NUREG-0032, Fuel Performance of Licensed Nuclear Power Plants Through 1974

NUREG-0032, Fuel Performance of Licensed Nuclear Power Plants Through 1974

ML20209A729
Person / Time
Issue date:	01/31/1976
From:	Bobe P NRC OFFICE OF MANAGEMENT AND PROGRAM ANALYSIS (MPA)
To:
References
NUREG-0032, NUREG-32, PB-249-788, NUDOCS 7908170429
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NUREG-0032 L l l i l l FUEL PERFORMANCE OF LICENSED { l NUCLEAR POWER PLANTS THROUGH 1974 'Y ,N P. E. Bobe 4 January, 1976 Office of Management Information and Program Control Performance Evaluation Branch UNITED STATES NUCLEAR REGULATORY COPetISSION

. _ = ABSTRACT i General aspects of fuel element design and specific design data for typical Pressurized and Bolling Water Reactors are presented. Based on a literature [ search, failure modes and specific failures incurred through December 31,1974 are described, together with a discussion of those problems which have had a significant impact upon plant operation. The relationship between fuel element failures and the resultant coolant activity / radioactive gaseous effluents upon radiation exposure, plant availability and capacity factors, economic impact, and waste management, are discussed. An assessment was made regarding the generation, availability, and use of fuel perfonnance data. i 4 1 1 i i s b 1 l 1 1.

Table of Contents Page Abstract................................................................. i

1.0 INTRODUCTION

I 1.1 Purpose...................................................... I 1.2 Scope.......................................................... 1 2.0 FUEL ELEMENT DESIGN................................................. 2 2.1 Introduction................................................... 2 2.2 General Element Design......................................... 2 2.3 Speci fic El ement De s1 gn............................ 4 3.0 FUEL ELEMENT PERFORMANCE.......................................... 7 3.1 Introduction........................ 7 l 3.2 Categori zation of Fuel El ement Failures........................ 7 3.3 Plant Experience................................... 13 3.4 Detection of Fuel Failures............... 19 3.5 Consequences of Fuel Failures.................................. 22 4.0 REPORTING REQUIREMENTS.................................... 23 4.1 Introduction................................................... 23 4.2 Review of Present Requirements.............................. 23

5.0 CONCLUSION

S....................................................... 27 BIBLIOGRAPHY........................ 29 Appendix A - Glossary.................................................. A-1 Appendix B - Design Data................................... B-1 Appendix C - Failure Experience......................................... C-1 Appendix D - Fuel Performance Reporti ng Requirements..................... D-1 ii

FUEL PERFORMANCE OF LICENSED NUCLEAR. POWER PLANTS THROUGH 1974 1.0 Introduction 1.1 Purpose A study was made of fuel element performance in licensed nuclear power plants in commercial operation through December 31, 1974 This report is a summary and qualitative evaluation of such performance, based upon that study. Identification of fuel element failures, their causes and effects, and present fuel performance reporting requirements are assessed. 1.2 Scope On December 31, 1974, there were 53 licensed plants 42 in comercial operation I and 10 in the power ascension phase. One plant, Indian Point 1, was shut down until 1977 for emergency core coolant system modifications. Forty nine plants I (21 BWRs and 28 PWRs) had generated electrical power. The data excluded Peach Bottom 1 (a high temperature gas cooled reactor) which was shutdown and in the process of being decomissioned. Several plants had not been in operation long enough for a refueling and core inspection. Fuel performance data was available j for 33 plants (14 BWRs.and 19 PWRs) and are included in this report. l Licensee reports, trade journals, transactions of professional society and special technical meetings, and core supplier topical reports, were the primary sources of data pertaining to fuel element performance. Most data for the last half of 1974 were obtained by persunal communications with licensee personnel and i= NRC regional office inspectors. Reporting requirements for fuel performance data are contained in license technical specifications, NRC regulations, and NRC Regulatory Guides. Two diametrically opposite approaches may oe taken in evaluating fuel performance. One approach is to stress successful performance while the other is to focus attention upon failures and other problem areas. The latter approach has been taken in this study since it not only provided an historical compendium of failure data, but also is the most likely to identify areas of potential improvement. l Despite the inherently negative tone of the presentation, it should be kept in mind that the success rate of fuel has been generally about 99% or greater and that there has been no impact on public health or safety from the failures which I have occurred. r-I l The categories of fuel element failures are defined in Section 3.2. Other terms which are frequently utilized in this report are defined in the Glossary of Appendix A. ! i

i Frequent reference is made in this report to the United States Atomic Energy - Conenission (AEC). In view of the historical nature of the report, it was con-sidered to be less confusing not to change the acronym AEC to NRC throughout the text. Licensee reports may be reviewed at the NRC's Public Cocument Room, located at 1717 H Street, N.W., Washington D.C.; documents pertaining to specific plants also are available at local public document rooms located in the vicinity of each plant. NRC rules and regulations are for sale by the Superintendent of Documents. U.S. Government Printing Office Washington, D.C., 20402. Regulatory Guides are availaale from the U.S. Nuclear Regulatory Commission, Washington D.C. 20555, attention: Director. Office of Standards Development. 2.0 Fuel Element Design 2.1 Introduction The evolution of the nuclear power industry has resulted in a variety of designs. Appendix B. Table B-1, is a compilation of miscellaneous plant data for the 49 plants which have genarated electricity as of December 31, 1974. This variety would be expected due to competition among several nuclear steam systems suppliers, the advancing state of technology, and tailoring of the design to specific utility requirements. In any case, certain safety and general design criteria must be satisfied. As the nuclear industry matures, the trend is to standardize designs as much as possible. Each of the four U.S. light water reactor nuclear steam system suppliers have submitted standarized designs for NRC review; even though one supplier withdrew his application in December, 1974, it is included in this report for completeness and comparison purposes. 2.2 General Element Design The design of fuel elements is based not only on safety considerations, but also on the following general considerations: a) a choice of fuel and cladding materials which are metallurgically compatible, and which will 1) efficiently transfer the heat generated to the coolant, 2) contain the radioactive fission products 3) attain the desired fuel burnup and corrosion lifetime.

4) be capable of thorough evaluation and testing prior to use, and

5) yield optimum physics characteristics, compatib!c to the rest of the design.

b) ease of refueling to replace spent or damaged fuel elements. c) reprocessing considerations. d) economics of the design, fabrication, and operation of the fuel elements. --

From the standpoint of safety, fuel element design must meet the requirements of 10 CFR 50. Most pertinent to fuel elements are Paragraph 50.46 (Acceptance Criteria for Emergency Core Cooling Systems [ECCS] for Light Water Nuclear Power Plants), Appendix A (General Design Criteria for Nuclear Power Plants) Appendix B (Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants), and Appendix K (ECCS Evaluation Models). 10 CFR 50 does not dictate specific parameters such as fuel enrichments, loadings, element dimensions, number of elements, core sizes, etc. The reactor designer has.the freedom to choose these parameters, provided the overall reactor design (including the engineered safety systems, such as ECCS) meets the perfomance limits specified in 10 CFR 50; therefore, certain restraints may indirectly be placed upon the choice of core parameters. The design of the fuel elements may also be influenced by the requirements of 10 CFR 20, 50, and 100, pertaining to radioactive effluent releases. Again, as described above, to satisfy perfomance limits specified in these requirements, certain restraints may indirectly be placed upon the choice of core parameters. A typical safety evaluation of the fuel element design includes: (a) materials adequacy throughout lifetime, including the effects of vibration, fuel element internal pressure and cladding stresses during normal and accident conditions with particular emphasis upon temperature transients or depressurization accidents. (b) potential for a waterlogging rupture. (c) potential for a chemical reaction, including hydriding effects. (d) fretting corrosion. (e) cycling and fatigue. l (f) dimensional stability of the fuel and critical components during l design lifetime. The evaluation should include discussions of failure and burnup experience, and l the thermal conditions for which the experience was obtained for the type of fuel to be used, and the results of long term irradiation testing of production fuel and test specimens. Inherent in the safety evaluation considerations described above is the potential failure of the fuel elements, such as a perforation or defect of the cladding which can permit release of fission products into the primary system. While there are other possible sources of radioactive material present, e.g., activation 1

i l products in various immobile positions and radioactive material in mobile foms, the fission products contained within the irradiated fuel far outweigh any other source. The reactor designer must necessarily assume that there will be some fuel failures, and factor this into the design of the shielding, containment, cleanup system and radioactive treatment system. Considerable conservatism is built into these systems to assure that the requirements of 10 CFR 20, 50, and 100 are met. Even if sufficient fuel failures occurred which could cause the radioactivity - discharge limits to be approached, the reactor operator has at his disposal operating procedures to reduce the leakage of fission products from the fuel, e.g., redaction of reactor power level, selective positioning of control rods to reduce the flux in the vicinity of the failed fuel (generally applicable to EWR plants), and limits on the rates of power change. However, use of these procedures generally result in operating inefficiencies and/or reduction in plant capacity factors. Excessive fuel failures may also necessitate earlier and/or more fuel replacements than originally planned, with the attendent relatively lengthy plant downtime and extra costs associated not only with the loss of plant operation and purchase of new fuel, but also refueling, inspection, storage, and reprocessing. The latter also increases the burden of waste disposal. In keeping with the requirements of "as low as practicable" radiation releases (described in Paragraph 20.1 of 10 CFR 20 and Paragraph 50.34a of 10 CFR 50), the reactor designer (or operator) must strike a balance between the costs required to reduce fuel element failure rates versus the costs of measures to cope with a given failure rate. In essence, it's a matter of treating the symptom vs. correcting the source of the problem. The fuel failures which have been statistically significant in the commercial LWR's, together with plants affected, are described in Section 3.0. 2.3 Specific Element Design The general design aspects of the fuel elements are similar for both BWRs and PWRs and have not changed significantly over the years. Essentially, for current generation cores, the fuel element (fuel rod) is a hollow tube (cladding) into which cylindrical fuel pellets are stacked end to end. The tubes are sealed by welding end plugs into each end of the tube. The fuel is slightly enriched which has been compacted into pellet fom and then (typically 1 to 4% U-235) UO2 sintered at high temperatures. The result is a pellet with a density somewhat less than the maximum possible (designated theoretical density). -

( Typical fuel elements are about 13 feet long containing about a 12-foot column of stacked pellets. The pellet columns are pressed downward by a spring in the top end (plenum space) of the tube; the springs hold the pellet column down during shipping and handling. After installation into the reactor. the springs are not needed and in fact, after relatively short periods of operation, the spring tension relaxes. The space within the fuel elements not occupied by fuel is filled with a gas, generally helium. Fuel elements are characterized as "unpressurized" or "prepressurized" depending on the backfill pressure of the gas at fabrication. The initial gas pressure is about atmospheric and up to several hundred pounds per square inch for unpressurized and pressurized elements, respectively. Additional design details are discussed in Sections 2.3.1 and 2.3.2 for BWRs and PWRs, respectively. 2.3.1 BWR Fuel Element Desian l As of December 31. 1974, there were 21 BWR licensed plants which had generated electrical power. One (Genca - also called Lacrosse or LACBWR) was designed by Allis-Chalmers while the remainder were designed by General Electric. Some typical BWR fuel element design parameters are shown in Appendix B Table B-2. l 'These data are applicable to the original cores installed in these plants, as described in their respective Final Safety Analysis Reports (FSARs); for the older plants the data are not necessarily indicative of the fuel elements presently installed. However, the data does show the evolution of fuel element design over a number of years. General Electric has utilized both stainless steel and Zircaloy cladding in their fuel rod designs; however, due to the advantages of imprgygd neutron econorny of Zircaloy and the inherent problem of stainless steel cladding in the BWR core-steam environment, GE has concentrated primarily upon Zircaloy cladding. Zircaloy cladding was used as early as 1960 in the Dresden 1 plant. The geometry of the Zircaloy rods has remained within a fairly narrow band. Some of the more significant changes have been conversion from segmented to full length fuel rods, increased void volume in the rods to accommodate fission gases and fuel expansion. shorter chamfered fuel pellets, and different clad heat treatment. The chan je to Zircaloy was not without penalty, however, both from the standpoint of fabrication costs and introduction of a failure mechanism called hydriding. Failure mechanisms are discussed further in Section 3.2. l It should be noted that the later GE designs utilized a 7x7 rod array. In order to reduce the linear heat generation per rod, the trend is to increase the total l linear footage of fuel within the core. Therefore, the latest GE designs utilize an 8x8 rod array; this type is also being used as reload fuel for most BWRs. Typical parameters for their "standarized" design are included in Appendix B Table B-4. - L l

