ML20031D715
| ML20031D715 | |
| Person / Time | |
|---|---|
| Site: | Allens Creek File:Houston Lighting and Power Company icon.png |
| Issue date: | 07/31/1981 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20031D662 | List: |
| References | |
| NUREG-0800, NUREG-0800-04.02, NUREG-800, NUREG-800-4.02, SRP-04.02, SRP-4.02, NUDOCS 8110140049 | |
| Download: ML20031D715 (21) | |
Text
N U REG-0800 (Formerly NUREG 75/087)
@/ g%
f U.S. NUCLEAR REGULATORY COMMISSION i STAlYDARD MV03W PLAN E
OFFICE OF NUCLEAR REACTOR REGULATION N.~...e o
4.2 FUEL SYSTEM DESIGN REVIEW RESPONSIBILITIES Primary - Core Performance Branch (CPB)
Secondary - None I.
AREAS OF REVIEW The thermal, mechanical, and materials design of the fuel system is evaluated by CPB.
The fuel system consists of arrays (assemblies or bundles) of fuel rods including fuel pellets, insulator pellets, springs, tubular cladding, end closures, hydrogen getters, and fill gas; burnable poison rods including com-ponents similar to those in fuel rods; spacer grids and springs; end plates; channel boxes; and reactivity control rods.
In the case of the control rods, this section covers the reactivity control elements that extend from the coupling interface of the control rod drive mechanism into the core.
The Mechanical Engineering Branch reviews the design of control rod drive mechanisms in SRP Section 3.9.4 and the design of reactor internals in SRP Section 3.9.5.
The objectives of the fuel system safety review are to provide assurance that (a) the fuel system is not damaged as a result of normal operation and antic-ipated operational occurrences, (b) fuel system damage is never so severe as to prevent control rod insertion when it is required, (c) the number of fuel rod failures is not underestimated for postulated accidents, and (d) coolability is always maintained.
"Not damaged," as used in the above statement, means that fuel rods do not fail, that fuel system dimensions remain within operational tolerances, and that functional capabilities are not reduced below those assumed in the safety analysis.
This objective implements General Design Criterion 10 (Ref. 1), and the design limits that accomplish this are called Specified Acceptable Fuel Design Limits (SAFDLs).
" Fuel rod failure" means that the fuel rod leaks and that the first fission product barrier (the cladding) has, therefore, been breached.
Fuel rod failures must be accounted for in the dose analysis required by 10 CFR Part 100 (Ref. 2) for postulated accidents.
"Coolability," in general, means that the fuel assembly retains its rod-bundle geometry with adequate coolant channels to permit removal of residual -heat even after a severe accident.
The general requirements to maintain control rod l
811014o049 811009 Rev. 2 - July 1981
- PDR ADOCK 05000466 T
PDR USNRC STANDARD REVIEW PLAN Star.dard review plans are prepared for the guidance of the Office of Nuclear Reactor Regulation staff responsit,:e for the review of apphcations to construct and operate nuclear power plants. These documents are made available to the public as part of the Commission's policy to inform the nuclear industry and the general public of regulatory procedures and policies. Standard review plans are not substitutes for regulatory guides or the Commission's regulations and compliance with them is not required. The standard review plan sections are keyed to the Standard Format ar d Ccra.nt of Safety Analysis Reports for Nuclear Power Plants.
Not all sections of the Standard Format have a corresponding review plan.
Published standard review plans will be revised periodically, as appropriate, to accommodate comments and to reflect new inf orma-tion and experience.
Comments and suggestions for improvement will be considered and should be sent to the U.S. Nuclear Regulatory Commission.
Of f ace of Nuclear Reactor Regulation. Washington, D.C. 20r 55.
J
om
_ aaJ cora cooMoility appear repeatedly in the General Design Criter;a (e.g., GDC 27 and 35).
Specific coolability requirements for the loss-of-coolant accident are given in 10 CFR Part 50, 650.46 (Ref. 3).
All fuel damage criteria are described in SRP Section 4.2.
For those criteria that involve DNbrior CPR limits, specific-thermal-hydraulic criteria are given in SRP Section 4.4.
The a,ailable radioactive fission product inventory in fuel rods (i.e., the gap inventory expressed as a release fraction) is provided to the Accident Evaluation Branch for use in estimating the radiological consequences of plant releases.
The fuel system review covers the following specific areas.
A.
Design Bases Design bases for the safety analysis address fuel system damage ricchanisms and provide limiting values for important parameters such that damage will be limited to acceptable levels.
The design bases should reflect the safety review objectives as described above.
B.
Description and Design Drawings The fuel system description and design drawings are reviewed.
In general, the description will emphasize product specifications rather than process specifications.
C.
Design Evaluation The performance of the fuel system during normal operation, anticipated operational occurrences, and postulated accidents is reviewed to determine if all design bases are met.
The fuel system components, as listed above, are reviewed not only as separate components but also as integral units such as fuel rods and fuel assemblies.
The review consists of an evaluation of operating experience, direct experimental comparisons, detailed mathematical analyses, and other information.
i D.
Testing, Inspection, and Surveillance Plans Testing and inspection of new fuel is performed by the licensee to ensure that the fuel is fabricated in accordance with the design and that it reaches the plant site and is loaded in the core without damage.
On-line fuel rod failure monitoring and postirradiation surveillance should be performed to detect anomalies or confirm that the fuel system is performing as expected; surveillance of control rods containing B C should be performed 4
to ensure against reactivity loss.
The testing, inspection, and surveil-lance plans along with their reporting provisions are reviewed by CPB to ensure that the important fuel design considerations have been addressed.
l II.
ACCEPTANCE CRITERIA l
Specific criteria necessary to meet the requirements of 10 CFR Part 50, 950.46; General Design Criteria 10, 27, and 35; Appendix K to 10 CFR Part 50; and 10 CFR Part 100 identified in subsection I of this SRP section are as follows:
l 4.2-2 Rev. 2 - July 1981 l
l
~
a.
