ML19221B059
| ML19221B059 | |
| Person / Time | |
|---|---|
| Issue date: | 03/31/1979 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-75-087, NUREG-75-087-04.2, NUREG-75-87, NUREG-75-87-4.2, SRP-04.02, SRP-4.02, NUDOCS 7907120330 | |
| Download: ML19221B059 (15) | |
Text
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NUREG 75'087 f p nac %
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0 yi._'6Ic'l U.S. NUCLEAR REGULATORY COMMISSION
,q-o WciQ'#\\.~...j PTANDARD REVIEW PLAN OFFICE OF NUCLEAR REACTOR RMULATION SECTION 4.2 FUEL SYSTEM CESIGN REVIEW RESPCNSIBILITIES P aary - Core Performance Branch (CPB)
Secondary - None I.
AREAS OF REVIEW The thermal, mechanical, and materials design of th.
fuel system is evlluated by CPB.
The fuel system consists of: arrays (a<semblies or 'undles) of fuel rods includina fuel pellets, insulator pellets, springs, tubular claddi g, end closures, hydrogen octters, and fill gas; bornakle poison rods including coi"p'nents 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 cxtend from the coupiing interface of the control rod drive o hanism i..o the core.
The Mechanical Engineering Branch reviews the design of con ;; 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 ovide assurance that (a) the fuel system is not damaged as a result of normal operatian and anticipated operational occurrences, (b) fuel system damage is never so severe as to prevent control rcd inser-tion when it is required, (c) the number of fuel rod failures is r at underestimated for postulated accidents, and (d) coolability is always maintained.
"Not damaged," as used in the above statement, means that fuel rods da not fail, that f'
. system dimensions remain within operational tolerances, and that functional capabilities are not reduced below those assumed in the safety analysis. " Fuel rod failure" means that the fuel rod leaks and that the first fission product barrier (the cladding) has, therefore, been breached. Ccolability, 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.
Fuel failure criteria and coolability criteria that involve thermal-hydraulic considera-tions are provided by the Core Performance Branch to the Analysis Branch for implementa-tion in 5RP Section 4.4.
The Analysis Branch provides hydraulic loads under SRP Section 4.4 to the Core Performance Branch for evaluation (in SRP Section 4.2) of fuel assembly mechanical response under normal and accident conditions. The avaiiable radio-active fission product inventory in fuel rods (i.e., the gap inventory expressed as a release frastion) is provided to the Accident Analysis Branch for use in estimating the radiological consequenct.
.f plant releases.
USNRC STANDARD REVIEW PLAN
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The fuel system review covers the following specific areas.
A.
Design Bases The principles and related assumptions of the fuel system design should be reviewed.
These bases may be expressed as explicit numbers or as general criteria. The bases will include traditional fuel design limits, ir.dustry codes and standards, and limits related to the safety analysis (i.e., related to fuel damage, rod failure, or coolauility requirements). Once such limits are approved in the safety evaluation re p', r t, they become the specified acceptable fuel design limits referred to in General Design Criterion 10 (Ref. 1).
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 produ~.t specifications rather than process specifications.
C.
Design Evaluation The performance of the fuel system during normal operation, anticipitated opera-tional occurrences, and postulated accidents is reviewed to determine if ail design bases are met.
Th 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 anaiyses, and other information.
D.
Testing, In:nection, and Surveillance Plans Testing and inspection of new fuel is performed by the licensee to ense e that the fuel is fabricated in accordance with *le 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 ccatrol rods con-taininc B C should be performed to ensure against reactivity loss. The testing, 4
inspection, and survei' lance plans alnnq with their reporting provisions are reviewed by CPB to ensure that the important fuel design considerations have been addressed.
11.
ACCEPTINCE CRITERIA A.
vesign Bases The fuel system design bases must reflect the four objectives described in Subsection I, Areas of Reviaw.
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:
}
b Rev. 1 4.2 -2
1.
Fuel System Damage Fuel system damage includes fuel rod failure, which is discussed below in Subsection II-A-2.
In addition to preci ding fuel rod failure, fuel dam a e m
criteria should assure that fuel syster dimensions remain within operational tolerances and that functional capabilities are not reduced below those usumed in the safety analg is Such iamage criteria should include the following to be completa.
