ML19221B063
| ML19221B063 | |
| Person / Time | |
|---|---|
| Issue date: | 11/24/1975 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-75-087, NUREG-75-087-04.4, NUREG-75-87, NUREG-75-87-4.4, SRP-04.04, SRP-4.04, NUDOCS 7907120340 | |
| Download: ML19221B063 (11) | |
Text
/ga rec NU R EG-75/087
+
o dy },,7i j STANDARD REVIEW PLAN U.S. NUCLEAR REGULATORY COMMISSION
'%..,* /
OFFICE OF NUCLEAR REACTOR REGULATION SECTION 4.4 THERMAL AND HYDRAULIC DESIGN PEVIEW RESPONSIBILITIES _
Primary - Reactor Systems Branch (RSB)
Secondary - Core Perfomance Branch (CFB)
I.
AREAS OF REVIEW The objectives of the review are to confim that the thermal and hydraulic design of the core and the reactor coolant system (RCS) has been acconplished using acceptable analytical methods, is equivalent to or is a justified extrapolation from proven designs; provides acceptable margins of safety from conditions which would lead to fuel damage during nomal reactor operation and anticipat<< operational transients; and is not susceptible to thermal-hydraulic instability. This plan describes the normal review of themal and hydraulic design, i.e., that for a plant similar in core and primary coolant system design to previously reviewed plants. The review of new prototype plants or new design methods require in addition the independent audit analyses discussed in the appendix to this plan.
The review includes evaluation of the propnsed techr ' cal specifications regarding safety limits and limiting safety system settings, to ascertain that these are consistent with the power-flow operating map for boiling water reactor (BWR) plants or the temperature-power operating map for pressurized water reactor (PWR) plants.
The review also includes detemination of the largest hydraulic loads on core and reactor coolant system cowonents during normal operation and postulated accident conditions. This infemation is provided to the Mechanical Engineering Branch for use in the review of reactor components and structures.
To accomplish the cbjectives, the reviewer examines features of core and RCS components, key process variables for the coolant system, calculated paraTters characterizing themal perfomance, data serving to support new correlations or changes in accepted correlations, and assumptions in the equations and solution techniques used in the analyses. The reviewer determines that the applicant has used approved analysis methods in the manner specified by topical reports describing the rethods and by staf f reports approving the methods. If an applicant has used previously unapproved correlations or analysis methnds, the reviewer initiates a generic evaluation.
USNRC STAND ARD REVIEW PLAN stende,d iew,=n. e,e,, eye,.d for two g ance of the o++6ce of Nwaeor n.e. tot nego ee on seee* r pone. bee for in. re ew of ec,4.cet.on. to construct and operete awcher power paonte These documente er, mede evadebee to the pwbhc es part of the Com miseson a pokey to 6pform the nucteer 6adostry end the generel pubhc of regw6etory procedures end pnhtsee Eteriderd review p6ene are not authtstutes for regasetory gusdee er the Commiseson a regutetsono and Con %pitenCe wfth theeft is rhet requered The standard revoew plan eertione are b eyed to Aewre.on 2 of the Etenderd Formet end Content of Safety Anelvese Reporte for Nwcieer Poese Plante Not en eactig ge of tne Stenderd Formet beve e corroependeng reweew plan Published stenderd rev6ew p4 ens verit be rewteed pe<6odice6fy. es approprSete to secommodate commente end to reflec+ new informet6on and espertence Commerne end suggaetteae for improvemeat wm be cone.dered and show6d be sont to the U $ Nedear peguietory Commessaan Office of Nwdeer R$erter Regweefsen VWeebngton D C 20ME
,.o 7907120340
The CPB, as described in Standard Review Plan (SRP) 4.3, provides technical ccasultati: > on matters related to core physic calculations and their integration with power distribution assumptions made for the core thermal and hydraulic analysis. The CPB, on request, partici-pates in generic evaluation of new thermal and hydraulic analysis methods.
II.
ACCEPTANCE CRITERIA 1.
General Design Criterion 10 (Ref.1) requires that the reactor core be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of ar.ticipated operational occurrences.