Design das The fuel system design bases must reflect the four objectives described in subsection I, Areas of Review.
To satisfy these objectives, acceptance criteria are needed for fuel system damage, fuel rod failure, and fuel coolability.
These criteria are discussed in the following-1.
Fuel System Damage This subsection applies to normal operation, and the information to be reviewed should be contained in Section 4.2 of the Safety Analysis Report.
To meet the requirements of General Design Criterion 10 as it relates to Specified Acceptable Fuel Design Limits for normal operation, including anticipated operational occurrences, fuel system damage criteria should be given for all known damage mechanisms.
Fuel system damage includes fuel rod failure, whish is discussed below in subsection II.A.2.
In addition to precluding fuel rod failure, fuel. damage criteria should assure chat fuel system dimen-sions remain within operational tolerances and that functional capabilities are not reduced below those assumed in the safety analysis.
Such damage criteria should address the following to be complete.
(a) Stress, strain, or loading limits for spacer grids, guide tubes, thimbles, fuel rods, control rods, channel boxes, and other fuel system structural members should be provided.
Stress limits that are obtained by methods similar to those given in Section III of the ASME Code (Ref. 4) are acceptable.
Other proposed limits must be justified.
(b) The cumulative number of strain fatigue cycles on the structural members mentioned in paragraph (a) above should be significantly less than the design fatigue lifetime, which is based on appro-priate data and includes a safety factor of 7 on stress amplitude or a safety factor of 20 on the number of cycles (Ref. 5).
Other proposed limits must be justified.
(c) Fretting wear at contact points on the structural members mentioned in paragraph (a) above should be limited.
The allowable fretting wear should be stated in the Safety Analysis Report and the stress anu fatigue limits in paragraphs (a) and (b) above should presume the existence of this vear.
(d) Oxidation, hydriding, and the buildup of corrosion products (crud) should be limited.
Allowable oxidation, hydriding, and crud levels should be discussed in the Safety Analysis Report and shown to be acceptable.
These levels should be presumed to exist in paragraphs (a) and (b) above. The effect of crud on thermal-hydraulic considerations is reviewed as described in l
SRP Section 4.4.
(e) Dimensional changes such as rod bowing or irradiation growth of fuel rods, control rods, and guide tubes need not be limited to 4.2-3 Rev. 2 - July 1981
set values (i.e., damage limiWj, but they must be included in the cesign analysis to establish operational tolerances.
(f) Fuel and burnable poison rod internal gas pressures should remain below the nominal system pressure during normal opera-tion unless otherwise justified.
(g). Worst-case hydraulic loads for normal operation should not exceed the holddown capabHit gravity or holddown springs).y of the fuel assembly (either Hydraulic loads for this evaluation are reviewed as described in SRP Section 4.4.
l (h) Control rod reactivity must be maintained.
This may require the control rods to remain watertight if water-soluble or leachable materials (e.g., B C) are used.
4 2.
Fuel Rod Failure This subsection applies to normal operation, anticip(a)ed operational at occurrences, and postulated accidents.
Paragraphs through(c) address failure mechanisms that are more limiting during normal operation, and the information to be reviewed should be contained in Section 4.2 of the Safety Analysis Report.
Paragraphs (d) through (h) address failure mechanisms that are more limiting during anticipated operational occurrences and postulated accidents, and the information to be reviewed will usUally be contained in Chapter 15 of the Safety Analysis Report.
Paragraph (i) should be addressed i:1 Section 4.2 of the Safety Analysis Report because it is not addressed elsewhere.
To meet the requirements of (a) General Design Criterion 10 as it relates to Specified Acceptable Fuel Design Limits for normal opera-tion, including anticipated operational occurrences, and (b) 10 CFR Part 100 as it relates to fission product releases for postulated accidents, fuel rod failure criteria should be given for all known fuel rod failure mechanisms.
Fuel rod failure is defined as the loss of fuel rod hermeticity.
Although we recognize that it is not possible to avoid all fuel rod failures and that cleanup systems are installed to handle a small number of leaking rods, it is the objective of the review to assure that fuel does not fail due to specific causes during normal operation and anticipated operational occurrences.
Fuel rod failures are permitted during postulated accidents, but they must be accounted for in the dose analysis.
Fuel rod failures can be caused by o m heating, pellet / cladding interaction (PCI), hydriding, claddiro collapse, bursting, mechanical fracturing, and fretting.
Fuel fa N re criteria should address the following to be complete.
(a) Hydridin :
Hydriding as a cause of failure (i.e., primary l
hydridin ) is prevented by keeping the level of moisture and other hy rogenous impurities very low during fabrication.
Acceptable moisture levels for Zircaloy-clad uranium oxide fuel should be no greater than 20 ppm.
Current ASTM specifications (Ref. 7) for U02 fuel pellets state an equivalent limit of 2 ppm of hydrogen from all sources.
For other materials clad in 4.2-4 Rev. 2 - July 1981
z.;..ai_
m
,. aa.qt::va ient quanci t.y vi rooisture or hydrogen can be tolerated.
A moisture level of 2 mg H O per cm3 of hot 2
void volume within the Zircaloy cladding has been shown (Ref. 8) to be insufficient for primary hydride formation.
(b) Cladding Collapse:
If axial gaps in the fuel pellet column
__l occur due to densification, the cladding has the potential cf collapsing into a gap (i.e., flattening).
Because of the large local strains that accompany this process, collapsed (flattened) cladding is assumed to fail.
(c) Fretting:
Fretting is a potential cause of fuel failure, but 1
it is a gradual process that would not be effective during the brief duration of an abnormal operational occurrence or a postulated accident.
Therefore, the fretting wear requirement in paragraph (c) of subsection II.A.1, Fuel Damage, is syfficient to preclude fuel failures caused by fretting during transients.
(d) Overneating of Cladding:
It has been traditional practice to assume that failures will not occur if the thermal margin criteria (DNBR for PWRs and CPR for BWRs) are satisfied.
The review of these criteria is detailed in SRP Section 4.4.
