution. reactivity. A measure of the deviation of k from unity. Specifically, (Eq.3) reactivity worth. The change in reactivity due to a change in a single parameter (e.g., temper ature or control rod movement): (Eq.4) where kl and are

28、 the effective multiplication factors for reactor states 1 and 2, respectively. Reactivity worth is sometimes given in pcm (percent milli-rho), where Ipcm = 10-5 ap . (Eq.5) regulating group (or control bank). The con trol rod group that is partially inserted in the core during full power operation

29、to provide shaping of the core power distribution and to reduce boron shim requirements. shall, should, and may. The word “shall“ is used to denote a requirement, the word “should“ to denote a recommendation, and the word “may“ to denote permission, neither a requirement nor a recommendation. In ord

30、er to conform with this standard, all requirements shall be satisfied. statepoint. For the purposes of this standard, a statepoint is a condition in which data acquisition is completed before there is a significant change in the value of any operating parameter. technical specifications. For the pur

31、poses of this standard, this term includes both the reactor specific Technical Specifications and the Core Operating Limits Report (COLR). test criterion. The predetermined value for evaluating the result of each test. total temperature coefficient of reactivity (TIC). The sum of the Doppler coeffic

32、ient ofre activity and the moderator temperature coeffi cient of reactivity when the change in the average fuel and average moderator temperatures is the same. worth. See reactivity worth. American National Standard ANSIIANS-19.11-1997 3.2.2 Definition of Acronyms AFD BOC COLR DBW DTC DRW EOC FP HFP

33、 HZP ITC Mpa MTC NI PC pcm PDC ppm psig PWR RCS TrC VCT WD Axial flux difference Beginning of cycle Core operating limits report Differential (soluble) boron worth Doppler temperature coefficient of re activity Differential (control) rod worth End of cycle Full power Hot full power Hot zero power Is

34、othermal temperature coefficient of reactivity Megapasca1s Moderator temperature coefficient of reactivity Out-of-core nuclear instrumentation Power coefficient of reactivity Percent milli-rho (used as a measure of reactivity; see reactivity worth in 3.2.1) Power Doppler coefficient of reactivity Pa

35、rts per million, by weight Pounds per square inch, gauge Pressurized water reactor Reactor coolant system Total temperature coefficient of reac tivity Volume control tank Withdrawn 4. Moderator Temperature and Re activity Although the definition of the MTC might seem obvious, two subtle aspects of t

36、hat definition can lead to confusion or misunderstanding. The first of these aspects is the definition of the “average“ moderator temperature, while the other is the contribution from secondary, indirect effects to changes in the reactivity. 4.1 Moderator Temperature. The MTC was defined in Section

37、3 as the change in reactivity due to a change in the average moderator tem perature. Unfortunately, several different defini tions of the average moderator temperature are used within the commercial nuclear industry. For example, it may be defined in terms of a vol umetric average, an enthalpy avera

38、ge, or simply the average of the inlet and outlet temperatures. The latter definition, for example, is widely used on-site at operating reactors. The methods dis cussed in this standard can be used in conjunc-3 .,-.- -American National Standard ANSIIANS-19.11-1997 tion with any of these definitions.

39、 It is impera tive, however, that the same definition of average moderator temperature be employed when com parisons are made between calculations, meas urements, and limits. 4.2 Moderator Temperature Coefficient of Reactivity. Just as different definitions for the average moderator temperature are

40、used in dif ferent portions of the commercial nuclear indus try, different definitions of the reactivity compo nent of the MTC are used as well. It is generally agreed that the MTC should include the direct re activity effect of the change in moderator temper ature as well as the reactivity effect o

41、f the accom panying change in moderator density. However, there is not universal agreement about whether it also should include reactivity changes related to secondary effects associated with a change in moderator temperature (such as a change in the axial flux distribution). For the purposes of thi

42、s standard, the MTC includes all reactivity effects associated with a change in the average modera tor temperature, whether direct or indirect; i.e., (Eq.l) Most of the methods discussed in this standard can be used in conjunction with either of the rep resentations for the MTC discussed above. It i

43、s imperative, however, that the same definition of MTC be employed when comparisons are made between calculations, measurements, and limits. 5. Calculation of the Moderator Tem perature Coefficient of Reactivity Calculations of the MTC are performed for a number of reasons: (1) To demonstrate that t

44、he calculational model produces results that are consistent with measurements. (2) To provide input for transient calculations. (3) To show that licensing limits are met. The calculations should match the intended use. For example, if the MTC is to be used in transient analyses, it is important that

45、 the definition of the temperature change used in the application be the same as that used in the calculation (see 4.1). 4 As discussed in Section 1, the MTC is a strong function of certain instantaneous operating con ditions. These conditions influence the value of the MTC directly by changing neut

46、ron absorption (such as the change in the soluble boron density that accompanies a change in the moderator density) and indirectly through changes to the neutron energy spectrum. Therefore, the MTC calculation should account for the instantaneous impact on neutron cross sections of the following con

47、ditions: (1) Moderator density (2) Soluble boron concentration (3) Moderator temperature (4) Fuel temperature. The value of the MTC also is affected by the burnup-dependent isotopic composition of the fuel, including the impact of control rods and burnable poisons. Evaluations of the effects of spec

48、tral histories (such as moderator density history and boron history) have shown that they have negligible impact on the MTC. Although modeling of these effects may be desirable for other calculated core parameters, it is not necessary for MTC calcula tions. For an accurate calculation of the total t

49、emper ature coefficient of reactivity (TlC) and the MTC, not only does the reactor itself need to be modeled properly, but the core reflector regions must be accounted for as well. Modeling of the reflector regions by cross sections or albedos should include consideration of the structural material and the density of the moderator in the reflector region. 5.1 Numerics and Convergence. The magni tude of the MTC, at least at beginning of cycle (BOC), typically is small. Consequently, the ac curacy of the calculated MTC can be affected by numerics and convergence. There are