_ ~- The fuel rods are of the unpressurized type. Compared to PWRs. the cladding of BWRs is subjected to lower external pressure and operate at a lower temperature. These features reduce the expected creep rate of the Zircaloy enough such that internal pressurization has not been deemed necessary. The Genoa fuel rod design in the replacement core did not differ significantly from the original core. J. 2.3.2 PWR Fuel Element Design As of December 31, 1974, there were 28 PWR licensed plants which had generated electrical power. Of these, 17, 7, and 4, were designed by Westinghouse, Babcock & f Wilcox, and Combustion Engineering, respectively. i Some typical PWR fuel element design parameters are shown in Appendix B. Table B-3. These data are applicable to the original cores installed in these plants, as described in their respective Final Safety Analysis Reports (FSARs). As discussed in Section P.3.1, the data is not necessarily indicative of the fuel elements presently installed in the older plants; however, the data does show the evolution of fuel element designs over a number of years. 1 In the earlier cores, Westinghouse utilized stainless steel fuel rod cladding, j dictated primarily by design considerations and uncertainties over possible fuel failures for these relatively short-lifetime cores. Thick stainless steel tubes are relatively inexpensive and provide high integrity. Later, fuel economy became more important to reduce overall fuel cycle cost. The use of Zircaloy for tubes, with its reduced neutron absorbing properties in high temperature water was initi-ated for plants placed into operation in 1968 and 1969. The use of Zircaloy, however, imposed the same type of penalties as for BWR cores discussed in Section 2.3.1. Since the introduction of Zircaloy, the fuel element design parameters have not differed greatly over the past several years. In order to reduce the linear heat generation rate, the typical 14x14 and 15x15 rod arrays are being charged to a 17x17 array in the latest standardized design. Typical parameters for the 17x17 array are shown in Appendix B. Table B-4. The B&W design (Table B-3) has also changed from a stainless steel cladding to Zr-4. The active length of the core has been increased and the number of fuel rods per assembly increased to reduce the thermal duty (linear heat generation rate). In their proposed standardized design (Table B-4), which was later withdrawn, a 17x17 fuel rod array is utilized; the fuel rod parameters are similar to the Westinghouse standardized design. The original core of Indian Point I cannot be directly com-pared to later B&W designs since the former was a thorium-oxide closed channel, rodded core. The replacement core (supplied t'y Westinghouse) contained no thorium; its parameters are also shown in Table B-3 for comparison purposes. 4 - . -. - - ~ - -

The Combustion Engineering design (Table B-3) has not changed significantly over the past few years. In the standardized design (Table B-4), the number of fuel rods per assembly has been changed to a 16x16 array which will reduce the thermal duty of the fuel rods. Again, it should be noted that the fuel rod parameters for this design are similar to the Westinghouse and B&W standardized designs. Both unpressurized and pressurized fuel rods have been used in PWR designs. Several years ago. PWR fuel manufacturers began to pressurize fuel rods and all current PWR fuel production is of the pressurized type. This change was initiated to decrease the " creep" of irradiated Zircaloy at reactor operating conditions. The pressuriza-tion partially offsets the large external pressure during operation and thereby reduces the inward creep of the cladding. Another advantage of pressurizing the fuel rods is improved gap conductance. 3.0 Fuel Element Performance 3.1 Introduction As described previously in Section 1.2. this report will concentrate primarily upon fuel element failures rather than upon successful performance. The success rate is ~ quite high (about 99% or greater) and the failures which have occurred have had no adverse affect on public health or safety. As described in Section 1.2. information was obtained from a number of sources. While considerable effort was expended in making the listing as complete as possible. it is expected that some data have not been included..This is due primarily to [ existing weaknesses in the reporting and compiling of data; this is discussed in more depth in Section 4.0. l The definition of a failure is based on that utilized by the Nuclear Plant Reliability Data (HPRD) System.(2) However, as described in Section 4.2.5. the definition has been broadened for use in this report. Inclusion of structural changes which result in plant operating restrictions or are beyond the predicted limits of performance, broadens the data base. However, it can also result in j several failures not being included since many such failures have probably not been i reported. Research to obtain all such data would have been a formidable task and was beyond the resources available for this study. Appendix C. Table C-1. presents fuel failure data for commerical light water reactors, together with the approximate failure frequency and type, through December 1974. 3.2 Categorization of Fuel Element Failures For purposes of this report, the causes of failure have been categorized as follows: Internal contamination Manufacturing defects .. _. - ~ _ _

Mechanical damage Fuel cladding interactions Accelerated corrosion Fuel Rod Bowing Cladding collapse Other Categorizing fuel element failures is somewhat arbritrary since there is no commonly ae;epted criteria available. Additional problems arise when there is difficulty in separating cause and effect, e.g., whether hydriding should be 1 considered a cause rather than a manufacturing defect, a design deficiency or the result of another failure mechanism; or as an another example, whether fretting and wear is the cause rather than a design deficiency. Finally, there have been several failures in which the cause is either unknown or the type not reported. This grouping is based on failures which are or were considered generic in nature, unless the failure could be specifically traced to be the result of another failure mechanism. For example, if a hydriding failure was specifically caused by a manufacturing defect, the failure would be listed under the latter; otherwise, it would be listed under internal contamination. Cladding collapse has been listed as a separate catagory, even though most, if not all, could be considered design deficiencies; this distinction was made since most cladding collapses have been due to a phenomena called "densification",(3) a subject which has received signifi-cant attention over the past few years, and thereby merits a category of its own. The categories of failure are described in greater detail below. Appendix C. Table C-2, presents a summary of the failure categories together with their associated applicable items from Table C-1. 3.2.1 Internal Contamination Internal contamination is the introduction of a foreign material into the fuel rod, which may attack the cladding. For claddings made of some form of zirconium, the predominant mechanism is hydriding, resulting from localized attack of the cladding by extraneous hydrogen. This hydrogen may be introduced in fabrication either through impurities or from moisture in the fuel itself. Moisture could also be introduced into the fuel rod during reactor operation via a fabrication defect. The hydrogen impurity attacks the inside surface of the cladding at various points, forming blisters (generally 1/8 to 1/4 inch in diameter) which are observable on the outer surface of the clad. Such blisters may result in rod perforation; such failures generally occur relatively early in the life of the rod. Measures taken to miminize such failures include new low-moisture specifications for fuel, procedural controls during manufacture to prevent introduction of

e s - hydrogenous impurities, drying of pellits prior to loading, hot vacuum drying of. the loaded fuel rod just prior t3 final end plug welding, and use of hydrogen ' getters in the fuel rods. As seen in Table C-2, hydriding has accounted for a significant percentage of all l the failures reported, especially for BWRs; a certain number attributed to hydriding, however, may have been initially caused by fuel-cladding interactions. For PWRs the most significant number of hydriding failures eccurred early in the operation of the Deznau 1 (Switzerland) and R.E. Ginna 1. ' Prior to the'se discoveries l in late 1969 and early 1970. Westinghouse had instituted 'more stringent specifica-tions on moisture content and quality control proceiures, based on results of tests from the experimental Saxton reactor. Fuel manufactured after that point in time has shown a very low incidence of hydride failures. l l For BWRs, hydriding failures were first encountered in Zircaloy clad UO2 pellet fuel in the early operation of the German KRB reactor. A significant number also occurred in a special reload batch of compacted UO2 powder in the Big Rock Point reactor. Therefore, General Electric initiated an extensive development program to study this failure mechanism. As a result of the program, changes in manufacturing and quality I control procedures were made. However, many cores were manufactured before these changes were made, and hydriding failures continued to be statistically significant. There is evidence that the changes have been effective'for the later cores, though the amount of data available is limited at this time. 3.2.2 Hanufacturing Defects The number of failures which can be specifically traced to manufacturing defects has been quite small, as seen in-Table C-2. Such defects can include defective material, welding flaws (e.g., defective welding of the end plugs), inadequate l inspection, enrichaent mixups and mechanical handling damage. Failures due to i such defects can also be the source of secondary failures, predominantly hydriding, if moisture penetrates into the rod. Failures due to manufacturing defects cannot be totally eliminated, but should remain very low in number provided stringent fuel manufacturer and utility quality assurance programs are maintained and continually upgraded as ertessary. l l 3.2.3 Mechanical Damage l For purposes of this report, this category includes such items as damage cue to mishandling of fuel at the reactor plant, failure of handling equipment, and fretting and wear of fuel rods. i

Fretting and wear have resulted from such causes as foreign material loose in the reactor coolant and vibration of fuel rods against grid springs. Lessons learned from experience and the use of flow loops for pretesting fuel designs have essentially eliminated failures of this type, l 3.2.4 Fuel Cladding Interactions ' l Fuel cladding interactions, more comonly referred to as fuel pellet-clad interactions (PCI), have been identified as a failure mechanism as early as 1971. Such failures have been predominately associated with BWR cores. The PCI mechanism involves localized mechanical loading of the cladding adjacent to cracks in the pellets and at pellet interfaces. The strain imposed by the fuel on the clad is primarily a function of the differential thermal expansion and fuel swelling and is dependent upon burnup, local power level, rate of power change, design character-istics, and manufacturing variables. The greatest potential for failure is later in life when the cladding has reduced ductility. After the PCI mechanism was identified in late 1971. General Electric (GE) provided BWR owners operational recomendations for reactor power level changes. In 1973, these were modified. The recommendations are implemented through a slow ascent to full power which is intended to precondition the fuel for subsequent normal full power operation. Unfortunately, such slow ascent to power results in loss of power generation and plant capacity. As a result of the PCI failure mechanism, GE made changes to their basic fuel rod design for fuel fabricated in 1972 and later (modified 7x7 assembly design). GE's latest design, consisting of an 8x8 array, should help reduce strain, due to the lower thermal duty of the rods (reduced linear heat generation). At the time of this report, there was insufficient data available to determine whether the various design changes have resulted in a significant reduction in PCI type failures. PWR fuel is not immune to pellet-clad interactions. As one example, defects, tentatively attributed to this failure mechanism, resulted during the Cycle 3 startup of Point Beach 1, in conjunction with a rapid rate of reactor power increase after the refueling shutdown. Cycle 3 startup of H.B. Robinson 2, conducted with fuel of similar design and burnup, was accomplished at more gradual rates of power increase andwithoutdefectindications.I4) 3.2.5 Accelerated Corrosion Some early commercial BWRs (Big Rock Point Dresden 1. Humboldt Bay 3) had fuel failures due to crud (scale) buildup on the fuel rods. Crud buildup can cause overheating of the cladding at the peak power locations, accompanied by accelerated oxidation and hydriding, which leads to eventual rod failures. The corrosion product buildup is highly dependent on the extraneous material in the primary ,s N* l g d)" ' s coolant. For the plants above, the failures were caused by excessive amounts of feedwater system corrosion products, high in copper content, which ' deposited on the fuel rod surfaces. These plants had cupro-nickel or monel feedwater heaters'. 5. Careful design selection of materials'in the feedwater system and improved primary D coolant water chemistry control appear to have eliminated failures due to crud buildup. A 3.2.6 Fuel Rod Bowing Occasionally, examination of irradiated fuel assembles reveals the presence of bowed fuel" rods. Rod bowing can generally be considered a design deficiency. ~ Bowin'g has rarely resulted in fuel rod perforation or breakage. One such case

• was antearly Dresden 1 core (Item 3a of Table C-1).

