ML20094C582
| ML20094C582 | |
| Person / Time | |
|---|---|
| Issue date: | 11/24/1975 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-75-087, NUREG-75-087-15.4.3, NUREG-75-87, NUREG-75-87-15.4.3, SRP-15.04.03, NUDOCS 9511020268 | |
| Download: ML20094C582 (5) | |
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k U.S. NUCLEAR RE2ULATORY COMMIS ISN f's STANDARD REVIEW PLAN OFFICE OF NUCLEAR REACTOR REGULATION SECTION 15.4.3 CONTROL R00 MISOPERATION (SYSTEM MALFUNCTION OR OPERATOR ERROR)
REVIEW RESPONSIBILITIES Primary - Core Performance Branch (CPB)
Secondary - Electrical, Instrumentation and Control Systems Branch (EICSB)
Reactor Systems Branch (RSB) 1.
AREAS OF REVIEW CPB reviews the following:
1.
The types of control rod misoperations that are assumed to occur. For a pressurized water reactor (PWR), this may include one or more rods moving or displaced from normal or allowed control bank positions (such as dropped rods and rods left behind when inserting or withdrawing banks, or single rod withdrawal) and may include the automatic control system attempting to maintain full power. For a boiling water reactor (BWR) with current modes of control rod operation, limiting anomalies have been reviewed in Standard Review Plans (SRP) 15.4.1 and 15.4.2, and no additional areas are considered here.
2.
Descriptions of rod position, flux, pressure, and temperature indication systems, and those actions initiated by these systems (e.g., turbine runback, rod withdrawal pro-hibit, rod block) which can mitigate the effects or prevent the occurrence of various misoperations.
E!CSB in SRP 7.2 and 7.7 reviews those safety systems required to prevent misopera-tions, as required by General Design Criterion 25 (Ref.1), as well as the control rod system. The purpose of the review is to determine what events are to be included as single error malfunctions (e.g., single rod withdrawal).
3.
The course of transients involved, e.g., rod drop followed by automatic return to full power with possible power overshoot.
4.
Descriptions of the calculational models used and justification of their validity and adequacy.
5.
The input to the calculations, including rod worths, power distributions, and feedback coefficients, and evidence of the conservatism of the input.
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Descriptions of the sgquence of events occurring during each transient, includirig the effects of important foedback mechanisms and trips.
7.
Results of the analyses including, for each of the transients confidered, plots of the time history of reactor power, reactor vessel pressure, critical heat flux for the limiting fuel rod, and maximum fuel centerline temperature.
II. ACCEPTANCE CRITERIA 1.
The following general design criteria (Ref. 1) apply:
Criterion 25, which requires that the reactor protection system be designed to a.
assure that specified fuel design limits are not exceeded in the event of a single malfunction of the reactivity control system, b.
Criterion 20, which requires that the protection system action be initiated ~
automatically.
2.
The following fuel design limits serve as the acceptance criteria for this eventi.
a.
Critical heat flux should not be exceeded. An example of limits used previously to satisfy this criterion is that the minimum departure from nucleate boiling ratio (DNBR) should not be less than 1.3, using the W-3 correlation (Ref. 2), or less than 1.32, using the B&W-2 correlation (Ref. 3). If the application under review does not use one of these limits, then the reviewer must assure that an acceptable correlation is established under SRP 4.4 and is used here, b.
Fuel temperature and fuel clad strain limits consistent with the acceptance cri-teria of SRP 4.2 should not be exceeded. For steady-state or nearly steady-state conditions, this can be referred to a linear heat generation rate (usually expressed in kW/ft). An example of this criterion is a linear heat generation rate of 20 to 22 kw/ft, which would result in a centerline fuel temperature equal or less than the melting point of UO. The specific value of linear heat generation rate for 2
this criterion is established in a manner consistent with the acceptance criteria of SRP 4.2 (Ref. 4). For non-equilibrium states, the calculated transient temper-atures and strains corresponding to these steady-state limits should not be exceeded.