There are two acceptable approaches to meet this criterion:
For departure from nucleate boiling ratio (DNBR) or critical heat flux ratio (CHFR) a.
correlations there should be a 95% probability, at the 95% confidence level, that no fuel rod in the core experiences a departure from nucleate boiling condition during norral operation or transients that are anticipated to occur with moderate frequency, b.
For critical power ratio (CPR) correlations, the limiting (minimum) value of CPR is to be established such that 99.97 of the fuel rods in the core would not be expected to experience boiling transition during core wide transients. For transients that effect only a portion of the core, the same value of CPR is u:eJ to provide additional conservatism.
Correlations of critical heat flux are continua ly being revised as a resu' of additional experimental data, changes in fuel assembly design, and improved calcula-tional techniques involving coolant mixing ard the effect of axial power distributior.i.
As guidance to the reviewer, the correlations listed below have been found acceptable for previously reviewed plants.
a.
BWR's - The minimum CHFR calculatej with the Hench - Levy correlation (Ref. 8) should exceed 1.0 at all times. The value of the ninimum CPR calculated with the GETAB analysis will vary for different plants and/or product lines. Typically, the value will exceed 1.C6, b.
PWR's - For 14 x 14 or 15 x 15 rod arrays the minimum CNBR calculated with due allowance for mixing grids (Refs. 3, 4, and 5) should exceed 1.32 using the BAW-2 correlation (Ref. 6) and 1.30 using the W-3 correlation (Ref. 7).
2.
As problems affecting DNBR or CPR limits arise, such as fuel densifit.otivn ur rud buw-ing, experimental and/or analytical methods for determining the appropriate design penalties are included in the review. Subchannel hydraulic analysis codes such as those described in References 8, 9, and 10 should be used to calculate local fluid conditions within fuel-assemblies for use in PWR DNB correlations. The acceptability of such codes must be demonstrated by measurements made in large lattice expt-iments 4.4 -2 147 019-
or power reactor cores. Calculations of BWR fluid conditions for use in CHF correlations have been in accordance with the models specified in Reference 11.
3.
The maximum value of the linear heat oeneration rate (LHGR) anywhere in the core, including all hot spot and hot channel factors, should be such that the centerline temperature of the f uel is below the melting temperature. For most core designs, full power steady-state operation is not the operating node which is most limiting in regard to LHGR. Rather, ECC3 performance following a postulated loss-of-coolant accident or various anticipated transients is more limiting. As guidance for the reviewer, the following values of LHbR have been found acceptable for previously reviewed designs:
LHGR (kW/ft)
BWR P_WR 17.5 kW/ft for 1965 product line 18.5 kW/ft for 15 x 15 array 18.5 kW/f t for 1967 product line 18.5 kW/f t for 14 x 14 array 18.5 kW/ft for 1969 product line 13.0 kW/f t for 17 x 17 array 13.4 kW/ft for BWR-6 13.0 kW/ft for 16 v 16 array While these values do not constitute criteria for acceptance, any design in which they are exceeded nust be supported by sufficient analysis to demonstrate that all acceptance criteria are ret. Ctter operating conditions such as fuel densification nay reduce thase values.
4.
For PWR and BWR fuel, tne maximum clad st ain calculated for operational transients and at end-of-life should be less than one percent. These analyses should consider the pressure associated witi gaseous fission products.
5.
The reactor should be demonstrated to be frae cf undamped oscillations or other hydrau-lic instabilities for all conditions of steady-state operation, fur a:1 operational transients, for all load-following maneuvers, and for partial loop operation. Typical nethodologies are described in References 12 and 13.
6.
Methods for calculating single-phase and two-phase fluid flow in the RCS piping and other components should include classical fluid nechanics relationships and appropriate empirical correlations. For components of unusual geometry, such as the following, these relationships should be confirred empirically:
a.
Steam generator (Re' 14).
b.
Reactor vessel (Ref. 15).
c.
Jet pump (Ref.16).
d.
Core flow c utribution (Refs. 15, 17, and 18).
4.4-3
7.
The proposed technical specifications should be established such that the plant can be safely operated at steady state conditions under all of the expected combinations of The safety limits and limiting safety settings must be established system parameti ;.
for each parameter, or combinations of parameters, such that acceptance criterion 1, above, is satisfied. The limiting csnditions of operation must provide app.'opriate operating restrictions. For example, the limiting conditions of operation must assure that the reacter coolant pumps have adequate net positive suction head for all expected modes of operation.