(a) Stress, strain, or loading limits for spacer grids, guide tubes, thimbles, fuel rods, co.ntrol rods, channel boxes, and other fuel system structural n mbers should be provided. Stress limits that are obtained by methods similar to those given in Section III of the ASME Code (Ref. 2) are acceptable. Other proposed limits must be justified.
(b) The cumulative number of strain fatiguo cycles on the structural members mentioned in paragraph (a) abi.e should be significantly less than the design fatigue lifetime, whicn is based on appropriate data and includes a safety factor of 2 on scress amplitude or a safety factor of 20 on the number of cycles (Ref. 3).
dther proposed limits must be justified.
(c) Fretting wear at contact points on the structural mernbers mentioned in paragraph (a) above should be limited. The allowable fretting saar should be stated in the safety analysis report and the stress ar.d fatigue limits in paragraphs (a) and (b) above should presume the exister.ce of this wear.
(d) 0xidation, hydriding, and the buildup of corrosion products (crud) should be limited. Allowable oxidation, hydriding, ar.d crud levels should Le 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 by the Analysis Branch as tscribed in SRP Section 4.4.
(e) Dimensional changes such as rod bowirig or irradiation growth of fuel rods, control rods, ar.d guide tubes need not be limited to set values (i.e., damage limits), but they must be included in the design analysis to establish operational tolerances.
(f) Fuel and burnable poison red internal gas pressures should remain below the nominal system pressure during normal operation unless otherwise justified.
(g) Worst-case hydraulic loads for normal operation sh.uld not exceed the holddown capability of the fuel assembly (either g'avity or holddown springs). Hydraulic loads for this eva uation are provided by the l
Analysis Branch as described in SRP Section 4.4.
146 346 4.2-3 Fev, I
(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 fuel rod failure is defined as the loss o' fuel rod he mesicity.
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 known failure mechanisms during.iormal operation and anticipated operational occurrences. Fuel rod failures can be caused by overheating, pellet / cladding interaction (PCI), hydriding, cladding collapse, burstirig, mech.
al fractur-ing, and fretting. A fuel failure criterion should be given for each known failure mechanism. Such criteria shculd address the following to be complete.
(a) Overheating: No useful mechanistic criteria exist at present for fuel rod failure due to overheating. However, to show that overneating will be avoided, it will be sufficient to show that (1) claddinc, temperatures do not greatly exceed the coolant temperature and (2) fuel melting does not occur.
Adequate cooling is assumed to exist when the thermal margin criterion to limit the departur, from nucleate boiling (DNB) or boiling transition condition in the core is satisfied. The review of this criterion is detailed in SRP Section 4.4.
For a severe reactivity initiated accident (RIA), Regulatory Guide 1.77 (Ref. 4) relies on a DNB criterion for determining failures in PWRs, whereas a radial average energy density c3 170 cal /g is accepted for BWRs under zero and low power conditions. Other limits may be more accurate for an RIA, but continued approval of these limits may be given until generic studies yield improvements.
AlthouF a DNB criterion is sufficie t to demonstrate the avoidance of overheating from a deficient cooling mechanism, it is not a necessary condition (i.e., DNB is not a tailute mechanism) and other mechanistic methods may be acceptable. Although there is at present little experience with other approachee, positions recommending differe...
criteria should address cladding temperature, pressure, time duration, oxidation, and embrittlement.
The second criterion used to assure that the cladding does not overheat is that fuel melting will not occur. There n uld otherwise be cor.cern.
that molten fuel night contact the cladding and cause local hotspots.
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Pev. 1 4,p_4
This criterion also avoids the axial relocation of molten fuel that could cause local overheating.
(b) Pellet / Cladding Interaction (PCI): There is no current criterion for fuel failure resulting from PCI, and the design basis can onl, be stated generally. Two related criteria shoulu Le applied, but they are not sufficient to preclude PCI failures: (1) the uniform strain of the cladding should not exceed 1%.
In this context, uniform strain (e'astic and inelastic) ' seiined as transient-induced deformation with gage lengths corresponding to cladding dimensions; steady-state creepdown and irradiation growth are excluded. Although observing this stidin limit may preclude some PCI failures, it will not preclude the corrosion-assisted failures that occur at low strains, nor will it preclude highly
?ocalized overstrain failures. (2) Fuel melting should be avoided. The large volume increase associated with melting may cause a pellet with a molten core to exert a stress on the cladding. Such a PCI is avoided by avoiding fuel melting. Note that this same criterion was invoked in paragraph (a) to ensure that overheating of the cladding would not occur.