For normal operation and anticipated operational occurrences, violation of the thermal margin criteria is not permitted.
For postulated accider.t.s, the total number of fuel rods that exceed the criteria has been assumed to fail for radiological dose calculation purposes.
Al'. hough a thermal margin criterion is sufficient to demonstrate the avoidance of overheating from a deficient cooling mechanism, it is not a necessary c udition (i.e., DNB is not a failure mechanism) and other mechanistic methods may be acceptable.
There is at present little experience with other approaches, but new positions recommending different criteria should address cladding temperature, pressure, time duration, oxidation, and embrittlement.
(e) Overheating of Fuel Pellets:
It has also teen traditional practice to assume that failure will occur if centerline melting takes place.
This analysis should be performed for the maximum linear heat generation rate anywhere in the core, including all hot spots and hot channel factors, and should account for the effects of burnup and composition on the melting point.
For normal operation and anticipated operational occurrences, centerline melting is not permitted.
For postulated accidents, the total number of rods that experience centerline melting should be assumed to fail for raciological dose calculation purposes.
The centerline melting criterion was established to assure that axial or radial relocation of molten fuel would neither allow molten fuel to come into contact with the cladding nor produce local hot spots.
The assumption that centerline melting results in fuel failure is conservative.
(f) Excessive Fuel Enthalpy:
For a severe reactivity initiated accident (RIA) in a BWR at zero or low power, fuel failure is assumed to occur if the radially averaged fuel rod enthalpy is 4.2-5 Rev. 2 - July 1981
greater than 170 cal /g at any axial location.
Fcr full power RIAs in a BWR and all RIAs in a PWR, the thermal margin critaria (DNBR and CPR) are used as fuel failure criteria to meet the l
guidelines of Regulatory Guide 1.77 (Ref. 6) as it relates to f
fuel rod failure.
The 170 cal /g enthalpy criterion is primarily intended to address cladding overheating effects, but it also indirectly addresses pellet / cladding interactions (PCI).
Other criteria may be more appropriate for an RIA, but continued approval of this enthalpy criterion and the thermal margin criteria may be given until generic studies yield improvements.
(g) Pellet / Cladding Interaction:
There is no current criterion I
for fuel failure resulting from PCI, and the design basis can only be stated generally.
Two related criteria should be applied, but they are not sufficient to preclude PCI f,ailures.
(1) The uniform strain of the cladding should not exceed 1%.
In this context, uniform strain (elastic and inelastic) is defined as transient-induced deformation with gage lengths corresponding to cladding dimensions; steady-state creepdown and irradiation growth are excluded.
Although observing this strain limit may preclude some PCI failures, it will not preclude the corrosion-assisted failures that occur at low strains, nor will it preclude highly localized overstrain failures.
(2) Fuel melting should be avoided.
The large volume increase associated with melting may cause a pellet with a molten center to exert a stress on the cladding.
Such a PCI is avoided by avoiding fuel melting.
Note that this same criterion was invoked in para-graph (e) to ensure that overheating of the cladding would not occur.
(h) Bursting:
To meet the requirements of Appendix K of 10 CFR Part 50 (Ref. 9) as it relates to the incidence of rupture during a LOCA, a rupture temperature correlation must be used in the LOCA ECCS analysis.
Zircaloy cladding will burst (rupture) under certain combinations of temperature, heating rate, and differential pressure. Although fuel suppliers may use different rupture-temperature vs differential pressure curves, an acceptable curve should be similar to the one described in Ref. 10.
l (i) Mechanical Fracturing:
A mechanical fracture refers to a l
defect in a fuel rod caused by an externally applied force such as a hydraulic load or a load derived from core plate motion.
Cladding integrity may be assumed if the applied stress is less than 90% of the irradiated yield stress at the appropriate
{
temperature.
Other proposed limits must be justified.
Results from the seismic and LOCA analysis (see Appendix A to this SRP section) may show that failures by this mechanism will not occur for less severe events.
3.
Fuel Coolability This subsection applies to postulated accidents, and most of the information to be reviewed will be contained in Chapter 15 of the Safety Analysis Report.
Paragraph (e) addresses the combined effects 4.2-6 Rev. 2 - July 1981
of two accidents, however, 2nd that information should be contained in Section 4.2 of the Safety Analysis Report.
To meet the require-ments of General Design Criteria 27 and 35 as they relate to cor 'rol rod insertability and core coolability for postulated accidents, fuel coolability criteria shculd be given for all severe damage mechanisms.
Coolability, or coolable geometry, has traditionally implied that the fuel assembly retains its rod-bundle geometry with adequate coolant channels to permit removal of residual heat.
Reduction of coolability can result from cladding embrittlement, violent expulsion of fuel, generalized cladding melting, gross structural deformation, and extreme coplanar fuel rod ballooning.
Control rod insertability criteria are also addressed in this subsection.
Such criteria should address the following to be complete:
(a) Cladding Embrittlement:
To meet the requirements of '10 CFR Part 50, 650.46, as it relates to cladding embrittlement for a LOCA, acceptance criteria of 2200 F on peak cladding temperature and 17% on maximum cladding oxidation must be met.
(Note:
If the cladding were predicted to collapse in a given cycle, it would also be predicted to fail and, therefore, should not be irradiated in that cycle; consequently, the lower peak cladding temperature limit of 1800 F previously described in Reference 11 is no longer needed.) Similar temperature and oxidation criteria may be justified for other accidents.
(b) Violent Expulsion of Fuel:
In severe reactivity initiated accidents, such as rod ejection in a PWR or rod drop in a BWR, the large and rapid deposition of energy in the fuel can result in melting, fragmentation, and dispersal of fuel.
The mechanical action associated with fuel dispersal can be sufficient to destroy the cladding and the rod-bundle geometry of the fuel and to pro-duce pressure pulses in the primary system.
To meat the guide-lines of Regulatory Guide 1.77 as it relates to preventing wide-spread fragmentation and dispersal of the fuel and avoiding the generation of pressure pulses in the primary system of a PWR, a radially averaged enthalpy limit of 280 cal /g should be observed.