8.
Any changes to accepted codes, correlations, and analytical procedures, or the addition of new ones must be juetified on theoretical or e..?i. cal grounds.
9.
Preoperational and initial startup test programs should follow the recommendations of Regulatory Guide 1,68 (Ref.19), as regards measurements, and confirmation of themal hydraulic design aspects.
III.
REVIEW PROCEDURES The procedures below are used during the construction permit (CP) review to assure that the design criteria and bases and the preliminary design as set forth in the preliminary safety analysis report meet the acceptance criteria given in Section II of this review plan. For operating license (UL) applications, the procedures are utilized to verify that the initial design criteria and bases have been appropriately imolemented in the final design as set forth in the final safety analysis report. The OL review also includes the proposed technical specifications, to assure that they are adequate in regard to safety limits, limiting safety system settings, ard conditions of operation.
The reviewer must begin with an understanding of currently acceptable thermal and hydraulic design practice for the reactor type under review. This understanding can be most readily gained from topical reports describing CHF correlations. system hydraulic models and tests, and core subchannel analysis methods; from standard texts and other technical literature which establish the methodology and the nomenclature of this technology; and from documents which sumarize current staff positions concerning acceptable design methods.
Much of the review described below is generic in nature and is not performed for each plant. That is, the RSB reviewer is to compare the core design and operating parameters to those of previously reviewed plants. He then devotes the major portior, of his review effort to those areas where the application is not identical to previously reviewed plants.
The reviewer is to comoare the infomation in the applicant's safety analysis report (SAR) to the documents referenced bv the applicant or in this plan to determine conformance to the bounds established by such documents. The reviewer must confirm that void, pressure drop, and heat transfer correlations used to estimate fluid conditions (flow, pressure, quality) are within the ranges of applicability specified by their authors or in previous staff reviews, that the analysis methods are used in the nanner specified by the developers or in previous staff reviews, that the reactor design falls within the ranges of applicability 4.4-4 147 021
specified for accepted analysis methods, and that the design is within the criteria specified in II, above, and is not an unexplained or unwarranted extrapolation of other themal-hydraulic designs.
The review does not routinely involve calculations by the staff. On occasion, e i., if a new model or correlation is proposed, independent analyses are perfomed by the staff or by consultants under the direction of the RSB. These analyses establish the range of appli-cability and associated accuracy of the new model or correlation and the reviewer ensures it is applied accordingly.
The reviewa is to establish that the themal-hydraulic design and its characterization by MCHFR or DNBR have been accomplished and are presented in a manner which accounts for all possible reactor operating states as detemined from operating maps. In this regard, the reviewer must confim, with the aid of the CPB, that the power distribution assumptions of SAR Section 4.4 are a conservative (i.e. worst-case) accounting of the power distributions derived in Section 4.3 from core physics analyses, and that the latter analyses include an acceptable calculation of local void fractions. He must also confim that the mass flux used in these calculations takes into account the core flow distribution (including that for partial loop operation) and the worst case of core bypass flow. The reviewer confims that the primary coolant flow 'ange shown in the operating map will be verified by pre-startup measurements.
The reviewer ensures that adequate account is taken of the effect of crud in the primary e_
coolant system, such as in the calculation of CHF in the core, heat transfer in the steam gencrators, and pressure drop throughout the RCS.
The reviewer is to examine the calculation of hydraulic loads for nomal operations and postulated accidents, ensure they are accurately estimated for the worst cases, and supply the worst case values to the Mechanical Ergineering Branch for their review of reactor components and supports.
The reviewer should be aware of the vibration and loose-parts monitoring equipment and procedures used on other comparable plants and, taking into account pertinent differences, ensure that an adequate system is provided for the plant under review.
The applicant's proposed preoperational and initial startup test programs are reviewed to determine that they are consistent with the intent of Regulatory Guide 1.68 (Ref. 19). At the OL stage, the reviewer is to assure that sufficient infomation is provided by the applicant to identify clearly the test objectives, methods of testing, and acceptance criteria. (See par. C.2.b of Reference 19.)
The reviewer evaluates the proposed test programs to detemine if they provide reasonable a surance that the core and reactor coolant system will satisfy functional requirements.