In this case, bowing led to accelerated corrosion which eventually caused some rods to fall. L 3 In general, the effects of rod bowing are factored into the design methods of the core. Some cores (such as the initial Lacrosse core) have continued to operate satisfactorily, while containing known bowed rods. 3.2.7 - Cladding Collapse [ Cladding collapse can also be generally considered as a design deficiency. Collapse-does not necessarily result in an actual perforation of the fuel rod; however, the potential exists and must be consispred in the safety analysis of the core. L l The most significant number of collapsed tubes,has resulted from a phenomenon i designated as " fuel densification". q m During the first refueling of.Beznau 1. cladding collapses were observed in a l-significant number of rods. ' The flattened sections ranged from 0.5 to 2.5 inches in length. Later, during refueling of R. E. Ginna 1 in April and May 1972, similar flattened rod sections were discovered. varying in length from 0.5 to 4.0 inches. It was found that the collapses resulted from the occurrence of gaps in the fuel B pellets column inside the rods. The Staff of the former AEC initiated an intensive review of the probable causes of gap occurrence and their effects on reactor operation. As described in the technical report issued by the Staff (3). the axial gaps arise from irradiation-induced densifica-tion of the uranium oxide fuel. The fuel pellets shrink both axially and radially. The axial shrinkage, combined with the hangup of a pellet (due primarly to creep down of the cladding) and intermittent pellet-stack settlement, result in axial gaps. The fuelrodcladdingcre'epiinwardwhensubjectedtothehydrostaticcompressivestresses of the water coolant. Ultimately.ithe cladding will contact the fuel pellets or continue to creep inward in the regions of axial gaps. The time required to reach ( l > ?, s ~ - -

cladding-pellet contact or cladding collapse depends on many fac+ ors, including the diameter, wall thickness, initial ovality, and temperature of the cladding, fabrica-tion history, coolant pressure, initial fuel rod internal gas pressure, additional internal gas pressure from fission gases and temperature changes, and neutron flux - level. If collapse does occur, the probability of leaking of individual fuel rods is increased, and local power peaking occurs. Even if collapse does not occur, fuel densification leads to increased stored energy in the fuel rod, increased linear heat generation rate of the pellet, decreased heat transfer capability of the fuel rod, and increased potential for j local power spikes. As seen in Table C-2, several reactor plants have experienced rod collapses. It should be noted that collapse was confined to PWR plants. As described in the Regulatory Staff report, data on irradiated BWR fuel rods shows that densification also occurs, but to a much lesser degree - probably due to higher initial pellet densities. Some instances of gaps between pellets had been found, but were too small to be of significance with regard to cladding behavior or power spiking. The expected creep rates of BWR rods are lower due to lower external system pressure on the rod and lower cladding temperature. In addition, the BWR free-standing cladding with a relatively large radial clearance gap, reduces the chance that pellet movement would be prevented by cladding creepdown, friction or other mechanisms. In this case, densification would most likely result in an overall shrinkage of the fuel pellet column as a unit. Nevertheless, due to the relatively little data available at the time, the Staff concluded that until sufficient data became available to permit an unambiguous understanding of the densification mechanism (and of possible methods to control or eliminate the mechanism), the Staff would assume that irradiation induced densification occurs in all light water reactors. Later experimental data showed that the Staff position was prudent; for example, gamma scanning of Oyster Creek 1 (BWR) fuel rods in 1973 showed significant axial gaps. While utilities and reactor designers developed sufficient models to account for densification, and made changes to the fuel and fuel rod design, several reactors were subject to operating restriction, reducing their performance. This is discussed further in Section 3.3.2.3. 3.2.8 Other Types of Failure For purposes of this report, "Other" types of failure are those which are not conveniently classified under one of the other catagories, such as miscellaneous design deficiencies; also included are those in which the mode of failure was either not known, or unreported. One of the more important problems encountered under the former category was fuel assembly channel box wear, described in Section 3.3.2.4. 3.3 Plant Experience Fuel failure sununary categorization is presented in Table C-2. The categories are those described in Section 3.2 above.. The item numbers correspond to those in Table C-1. A sizeable amount of the data through 1973 presented in these tables was taken from a special report generated by Battelle Pacific Northwest Laboratories for the AEC Directorate of Licensing, under Contract AT(45-1)-1830, issued in June 1974(5) Additional data was obtained from startup reports, operating reports, special reports / letters from the utilities to the AEC, and various com-mercial publications. References for the data are given in the last column of Table C-1. 3.3.1 Discussion of Plant Experience The data presented in Tables C-1 and C-2 are for fuel elements which are not necessarily representative of current production. The design of the elements and the manufacturing / quality assurance technologies have been continually evolving. Data for relatively current production fuel is sparse and probably will not be available until some time in 1975. In addition, it should be noted that a signifi-cant number of failures are listed in Table C-2 as unknown or unreported; however, for BWR plants, several of these are most probably hydriding and/or pellet-clad interaction failures. Therefore, a comparison of the number of fuel problems of each category would be of limited value. Nevertheless, some general observations can be made. BWR plants will probably continue to experience hydriding and PCI failures, at least until the older fuel has been replaced by the later fuel designs. Significant power generation has been lost due to fuel related problems. Manufacture and use of many cores of a similar design, before the design has been fully tested, have resulted in severely compounding the effects, when a problem is discovered. The cause of fuel problems are far too often unreported, or reported as unknown by the licensees. In addition, there is a wide variation in the reporting of fuel performance. This item is discussed further in Section 4.0. Since many or most of the irradiated fuel examinations are conducted by the core designer / manufacturer, periodic reports, issued not less than yearly and summarizing their fuel experience, would be invaluable. At the end of 1974, the last topical report by General Electric covered experience through September, 1971(6) This item is discussed further in Section 4.2.6. 1 4

7 1 Due to the rapidly increasing number of plants and the different types of fuel in each, within each cycle and cycle to cycle, it becomes increasingly I important to utilize a centralized file for data evaluation. This is j discussed further in Section 4.0. J ) It should be noted that the data presented in Table C-1 was gleaned from a large number of sources, including licensee reports to the AEC, topical reports by some core designer / manufacturers, professional society transactions, and various trade' journals. None of these sources may be considered all inclusiv'e. It is also likely that some data was not included in this review, due to the large number of potential sources for such data. The need for better reporting of data is discussed further in Section 4.0. 3.3.2 problems Having Significant impact Upon Plant Operation Based on data presently available to the Government, it is impossible to estimate with any degree of accuracy the amount of power generation lost due to fuel problems. The licensees submit monthly the average daily operating power levels and a sumary of plant outages (shutdowns); however, no data is included on partial outages (power reductions - such as those associated with fuel problems, startups, shutdowns, training, equipment related problems, etc.). Therefore, no accurate analysis can be made. However, it is of interest to note that if at any given time there is an average loss of only 2% power generation for the 49 plants which have generated electricity (see Table B-1), this corresponds to a loss of 648 MWe-net (based on a total potential capacity of 32,403 MWe-net). This is equivalent to having a plant of a size such as H.B. Robinson 2 or Pilgrim I shutdown nearly 100% of the time. While nearly all the problems listed in Tables C-1 and C-2 did restrict or had the potential for restricting plant operation, the three most significant are hydriding, pellet-clad interactions, and densification. During the time this report was written, the first two continue to be either a significant source of failures, leading to power restriction to limit release of radioactive fission products and in some cases, refuelings earlier than anticipated, or restriction on plant operation to preclude their occurrence. 3.3.2.1 Hydriding Some examples of power restrictions or early refuelings due to hydriding failures (including those cases in which the most probable cause of failure is hydriding) are shown in Table 3-1. This by no means should be considered a complete list. Included are only those cases in which hydriding has,been reported as a predominant failure mechanism. The lack of complete data precludes a determination of the loss of plant availability and capacity due to these power restrictions and/or early refuelings. In addition, for Quad Cities 1 and Dresden 3, pellet-clad interactions also contributed a significant amount of failures. - -~

Table 3-1 EXAMPLES OF PLANT RESTRICTIONS OR EARLY REFUELINGS CAUSED BY HYDRIDING Reactor Plant Type Conments Dresden 2. BWR During Cycle II. (5/29/71 to 2/19/72), the unit was base loaded at approxi-mately 60% power, to minimize off-gas activity release rate. Dresden 3 BWR As a result of fuel failures and off-gas limitations, the first refueling (originally scheduled for Fall of 1973) occurred in the Spring of 1973. Monticello BWR During late 1973 and until refueling shutdown in the first quarter of 1974, power was restricted at times to reduce off-gas release ra'.e. Nine Mile Point 1 BWR During first part of 1971, above normal off gas activity caused the reactor power to be limited until the fuel was replaced. Quad-Cities 1 BWR The plant was administratively limited in power level at times starting in last half of 1973 to maintain stack rates at acceptable levels. Vermont Yankee BWR During 1972, excessive gaseous release activity levels resulted in power reductions until the scheduled mid-January 1973 shutdown. The plant was administrative 1y limited to lower power during 1974 until refueling shutdown in October due partly to excessive off-gas activity at the g steam jet air ejector. i

The examples in Table 3-1 are for cases in which failures have already occurred in sufficient numbers to cause plant restrictions on power level. Further restrictions may be necessary at times to limit the rate of power changes, to limit the stress on the fuel elements. 3.3.2.2 Pellet-Clad Interactions As described in Section 3.2.4, mitigation of the effects of pellet-clad interac+1ons - has been approached from the standpoint of design changes to the fuel assembly, and the use of " fuel preconditioning". The latter is in the form of procedures for a periodic, slow ascent to full power, which preconditions the fuel for subsequent - nomal full power generation. These procedures, for General Electric designed BWRs, are designated Preconditioning Interim Operating Management Recommendations (PCIOMR's). From late 1971 to 1973, GE provided the utilities with Interim Operating Management Recomendations (IOMR's); however, as one utility pointed out in a semiannual opera-tions report (7), "these reconsnendations were of limited value in reducing failure rates. A signifidant amount of generation was lost in following the recommenda-tions." Results using PCIOMR's appear to be better, but are not yet definitive. Unfortu-nately, these procedures result in even greater losses of power generation (capacity factor) as compared to ' the 10MR's. Clearly, use of such special procedures is not as satisfactory as better fuel rod design. At the time of preparing this report (early 1975), it is not known whether the design changes incorporated to mitigate PCI failures will be effective. Some data should be available by late 1975. 3.3.2.3 Densi fication As described in Section 3.2.7 fuel densification can lead to fuel rod collapse, with a potential for cladding perforation and power spikes. The concern over densification resulted in not only the expenditure of considerable manpower on the part of the Government, utilities, and core designers, but also operating restrictions (either Government or utility imposed) for varying lengths of time. These restrictions are discussed separately for PWRs and BWRs below. 3.3.2.3.1 PWR Plants Several PWRs operated at derated power levels for a part of 1973 until detailed analyses were submitted and reviewed by the AEC. In some cases the restriction on power level decreased with fuel burnup. During operation, the pressure dif-ferential across the clad results in a tendency for the clad to creep toward the fuel pellet stack, narrowing the fuel to clad gap, and increasing the fuel to clad heat transfer coefficient. This can result in higher power densities with the same peak centerline temperature, thereby pemitting higher power levels with increasing core - burnup.