(c) Hydriding: Hyoriding as a cause of failure (i.e., primary tdriding) is prevented by keeping the level of moisture and other hydroc,enous impuri-ties very low during fabrication. Acceptable muisture ievels for Zircaloy-clad uranium oxide fuel should be no greater than 20 ppm.
Current ASTM specifications (Ref. 5) for U0 fuel pellets state an equiva-2 lent limit of 2 ppm of hydrogen from all sources. For other materials clad in Z:rcaloy tubing, an equivalent quantity of moisture or hydrogen 3
can be tolerated. A moisture level of 2 mg H O per cm of hot void 2
volume within the Zircaloy cladding has been shown (Ref. 6) to be insufficient for primary hydride formation.
(d) Cladding Collapse: If axial gaps in the fuel pellet column occur due to densification, the cladding has the potential of collapsing into a gap (i.e., flattening). Because of the large local strains thF 3ccompany this process, collapsed (flattened) cladding is assumed tc fail.
(e) Bursting: Zircaloy cladding will burst (rupture) under certain combina-tions of temperat re, heating rate, and differential oressure. Although fuel suppliers may use differ >nt rupture-temperature vs differential-pressure curves, an acceptable curve should be similar t3 the one deter-mined by Oak Ridge National Laboratory (Ref. 7).
This criterion is included in the ECCS evaluation model required by Appendix K (Ref. 8).
(f) Mechanical Fracturing: A mechanical 'racture r?fers to a defect in a fuel roo caused by an externally applied force such as a hydraulic. load or a load derived from core plate motion. Cladding integrity may be 146 348
"'I 4.2-5
assumed if the applied stre, less than 90% of the irradiated yield stress at the appropriate ter.perature. Other proposed limits must be justified.
(g) Fretting: Fretting is a potential cause of fuel failure, but it is a gradual process that would not be effective during the brief duration of an abnormai operational occurrence or a postu;oted accident. Therefore, the fretting wear requirement in paragraph (c) of Subsection II-A-1, Fuel Damage, is sufficient to preclude fuel failures caused by fretting during transients.
3.
Fuel Coolability Coolability has traditionally implied that the fuel assembly retains its rod-bundle geometry with adequate coolant channels to permit removal of resid-ual heat. Reducticn of coolability can result from cladding embrittlement, violent expulsion of fuel, generalized cladding melting, gross str tural deformation, and extreme coplanar fuel rod ballooning. Coolability criteria should include the following to be complete:
(a) Cladding Embrittlement: Oxygen contamination and hydriding in Zircaloy cladding are the primary causes of cladding embrittlement. For the LOCA, Appendix K addresses these phenomena with a criterion of 2200 F peak cladding temperature and a criterion of 1/% maximum cladding oxidation.
(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; corscquently, the lower peak cladding temperature limit of 1800 7 previously described in Reference 9 is ne longer needed in CP and OL reviews.) Specific temperatuce and oxidation criteria have not been derived for other accidents, but should they be needed, Appendix K can be used as guidance.
(b) 'v'iolent Expulsion o f ruel: In severe reactivity initiated accidents (RIAs), such as rod ejectior. in a PWP or red drop in a BWR, the large and rapid deposition cf en?rgy in the fuel can result in melting, fragmenta-tion, 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 produce pressure pulses in the primary system. Observing the 28J cal /g limit specified by Regulatory Guide 1.77 pre w ts widespread fragmentation and dispersal of the fuel and avoids generating pressure pulses in the prin.ary system during an RIA. This 280 cal /g limit shoulc; be used for PWRs and BWRs.
(c) Generalized Cladding Melting-Generalized (i.e., non 'ocal) melting of the cladding could result in the loss of rod-bundle fuel geometry.
O Rev, 1 4.2-6
Criteria for cladding embrittlement in paragrap7 (a) above are more stringent than melting criteria would be; therefore, additional specific criteria are not used.
(d) Structural Deformation: It is sufficient, but not necessary, to show that no permanent teformation of spacer grids, channel boxes, fuel rods, guide tub n, and other structural members results from mechanical loads.