As an alternative to this detailed evaluation, the reviewer may compare the core and reactor 8
coolant system design to that of previously reviewed plants. If the design is essentially 4.4-5
identical and if the proposed test programs are essentially the same as perfomed previously on other plants, the reviewer may conclude that the proposed test programs are adequate for the core and reactor coolant system.
If the core or the reactor coolant system differs significantly from that of previously reviewed designs, the impact of the proposed changes on the preoperational and initial startup testing programs are reviewed at the construction permit stage. This effort should particularly evaluate tne need for any special design features required to perform acceptable test programs.
The proposed technical specifications that relate to the core and the reactor coolant system are evaluated. This evaluation is to cover all of the safety limits and bases that could affect the thermal and hydraulic perfomance of the core. The limiting safety system settings are reviewed to ascertain that acccptable margins exist between the values at which reactor trip occurs automatically for each parameter (or combinations of parameters) and the safe'.y limits. The reviewer confirms that the limiting safety system settings and limiting con-ditions for operation, as they relt.te to the reactor coolant system, do not permit operation with any expected combination of parameters that would not satisfy criterion 1 of Section II.
IV.
EVALUATION FINDINGS The reviewer verifies that the SAR contains sufficient information and his review supports the following kinds of statenents and conclusions, which should be included in the staff's safety evaluation report:
9 "The thermal-hydraulic design of the core for the _ _
plant was reviewed. The scope of review included the design criteria, preliminary ccre design, and the steady-d-
elic perfomance.* The review concentrated on state analysis of the core thermal-hy the differences between the proposed cv.e design (and criteria) and those designs and criteria that have been previously reviewed and found acceptable by the staff. It was found that all such differences were satisfactorily justified by the applicant.
The applicant's themal-hydraulic analyses were perfor.ned using analytical methods and correlations that have been previously reviewed by the staff and found acceptable.
"The staff concludes that the thermal-hydraulic design of the core conforns to the Comission's regulations and to applicable regulatory guides and staff technical positions and is acceptable."
V.
REFERENCES 1.
10 CFR Part 50, Appendix A, General Design Criterion 10. " Reactor Design."
2.
" General Electric BWR Thermal Analysis Basis (GETAB): Data, Correlation and Design Application," NED0-10958, General Electric Company (1973).
- For an OL review this sentence should be modified to include the implementation of the design criteria as represented by the final core design.
t7 023 4.4-6 1
3.
F. F. Cadek, F. E. Motley, and D. P. Dominicis, "Effect of Axn acing on Interchannel Thermal Mixing with the R Mixing Vane Grid, WCAP-7941-L (propr
.ary), Westinghouse Eleccric Corporation, June 1972.
E. Motley and F. F. Cadek, "0NB Test Results for New Mixing Vane Grids (R)," WCAP-4.
7695-L (proprietary) Westinghouse Electric Corporation, July 1972.
5.
F. E. Motley and F. F. Cadek, " Application of Modified Spacer Factor to L Grid Typical and Cold Wall Cell DNB," WCAP-7988, Westinghouse Eicctric Corporation, October 1972.
(See also WCAP-8030.)
6.
S. Gellerstedt, R. A. Lee, W. J. Oberjohn, R. H. Wilson, and L. J. Stanek, "Cor-c.
relation of Critical Heat Flux in a Bundle Cooled by Pressurized Water," in "Two-Phase Flow and Heat Transfer in Rod Bundles," American Society of Mechanical Engineers, New York (1969). (See also BAW-10000 and BAW-10036.)
7.
L. S. Tong, " Prediction of Departure from Nucleate Boiling for an Axially Non-Unifom Heat Flux Distribution," Journal of Nuclear Energy, Vol. 21, 241-248 (1967).
8.
" TEMP - Themal Enthalpy Mixing Program," BAW-10021, Babcock and Wilcox Company, April 1970.
9.
H. Chelener, P. T. Chu, and L. E. Hochreite, "THINC-IV-An Irrproved Program for Themal-Hydraulic Analysis of Rod Bundle Cores," WCAP-7956, Westinghouse Electric Corpc ation, June 1973 (under review). (See also WCAP-7359-L and WCAP-7838.)
10.