- ~ ~- Y [.. 4-Reduction of the pressure differential across the clad also decreases the creep [ rate of the clad toward the fuel pellet stack. Some utilities requested and received permission in 1973 to lower the primary system pressure to delay predicted 4 time of clad flattening..Such reductions-lower the plant efficiencies. I.. The PWR plants and the restrictions imposed by the densification concern initiated in 1973 are shown in Table 3-2. 3.3.2.3.3 BWR Plants As described in Section 3.2.7, densification effects were expected to be most 3 significant in PWR plants. However, in late 1972, relatively little data was available; therefore, the AEC assumed that all light water reactors could be affected to some degree. In reply to a November 1972 AEC request. the licensees provided analyses and data in early 1973 relating to determining the consequences of fuel densification. Included by reference was a General Electric report NEDM-10735(8); later, supplements to this report were submitted in response to requests for additional information from the AEC. Meanwhile, during the Spring 1973 refueling of Oyster Creek 1,' evidence of densifica-tion (axial gaps, etc.) was noted in post irradiation examinations. l Based upon the AEC staff's review, it was determined that changes in the operating conditions were necessary to assure that calculated' peak cladding temperature of the cores following postulated loss of coolant accidents would not exceed 2300*F taking into account fuel densification effects. On August 24, 1973 the Director of Regulation issued orders to the licensecs of 10 BWR plants modifying the licenses by amending the technical specifications. The 10 plants were Oyster Creek 1. Dresden i 2 and 3. Quad Cities 1 and 2, Nine Mile Point 1, Millstone 1 Monticello, Pilgrim 1. and Vermont Yankee. These new limitations were met at several reactors without a reduction from rated power through modification of control rod pattern, at some - decrease in cycle length. Several plants were limited in power by other considera-tions. At Oyster Creek 1, conformance to these limits proved operationally restric-tive and resulted in an approximate 51 loss in generation until the limits were relaxed on December 28, 1973. I9) In November 1973. General Electric submitted report NED0-20181 , which as a consequence, the AEC approved a modified GE model for fuel densifications. Limits on the maximum allowable planar LHGR (linear heat generation rate) and the peak LHGR were relaxed (increased) on December 28, 1973, for the affected BWRs. This allowed increased operational flexibility, including possible full power operation, for any BWR which had previously been limited to less than 100% power. i ,i, 3.3.2.4-Channel Wear This problem affected only two licensed reactors; therefore, cannot be considered as significant as the problems described in Sections 3.3.2.1 through 3.3.2.3.

However, e

it is included for completeness and general interest. i w 3-.. --,_,..-w-, w-

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j i 1 TABLE 3-2 ) PWR PLANT RESTRICTIONS DURING 1973 DUE TO DENSIFICATION 1 Plant Restrictions H.B. Robinson 2 Power level limited to 94.8% until' July. Point Beach 1* Power level limited to 75% from March to May. Point Beach 2 Approval of 2000 psia operation in December. R.E. Ginna 1 Power level limited to 83.3% until July. Approval of 2000 psia operation in October. San Onofre 1 Power level limited to 90% for part of July and August. Surry 1 and 2 Power level limited to 92% at beginning of life. By year end, they were allowed to operate to 97.6% and 96.2%. respectively, with both units reaching 100% in the first half of 1974. Both units received approval in July, 1973 for 2000 psia operation. Turkey Point 3 and 4 Power level limited to 931 at beginning of life. By year end. Unit 3 was allowed to operate up to 98%. reaching 100% in March 1974. Unit 4 reached 100% in March 1974. Unit 3 and Unit 4 operating pressures were reduced to 1885 psia in December 1973 and January 1974. respectively.

• Reactor coolant system reduced from 2235 psia to 2000 psia in May 1974 during beginning of Cycle 3.

In August 1973, extensive wear on the corners of some fuel assembly channel boxes was observed during an inspection of fuel at the BWR KKM reactor (Swiss). The most significant channel wear, including through-wall penetrations, corresponded to the location of temporary control curtain stiffeners. The severe wear was caused by rubbing of the curtains against the channel boxes due to the impingement on the curtains of high velocity jets of water flowing through the bypass flow holes in the lower core plate. A similar design arrangement was also utilized at Pilgrim 1 and Vermont Yankee. On September 28, 1973, Vermont Yankee was shutdown for reasons unrelated to fuel channel box damage. Fuel channels were inspected for excessive wear. Channels showing wear greater than 10 mils or having potential for wear (172 channels) were replaced. Testing was perfonned to develop a plug design for the bypass flow holes to eliminate the cause. Plugging of the bypass holes was approved by the AEC and accomplished with a. return to operaton authorized on November 16, 1973. 4 As a result of the observed degradation of fuel channel boxes at Vennont Yankee and [ KKM, General Electric (the reactor designer) reconnended reactor power and core flow reductions at Pilgrim 1. On October 6, 1973, Pilgrim 1 voluntarily reduced reactor power and core flow to 50% of the rated design values. By letter to Boston Edison Company dated October 16, 1973, the AEC approved operation of Pilgrim 1 for about 60 days subject to limitations of 50% in core flow and core power. On December 28, 1973, the reactor was shutdown in compliance with an AEC order to repair any channel box damage. Inspection showed damage ranging from slight to through-wall wear. All damaged channels were replaced. 3.4 Detection of Fuel Failures Detection of fuel failures may be classifed into two categories: while the fuel is not producing power (in-core and out-of-core tests) and while the fuel is_ producing power (more properly referred to as fuel perfonnance monitoring). These are discussed separately below. 3.4.1 In-Core and Out-of-Core Tests The in-core tests are generally of a more qualitative nature, due to the core's remoteness and the fact it may be highly radioactive. The tests available are limited and vary with the design of the plant and the type of test equipment available. Some plants have flux monitors installed which can detect flux peaking, possibly due to fuel gaps or clad collapse; however, the monitor cannot generally 1 indicate the specific rod with a problem, but only the general location in a group 1 of rods. Visual exaninations (such as binocular, TV, and boroscope) can only discern the periphery of the core, where the fuel with the least burnup is likely to be located; these tests can examine for pronounced rod bowing, damage to the assemblies and general appearance of the fuel clad. For BWRs, a sample of water, taken from any assembly, can be analyzed for fission products (in-core sipping), to determine the presence of clad perforation. This method has not been totally effective in detect-ing all assemblies containing clad failures; however, depending on the techniques used, as many as 90 - 95% of the leakers may be detected. '

Out-of-core tests are generally more definitive. The assembly (or individual rods, in the case of a BWR) may be removed to a facility with a hot cell for various nondestructive and destructive tests. However, before this is done, the assembly is generally wet-sipped (in or out-of-core) or dry-sipped. In the former case, a sample of water from an assembly can be analyzed for fission products either while it is still in the pressure vessel for BWRs (in-core sipping) or while it is in the spent fuel pool for either BWRs or PWRs (out of-core sipping). In the latter case (utilized at various BWR plants), a fuel bundle sampler located at the spent fuel pool introduces an air bubble into the bundle which voids the water from the bundle; the bundle heats up due to gamma decay heat and the air bubble is released back to the sampling station. A grab sample is taken directly for gamma spectroscopy. Dry sipping is quite useful when the fission product release is relatively small, e.g., for assemblies with short irradation exposure and for assembles which have been out of the core for long periods of time such that fission products have decayed to relatively low levels. Other out-of-core tests include ultrasonic and eddy current. These two tests, for GE BWR plants, can generally be perfomed at the plant site, since the fuel assemblies can be readily disassembled. For PWR and BWR (certain non-GE fuel) assemblies, which cannot be readily disassembled, the assemblies generally must be shipped off site. Both in-core and out-of-core tests are valuable as part of the overall fuel surveillance programs to determine how the fuel performed. Both, however, are of limited value in regard to knowledge of fuel perfomance during operation. 3.4.2 Fuel Perfomance Monitoring The tests described in Section 3.4.1 are primarily utilized at the end of reactor operating cycles. It would obviously be more desirable to predict fuel failures or detect specific failures as they occur. At present, the plants have no monitoring device which has this capability; neither is such a device commercially available which could be practically utilized. An indication that some number of rods have failed can be obtained from monitoring the primary system for fission products; for the PWR's, the reactor coolant is monitored, while for the BWRs the off-gas at the steam jet air ejector is monitored. Attempts to correlate the fission product level to a specific number of failed fuel rods have met only limited success, due primarily to the number of variables involved. The amount of fission product activity at the monitoring locations depends not only upon the number of failures and the specific design of the plant, but also upon such parameters as the size of the defects, the amount of burnup, the power level, amount of " tramp" uranium in the system, fuel temperature, and the efficiency of the cleanup system. 4 A discussion relating reactor coolant radiotodine concentration with fuel defect levels is contained in Westinghouse report WCAP-8253.00) Use of iodine. isotopes (I-131andI-133)andxenon(Xe-133andXe-135)areseparatelyexplored;however, it ns concluded that use of iodine presented loss uncertainties. Iodine activity can originate both within the fuel rod or from " tramp" uranium (generally con-sisting of uranium which adheres to the outer clad during the manufacturing pro-cess). The half lives of I-131 and I-133 are approximately 8 days and i day, respectively. For small cladding defects..the diffusion time to the surface may be on the order of days; therefore I-131 will be the dominant isotope. For any exposed uranium in the system, there is no diffusion barrier and fission products from this source are deposited directly into the reactor coolant by the recoil mechanism. A correction can then be made for this recoil mechanism by observing the ratio of I-131 to I-133. - Differential equations, which describe the fission product activity as a function of fuel defects and various plant parameters, are solved to yield the predicted I-131 concentration at the one percent defect level i using an assumed value of the escape rate coefficient for diffusion into the coolant. The measured I-131 concentration is compared to the predicted value to determine the apparent defects by a simple ratio. By applying a correction for the recoil component, a " corrected" defect level is obtained. J The Westinghouse report did not include any comparison of calculated and measured defect levels, but indications are that the variance could be wide due to the assumptions made. As the report points out, other errors are introduced by such items as a) power changes which cause fluctuations in primary coolant activity, b) fission product saturation being time dependent, and c) extrapolations to 100% power assumed to be linear while release of volatile fission products are likely to have a non-linear dependence on fuel temperature. Nevertheless, the method does provide an indication, however crude, of fuel performance. One licensee l (Surry 1 and 2) in their Semiannual Operating Report (87) for the period ending June 30, 1974, provided a comprehensive presentation of coolant activity together with the calculated defect level. Such information would also be valuable from other PWR licensees to enable the licensee and the Government to be aware of trends in fuel performance. It should be noted that the analysis presented by Westinghouse is not subject to a large source of potential error introduced by control rod perturbations. For normal operation, these reactors depend primarily on soluble poison control and the rods ere withdrawn. For the GE BWRs, the rods are manipulated considerably more. For leaking or suspected leaking assemblies, the adjacent control rod may be inserted, reducing the power level in those assemblies, and thereby, the fission 1 product release. A further complication is the fact that current BWR Technical Specifications and Regulatory Guide 1.21 require reporting of gaseous effluents released from the stack. There are variations in hold-up times, use of advanced off-gas treat-ment systems, and use of a single stack for multiple units. Data would be more useful I if the gaseous effluents were measured at the steam jet air ejector; however, even ) this data would be of limited value in the absence of an accurate technique to 3 correlate measured activity with defect level. ~ 3 --.-_- ~ - -, - - - -- ~ - - - - - --