This 280 cal /g limit should also be used for BWRs.
l (c) Generalized Cladding Melting:
Generalized (i.e., non-local) l melting of the cladding could result in the loss of rod-bundle fuel geometry.
Criteria for cladding embrittlement in paragraph (a) above are more stringent than melting criteria would be; therefore, additional specific criteria are not used.
(d) Fuel Rod Ballooning:
To meet the requirements of Appendix K of 10 CFR Part 50 as it relates to degree of swelling, burst strain and flow blockage resulting from cladding ballooning (swelling) must be taken into account in the analysis of core flow distribution.
Burst strain and flow blockage models must be based on applicable data (such as Refs. 10, 12, and 13) in l
such a way that (1) the temperature and differential pressure l
at which the cladding will rupture are properly estimated (see paragraph (h) of subsection II.A.2), (2) the resultant degree of cladding swelling is not underestimated, and (3) the asso-ciated reduction in assembly flow area is not underestimated.
4.2-7 Rev. 2 - July 1981 l
The flow blockage model evaluation is provided to the Reactor Systems Branch for incorporation in the comprehensive ECCS evaluation model to show that the 2200*F cladding temperature and 17% c.ladding oxidation limits are not exceeded.
The reviewer should also det. ermine if fuel rod ballooning should be included in the analysis of other accidents invciving system depressurization.
(e) Structural Deformation:
Analytical procedures are discussed in Appendix A, " Evaluation of Fuel Assembly Structural Response to Externally Arplied Forces."
B.
Description and Design Drawings The reviewer should see that the fuel system description an" design drawings are complete enough to provide an accurate repres~itation and to supply information needed in audit evaluations.
Completeness is a matter of judgment, but the following fuel system information and associated tolerances are necessary for an acceptable fuel system description:
Type and metallurgical state of the cladding Cladding outside diameter Cladding inside diameter Cladding inside roughness Pellet outside diameter Pellet roughness Pellet density Pellet resintering data Pellet length Pellet dish dimensions Burnable poison content Insulator pellet parameters Fuel column length Overall rod length Rod internal void volume Fill gas type and pressure Sorbed gas composition and content Spring and plug dimensions Fissile enrichment Equivalent hydraulic diameter Coolant pressure The following design drawing have also been found necessary for an acceptable fuel system description:
Fuel assembly cross section Fuel assembly outline Fuel rod schematic Spacer grid cross section Guide tube and nozzle joint Control rod assembly cross section Control rod assembly outline Control rod schematic Burnable poison rod assembly cross section Burnable poison rod assembly outline Burnable poison rod schematic Orifice and source assembly outline 4.2-8 Rev. 2 - July 1981
C.
Design Evaluation The methods of demonstrating that the design bases are met must be reviewed.
Those methods include operating experience, prototype testing, and analytical predictions.
Many of these methods will be presented generically in topical reports and will be incorporated in the Safety Analysis Report by reference.
1.
Operating Experience Operating experience with fuel systems of the same or similar design should be described. When adherence to specific design criteria can be conclusively demonstrated with operating experience, prototype testing and design analyses that were performed prior to gaining that experience need not be reviewed.. Design criteria for fretting wear, oxidation, hydriding, and crud buildup might be addressed in this manner.
2.
Prototype Testing When conclusive operating experience is not available, as with the introduction of a design change, prototype testing should be reviewed.
Out-of-reactor tests should be performed when practical to determine the characteristics of the new design.
No definitive requirements have been developed regarding those design features that must be tested prior to irradiation, but the following out-of-reactor tests have been performed for this purpose and will serve as a guide to the reviewer:
Spacer grid structural tests Control rod structural and performance tests Fuel assembly structural tests (lateral, axial and torsional stiffness, frequency, and damping)
Fuel assembly hydraulic flow tests (lift forces, control rod wear, vibration, and assembly wear and life)
In-reactor testing of design features and lead-assembly irradiation of whole assemblies of a new design should be reviewed.
The following phenomena that have been tested in this manner in new designs will serve as a guide to the reviewer:
Fuel and burnable poison rod growth Fuel rod bowing Fuel assembly growth Fuel assembly bowing Channel box wear and distortion Fuel rod ridging (PCI)
Crud formation Fuel rod integrity Holddown spring relaxation Spacer grid spring relaxation l
Guide tube wear characteristics In some cases, in-reactor testing of a new fuel assembly design or a new design feature cannot be accomplished prior to operation of a full core of that design.
This inability to perform in-reactor 4.2-9 Rev. 2 - July 1981 l
m testing may result from an incompatability of the new design with the previous design.
In such cases, special attention should be given to the surveillance plans (see subsection II.D below).
3.
Analytical Predictions Some design bases and related parameters can only be evaluated with calculational procedures.
The analytical methods that are used to make performance predictions must be reviewed.
Many such reviews have been performed establishing numercus examples for the reviewer.
The following paragraphs discuss the more established review patterns and provide many related references.
(a) Fuel Temperatures (Stored Energy):
Fuel temperatures and stored energy during normal operation are needed as. input to ECCS performance calculations.
The temperature calculations require complex computer codes that model many different phenomena.
Phenomenological models that should be reviewed include the following:
Radial power distribution Fuel and cladding temperature distribution Burnup distribution in the fuel Thermal conductivity of the fuel, cladding, cladding crud, and oxidation layers Densification of the fuel Thermal expansion of the fuel and cladding Fission gas production and release Solid and gaseous fission product swelling Fuel restructuring and relocation Fuel and cladding dimensional changes Fuel-to-cladding heat transfer coefficient Thermal conductivity of the gas mixture Thermal conductivity in the Knudsen domain Fuel-to-cladding contact pressure Heat capacity of the fuel and cladding Growth and creep of the cladding Rod internal gas pressure and composition Sorption of helium and other fill gases Cladding oxide and crud layer thickness Cladding-to-coolant heat transfer coefficient
- Because of the strong interaction between these models, overall code behavior must be checked against data (standard problems or benchmarks) and the NRC audit codes (Refs. 14 and 15).