"Systen 80 Standard Safety Analysis Report (CESSAR)," Combustion Engineering, Inc.,
August 1973 (under review).
11.
J. C. Slifer and J. E. Hench, " Loss of Coolant Accident and Emergency Core Cooling Models for General Electric Boiling Water Reactors," NEDO-10329, Appendix C, General Electric Company, April 1971.
12.
L. A. Carmichael and G. J. Scatena, " Stability and Dynamic Perfomance of
'e General Electrf Boiling Water Reactors," APEG-5652, General Electric Company, April 1969.
13.
R. L. Rosar. thal, "An Experimental Investigation of the Effect of Open Channel Flow on Thermal-Hydrodynamic Flow Instability," WCAP-7966, Westinghouse Electric Corporation, October 1968.
14.
B. N. Mcdonald, R. C. Post, and J. 5. Scearce, "Once Through Steam Generator Research and Development Report," BAW-10027, Suppl. 1 (non-proprietary version of BAW-10002),
Babcock and Wilcox Company, April 1971.
15.
B. S. Mullanax, R. J. Walker and B. A. Karrasch, " Reactor Vessel Model Flow Tests,"
BAW-10037 (non-proprietary version of BAW-100l?), Revision 2, Babcock and Wilcox Company, September 19.8.
024 7
4.4-7 i
16.
" Design and Perfomance of General Electric Boiling Water Reactor Jet Pumps," APED-5460, General Electric Company, September 1968.
17.
H. T. Kim, " Core Flow Distribution in a Modern Boiling Watar Reactor as Measured in Monticello," NED0-10299, General Electric Company, January 1971.
18.
F. D. Carter, " Inlet Orificing of Open PWR Cores," WCAP-9004, Westinghcuse Electric Corporation, January 1959.
19.
Regulatory Guide No.1.68, "Preoperational and Initial Startup Test Programs for Water-Cooled Power Reactors."
O 4.4-8
APPENDIX STANDARD REVIEW PLAN 4.4 INDEPENDENT AUDIT ANALYSIS The Lore Performance Branch may be requested to perform independent analyses of thermal-hydraulic and physics phenomena for both steady-state and transient conditions. These analyses may be requested by various technical groups within thE< '.aff.
The required analyses may be in the fc!!cd r.; fc:= :
1.
Independent computer calculations to substantiate reactor vendor analyses of steady-state or transient events.
2.
Evaluations of vendor computer programs ano analysis methods.
3.
Reduct ens and correlations of experimental data to verify processes or phenomena which are appiied to reactcr design These independent audit analyses. ray also be undertaken in support of Standard Review Plans (SRP) 4.2 and 4.3. in addition to the i.. ependent analysis discussed in SRP 4.4.
TYFES OF ANALYSES The types of analyses that are performed are the following:
1.
Steady-State Analyses a.
Tha steady-state reactor core flow distribution, steam quality, void d stribu-tion, and pressure droo have been calculated for PWR-type fuel assemblies usrm the multichannel boiling code, COBRA III-C (Ref. 1).
From these quantities COBRA III-C also calculates the fuel thermal margin in terms of the ratio of the local predicted critical heat flux to the operating heat flux. The W-3 and B&W-2 critical heat flux correlations (Ref s. 2 and 3) are used in the code. From these results, the thermal margin and fuel clad temperature calculated by the vendor's computer progran can be verified. To the extent possible, inputs to ccmputer programs used by the staf f correspond to those used by the reactor vendors.
b.
Through the use of consultants, independent comparisons and correlations are made of data frcm experimental programs. These reviews also include analyses of experimental techniques, test repeatability, and data reduction methods.
2.
Transient _Ana.yses Independent computer calculations are parformed to provide an audit on the adequacy of a particular analysis performed by an applicant. The thermal-hydraulic phenomena associated with the transient are calculated with the HELAP-3 (Ref. 4) or RE AP-4 (Ref.
- 5) computer programs. The fuel performance is calculated by the COBRA I!i-L program, which obtains the necessary thermal-hydraulic parameters from the abcve programs.
4.4-9
3.
Loss-of-Coolant Accident (LOCA) Analyses Independent calculations are performed by the staff to verify the LOCA analyses submitted by applicants in accordance with the requirements of 10 CFR s50.46 and Appendix K of 10 CFR Part 50.