This can be accomplished by demonstrating that the expected load is less than the grid crushing strength or that the load produces 3 maximum stress that is less than the material's irradiated yielt' stress at the appropriate temperature.
Some structural deformation may not always be avoidab!e. In those cases, the legree of deformation must be determined before establishing coolability. For the LOCA, structural deformation should not cause the 2200 F cladding temperature and 17% cladding oxidation limits to be exceeded. 5)ecific criteria have not been derived for other accidents, but, should they be needed, Appendix K should be used as guidance.
A Branch Technical Position is under development to quantify margins and other aspects of structura' a formation criteria. Reference 10 will also provide some gaidance to the reviewer on this matter.
(e) Fuel Rod Balloonirg: For the LOCA analysis, Appendix K requires that flow blockage resulting from cladding ballooning (swelling) be taken into account in the analysis of core flow distribution. Flow blockage models must be based on applicable data (Refs. 7, 11, and 12) in such a way that (1) the temperature and differential pressure at which the cladding will rupture are properly estimated (see paragraph (e) of Subsection II-A-2),
(2) the resultant degree of cladding swelling is not underestimated, and (3) the associated reduction in assembly flow area is not underestimated.
The flow blockage model ev mtion is provided to the Analysis Branch for incorporation in the comprehensive ECCS model evaluation to show that the 2200 F cladding temperature and 17% cladding oxidation limits are not exceeded. The reviewer should also determine if fuel rod ballooning shoalo be included in th2 analysis of other accidents involving system depre N urization.
B.
Description and Design Drawings The reviawer should see that the fuel system description and design drawings are complete enough to provide an accurate representation and to supply information needed in audit evaluations. Ccmpleteness is a matter of judgment, but the follor ing fuel system information and associated tolerances are necessary for an accept-able fuel system description:
i' 7[O l+ Q JJU 4.2-7 Rev-I
Type and metallurgical state of the cladding Cladding outside diameter Cladding inside diameter Cladding i.d..oughness Pellet outside diameter Pellet roughness Pellet density Pellet resintering data Pellet length Pellet dish dimensions Burnable poison content Insulator pellet parameters Fuel column length length Overall rm Rad int (:nal void volume Fill gas type and pressure Sorbed gas composition and content Spring and plug dimension Fissile enrichment Equivalent hydraulic diameter Coolant pressure The following design drawings 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 nczzle 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 C.
Design Evaluation The methods of demonstrating that the design bases are met * ;t be reviewed. Those methods include operating experience, prototype testing, -
nalytical predictions.
Many of thes< methods will be presented generically in topi 21 reports and will be incorporate oSARs and FSARs 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 experi.nce, prototype testing and design analyses that were performed prior to gaining that experience needed not be reviewed.
Design criteria for fretting wear, oxidation, hydriding, and crud buildup.
raight be ' dressed in this manner.
2.
Prototype Testing khen conclusive operating experience is not available, as w; che introduc-tion of a design change, prototype testing should be reviewed. Out-of reactor tests should be performed when practical to determine the characteristics of Rey. 1 4.2-8 146 351
the new design. No definitive requirements have been developed regarding those design features that must be tested prior to irradiation, but the follow-ing out-of-reactor tests have been performed for this purpose and will serve 9
as a guide to t' e reviewer:
Spacer grid structural tests Control rod structural and performance tests Fuel assembly structural tests (lateral, axial and torsional stif fness, frequ2ncy, and dampir.g)
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 relax > tion Spacer grid spring reiaxation Guide tube wear characteristics In some cases, in-reactor testing of a new fuel assembly desi n or a new 9
design feature cannot be accomplished prior to operation of a full core of that design. This inability to perform in-rcactor 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 ( ee Subsection II-D below).
3.
Analytical Predictions Some design bases and related parameters c3n only be evaluated with calcula-tional procedures. The analytical methods that are used to make performance predictions must be reviewed. Many such reviews have been performed establish-ing numerous examples for the reviewer. The following paragraphs discuss the more established review patterns and provide many related references.