Examples of previous fuel performonce code reviews are given in References 16 through 20.
(b) Densification Effects:
In addition to its effect on fuel temperatures (discussed above), densification affects (1) core A
Although needed in fuel performance codes, this model is reviewed as described in SRP Section 4.4.
4.2-10 Rev. 2 - July 1981
power distributions (power spiking, see SRP Section 4.3),
(2) the fuel linear heat generation rate (LHGR, see SRP Section 4.4), and (3) the potential for cladding collapse.
Densification magnitudes for power spike and LHGR analyses are discussed in Reference 21'and in Regulatory Guide 1.126 (Ref. 22). -
To be acceptable, densification models should follow the guide-
_ lines of Regulatory Guide 1.126.
Models for cladding-collapse times must also be reviewed, and previous review examples are given in References 23 and 24.
(c) Fuel Rod Bowing:
Guidance for the analysis of fuel rod bowing is given in Reference 25.
Interim methods that may be used prior to compliance with this guidance are'given in Reference 26.
At this writing, the causes of fuel rod bowing are not well understood and mechanistic analyses of rod bowing are.not being approved.
(d) Structural Deformation:
Acceptance Criteria are discussed in Appendix A, " Evaluation of Fuel Assembly Structural Response to Externally Applied Forces."
(e) Rupture and Flow Blockage (Ballooning):
Zircaloy rupture and flow blockage models are part of the ECCS evaluation model and should be reviewed by CPB.
The models are empirical and should be compared with relevant data.
Examples of such data and previous reviews are contained in References 10, 12, and 13.
l (f) Fuel Rod Pressure:
T N thermal performance code for calculating temperatures discussed in paragraph (a) above should be used to calculate fuel rod pressures in conformance with fuel damage criteria of Subsection II.A.1, paragraph (f).
The reviewer should ensure that conservatisms that were incorporated for calculating temperatures do not introduce nonconservatisms with regard to fuel rod pressures.
(g) Metal / Water Reaction Rate:
To meet the requirements of Appendix K of 10 CFR Part 50 (Ref. 9) as it relates to metal / water reaction rate, the rate of energy release, hydrogen generatico, and cladding oxidation from the metal / water reaction should be cilculated using the Baker-Just equation (Ref. 27).
For non-LOCA l
applications, other correlations may be used if justified.
(h) Fission Product Inventory:
To meet the guidelines of Regulatory Guides 1.3, 1.4, 1.25 and 1.77 (Refs. 6, 28-30) as they relate to fission product release, the available radioactive fission product inventory in fuel rods (i.e., the gap inventory) is presently specified by the assumptions in those Regulatory Guides.
These assumptions should be used until improved calculational methods are approved by CPB (see Ref. 31).
D.
Testing, Inspection, and Surveillance Plans Plans must be reviewed fer each plant for testing and inspection of new fuel and for monitoring and survcillance of irradiated fuel.
4.2-11 Rev. 2 - July 1981
1.
Testing anJ fnspection of New Fuel Testing and inspection plans for new fuel should include verification of cladding integrity, fuel system dimensions, fuel enrichment, burnable. poison concentration, and absorber composition.
Details of the manufacturer's testing and inspection programs should be documented in quality control reports, which should be referenced and summarized in the Safety Analysis Report.
The program for onsite inspection af new fuel and control assemblies after they have bten delivered to the plant should also be described. Where the overall testing and inspection programs are esse. 111y the same as for previously approved plants, a statement to that effect should be made.
In that case, the details of the programs need not be included in the Safety Analysis 'enn t, but an appropriate reference should be cited and a (tabular) summary should be presented.
2.
On-line Fuel System Monitoring The applicant's on-line fuel rod failure detection methods should be reviewed.
Both the sensitivity of the instruments and the applicant's commitment to use the instruments should be evaluated.
References 32 and 33 evaluate several common detection methods and should be utilized in this review.
'urveillance is also needed to assure that B C control rods are not 4
osing reactivity.
Boron compounds are susceptible to leaching in the event of a cladding defect.
Periodic reactivity worth tests such as described in Reference 34 are acceptable.
3, Post-irraciation Surveillance A ost-irradiation fuel surveillance program should be described for each plant to detect anomalies or confirm expected fuel performance.
The extent of an acceptable prograc will depend on the history of the fuel design being considered, i.e., whether the proposed fuel design is the same as current operating fuel or incorporates new design features.
For a fuel design like that in other operating plants, a minimum acceptable program should include a qualitative visual examination of some discharged fuel assemblies from each refueling.
Such a program should be sufficient to identify gross problems of structural integrity, fuel rod failure, rod bowing, or crud deposition.
There should also be a enmmitment in the program to perform additional surveillance if unusual behavior is noticed in the visual examination or if plant instrumentation indicates gross fuel failures.
The surveillance program should address the disposition of failed fuel.
In addition to the plant-specific surveillance program, there t ould h
exist a continuing fuel surveillance effort for a given type, make, or class of fuel that can be suitably referenced by all plants using similar fuel.
In the absence of such a generic program, the reviewer should expect more detail in the plant-specific program.
For a fuel design that introduces new features, a more detailed surveillance program commensurate with the nature of the changes 4.2-12 Rev. 2 - July 1981 i
should be described.
This program should include appropriate qualitative and quantitative inspections to be carried out at interim and end-of-life refueling outages.
This surveillance program should be coordinated with prototype testing discussed in subsection II.C.2.
When prototype testing cannot be performed, a special detailed surveillance program should be planned for the first irradiation of a new design.
III. REVIEW PROCEDURES For construction permit (CP) applications, the review shoulo assure that the design bases set forth in the Prelisinary Safety Analysis Report (PSAR) meet the acceptance criteria given in subsection II.A.
The CP-review should further determine from a study of the preliminary fuel system design that there is reasonable assurance that the final fuel system design will meet the design bases.
This judgment may be based on experience with similar design's.