f in summary, it may be stated that present monitoring techniques and calculational models are of limited use in the evaluation and prediction of fuel integrity during reactor operation. In addition, calculational models presently used, and any such models developed in the future, would be of more value if compared to measurements to determine the accuracy of such models. 3.5 Consequences of Fuel Failures Fuel failures have a direct influence upon the radiation exposure of plant personnel and the public, cost of plant operation (and consequently the cost of power genera-tion), plant availability factor, plant capacity factor, and waste management. It is 3 beyond the scope of this report to quantitatively evaluate these aspects. The necessary data for such an analysis was not generally available. However, some i general observations may be made. Fuel performance has, and in some cases continues to be, a significant factor in lower plant availability and capacity factor, particularly for boiling water reactor plants. Failures lead to shorter cycle length, increasing plant outages, with a consequent reduction in plant availability. Power reductions to reduce gaseous radioactive effluents, fuel preconditioning, and past power reductions due to the densification problem have resulted in reduced electrical generaticn (plant capacity factor). Replacement of fuel rods or assemblies before the end of their design life-time results in increased waste reprocessing and disposal. The problem is compounded due to the present lack of reprocessing facilities, necessitating storage of more irradiated fuel for longer periods of time. Replacement of fuel before the end of the design lifetime increases the cost of plant operation and consequently the cost to the public of electrical power. The increased costs are reflected in the cost of the fuel reloads themselves, plus refueling operations, and storage and/or reprocessing of irradiated fuel. Additional costs are associated with the inefficiences associated with lower plant availability and operating restrictions reducing the power generation due to fuel performance limitations. Even though plants are designed to operate with some failed fuel rods (use of shielding and various sophisticated treatment systems for keeping fission product i release within prescribed limits), reductions in radiation exposure levels could be achieved by reducing or eliminating fuel failures. This is consistent with the requirements of reducing radiation to "as low as practicable levels." Significant advances have been made in radioactive waste disposal systems; however, in some instances it seems to be a case of treating the symptoms rather than the cause of the problem. Meeting limits of fission product release should not, and cannot, rely solely upon the reliability of the disposal systems, cleanup systems and shielding i ns talled. Finally, even though the exposure to the general public (outside the boundaries of the plant) is generally considerably below permitted limits, the exposures to plant personnel appear to be increasing. (24 While a large fraction of this exposure is due to maintenance and repair of equipment contaminated with _ _ .,, - ~

corrosion products and crud (tiping, valves, PWR steam generators, BWR recirculation pumps and clean-up systems) rather than from fission products, any reduction in exposure from any source is desirable. It should also be noted that exposure to personnel conducting inservice inspections is increasing as a result of contamination of primary piping, valves, etc., due to fuel failures. As a result, decontamination of components subject to inservice inspection may be required. 4.0 Reporting Requirements 4.1 Introduction As a result of the research performed for this report, it soon became evident that there is a wide variety in the means, and quality, of reporting fuel perfomance data. Data was found in such sources as various licensee reports to the Government, (AOR's, Semiannual Reports, special reports requested by the Government), trade journals, transactions of professional society meetings, core manufacturer topical reports, and transactions of special technical meetings. In many instances, con-siderable time transpires before the data is published. Data are available at the licensees, core manufacturers and/or designers, which have not been published at all. Adequate data are important to pemit a realistic evaluation of the predicted behavior of fuel, changes to fuel element design, and actions taken by licensees and their contractors to correct problem areas as well as to keep the Government and the public infomed. 4.2 Review of Present Requirements 4.2.1 Technical Specifications Section 50.36. " Technical Specifications," of 10 CFR Part 50, " Licensing of Produc-tion and Utilization Fact 11 ties," requires that each applicant for a license authorizing operation of a nuclear power plant include in its application proposed technical specifications. The latter are reviewed by the NRC, modified and/or supplemented as necessary, incorporated into the facility license and are conditions of the license. These technical specifications include a section on reporting requirements. In addition to the reporting requirements necessary for compliance with technical specifications, specific reporting requirements are included in Part 50, as well as in other Parts of Title 10. Chapter I, Code of Federal Regulations. Additional guidance in regard to reporting requirements is given in Regulatory Guide 1.16, which is discussed separately in Section 4.2.2. While the Guide is not manda-tory, Itcensees generally incorporate its intent into their technical specifications. A perusal of the technical specifications for several plants (Appendix 0. Table D-1) shows that requirements pertaining to reporting of fuel perfomance are vague. Indeed, a review of Startup Reports, First Year Operation Reports, and Semiannual Operating Reports shows a wide variety in the type and depth of infomation provided by the licensees. To properly access the performance of fuel, an adequate data base is essential. This is discussed further in Sections 4.2.2 through 4.2.6. __

4.2.2 Regulatory Guides Regulatory Guides are issued to describe methods considered acceptable to the NRC Regulatory Staff of implementing specific parts of the NRC's regulations, to delineate techniques used by the Staff in evaluating specific problems or postulated accidents, or to provide guidance to applicants. The Guides are not substitutes for regulations and compliance with them is not required. Regulatory Guide 1.16 is applicable to the reporting of operating information. Revision 1, in effect during most of 1974, had only one s _ecific reference to fuel j perfomance. In the Semiannual Operating Reports, under Operations Summary (a sunrnary of operating experience occuring ciuring the reporting period that relates to the safe operation of the facility), a summary should be included in regard to " performance characteristics (e.g. equipment and fuel perfomance)". No guidance was presented in regard to the specifics to be included. Regulatory Guide 1.16 Revision 2 issued in September 1974, contained a considerably expanded list of reporting guidelines pertaining to fuel performance; however, on further review, several of these guidelines were considered either unnecessary or impractical and were eliminated in the next revision of this Guide. Regulatory Guide 1.16 Revision 3 was issued in January 1975. This revision states that the Annual Operating Report should include " indications of failed fuel resulting ] from irradiated fuel examinations, including eddy current tests, ultrasonic tests, or visual examinations completed during the report period". While this change is a considerable improvement compared to Revision 1 it should be noted that the l Operating Report was changed from semiannual to annual. Data could be up to a year old before it is submitted. To strengthen the reporting of fuel performance data, the following supplementary data, not included in Revision 3 of Regulatory Guide 1.16, would be of value: ( Submittal of a preliminary report of fuel examinations performed at the licensees's facility (including sipping, visual, eddy current, ultrasonic, etc.) within 30 days af ter completion of these examinations. After all examinations are completed (including destructive tests), a final report to be submitted within 60 days. These reports would include the infomation listed below. Description of the fuel assemblies inspected, including geometric and material parameters. Specific references to previous submittals containing this data would be pemissible. Sipping results. (including the sipping technique employed) and any other inspections performed, the number of assemblies tested, their designation and loca-tion in the core, and the specific ones which failed. All inspections completed to be included, whether performed by the licensee, its contractors, or fuel suppliers. 24 ~ _ __

Most probable cause of failure. together with the reasons for the conclusion. The Annual Operating Reports should include: a summary description of power generation lost by plant shutdown or operating restrictions (limiting radioactive gaseous effluents, fuel preconditioning, etc.) due to fuel problems. any unusual problems incurred, associated with fuel, whether or not there were any effects on plant operation. reference to any special fuel inspection reports issued in the interim. The results of the latter reports to be briefly summarized in the Annual Operating Reports. an estimate on a month by month basis of the number of failed fuel rods in the core, together with the basis for the estimate. The changes described above would represent a short term goal to improve reporting requirements. To be fully effective, a longer term goal could be established to set up a complete data bank on fuel installed into the cores. This could be in the form of a new regulatory guide devoted entirely to the reporting of fuel performance information. Such a guide could include not only the infomation described above, but also the fuel suppliers name core loading maps, goal burnups for each assembly, design and fabrication data for each variety of fuel rod. operating data, performance data. etc. Questions regarding proprietary data could be handled under existing rules and regulations. Much of the information could be submitted by appropriate data forms or punched cards ready for data processing. As described in Section 4.2.6. a computer program which could be used to handle such data has already been developed. 4.2.3 Abnomel Occurrence Reports The Technical Specifications and Regulatory Guide 1.16 Revision 3 require Abnomal Occurrence Reports (AOR's) for certain types of events. These AOR's were reviewed for any information pertaining to fuel performance; however, they are only a very limited source of information since they are not designed for the reporting of rou-tine data. For example. Revision 3 of Regulatory Guide 1.16 would require an AOR only for such fuel related problems as (1) an " abnormal degradation discovered in fuel cladding, reactor coolant pressure boundary, or primary containment" and (2) certain " reactivity anomalies". Revision 3 revised the Abnormal Occurrence Report section to delete radiological effluent releases from Appendix A technical specifi-cations reporting requirements. In view of the fact that A0R's were not designed for the reporting of routine data, no changes will be necessary in A0R reporting requirements pertaining to fuel performance, especially in view of the comments presented in Section 4.2.2. and 4.2.6. 4.2.4. NRC Request for Fuel Perfomance Data From time to time, the NRC has requested fuel performance data from the licensees and fuel suppliers. Also, of course, considerable data are available at the licensee and fuel supplier plants for Government perusal. While these methods can provide a valuable source of information it would be more desirable to have it made available on a routine basis, together with an analysis of the data. Inprove-ments to acquiring data and analyses are covered elsewhere in this report. 4.2.5 Nuclear Plant Reliability Data System The Nuclear Plant Reliability Data (NPRD) System (2) was developed by the Equipment Availability Task Force of the Edison Electric Institute (EEI). The system is designed to complie data (provided by organizations operating nuclear power generating facilities in the United States) pertaining to equipment failures and to generate reliability and failure statistics for this equipment. Though it is not mandatory for the organizations to provide such data, and the accuracy of l the reports are strongly dependent upon the completeness of the input data, the NPRD System can provide a valuable tool in improving the reif ability of equipment. l In regard to fuel assemblies the NPRD System defines a failure as any breach of l the cladding or any structural change, such as distortion or break, that for safety considerations causes abnormal maintenance or early replacement of a fuel assembly. 1 This definition would cover most fuel failures; however inclusion of the words "for safety considerations" could introduce some degree of interpretation in whether some failures should be included or not. In addition. it could be argued that perfomance outside of predicted limits can also be considered a failuret these would generally be related to failures in design. The limitations of this defini-tion would generally be most pertinent for an assembly in its last cycle. The fuel rod could contain bulges, collapsed sections, and other irregularities, without a breaching of the cladding. Severe bowing could also be present. but requiring no abnormal maintenance (in this case, maintenance being defined as removal of the assembly from the core). For purposes of this report, fuel failure is defined as (1) any breach of the fuel rod cladding. (2) any structural change of the assembly or its component parts. Such as distortion or break, which causes abnomal maintenance or early replacement of the assembly or plant operating restrictions, and (3) any structural changes of the assembly or its component parts which exceed predicted limits of performance. 4.2.6 Olscussion of Fuel Perfomance Oasa l l Strengthening the reparting of fuel perfomance data, as described in Section 4.2.2. l would still leave one limitation which can be significant. Fuel suppliers fre-l quently perform inspections and analysis of data which may not be available to the I l -