These calculations are performed to check the blowdown
/
phenor'ena, ECCS response, and fuel cladding temperature transients. The CCS per-fomance criteria are Specified in.ippendix K.
Also, sen>itivity studies are performed to verify the convergence of analytical techniques, and the sensitivity to various postulated break sizes, types, and locations.
Evaluations are also made of the computer programs used by the vendors to perfonn ECCS evaluations. These computer programs are checked to determine confonnance with the required features specified in Appendix K.
In addition, the analysis methods and heat transfer correlations are evaluated by comparisen with existing experimental data.
4 Reactivity Analyses Independent analyses are perfortnec by consultants to provide a check on the adequacy of a particular analytical method and the basic assumptions. These include items such as maximum control rod worth, power distribution, and reactivity coefficients such as the Doppler and moderator temperature coefficients.
Staff consultants assess the conservatism of the vendors' r.odels, either by compar ison with exper; ment or with more sophisticatcd models. In particular, the importance of two-or three-dimensioral flux characteristics and changes in flux shapes are investi-gated and the conservatism of the flux shapes used for reactivity input and feedback, peak energy deposition, total energy, and gross heat trarsfer to the coolant v/e investigated.
5.
ATWS (Anticipated Transients Without ram) Analyses Independent audit analysis for ATWS serves three specific functions:
a.
To confirm the vendors' and applicants' interpretations of WASH-1270 (Ref. 6).
b.
To evaluate the adequacy of backup protection systems.
To confirm the accuracy of the calculation of consequences of ATWS and of the c.
models used in the analysis.
The RELAP 3B computer code (Ref. 7) will be used by the staff and its consultants for the ATWS studies. The preparation of data for an independent audit computation recuires a careful review of all reactor systems to ascertain if operational credit can be taken for then, in the analysis.
The process will then serve as a means of confirming the vendors' and applicants' interpretations of WASH-1270.
An evaluation of the effectiveness and the response of a backup protection system is achieved by an audit computation. The degree of protection can be evaluated by conducting analyses with and without the backup system.
147 027 4.4-10
A calculation of the conser;uences of an ATWS transient serves as c reans of ev luating vendors' analytical models and the accuracy of the results. For example, an ATWS loss of load event ive e
a PWR can result in very high primary pressure.
The m nitude of the pressure response is a function of the performance of the pressurizer safety valves in discharging water. An indap-endent audit computation would verify the analytical rodel used for discharging water and the magnitude of the pressure response. The pressure response is important in evaluating the integrity of the reactor vessel.
R_E_F E R E N C E S l.
D. S. Rowe, COBRA-IIIC: A digital Computer Program foi Steady-State and Transient Thermal-Hydraulic Analysis of Rod Bundle Nuclear Fuel Elements, BNWL-1965.
'>aru.973.
2.
L. S. Tong, " Prediction of Departure from NJcleate Boiling for an Axially Ncn-Uniforn Heat Flux Distribution,' Journal of Nuclear Energy, Vol 21, 241-248 (1967).
3.
e S. Gellerstedt, R. A. Lee, W. J. Oberjohn, R. H. Wilson, and L. J. Stanek, 'Cerrelation of rritical Heat Flux in a Rundle Coolea by Pressurized Water," in "Two Phase Flow and Heat Transfee in Red Bundles, Arerican Society of 'dechanical Engineers, New York (1969). (See also BAW-10000 and BfW-lC036.)
4 4
W. H. Rettig, G. A. Jayne, K. V. Moore, C. E. Slater, M L. Upt cre, RELA? 3 - A Co puter 9;
Progra-for Reactor Bluwdown Analvsis, !N-1321 (June 1970).
5.
K.
'>o c re, W. H. Rettig, RELA? A Computer Progra for Transient Therral H draulic Analysis, UC-32
'CR 1127 (Dece cer 1973).
6.
squlatory Sta f f, m hnical Reper'. cn Anticipated Transients Without Scran for Water-Cooled c
Do,er Reactors, WASH-1270, U.S. Atomic Ereray Com ission, Sept. 1973.
7.
RELAp-1B ' Manual, A Peactor System Transient Code, Brook haven National oboratory PD !O35 (Cecedaer 1974).
h 147 028 4 4-11