For operating license (0L) applications, the review should confirm that the design bases set forth in the Final Safety Analysis Report (FSAR) meet the acceptance criteria given in Subsection II.A and that the final fuel system design meets the design bases.
Much of the fuel system review is generic and is not repeated for each similar plant.
That is, the reviewer will have reviewed the fuel design or certain aspects of the fuel design in previous PSARs, FSARs, and licensing topical reports.
All previous reviews on which the current review is dependent should be referenced so that a completely documented safety evaluation is contained in the plant safety evaluation report.
In particular, the NRC safety evaluation repnrts for all relevant licensing topical reports should be cited.
Certain generic reviews have also been performed by CPB reviewers with findings issued as NUREG-or WASH-series reports.
At the present time these reports include Refereaces 9, 11, 21, 31, 32, 35, and 36, and they shoula all be I
appropriately cited in the plant safety evaluation report.
Applicable Regulatory Guides (Refs. 6, 22, 28-30, and 41) should also be mentioned in the plant
[
safety evaluation reports.
Deviation from these guides or positions should be explained.
After briefly discussing related previous reviews, the plant safety evaluation should concentrate on areas where the application is not identical to previously reviewed and approved applica*:ons and areas related to newly discovered problems.
Analytical predictions discussed in Subsection II.C.3 will be reviewed in PSARs, FSARs, or licensing topical reports.
When the methods are being reviewed, calculationt by the staff may be performed to verify the adequacy of the analytical methods.
Thereafter, audit calculations will not usually be performed to check the results of an approved method that has been submitted in a Safety Analysis Report.
Calculations, benchmarking exercises, and additional reviews of generic methods may be undertaken, however, at any time the clear need arises to reconfirm the adequacy of the method.
IV.
EVALUATION FINDINGS The reviewer should verify that sufficient information has been provided to satisfy the requirements of this SRP section and that the evaluation supports conclusions of the following type, to be included in the staff's safety evaluation report:
4.2-13 Rev. 2 - July 1981
" Sat the fuel system of the plant has l
The been m
.,.x w that (a) the fuel system will not be damaged as a result of normal operation and anticipated operational occurrences, (b) fuel damage during postulated accidents would not be severe enough to prevent control rod insertion when it is required, and (c) core coolability will always be main-tained, even after severe postulat v accidents and thereby meets the related a
requirements of 10 CFR Part 50, 95u.46; 10 CFR Part 50, Appendix A, General Design Criteria 10, 27 and 35; 10 CFR Part 50, Appendix K; and 10 CFR Part 100.
This conclusion is based on the following:
1.
The applicant has provided sufficient evidence that these design objectives will be met based on operating experience, prototype testing, and analytical predictions.
Those analytical predictions dealing with structural response, control rod ejection (PWR) or drop (BWR), and fuel densification have been performed in accordance with (a) the guidelines of Regulatory Guides 1.60, 1.77, and 1.126, or methods that the staff has reviewed and found to be acceptable alternatives to those Regulatory Guides, and (b) the guidelines for
" Evaluation of Fuel Assembly Structural Respor.se to Externally Applied Forces" in Appendix A to SRP Section 4.2.
2.
The applicant has provided for testing and inspection of new fuel to ensure that it is within design tolerances at the time of core loading.
The applicant has made a commitment to perform on-line fuel failure monitoring and postirradiation surveillance to detect anomalies or confirm that the fuel has performed as expected.
The staff concludes that the applicant has described methods of adequately predicting fuel rod failures during postulated accidents so that radioactivity releases are not underestimated and thereby meets the related requirements of 10 CFR Part 100.
In meeting these requirements, the applicant has (a) used the fission product release assumptions of Regulatory Guides 1.3 (or 1.4),
1.25, and 1.77 and (b) performed the analysis for fuel rod failures for the rod ejection eccident in accordance with the guidelines of Regulatory Guide 1.77 or with methods that the staff has reviewed and found to be an acceptable alternative to Regulatory Guide 1.77.
V.
IMPLEMENTATION The following is intended to provide guidance to applicants and licensees regarding the NRC staff's plans for using this SRP section.
Except in those cases in which the applicant proposes an acceptable alternative method for complying with specified portions of the Commission's regulations, the method described herein will be used by the staff in its evaluation of conformance with Commission regulations.
Implementation schedules for conformance to parts of the method discussed herein are contained in the referenced regu'atory guides and NUREGs.
V I'.
REFERENCES 1.
10 CFR Part 50, Appendix A, " General Design Criteria for Nuclear Power Plants."
s 4.2-14 Rev. 2 - July 1981
2.
10 CFR Part 100, "Reactu.' site Criteria."
3.
10 CFR Part 50, 950.46, " Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors."
4.
" Rules for Construction of Nuclear Power Plant Components," ASME Boiler and Pressure Vessel Code,Section III, 1977.
5.
W. J. O'Donnel and B. F. Langer, " Fatigue Design Basis for Zircaloy Components," Nucl. Sci. Eng. 20, 1 (1964).
6.
Regulatory Guide 1.77, " Assumptions Used for Evaluating a Control Rod Ejection Accident for Pressurized Water Reactors."
7.
" Standard Specification for Sintered Uranium Dioxide Pellets," ASTM Standard C776-76, Part 45, 1977.
8.
K. Joon, " Primary Hydride Failure of Zircaloy-Clad Fuel Rods," Trans. Am.
Nucl. Soc. 15, 186 (1972).
9.
10 CFR Part 50, Appendix K, "ECCS Evaluation Models."
10.
D. A. Powers and R. O. Meyer, " Cladding Swelling and Rupture Models for LOCA Analysis," USNRC Report NUREG-0630, April 1980.
11.
" Technical Report on Densification of Light Water Reactor Fuels," AEC Regulatory Staff Report WASH-1236, November 14, 1972.
12.
F. Erbacher, H. J. Neitzel, H. Rosinger, H. Schmidt, and K. Wiehr, " Burst Criterion of Zircaloy Fuel Claddings in a LOCA," ASTM Fifth International Conference on Zirconium in the Nuclear Industry, August 4-7, 1980, Boston, Massachusetts.