licensees. Fuel suppliers could issue topical reports at least once a year presenting results pertaining to their fuel and a discussion of their significance. Results and analyses of special testing and actions being taken to improve their fuel could also be included. It is recognized that suppliers may be reluctant to release data which may be of value to their competitors. With the increasing number of nuclear power plants operating each year, together with their periodic replacement of fuel assemblies, the data pertaining to fuel performance becomes increasingly difficult to handle. AEC Regulatory-Licensing placed a contract in early 1973 with Battelle, Pacific Northwest Laboratories for a preliminary study of a fuel performance analysis program. The program, using a computer, involves collecting, analyzing, and interpreting infonnation relating to fuel performance. The infonnation would be collected from the literature, reactor operators, fuel vendors AEC, and from other sources throughout the world. The report, issued in June 1973 UI, points out that performance information is available on thousands of fuel bundles; however, such information has not been systematically complied, analyzed and correlated with goal exposures achieved, failures incurred, or indentified failure causes. Also, the report states that analyses of fuel performance are essential for: Evaluating comparative performance of current fuels versus those in reactors of similar type, Assessing realistically the probable future performance and problems of fuel loadings, and Providing a key information base for the improvement of technical bases for design. Since the end of 1973, no further work has been performed in developing this particular program, due to need for funds and manpower on higher priority work. However, a program has been developed which has the capability of compiling fuel element design data, fabrication data, operating data and performance data. The program, at present, is of limited use due to a lack of fuel input data. Adoption of the comments presented in Section 4.2.2 would permit the program to become a valuable tool in fuel performance analysis. 5.0 Conclusions Fuel element failures have been caused by internal contamination (hydriding), manufacturing defects, mechanical damage, fuel pellet-cladding interactions, accelerated corrosion, fuel rod bowing, cladding collapse (densification of fuel), and by other miscellaneous causes. The failures experienced have not had any j adverse effect on the public health or safety. The problems which have had the most impact on plant performance are hydriding, fuel pellet-cladding interactions, and denstitcation. There have been a significant nunber of failures in which the cause is either unknown or unreported (primarily the latter). Plant restrictions, due to fuel failures or actions required to mitigate failures, have resulted in -27 .~ _

lowered plant availablility and capacity factors. Fuel failures have the potential for (1) increasing radiation exposure to plant personnel and the public, (2) com-pounding the problems of storage, waste reprocessing, and disposal, (3) increasing the cost of power generation, and (4) increasing the number and complexity of plant radioactive waste disposal systems. Most of the failure data in this report is for fuel manufactured several years ago. Current production fuel have incorporated several improvements in the design, in the manufacturing process, and in the quality assurance programs. PWR fuel is presently experiencing a very low failure rate. However, BWR fuel will probably continue to be subject to hydriding and pellet-clad interaction failures at least until all the older design fuel has been phased out; similarly, power generation losses due to fuel preconditioning and power restrictions will continue at least until the older fuel is phased out. Nuclear power plants are designed under the assumption that some fuel rods will fail and release fission products to the primary coolant system. Present monitoring systems are capable of detecting these activity levels and the radwaste systems are sized to safely handle the radioactivity released. When necessary, plant operating procedures (such as reduced power generation and control rod manipulation) can and have been employed to keep radioactive gaseous effluents from the plant well below permissible levels. While present monitoring systems are capable of detecting increased activity levels for even a few fuel rods, the monitors and associated l I analytical techniques cannot accurately determine the specific number or location of failed elements or assemblies during plant operation. Analytical techniques based on measured coolant activity or gaseous effluents during operation would be more accurate if the predictions were compared to in-core and out-of-core fuel inspection results. Present requirements for reporting fuel perfomance are not adequate in regard to consistency, depth and timeliness of such data. Some improvements were included in the recently issued Regulatory Guide 1.16 Revision 3; however, further improve-ments would be desirable in order to properly compile, analyze and utilize the data avainble at the utilities and fuel suppliers. Present technical specifica-tions, in regard to fuel performance, vary considerably from licensee to licensee. Topical repur o, by most fuel suppliers, are issued infrequently, if at all, and generally could be improved to increase their usefulness. The amount of performance data available is already large and is expected to grow l l at a significant rate. However, the information has not been systematically compiled, analyzed, and correlated with goal exposures achieved, failures incurred, I or identified failure causes. The creation of a centralized data bank would be invaluable. Analysis of fuel perfomance is essential for (1) evaluating compara-tive perfomance of current fuels versus those in reactors of similar type, (2) assessing effectiveness of past corrective actions, (3) assessing realistically the probable future performance and problems of fuel loadings, (4) providing a key information base for the improvement of technical bases for design, and (5) assessing the adequacy of quality assurance requirements. - - - -

BIBLIOGRAPHY 1. Regulatory Staff, U.S. A.E.C., Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, Regulatory Guide 1.70 Revision 1, October 1972. 2. Southwest Research Institute, San Antonio, Texas Reporting Procedures Manual For the Nuclear Plant Reliability Data System. Revision 2. April 1974. Prepared under the direction of ANSI Subconsnittee N18-20 for Anerican National Standards Institute American Public Power Association, Edision Electric Institute, and- .U.S.A.E.C. 4 3. Regulatory Staff, U.S. Atomic Energy Consnission. Technical Report on Densification of Light Water Reactor Fuels, WASH-1236. November 14, 1972. 4 4. V. J. Plocido and R. E. Schreiber. Operational Experience with Westinghouse Cores (Up to October 1974). WCAP 8183. Revision 2 November 1974. l 5. W. J. Bailey, Fuel Failures in Commercial Nuclear Power Reactors June 1974. A study performed by Battelle Pacific Northwest Laboratories for the Quality Assurance Branch. Directorate of Licensing, U.S.A.E.C., under contract AT(45-1) - 1830. 6. H. E. Williamson and D. C. Ditmore, Experience With BWR Fuel Through September 1971, NED0-10505, May 1972. 7. Jersey Central Power and Light Company, Oyster Creek Station Semiannual Report No. 8 for Period January 1.1973. to June 30. 1973 Docket 50-219 August 31, 1975. 8. D. C. Ditmore and R. B. Elkins, Densification Considerations in BWR Fuel Desian and Performance, NEDM-10735. December 1972. Various supplements to the report were issued during 1973. i 9. Boiling Water Reactor Systems Department General Electric Company, San Jose. 1 California GEGAP III: Model for the Prediction of Pellet - Claddino Thermal J Conductance in BWR Fuel Rods, NED0-20181. Revision 1, November 1973.

10. Westinghouse Electric Corporation, Nuclear Energy Systems. WCAP-8253, Source Term Data for Westinghouse Pressurized Water Reactors, May 1974.

i 11. W. J. Bailey, Fuel Performance Analysis Program - A Preliminary Study June 1973. A study performed by Battelle Pacific Northwest Laboratories for the Quality Assurance Branch, Directorate of Licensing, U.S.A.E.C.. under contract AT(45-1)- 1830.

12. Letter, Arkansas Power & Light Company. Arkansas Unit 1. J. D. Phillips to AEC-RO,

Subject:

Fuel Assembly - Orifice Rod Assembly Binding Docket 50-313, dated i February 13, 1974. i l 13. H. E. Williamson and D. C. Ditmore. " Current BWR Fuel Design and Experience." Reactor Technology, Vol. 14-1, 68-98, Spring 1971.

14. Consumers Power Company, Report of Operation of Big Rock Point Nuclear Plant, j

January 1. 1971 through June 30. 1971, Docket 50-155, August 13. 1971.

15. Consumers Power Company. Big Rock Point Semiannual Report for Period I

January 1 Through June 30. 1972 Docket 50-155. August 30, 1972. l

16. Letter (TWX) Consumers Power Co. to AEC-DOL. Subject-Removal of Cobalt Target Rods From Fuel Assemblies at Big Rock Point, dated March 26, 1973, (Docket 50-155).

l l

17. Letter, Consumers Power Company to AEC-DOL. Subject-Modification to Fuel Assemblies j

at Big Rock Point Point, dated April 10, 1973, (Docket 50-155). ] i i i ! p L

18. Letter. Consumers Power Company to AEC-DOL, Subject-Spent Fuel Rod Found on Pool Floor, dated November 20, 1973. (Docket 50-155).

19. Consumers Power Company, Big Rock Point Nuclear Plant Special Report (SR-14),

dated June 22. 1973. Docket 50-155, July 5,1973.

20. Consumers Power Company. Big Rock Point Semiannual Report of Operations for Period January 1.1974 Through June 30, 1974, Docket 50-155, August 29, 1974.

21. Consumers Power Company, Big Rock Point Plant Operations. Special Report No.18, Docket 50-155. August 2, 1974.

22. Consumers Power Company, Big Rock Point Unusual Event No.13-74 of October 22, 1974. Docket 50-155, November 21, 1974.

23. H. H. Klepfer R. B. Richards, and T. Trocki, " Fuel Perfonnance in Boiling Water Reactors," American Power Conference, April 1972.

24. Nuclear Power Plant Operating Experience During 1973. 00E-ES-004, Office of Operations Evaluation, Directorate of Re9ulatory Operations. USAEC. December 1974. (Also designated as WASH-1362).

25. Letter. Comonwealth Edison Co. to AEC-DRL. Subject-Additional Information Relative to Provisional Operating License DPR-19 for Dresden Unit 2, dated July 31, 1970, (Docket 50-237). Exhibit I includes "Dresden II-Investigation of Fuel Defects" dated July 1970.

26. Comonwealth Edison Co.. Report No. 5 Fuel Inspection and Evaluation. Dresden Unit 2. Docket 50-237, June 4,1971.

27. Comonwealth Edison, Dresden Station Unit 2-Special Report No. 34, Fuel Perfonnance Report, Cycles I-III." Docket 50-237. November 13, 1973.

28. Telephone conversations with various NRC regional and licensee personnel (primarily for plants refueled between 6/30/74 and 12/31/74) by P. E. Bobe. Office of Operations Evaluation Directorate of Regulatory Operations. USAEC, on January 8 and 9. 1975. The data must be considered preliminary at this time since the data and its interpretation have not been completed or formally published by the licensees in most cases.

29. Letter, Commonwealth Edison Company, Dresden Station Unit 3. B. Lee, Jr. to AEC, Docket 50-249, dated April 28. 1971.

30. Comonwealth Edison, Dresden Station Unit 3 - Fuel Perfonnance Report. End of Cycle 1. Special Report No. 35. Docket 50-249, April 29, 1974.

31. Comonwealth Edison, Dresden Station Units 1. 2. and 3 Semiannual Report for Period January 1. 1974 Through June 30. 1974 Dockets 50-10, 50-237, and 50-249, respectively, August 21, 1974.

32. Comonwealth Edison, Dresden Unit 3 Abnonnal Occurrence Report No. 50-249/74-38 of October 31, 1974 Docket 50-249, January 17. 1975.

