ML19260C455
| ML19260C455 | |
| Person / Time | |
|---|---|
| Issue date: | 09/18/1979 |
| From: | Meyer R Office of Nuclear Reactor Regulation |
| To: | Hosford S Office of Nuclear Reactor Regulation |
| References | |
| REF-GTECI-A-01, REF-GTECI-RV, TASK-A-02, TASK-A-2, TASK-OR NUDOCS 8001030494 | |
| Download: ML19260C455 (11) | |
Text
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UNITED STATES y
g NUCLEAR REGULATORY COMMISSION g
,E WASHINGTON, D. C. 20555 g,...../
SEP 181979 MEMORANDUM FOR: Steve Hosford, Task Manager (A-2),
Engineering Branch, 00R FROM:
Ralph 0. Meyer, Leader, Reactor Fuels Section, Core Performance Branch, DSS
SUBJECT:
COMMENTS ON A-2 DRAFT REPORT The draft report, "Assymetric Blowdown Loads on PWR Primary Systems,"
includes a section on the fuel analysis. Enclosed is a revision of that section, which is comprised of a few introductory remarks and the latest version of our draft appendix to SRP Section 4.2.
I expect our draft SRP appendix to be formally issued for connent (in its present form) be-fore your report could be published.
Two other mir.or changes are needed.
(1) Delete subsections 5.1 and 5.2 from the Table of Contents, and (2) insert the date of issuance (I will advise you when it happens) of our draft SRP appenoix.
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fI'-ch,c)
Ralph 0. Mey #
er, Leader Reactor Fuels Section Core Performance Branch Division of Systems Safety
Enclosure:
As stated cc:
S. Hanuaer C ycock J. Knight R. Denise W. Butler P. Check R. Bosnak Z. Rosztoczy R. Gamble R. Mattu E. Throm C. Tinkler D. Eisenhut V. Noonan L. Shao A. Schwencer D. Ziemann D. Crutchfield T. Ippolito R. Reid G. Lainas S. Varga U. Stolz
- 0. Parr F. Schros. der
Contact:
R. O. Meyer, x27603 Y
8001030 ;l 1700 028
5.0 FUEL EVALUATION Assymetric loads on reactor vessel supports, reactor internals, and other components will transmit loads to the fuel assemblies. A detailed structural analysis of the fuel assembly is thus required to evaluate loads on fuel assembly components such as spacer grids, fuel rods, and guide tubes in order to ascess design adequacy.
Similar requirements for analysis exists for postulated earthquakes, and the procedures for doing the fuel assembly analysis are nearly identical for seismic and LOCA loads. A single description of pro-cedures for both seismic and LOCA loads has been developed recently as an appendix to the Standard Review Plan, Section 4.2, " Fuel System Design." This draft appendix (issued for comment in 1979) appears in its entirety on the following pages and includes all procedures and acceptance criteria 'or the analysis of assymetric loads in fuel assemblies.
Typist's Note: This job is on Mag Card. See Patty Vitale, x27577 1700 029
APPENDIX EVALUATION OF FUEL ASSEMBLY STRUCTURAL RESPONSE TO EXTERNALLY APPLIED FORCES 1.
BACKGROUND Earthquakes and postulated pipe breaks in the reactor coolant system would result in external forces on the fuel assembly. SRP Section 4.2 states that fuel system coolability should be maintained and that damage should not be so severe as to prevent control rod in-sertion during these low probability accidents.
This appendix describes the review that should performed of the fuel assembly structural response to seismic and LOCA loads.
Background raterial for this appendix is given in Refs. 1-3.
2.
ANALYSIS OF LOADS a.
Input Input for the fuel assembly structural analysis comes from results of the primary coolant system structurai analysi,, which is reviewed by the Mechanical Engineering Branch.