13.
R. H. Chapman, "Multirod Burst Test Program Progress Report for January-June 1980," Oak Ridge National Laboratory Report NUREG/CR-1883, March 1981.
14.
C. E. Beyer, C. R. Hann, D. D. Lanning, F. E. Panisko and L. J. Parchen,
" User's Guide for GAPCON-THERMAL-2: A Computer Program for Calculating the Thermal Behavior of an Oxide fuel Rod," Battelle Pacific Northwest Laboratory Report BNWL-1897, November 1975.
15.
C. E. Beyer, C. R. Hann, D. D. Lanning, F. E. Panisko and L. J. Parchen, "GAPCON-THERMAL-2:
A Computer Program for Caculating the Thermal Behavior of an Oxide Fuel Rod," Battelle Pacific Northwest Laboratory Report BNWL-1898, November 1975.
- 16. R. H. Stoudt, D. T. Buchanan, B. J. Buescher, L. L. Losh, H. W. Wilson and P. J. Henningson, " TACO - Fuel Pin Performance Analysis, Revision 1,"
Bacock & Wilcox Report BAW-10087A, Rev. 1, August 1977.
17.
" Fuel Evaluation Model," Combustion Engineering Report CENPD-139-A, July 1974 (Approved version transmitted to NRC April 25, 1975).
18.
" Supplement I to the Technical Report on Densification of General Electric Reactor Fuels," AEC Regulatory Staff Report, December 14, 1973.
4.2-15 Rev. 2 - July 1981
19.
"Tecruier nepon on sann fication ur Em. duclear PWR Fuels," AEC Regu: 3 tory Staff Report, February 27, 1975.
20.
Letter from J. F. Stolz, NRC, to T. M. Anderson, Westinghouse,
Subject:
Safety Evaluation of WCAP-8720, dated February 9, 1979.
21.
R. O. Meyer, "The Analysis of Fuel Densification," USNRC Report NUREG-0085, July 1976.
22.
Regulatory Guide 1.126, "An Acceptable Model and Related Statistical Methods for the Analysis of Fuel Densification."
23.
Memorandum from V. Stello, NRC, to R. C. DeYoung,
Subject:
Evaluation of Westinghouse Report, WCAP-8377, Revised Clad Flattening Model, dated January 14, 1975.
24.
Memorandum f rom D. F. Ross, NRC, to R. C. DeYoung,
Subject:
CEPAN --
Method of Analyzing Creep Collapse of Oval Cladding, dated February 5, 1976.
25.
Memorandum from D. F. Ross, NRC, to D. B. Vassallo,
Subject:
Request for Revised Rod Bowing Topical Reports, dated May 30, 1978.
26.
Memorandum from D. F. Ross and D. G. Eisenhut, NRC, to D. B. Vassallo and K. R. Goller,
Subject:
Revised Interim Safety Evaluation Report on the Effects of Fuel Rod Bowing in Thermal Margin Calculations for Light Water Reactors, dated February 16, 1977.
27.
L. Baker and L. C. Just, " Studies of Metal-Water Reactions at High Temperatures, III.
Experimental and Theoretical Studies of the Zirconium -
Water Reaction," Argonne National Laboratory Report ANL-6548, May 1962.
28.
Regulatory Guide 1.3, " Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Boiling Water Reactors."
29.
Regulatory Guide 1.4, " Assumptions Used for Evaluating the Pottntial Radiological Consequences of Loss-of-Coolant Accident for Pressurized Water Reactors."
30.
Regulatory Guide 1.25, " Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors."
31.
"The Role of Fission Gas Release in Reactor Licensing," USNRC Report NUREG-75/077, November 1975.
32.
B. L. Siegel and H. H. Hagen, " Fuel Failure Detection in Operating Reactors,"
USNRC Report NilREG-0401, March 1978.
33.
W. J. Bailey, et al., " Assessment of Currers Onsite Inspection Techniques for LWR Fuel Systems," Battelle Pacific Ncrthwest Laboratory Report NUREG/CR-1380, Vol.1, July 1980, Vol. 2, January 1981.
l i
l 34.
" Safety Evaluation Report Related to Operation of Arkansas Nuclear One, l
Unit 2," USNRC Report NUREG-0308, Supp. 2, September 1978.
l l
4.2-16 Rev. 2 - July 1981 l
l
35.
B. L. Siegel, "EvalJation of che Behavior of Waterlogged fuel Rod Failures in LWRs," USNRC Report NUREG 0308, March 1978.
36.
R. O. Meyer, C. E. Beyer and J. C. Voglewede, " Fission Gas Release from fuel at High,Burnup," USNRC Report NUREG-0418, March 1978.
37.
R. L. Grubb, " Review of LWR Fuel System Mechanical Response with Recommendations for Component Acceptance Criteria," Idaho National Engineering Laboratory, NUREG/CR-1018, September 1979.
38.
R. L. Grubb, " Pressurized Water Reactor Lateral Core Response Routine, FAMREC (Fuel Assembly Mechanical Response Code)," Idaho National Engineering Laboratory, NUREG/CR-1019, September 1979.
39.
R. L. Grubb, " Technical Evaluation of PWR Fuel Spacer Grid Response Load Sensitivity Studies," Idaho National Engineering Laboratory, NUREG/CR-1020, September 1979.
40.
S. B. Hosford, et al., " Asymmetric Blowdown Loads on PWR Primary Systems,"
USNRC Report NUREG-0609, January 1981.
41.
Regulatory Guide 1.60 " Design Response Spectra for Seismic Design of 3
Nuclear Power Plants e
b i
i 4.2-17 Rev. 2 - July 1981
U.S. Nuclear Regulatory Commission
- v..~..s of Nuclear Reactor Regulation APPENDIX A EVALUATION OF FUEL ASSEMBLY STRUCTURAL RESPONSE TO EXTERNALLY APPLIED FORCES TO STANDARD REVIEW PLAN SECTION 4.2 A.