(a) Fuel Temperatures (Stored Energy)- Fuel temperature-and stored energy during normal operation are needed as input to ECCS performance calcula-tions. The temperature calculations require complex computer codes that model many different phenomena. Phenomenolor' ;.al models that soould be reviewed include the following:
Radial power distribution Fuel and cladding temperature distribution
(
Zf Burnup distribution in the fuel l
U J3 Thermal conducti >ity of the fuel, cladding, cladding crud, and oxidation layers 4.2-9 p,. v. i
Densification of the fuel Thermal expansion of the fuel and cladd.ag Fission gac nroduction 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 gates Cladding oxide and crud layer thickbess Cladding-to-coolant heat tran3'er coef ficient*
Because of the strong interaction between these models, overall code behavior must be checked against data (standard problems or benchnarks) and the NRC audit codes (Refs. 13 and 14).
Examples of previous fuel performance code reviews are given in References 15 through 18.
(b) Densification Effects: In addition to its effect on fuel temperatures (Sscussed above), densification affects (1) core power distributions (pcwer spiking, see SRP Section 4.3). (2) the fuel linear heat generation r ate (tHGR, see SRP Section 4.4), and (3) the potential for cladding collapse. Densification magnitudes for power spike and LHGR analyses are discussed in Reference 19 and in Regulatory Guide 1.126 (Ref. 20).
Models for cladding collapse times must alsn be reviewed, and previous r(view examples are given in References 21 and 22 (c) Fuel Rod Bouing: Guidance for the analysis of fuel rod bcwing is given in Reference 23.
Interim methods that may be used prior to compliance with this guidance are given in Reference 24.
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: The mechanical response of fuel assemblies to impact loads should be analyzed with a multi-assembly spring / mass repre-sentation. Because of the complexity and related uncertainties of these analyses, independent generic audit calculations should be performej to compare analytical methods with the NRC audit code. An example of such a review and audit calculation is given in Reference 25.
(e) Rupture and Flow Blockage (Ballooning): Zircalcy rupture and flow block-age models are part of the ECCS evaluatic i model end should be reviewed by CPB. The models are empirical and should be comparec with relevant data.
Examples of such data and a previous review are contained in References 7, 11, 12, and 26.
'Although needed in fuel performance codes, this model is reviewed by the Analysis Branch as described in SRP Section 4.4.
Rev. I
} k b
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4.2-10
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(f) Fuel Rod Pressure: The thermal performance code for calculat 7g temperatures discussed in paragraph (a) above should be used t calculate fuel rod pressures in conformance with fuel damage criterio of Subsection II-A-1, paragraph (f). The reviewer should en'ure that conser-vatisms that were incorporated for calculating temperatu et do not intro-duce nonconservatisms with regard to fuel rod pressure.
(g) Metal / Water Reaction Rate: The rate of c.ergy release, hydrogen genera-tion, and cladding oxidation from the metal / water reaction should be calculated using the Baker-Just equation (Ref. 27) as required by Appendix K.
For non-LOCA applications, other correlations may be used if justified.
(h) Fission Product Inventory: The available radioactive fission product inventory in fuel rods (i.e., the gap inventory) is presently specified by assumptions in Regulatory Guides (Refs. 4, 28-30). These assumptions sh':uld be used until improved calculational methods are approved by CPB (see Ref. 31).
D.
Testing, Inspection, and Surveillance Plans Plans must be reviewed for each plant for testing and inspection of new fuel and for monitoring and surveillance of irradiated fuel.
1.
Testing and Inspection 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 test-ing and inspection programs should be documented in quality control reports, which should be referenced and summarized in the safety analysis report. The program for on-site inspection of new fuel and control assemblies after they have been delivere o the plant should also be described. Where the overall testing and inspeution programs are essentially the same as for previously approved plants, a statement te that effect should be made.
In that case, the details of the programs neeo not be included in the safety analysis report, but an rppropriate ref2rence should De 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 'nstruments and the applicant's commitnent to use the instruments should be evaluated. Reference 32 evaluates several common detection methods and should be utilized in this review.
Surveillance is also needed to assure that B C control rods are not losing 4
reactivity. Boron compounds are susceptible to leaching in the event of a 146 354 V-I 4.2-11
cladding defect. Periodic reactivity worth tests such as described in Reference 33 are acceptable.
3.