!!!. REVIEW PROCEDURES The reviewer, in determining whether the criteria are met, must determine the transients.that
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should be considered for this event. Generally, the list of errors should include: inadvert-ently withdrawing one or several rods; leaving one or several rods behind during bank withdrawal; and inserting one or several rods with power compensation in other portions of
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the core.
In addition to these events, the reviewer must also decide, by postulating single failures in equipment or errors in operation, whether additional single rod malfunctions can be created. Once the list of transients has been established, the reviewer must determine acceptability in accordance with the criteria of Section !! of this SRP.
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1.
For each failure event, the limiting (i.e., worst) fuel rod condition are determined by consideration of sensitivity calculGtions.
2.
For each event, the analytical methods used by the applicant are reviewed. Those steady-state and transient methods that are primarily based on reactor physics considerations are the responsibility of CPB, Where thermal-hydraulic methods are involved, assistance of RSB may be required. In either case, the reviewer must deter-mine whether the applicant's evaluation methods are acceptable. This may be done by using one or more of the following procedures:
Determine whether the method has been reviewed and approved previously, by con-a.
sidering past safety evaluation reports (SER's) and reports prepared in response to specific technical assistance requests (TAR's).
b.
Perform a de novo review of the method (usually described in a separate licensing topical report, and of ten handled outside the scope of the review for a particular facility).
Perform auditing-type calculations with methods available to the staff.
c.
d.
Require additional bounding calculations by the applicant to confirm the validity of those portions of the applicant's analytical method that have not already been fully reviewed and approved.
3.
For each event, the results are evaluated. In addition to verifying conformance to the acceptance criteria of Section II above, the reviewer determines that:
Input conditions (e.g., pressure, temperature, flow rate) are at the adverse a.
end of the range of values specified as the operating range, b.
Initial power is 102% of licensed core thermal power, unless a lower power level is justified by the applicant.
c.
Output signals (power, temperature, flux perturbation) provided adequate alarm or scram signals, d.
Nuclear conditions that interact with this event (e.g., Doppler coefficient, void coefficient) have been calculated as described in SRP 4.3.
IV. EVALVATION FINDINGS If the reviewer's evaluation shows that the applicant's analyses are acceptable, the following kinds of statements should be included in the staff's safety evaluation report:
15.4.3-3 11/24/75
O "The ' possibilities for single failures of the reactor control system which could result -
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in a movement or malposition of control rods beyond normal limits have been reviewed.
The scope of the review has included investigations of possible rod malposition con-figurations, the course of the resulting transients or steady-state conditions, and the instrumentation response to the transient or power maldistribution. The methods used to determine the peak fuel rod response, and the input to that analysis, such as power distribution changes, rod reactivities, and reactivity feedback effects due to moderator and fuel temperature changes. have been examined. (If check calculations have been done, they should be summarized).
r "The resulting extreme conditions of fuel power, temperature, and departure from nucleate boiling (DNB) have been compared to acceptance criteria for fuel integrity, which for this reactor are (insert criteria from SRP 4.2 and 4.4).
The analyses have shown that these limits are not exceeded.
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i "The basis for acceptance in the staff review is' that maximum configurations and tran-sients for single error control rod malfunctions have been analyzed, that the analysis methods and input data are reasonably conservative, and that fuel damage limits are not exceeded. The staff concludes that the calculations contain sufficient conserva-tism, in both input assumptions and models, to assure that fuel damage will not -
result from control rod malposition transients."
V.
REFERENCES 1.
.10 CFR Part 50,' Appendix A.' " General Design Criteria for Nuclear Power Plants."
2.
L. S. Tong, "Prodiction of Departure from Nucleate Boiling for an Axially Non-Uniform Heat Flux Distribution," Jour. Nuclear Energy, Vol. 21,241-248(1967).
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3.
J. S. Gellerstedt, R. A. Lee, W. J. Oberjohn, R. H.' Wilson, and L. J. Stanek, "Correla-tion 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).
4 Standard Review Plan 4.2, " Fuel System Design."
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