BACKGROUND Earthquakes and postulateo 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 contrrl rod insertion when required during these low pr ; ability accidents.
This Appendix describes the review that should oe performed of the fuel assembly structural tesponse to seismic and LOCA loads.
Background material for this Appendix is given in References 37-30.
B.
ANALYSIS OF LOADS 1.
Input Input for the fuel assembly structural analysis comes from results of the primary coolant system and reactor internals structural,
analysis, 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 loadings in the primary coolant system and reactor internals analysis, and (b) transient 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 in accordance with the guidelines of Regulatory Guide 1.60 (Refs 41).
2.
Methods Analytical metnods used in performing structural response analyses should 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) should also be reviewed.
Experimental verification of the analytical rapresentation of the fuel assembly components should be provided when practical.
4.2-18 Rev. O
l i
A sample problem of a simplified nature should i,e worked by the applicant and compared by the reviewer with either hand calculations or results generated by the reviewer with an independent code (Ref. 38).
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 i
be made (e.g., one might use a 3-assembly core region with continuous sinusoidal input).
The sample problem should be designed to exercist' 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, but rather to alert the reviewer to major discrepancies l
so that an explanation can be sought.
3.
Uncertainty Allc,wances f
i 4
The fuel assembly structural models and analytical methods are probably conservative and input parameters are also conservative.
However, to ensure that the fuel assembly analysis does not introduce any non-convervatisms, two precautions should be taken:
(a) If it a
i 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 assortated with steam flashing that affects only the PWR I
fuel assembly analysis.
(b) 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 t10% variations in input amplitude and frequency; variations in amplitude and frequency y
thould be made separately, not simultaneously.
A factor should be i
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 of 35%, the sensitivity factor 3
4 would be 1.2.
Since resonances and pronounced sensitivities may be i
plant-dependent, the sensitivity analysis should be performed on a plant-by plant basis until the reviewer is confident that further i
j sensitivity analyses are unnecessary or it is otherwise demonstrated j
that the analyses performed are bounding.
4.
Audit Independent audit calculations for a typical full sized core should be performed by the reviewer to verify that the overall structural representation is adequate.
An independent audit code (Ref. 38) should be used for this audit during the generic review of the analytical methods.
t 5.
Combination of Loads To meet the requirements of General Design Criterion 2 as it relates I
to combining loads, an appropriate combination of loads from natural phenomena and accident conditions must be made.
Loads on fuel assembly components should be calculated for each input (i.e.,
seismic and LOCA) as described above in Paragraph 1, and the resulting loads should be added by the square-root-of-sum-of-squares (SRSS) 4.2-19 Rev. O i
2
--y,
,-+,-,rw,.9we,y,-
m -wee rw--y-v,~, w, g g w + w w m, p,, mwm,wy,,,y-ee rw -, mm--- y m wmww w.,2----en--,*-t--r---,g---enn-w--g
= e v et +m--w
-- p e----e-e
- e w w w +e - es-- ev t
method.
These combined loads should be compared with the component strengths described in Section C according to the acceptance criteria in Section D.
C.
DETERMINATION OF STRENGTH 1.
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.
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 crushing load P(crit) has been suitably selected from the load vs-deflection curves.
Because of the potential for different test rigs to introduce measurement variations, an evaluation of the grid strength test equipment will be included as part of the revies of the test procedure.
The consequences of grid deformation are small. Gross deformation of grids in many PWR assemblies would be needed to interfere with control rod insertion during all SSE (i.<e., buckling of a few isolated grids could not displace guide tubes significantly from their proper location), 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 appropriate, and the allowable crushing load P(crit) should be the 95% confidence level on the true mean as taken from the distribution of measurements on unirraditted production grids at (or corrected to) operating temperature. While P(crit) will increase with irradiation, ductility will be reduced.
The extra margin in P(crit) for irradiated grids is thus assumed to offset the unknown deformation behavior of irradiated grids beyond P(crit).
2.
Components Other than Grids Strengths of fuel assembly components other than spacer grids may be deduced from fundamental material properties or experimentation.
Supporting evidence for strength values should be supplied.
Since structural 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 Vessel Code values and procedures may be used where appropriate for determining yield and ultimate strengths.
Specification of allowable values may follow the ASME Code require-ments and should include consideration of buckling and fatigue effects.
l 4.2-20 Rev. 0
D.
ACCEPTANCE CRITERTA 1.
Loss-of-Coolant Accident Two principal criteria apply for the LOCA:
(a) fuel rod fragmentation must not occur as a direct result of the blowdown loads, and (b) the 10 CFR Part 50, 950.46 temperature and oxidation limits must not be exceeded.
The first criterion is satisifed if the combined 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 combined loads on the grids remain below P(crit), as defined above, then no significant distortion of the fuel assembly would occur and the usual ECCS analysis is sufficient.
If combined grid loads exceed P(crit), then grid deformation must be assumed 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.
Loads from the worst-case LOCA that requires control rod insertion must be combined with the SSE loads, and control rod insertability must be demonstrated for that combined load.
For a PWR, if combined loads on the grids remain below P(crit) as defined above, then significant deformation of the fuel assembly would not occur and con-trol rod insertion would not be interfered with by lateral displacement of the guide tubes.
If combined loads on the grids exceed P(crit),
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 demonstrate control blade insertability:
(a) combined loads on the channel box must remain below the allowable value defined above for components other than griJs; otherwise, additional analysis is needrd to snow that deformation is not severe enough to prevent control blace insertion, and (b) vertical liftoff forces must not unseat the lower tieplate from the fuel support piece such that the resulting loss of lateral fuel bundle positioning could interfere with control blade insertion.
2.
Safe Shutdown Earthquake Two criteria apply for the SSE:
(a) fuel rod fragmentation must not occur as a result of the seismic loads, and (b) control rod inserta-bility must be assured.
The first criterion is satisfied by the criteria in Paragraph 1.
The second criterion must be satisfied for SSE loads alone if no analysis for combined loads is required by Paragraph 1.
4.2-21 Rev. 0
..