Post-irradiation Surveillance A post-irradiation fuel surveillance program should be described for each plant to detect anomalies or confirm expected fuel performance. The extent of an acceptable program will depend on the history of the fuel design being considered, i.e., whether the proposed fuel design is the same as current cperating 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 cf some discharged fuel assemblies from each refueling. Such a program should be sufficient to identify gross prcblems of structural integrity, fuel rod failure, rod bowing, or crud deposition. There should also be a commitment in the program to perform additional surveillance if unusual behavicr is noticed in the visual examination or if plant instrumentation indicates gross fuel failures. TM surveillance program should address the dispcsition of failed fuel.
In addition to the plant-specific surveil'.nce program, there should esist a fuel continuing fuel surveillance effort for a given type, make, or class or that can bc 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 p agram commensurate with the nature of the changes shoulc be described. This program should include appropriate qualitative and quantitative inspectio m to be carried out at interim and end-of-life refueling outages. Inis surveillance program should be coordinated with prototype testing discussed in Subsection II-C-2.
When prototype testing cannot be peri.,rmed, 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 should assure that the design bases set forth in the preliminary 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 designs.
For operating license (OL) applications, the review should confirm that the design bases set forth in the final safety analysis report (FSAR) meet the acceptar.ce criteria given in Subsection II-A and that the final fuel system design meets the design bases.
O Rev. 1 4.2-12 1 4 6 3 5.,.3
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 avaluation is contained in the plant safety evaluation report. In particular, the NRC safety evaluaticn reports for all relevant licensing topical reports should be cited. Certain generic reviews have also been performed i;y CPB reviewers with findings issued as NUREG or WASH-series reports. At the present time these reports include References 9, 19, 31, 32, 34 and 35, and they should all be appropriately cited in the plant safety evaluation report. Applicable Regulatory Guides (Re's. 4, 20, 28-30) and Branch Technical Positions (there are none at present) should also be mentioned in the plant safety evaluation reports. Deviation from these guides or positions should be explained. After briefly discu> sing related previous reviews, the plant safety csalua-t i o t. avuld concentrate on areas where the application is not identical to previously reviewed and approved applications 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, calculations by the staff may be performed to verify the adequacy of the Enalvtical methods. Thereafter, audit calculations will not esually be performed to check the results tf an approved method that has been submitted in a safety analysis report. Calculations. benchmarking exercises, and additional reviews of generic methods may De unce.rtaken, imwever, at any time the clear need arises to reconfirm the adequacy of the trethod.
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 f ollowing type, to be included in the staf f's safety evaluation repcrt:
"The FJel system of the plant has been designed 50 that (a) the fuel system will not be damaged as a resuit of normal operation and anticipatad operational occurre.ces, (b) fuel damage during postulated accidents would not be se,ere enough to prevent (ontrol rod insertion when it is required, and (c) core coolability will always be maintained, even after severe postulated accidents.
N -lirant has provided suf ficient evidence that these design objectives will be met based on cperating experience, prototype testing, and analyt4 cal predictions.
'The applicant has described methods of adequately predicting fuel rod failures during postulated accidents 50 that radioactivity rele3ses are not underestimated.
"The applicant his ala previud 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 post-irradiation surveillance to detect anomalies or confirm that the fuel has performed as expected.
146 %6 4.2-13
~
"On the basis of our review of the fuel system design, we conclude that the applicant has met all the requirements of the applicable regulations, current regulatory posi-tions, and good engineering practice."
V.
REFERENCES 1.
10 CFR Part 50, Appendix A, f.eneral Design Criterion 10, " Reactor Design."
2.
"Rulos for Construction of Nuclear Power Plant Components," ASME Boiler and Pressure Vessel Code,Section III, 1977.
3.
W. J. O'Donnel and B. F. Langer, " Fatigue Design Basis for Zircaloy Components,"
Nucl. Sci. Eng. 20, 1 (1964).
4.
Regulatory Guide 1.77, " Assumptions Used for Evaluating a Control Rod Ejection Accident for Pressurized Water Reactors."
5.
" Standard Specificatior, for Sintered Uranium Dioxide Pellets," ASlM Standard C776-76, Part 45, 1977.
6.
K. Joon, " Primary Hydride Failure of Zircaloy-Clad Fuel Rods," Trans. Am. Nucl.
Soc. 15, 18C (1972).
7.
R. H. Chapman, 'Maltirod Burst Test Program O r rtarly Progress Report for April -
Juna 1977," Dak Ridge National Laboratory Report ORNL/NUREG/TM-135, December 1977.