Input for the fuel assembly response to a LOCA should include (a) motions of the core plate, core shroud, fuel alignment plate, or other relevant structures; these motions should correspond to the break that produced the peak fuel assembly loacings in the reactor primary coolant system analysis, and (b) transient i700 030 A-1
pressure differences that apply loads directly to the fuel assembly. If the earthquake loads are large enough to produce a non-linear fuel assembly response, input for the seismic analysis should use structure motions corresponding to the reactor primary coolant system analysis for the SSE; if a linear response is produced, a spectral analysis may be used (see Regulatory Guide 1.60).
b.
Methods Analytical methods used in performing structural response analyses must be reviewed. Justification should be supplied to show that the numerical solution techniques are appropriate.
Linear and non-linear structural representations (i.e., the modeling) must also be reviewed.
Experimental verification of the analytical representation of the fuel assembly components should be provided when pratical.
A sample problem of a simplified nature must be worked by the applicant and compared by the reviewer with either hand calcula-tions or results generated by the reviewer with an independent code (2). Although the sample problem should use a structural representation that is as close as possible to the design in question (and, therefore, would vary from one vendor to another),
simplifying assumptions may be made (e.g., one might use a 3-assembly core region with continuous sinusoidal input).
A-2 1700 031
The sample problem should be designed to exercise various features of the' code and reveal their behavior. The sample problem comparison is not, however, designed to show that one code is more conservative than another, F t rather to alert the reviewer to major discrepancies so that an explanation can be sought.
c.
Uncertainty Allowances The fuel assembly structural models and analytical methods are probably conservative and input parametcrs are also conservative.
However, to ensure that the fuel assembly analysis does not introdu e any non-conservatisms, two precautions should be taken. (1) If it is not explicitly evaluated, impact loads from the PWR LOCA analysis should be increased (by about 30%) to account for a pressure pulse, which is associated with steam flashing that affects only the PWR fuel assembly analysis.
(2) Conservative margin should be added if any part of the analysis (PWR or BWR) exhibits pronounced sensitivity to input variations.
Variations in resultant loads should be determined for 110%
variations in input amplitude and frequency; variations in amplitude and frequency should be made separately, not simulta-neously. A factor should be developed for resultant load magnitude variations of more than 15%.
For example, if 110% variations in input magnitude or frequency produce a maximum resultant increase 1700 032 A-3
of 35%, the sensitivity factor would be 1.2.
Since resonances and pronounced sensitivities may be plant-dependent, the sen-sitivity analysis should be performed on a plant-by-plant basis until the reviewer is confident that further sensitivity analyses are unnecessary.
d.
Audit Independent audit calculations for a typical full-sized core must be performed by the reviewer to verify that the overall structural representation is adequate. An independent audit ccde (2) should be used for this audit during the generic review of the analytical method..
e.
Combination of Loads General Design Criterion 2 does not require the combination of loads from the most' severe natural phenomena and accident condi-tions, although these have been combined for some components to increase the margin against earthquakes causing LOCAs.
Inasmuch as the loss of structural capabilities in the core cannot lead to a LOCA, no combination of loads is needed.
Each event is covered by an adequate conservative margin in the fuel assembly analysis.
1/00 033
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A-4
3.
DETERMINATION OF STRENGTH a.
Grids All modes of loading (e.g., in-grid and through-grid loadings) should be considered, and the most damaging mode should be represented in the vendor's laboratory grid strength tests.
o Test procedures and results should be reviewed to assure that the appropriate failure mode is being predicted.
The review should also confirm that (a) the testing impact velocities correspond to expected fuel assembly velocities, and (b) the has been suitably selected from the crushing load Pcrit load-vs-deflection curves. Because of the potential for dif-ferent test rigs to introduce measurement variations, benchmark measurements should be made by each fuel vendor on specimen grids provided by NRC as part of the review of his test procedure.
The consequences of grid deformation are small. Gross deforma-tion of grids in many PWR assemblies would be needed to interfere with control rod insertion dcring an SSE (i.e., buckling of a few isolated grids could not displace guide tubes significantly from their proper locatinn). and grid deformation (without channel deflection) would not affect control blade insertion in a BWR.