33. Letter, Iowa Electric and Power Company Duane Arnold Energy Center C.W. Sandford to AEC-RO,

Subject:

Fuel Bundle Lower Tie Plate Orifice", Docket 50-331, dated December 26. 1973. 34. R. N. Duncan, P. G. Smerd. H. Knaab, and R. Manzel, " Fuel Performance Experience in CE and KWU Pressurized Water Reactors," Trans. Am. Nucl. Soc. Vol. 18, pp. 250-251. June 1974.

35. Letter, Allis Chalmers to AEC-DRL, Subject-Fuel Pin Bowing at Lacrosse, dated June 25,1969, (Docket 115-5). -. _.._.

36. Lacrosse-Boiling Water Reactor Monthly Operating Report for November 1969

.s DPC-851-1, (Docket 115-5).

37. Letter Dairyland Power Cooperative to AEC-DOL, Subject-Lacrosse Reports to DRL on Bowed Fuel Pins, dated April 24,1970,(Docket 115-5).

38. Dairyland Power Cooperative, Inspection of LACBWR Fuel in November 1970, January 29,1971,(Docket 115-5).

39. Dalryland Power Cooperative, Lacrosse Boiling Water Reactor Monthly Operating Report for August 1972. DPC-851-47, August 1972, (Docket 50-809).

40. Dairyland Power Cooperative, Lacrosse Boiling Water Reactor Monthly Operati.n,q, Report for December 1972, DPC-851-51 December 1972, (Docket 50-409).

41. Dalryland Power Cooperative, Lacrosse Boiling Water Reactor Monthly Operating Report for April 1973, April 1973 Docket 50-409.

42. Dairyland Power Cooperative, Lacrosse Boiling Water Reactor Monthly Operatig Report for November 1973. LAC-TR-004 November 1973, Docket 50-409.

~

43. Letter, Connecticut Yankee Atomic Power Company to USAEC, Subject-Preliminary Report of Two Failed Control Rod Cluster Assemblies, dated May 15, 1970, (Docket 50-213).

44. W. J. Dollard and F. W. Kramer, " Westinghouse Nuclear Fuel Operating Experience " American Power Conference, April 1972.

45. Letter, Carolina Power and Light Company to AEC-DRL, Subject-Items of Interest Found During Refueling Inspection at Robinson-2, dated May 25,1973,(Docket 50-261).

s 46. H. M. Ferrari, E. Roberts, and J. Scott, " Fuel Densification Experience in Westinghouse Pressurized Water Reactors " BNES Conference, October 15-19, 1973 (pp. 54.1-54.4).

47. Carolina Power and Light Company, H. B. Robinson Unit No. 2 Routine Operating Report No. 6 for Period January 1.1973 - June 30,1973 Docket 50-261, August 29, 1973.

48. Carolina Power and Light Co., H. B. Robinson, Unit 2-Report Re the Discovery of Several Sections of a Fuel Assembly Grid Clip in $9sm Generator C, Docket 50-261, i

December 27, 1973.

49. Carolina Power and Light Company H. B. Robinson Unit No. 2 Routine Operating Report No. 8 for Period January 1.1974 - June 30,1974 Docket 50-261, August 29, 1974.

50. Region IV Daily Report of 11/12/74 to Regulatory Operations - Headquarters.

51. Letter (TWX), Consolidated Edison Co. of New York to AEC-DOL, Subject-Failure of Fuel Assembly Nozzle, dated August 27, 1970, (Docket 50-003).

52. J. A. Pezzello and M. Lee " Fuel Perfomance of Indian Point Unit No.1. "Trans. Am. Nuclear Soc., Vol.16. pp.102-103, June 1973. d

53. Wisconsin Public Service Corporation, Kewaunee Nuclear Power Plant Abnomal Occurrence Report No. 50-305/74-15 dated September 13, 1974, Docket 50-305, September 13, 1974.

i I 54. J. LeBlanc and J. Gibbons. " Maine Yankee Startup and Initial Operations", Proceedings of American Power Conference, Volume 35 (1973). 4 4...

55. Maine Yankee Atomic Power Company, Maine Yankee Operation Report for September 1972.

Docket 50-309, November 7,1972.

56. Nucleanics Week, 15(18)12 (May 2,1974).

57. Maine Yankee Atomic Power Company, Maine Yankee Semiannual _ Operating Report for the Period January 1, 1974 to June 30, 1974 Docket 50-309, August 27, 1974.

58. Internal NRC connunications, January 7,1975.

59. Nuclear Assurance Corporation, Nuclear Power Plant Performance, p. 26, January 1973.

60. Letter, Millstone Point Company, Millstone Nuclear Power Station, W. G. Counsil to AEC-DOL, transmitting abnormal occurrence report A0-74/5, Docket 50-245, dated September 25, 1974.

61. Northern States Power Company, Monticello Nuclear Generating Plant Unit 1, Cycle 3 Startup Report and Summary Status of Fuel Report Docket 50-263. July 19, 1974.

62. Niagara Mohawk Power Corporation, Operating Report for January 1 through June 30, 1971 for Nine Mile Point Docket 50-220, June 30,1971.

63. Niagara Mohawk Power Corporation, Nine Mile Point Nuclear Station Semiannual Report of Operations for Period January 1,1974 - June 301974 Docket 50-220, August 26, 1974.

64. D. H. Roy and J. S. Tulenko, " Performance of the Oconee-1 Core," Trans. Am. Nucl. Soc., Vol. 18, pp. 249-250, June 1974. 65. E. R. Appleby (Battelle Pacific Northwest laboratories), Nuclear Power Reactor Summary, January 1974.

66. Letter, Jersey Central Power & Light Company, Oyster Creek Station I. R. Finfrock, Jr. to AEC-DOL,

Subject:

Fuel Assembly Loading Error, Docket 50-219, dated May 24, 1972. 67. B. H. Cherry, et al., " Nuclear Fuel Perfonnance in Oyster Creek," Trans. Am. Nucl. Soc., Vol.16, pp.101-102, June 1973.

68. Jersey Central Power and Light Company, Oyster Creek Station Semiannual Report No. 10 for Period January 1, 1974, to June 30, 1974, Docket 50-219, August 29, 1974.

69. Letter Consumers Power Company, Palisades Plant, R. L. Haueter to AEC-DOL, Docket 50-255, dated December 16, 1971.

70. Consumers Power Compay, Palisades Plant Semiannual Report of Operations No. 7 for Period January 1, 1974 Through June 30, 1974, Docket 50-255, August 30, 1974.

71. Letter, Boston Edison Company, Pilgrim Nuclear Power Station, J. Carroll to AEC-DOL, Docket 50-293, dated November 30, 1973. Also, AEC Release No. R-523, dated December 17, 1973.

72. Letter, Boston Edison Company, Pilgrim Nuclear Power Station, M.J. Feldmann to AEC-DOL, Docket 50-293, dated February 8, 1974.

73. Letter, Boston Edison Company Pilgrim Nuclear Power Station, G. D. Baston to AEC-DOL, transmitting abnormal occurrence report A0-74/3, Docket 50-293, dated January 30, 1974.

74. Pilgrim Nuclear Power Station Pilorim Semiannual Report No. 4 for Period January 1, 1974 to June 30, 1974 Docket 50-293 August 30, 1974. ___

75. Letter, Wisconsin Electric Power Co. to AEC-DOL, Subject-Collapsed Fuel Rods Found at Point Beach-1, dated October 24, 1972, (Docket 50-266).

76. Wisconsin Electric Co., Fuel Examination at Point Beach-1, Docket 50-266.

November 10, 1972. I

77. Letter, Commonwealth Edison to AEC-DOL, Subject-Release Rate for Iodine-131 Exceeded at Quad Cities -1 and -2, dated July 20, 1973, (Dockets 50-254 and 50-265, respectively).

78. Comonwealth Edison, Quad - Cities Station Semiannual Report for Period i

July 1, 1973 through December 21, 1973, Dockets 50-254 and 50-265, respectively. February 25, 1974.

79. Comonwealth Edison, Quad - Cities Units 1 and 2 - First Year Operating Report.

Dockets 50-254 and 50-265, respectively, May 10, 1974.

80. Comonwealth Edison, Quad - Cities Station Units 1 and 2 Semiannual Report for Period January 1,1974 through June 30, 1974 Dockets 50-254 and 50-265, respectively, August 22, 1974.

81. A Review of Fuel Rod Integrity at the Ginna Reactor, WCAP-7703, September 10, 1971, (Docket 50-244).

82. Letter, Rochester Gas and Electric Corp. to AEC-DOL, Subject-Stray Fuel-Rod End-Plug Delays Refueling for Four Days at Ginna, dated October 30, 1972, (Docket 50-244).

83. Rochester Gas and Electric Corp., Ginna Operations Report for June 26 through December 31, 1972, Semiannual Report No. 6 Docket 50-244. February 27, 1973.

84. "NUCLEX '72 - Report on Technical Meeting, Part 1," Nuclear Engineering International, pp. 1027-1030 December 1972.

85. Harry M. Ferrari, " Nuclear Fuel Experience in Westinghouse Pressurized Water Reactors," Trans. Am. Nucl. Soc., Vol. 16, p. 101, June 1973.

86. M. A. Rigdon, et al., " Babcock and Wilcox's Irradiation Program on Fuel Densification " BNES Conference, p. 59.1, October 15-19, 1973.

87. Virginia Electric and Power Company, Surry Units 1 and 2 Semiannual Operating Report for Period January 1,1974 through June 30, 1974 Dockets 50-280 and F0-281 respectively, August 30, 1974.

88. Virginia Electric and Power Company, Surry Unit 1 First Year Operation Report.

Docket 50-280, March 4,1974.

89. Letter, Florida Power and Light Co. to AEC-DRL, Subject-Fuel Cladding Defects at Turkey Point-3, dated July 20, 1973, (Docket 50-250).

90. Letter, Florida Power and Light Co. to AEC-DRL, Subject-Fuel Assembly Dropped at Turkey Point-4, dated April 20, 1973, (Docket 50-251).

91. Letter, Vennont Yankee Nuclear Power Corp. to AEC-DOL, Subject-Leaking Fuel Rods and Reduced Power Operation at Vermont Yankee, dated December 1,1972, (Docket 50-271).

92. Letter, Vermont Yankee Nuclear Power Corp. to AEC-DOL, Subject-Vermont Yankee Operation Report for February 1973, dated April 2,1973, (Docket 50-271).

93. USAEC, DRO Inquiry Report No. 50271/73-01Q to Vermont Yankee Nuclear Power Corp.,

Subject-Leaking Fuel Rods at Vermont Yankee, dated February 13, 1973. - _ _ _.

l

94. Vermont Yankee Nuclear Power Corporation, Vermont Yankee Semiannual Operatinq Report for the Period January 1,1974 to June 30, 1974 Docket 50-271, August 27, 1974.

95. Letter, Yankee Atomic Electric Co. to D. J. Skovholt (AEC-DRL), Subject-Conversion of Yankee to Zircaloy-Clad Fuel, dated March 24, 1969, (Docket 50-213).

96. Letter, Yankee Atomic Electric Co. to AEC-DRL, Subject-Removal of Zircaloy-Clad Test Elements, dated August 25, 1969, (Docket 50-029).

97. Yankeo Atomic Electric Co., Yankee Operation Report No.104 for August 1969.

Docket 50-029, September 23, 1969.

98. Letter, Yankee Atomic Electric Co. to Distribution Subject-Yankee Rowe Operation Report for October 1972, dated December 1,1972, (Docket 50-029).