8.
10 CFR Part 50, Appendix K, "ECCS Evali tion Models."
9.
" Technical Report on Densification of Light Water Reactor Fuels," AEC Regulatory Staff Report WASH-1236, November 14, 1972.
10.
" Safety Evaluation Report Related to Geration of North Anna Power Station," USNRC Report NUREG-0053, Supp. 7, August 1977.
11.
F. Erbacher, " Single and Multirod Tests, Transient and Steady State, Internal Conduction Heatiag," Fifth NRC Water Reactor Safety Rese3rch Information Meeting, Gaithersburg, Maryland, November l~,
1977.
12.
R. H. Chapman, "Some Preliminary Results of Single Rod and Multirod Tests With Internal Heaters," NRC Zircaloy Clidding Review Group Meeting, Silver Spring, Maryland, January 18, 1978.
13.
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.
14.
C. E. Beyer, C. R. Hann, D. D. Lanning, F. E. Panisko and L. J. Parchen, "GAPCON-THERMAL-2: A Computer Program for Calculating the Thermal Behavior of an Oxide Fuel Rod," Battelle Pacific Northwest Laboratory Report BNWL-1898, November 1975.
15.
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," Babcock &
Wilcox Report BAW-10087A, Rev. 1, August 1977.
16.
" Fuel Evaluation Model," Combustion Engineering Reprrt CENPD-139-A, July 1974 (Approved version transmitted to NRC April 25, 1975s.
17.
" Supplement I to the iechnical Report on Densification of General Electric Reactor Fuels," AEC Regulatory Staff Report, December 14, 1973.
18.
" Technical Report on Densification of Exxon Nuclear eWR Fuels," AEC Regulatory Staff Report, February 27, 1975.
Rev. 1 4.2-14
]/{[
19. R.0. Meyer, "The Analysis of Fuel Densification," USNRC Report NUREG-0085, July 1976. 20. Regulatory Guide 1.126, "An Acceptable Model and Related Statistical Methods for the Analysis of Fuel Densification." 'ubject: Evaluation of Westing-21. Memorandum from V. Stello, NRC, to R.C. CeYoung i house Report. WCAP-8377, Revised Clad Flattening Model, dated January 14, 1975. 22. Memorandum from D. F. Ross, NRC, to R. C. DeYoung,
Subject:
CEPAN -- Method of Analyzing Creep Collapse of Oval Cladding, dated February 5,1976. 23. Memorandum from D. F. Ross, NRC, to D. B. Vassallo,
Subject:
Request for Revised Rod Bowing Topical Reports, dated May 30, 19/8. 24. Memurandum from D. F. Ress and D. G. Eisenhut, NRC, to D. B. Vassallo and K. R. Coller,
Subject:
Revised Interim Safetv Fvaluation Report on the Effects of Fuel Rod Bowing in Thermal Margin Calculations for Light Water Reactors, date1 February 16, 1977. 25. R. L. Grubb, "PWR Fuel Assembly Mechanical Response Analysis, Idaho National Engineering Laboratory Report RE-5-76-164, September 1976; Amendment 1, RE-E-77-103, January 1977. 26. Letter from D. F. Ross, NRC, to A.E. 9erer, Combustion Engineering, dated March 22, 1978. 27. L. Baker and L. C. Just, "Stacies of Metal-Water Reactions at High Temperatures, III. Experimental and Theoretical Studies v' 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 toss of Coolant Accident for Boiling Water Reactors." 29. Regulatory Guide 1.4, " Assumptions Used for Evaluating the Potential Radiological Consequences of a Lods 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 "UREG-75/077, Noverrber 1975 32. B. L. Siegel and H. H. Hagen, "ruel Failure Detection in Operating Reactors," USNRC Report NUREG-0401, March 1978. 33. " Safety Esa1mation Report Related to Operation of Arkansas N'. clear One, Unit 2," USNRC Report NUREG-0308, Supp. 2 (to be issued). 34. B. L. Siegel, " Evaluation of tha Behavior of Waterlogged Fuel Rod failures in LWRs," USNRC Report NUREG-0303, Marcn 1978. 35. R. O. Meyer, C. E. Beyer and J. C. Voglewede, " Fission Gas Release from Fuel at High Burnup," U$NRC Report NUREG-0418, March 1978. 146 %8
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