In a LOCA, gross deformation of the hot channel in either a PWR or a BWR would result in only small increases in peak cladding temperature. Therefore, average values are appro-1700 034 A-5
priate, and the allowable crushing load P should be the 95%
crit confidence level on the true mecn as taken from the distribution of measurements on unirradiated production grids at (or corrected to) operating temperature. While P will increase with crit irradiation, ductility will be reduced. The extra margin in P
for irradiated grids is thus assumed to offset the unknown crit deformation behavior of irradiated grids beyond P crit" b.
Components other than Grids Strengths of fuel assembly components other than spacer grids may be deduced from fundamental properties or experimentation.
Supporting evidence for strength values should be supplied.
Since struc.tural failure of these components (e.g., fracturing of guide tubes or fragmentation of fuel rods) could be more serious than grid deformation, allowable values should bound a large percentage (about 95%) of the distribution of component strengths. Therefore, ASME Boiler and Pressure Vessle Code values and procedures may be used where aporopriate for determining yield and ultimate strengths.
Specification of allcwable values may follow the ASME Code requirements and should include con-sideration of buckling and fatigue effects.
A-6 1700 Ob,-
4.
ACCEPTANCE CRITERIA a.
Loss-of-Coolant Accident Two principal criteria apply for the LOCA:
(1) fuel rod fragmentation must not occur as a direct result of the blowdown loads, and (2) the 10 CFR 50.46 temperature and oxidation limits must not be exceeded. The first criterion is satisfied if the calculated loads on the fuel rods and components other than grids remain below the allowable values defined above. The second criterion is satisfied by an ECCS analysis.
If calculated loads on the grids remain below Pcrit, as defined above, then no significant distortion of the fuel assembly would occur and the usual ECCS analysis is sufficient.
If calculateo grid loads exceed Pcrit, then grid deformation must be assur..ed and the ECCS analysis must include the effects of distorted fuel assemblies. An assumption of maximum credible deformation (i.e., fully collapsed grids) may be made unless other assumptions are justified.
Control rod insertability is a third criterion that must be satisfied for the LOCAs that require insertion to assure sub-criticality. This may not be a major consideration for a PWR since rod insertion is not required for a large-break LOCA, which would produce the largest asymmetric loads. Although control blade insertion is required in a BWR, blowdown loads in a BWR are small so that this may not be a major consideration in BWRs either.
1700 036 A-7
b.
Safe Shutdown Earthquake Two criteria apply for the SSE:
(1) fuel rod fragmentation must not occur as a result of the seismic loads, and (2) control rod insertability must be assured. The first criterion is satisfied if the calculated loads on the fuel rods and components other than grids remain below the allowable values defined above. The second criterion is satisfied differently for PWRs and BWRs.
For a PWR, if calculated loads on the grids remain below P crit as defined above, then significant deformation of the fuel assembly would not occur and control rod insertion would not be interfered with by lateral displacement of the guide tubes.
If calculated loads on the grids exceed Pcrit, then additional analysis is needed to show that deformation is not severe enough to prevent control rod insertion.
For a BWR, several conditions must be met to satisfy the second criterion: (a) calcula ed loads on the channel box must remain below the allowable value defined above for components other than grids because a small amount of channel deformation could interfere with control blade insertion, and (b) vertical liftoff forces must not unseat the lower tieplate from the fuel support piece because the resulting loss of lateral fuel bundle positioning could also interfare with control blade insertion.
1700 037
5.
REFERENCES 1.
R. '.. Grubb, "" view of LWR Fuel System Mechanical Response with Re ammendationi er Component Acceptance Criteria," Idaho National Engineering Laboratory Report, NUREG/CR-1018, September 1979.
2.
R. L. Gr JHL " Pressurized Water Reactor Lateral Core Response Routine, FAMREC (Fuel Assembly Mechanical Response Code)," Idaho National Engineering Laboratory Report, NUREG/CR-1019, September 1979.
3.
R. L. Grubb, " Technical Evaluation of PWR Fuel Spacer Grid Response Load Sensitivity Studies," Idaho National Engineering Laboratory Report, NUREG/CR-1020, September 1979.
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