99. Letter, Yanhae Atomic Electric Co. to AEC-DRO, Subject-Damage to Fuel Element

~-- at Yankee Rcwe, dated November 20, 1973, (Docket 50-029). 100. Letter, AEC-DOL to Yankee Atomic Electric Co., Subject-Limited Rearrangement of Fuel Assemblies in Yankee Rowe Core X, dated March 14, 1973, (Docket 50-029). 101. Letter Yankee Atomic Electric Company, Yankee (Rowe) Core X-XI Refueling Report (YAEC-7075. Ju A1974) and Yankee (Rowe) Core X-XI Refueling Reactor Component l Inspection D'AEC-7076. July 1974), WYR-74-38 November 6,1974 (Docket 50-029). 102. R0 Inquiry Report No. 050-304/73-01Q from AEC to Commonwealth Edison Company, Zion Unit 2. Docket 50-304, dated December 5,1973. APPENDIX A - GLOSSARY This Glos *ary contains definitions of the terms utilized throughout this report. The definitions are generally based upon Edison Electric Institute (EEI) defini-nitions as well as accepted engineering and trade terminology. The definition of fuel failure has been broadened from its usual meaning i.e.. a perforation or defect in the fuel cladding. a. Fuel Failure: Generally defined as a perforation or defect of the fuel cladding, or any structural change. such as distortion or break, which causes abnonnal maintenance or early replacement of a fuel assembly (or its component parts). or plant operating restrictions. Also, for pur-poses of this report, a failure includes any structural change of the assembly or its component parts which exceed predicted limits of per-

formance, b.

Fuel Rod: A cylindrical metal (zirconium alloy or stainless steel) tube which contains ceramic uranium (or other fuel such as plutonium) pellets. The tube (cladding) constitutes the first barrier to fission product release. A fuel rod is sometimes referred to as a fuel element. There is one reference in this report to a " fuel segment" (Table C-1 Item 3a). This was an early design concept in which the fuel rod consisted of several connected segments. Each segment contained uranium dioxide fuel pellets enclosed in a tube caoped at both ends with connectors to act as fasteners. Later designs utilized the present concept of full length fuel rods, c. Fuel Assembly: A group or array of fuel rods (some rods may contain a poison material) structurally bound together. A fuel assembly is sometimes referred to as a fuel bundle. d. Leaker: A leaker fuel assembly is one which contains one or more Tuel rods with breached cladding, pennitting the release of fission products to the primary coolant. The leakers are identified by " sipping". e. mSiin.: Analysis of the water (wet sipping) or gas (dry sipping) from a fuel assembly, for the presence of fission products. (See Section 3.4.1). f. Plant Availability Factor: The quotient of time (hours) that the plant was operated with the main generator on-line during a given period, divided by the total time (hours) in the given period, expressed as a percent. g. Plant Capacity Factor: The quotient of the net electrical output produced by the plant in a given time period divided by the net electrical output the plant would have produced had it been operated at its design electrical capacity (net) for the given period, expressed as a percent. h. Radioactive Gaseous Effluents: For purposes of this report, the term will be limited to fission product gases, since these gases are an indirect indication of fuel rod failures. Part of the fission gases, generally quite small, can be contributed by fission recoil into the coolant from " tramp uranium." For all plents, these gases are monitored at the plant stacks. In BWR plants, these gases are released and carried out of the reactor with the steam, subsequently appearing in the air ejector off-gas on their way to the stack. In this report. "off-gas" is limited to those gases monitored at the BWR air ejectors, i A-1

1. Tramp Uranium: Generally defined as trace amounts of uranium in the structural materials of the core and residual uranium that may be left on the surfaces of the fuel rods from the manufacturing process. l l A-2

l l l APPENDIX B - DESIGN DATA I. TABLE B-1 This table is a compilation of miscellaneous plant data of the 49 plants which have generated electricity as of December 31,1974. The plants are listed in order of date of first electrical generation. The listing includes 21 BWR's i and 28 PWR's. !!. Table B-2 This table lists some typical BWR fuel element design pa*ameters for the I original cores installed in various plants. The data shows the evolution of fuel element design over a number of years. 7 !!!. Table B-3 This table lists some typical PWR fuel element design paramatirs for the original cores installed in various plants. The data shows the evolution of fuel element design over a number of years. l IV. Table B-4 i This Table lists some fuel element design parameters for the standarized designs submitted by the 4 LWR nuclear steam system suppliers. It should be noted that the design submitted by Babcock & Wilcox was subsequently withdrawn. + 3 d l I i 4 i 1 l l b i, I i B-1 i i ~

TABLE B-1 MISCELLANEOUS PLANT CATA (LISTED IN OCER OF CATE OF FIRST ELECTRICAL GENERATION) Frincipal Owner Reactor Nuclear Architech) Docket No. of Fuel ho. of Control Date of First Therrj}) Capacity {}) Electric Plant (Utility) Type Steam Engineer Nweer Assectlies Assemblies Electrical Power System Generation (MWt) (MWe-Net) Supplier {,) Dresden 1 Coeronwealth BWR GE Bechtel 50-010 464(max) 80 4/15/60 700 200 Edison Co. IdI Yankee (Rowe) Yankee Atomic PWR W S&W 50-029 76 24 11/10/60 600 175 Electric Co. 21 'I 9/16/62 615 265 I Ir.dian Point 1 Consolidated PWR B&W 0/ Vitro 50-003 120 Edison Co. of New York, Inc. i m j~ Sig Rock Point Consuners Power EWR GE Eechtel 50-155 84(r.ax) 32 12/8/62 240 75 4 Co. of Michigan Hu:toict Eay 3 Pacific Gas & EWR GE Eechtel 50-133 184(max) 32 4/18/63 240 65 Electric Co. i III San Onofre 1 Southern Calif. PWR W Bechtel 50-206 157 45 7/16/67 1347 430 Edison & San Diego Gas & Electric Co. III Hadcas Neck Connecticut PWR W S&W 50-213 157 45 8/7/67 1825 575 Yankee Power l Co. Genoa Dairyland Power BWR AC S&L 50-409 72 29 4/26/68 165 50 Cooperative Oyster Creek 1 Jersey Central BWR GE BAR 50-219 560 137 9/23/69 1930 640 Power & Light Co.

~ .s 's i <.,,.v TABLE B-1 (Cont'd) I i Principal Owner Reactor Nuclear Architeckb) Docket No. of Fuel No. of Control Date of First Thery)"Electrieg) Power Capacity Plant' (Utility) Type Steam Engineer Number Assemblies Assemblies Electrical System Generation (MWt) ir (MWe-Net) Supplier (,) Nine Mile Point 1 Niagara Mohawk BWR GE 'O 50-220 532 129 11/9/69 1850 625 ' Power Corp. I9) 12'/2/69' 1520 490 R.E. Ginna 1 Rochester Gas PWR W Gil 50-244 121 33 1 i & Electric Co. Dresden 2 Commonwealth BWR GE S&L 50-237 724 177 4/13/70 2527 809 Edison Co. I I9) H.B. Robinson 2 Carolina Power PWR W Ebasco 50-261 157 53 9/26/70 2200 00 7 & Light Co. Point Beach 1 Wisconsin Elec. PWR W Bechtel' 50-266 121 37 11/6/70 1518 497. -M Power Co. & Wisconsin-Michigan Power Co. Millstone 1 Northeast Nuclear BWR GE Ebasco 50-245 580 145 11/29/70 2011 652 Energy Co. Monticello Northern States BWR GE Bechtel 50-263 484 121 -.3/5/71 1670 545 Power Co. Dresden 3 Commonwealth BWR GE S&L 50-249 724 177 7/22/71 2527 809 Edison Co. f9) Palisades Consumers Power PWR Comb Bechtel 50-255 204 45 12/31/71 2200 700 Co. of Michigan 1

i TABLE B-1 (Cont'd) Principal Owner ewtor Nuclear Architecfb) Docket No. of Fuel No. of Control Date of First Thery) Electrieg) Power Capacity Plant (Utility) lype Steam Engineer Number Assemblies Assemblies Electrical System Generation (MWt) (MWe-Net) Supplier {,) Quad-Cities 1 Com. Edison-Iowa-BWR GE S&L 50-254 724 177 4/12/72 2511 800 Illinois Gas & Electric Quad-Cities 2 Com. Edison-Iowa-BWR GE S&L 50-265 724 177 5/23/72 2511 800 Illinois Gas & Electric I9) Surry 1 Virginia Electric PWR W S&W 50-280 157 53 7/4/72 2441 788 & Power Co. Pilgrim 1 Boston Edison Co. BWR GE Bechtel 50-293 580 145 7/19/72 1998 664 Point Beach 2 Wisconsin Elec. PWR W Bechtel 50-301 121 37 8/2/72 1518 497 Power Co. & Wisconsin-Michigan Power Co. Vermont Yankee Vermont Yankee BWR GE Ebasco 50-271 368 89 9/20/72 1593 514 Nuclear Power Corp. I9) Turkey Point 3 Florida Power & PWR W Bechtel 50-250 157 53 11/2/72 2200 693 Light Co. I9) Maine Yankee Maine Yankee PWR Comb. S&W 50-309 217 85 11/8/72 2440 790 Atomic Power Corp.

TABLE B-1 (Cont'd) Capacity}) Electricg Frincipal Owner Reactor Nuclear Architecfb) Docket No. of Fuel No. of Control Date of First Therg}) Plant (Utility) Type Steam Engineer Number Assemblies Assemblies Electrical Power System Generation (MWt) (MWe-Net) Supplier (3) I9) 3/10/73 2441 788 Surry 2 Virginia Elec. & PWR W S&W 50-281 157 53 Power Co. Oconee 1 Duke Power Company PWR B&W 0/Bechtel 50-269 177 69(h) 5/6/73 2568 886 I9) 6/21/73 2200 693 Turkey Point 4 Florida Power & PWR W Bechtel 50-251 157 53 Light Co. I9) 8/25/73 1420 457 Fort Calhoun Omaha Public PWR Comb. GHDR 50-285 133 49 Power District 59) 6/26/73 2758 873 Indian Point 2 Consolidated PWR W UE&C 50-247 193 61 ? Edison Co. of New York. Inc. 59) 6/28/73 2760* 893* Zion 1 Commonwealth PWR W S&L 50-295 193 61 Edison Co. Brown's Ferry 1 Tennessee Valley BWR GE O 50-259 764 185 10/15/73 3293 1065 Authority Oconee 2 Duke Power Company PWR B&W 0/Bechtel 50-270 177 69(h) 12/5/73 2568 886 I9) Prairie Island 1 Northern States PWR W PS&E 50-282 121 33 12/4/73 1650 530 Power Company I9) Zion 2 Commonwealth PWR W S&L 50-304 193 61 12/26/73 2760* 893* Edison Co.

• Represents 85% capacity.

.~ ' TABLE B-1 (Cont'd) 1 Principal Owner Reactor Nuclear Architeckb) Docket No. of Fuel No. of Control Date of First Therq Power}) Electrieg Capacity}) Plant (Utility) Type Steam Engineer Nisnber Assemblies Assemblies Electrical + System Generation (MWt) (MWe-Net) Supplier (a) Peach Bottom 2 Philadelphia BWR GE Bechtel 50-277 764 185 2/19/74 3293 1065 Electric Co. I9) Kewaunee Wisconsin Public PWR W PS&E 50-305 121 33 4/8/74 1650 541 Service Corp. Cooper Station Nebraska Public BWR GE B&R 50-298 548 137 5/10/74 2381 778 Power District Duane Arnold Iowa Electric BWR GE Bechtel 50-331 368 89 5/19/74 1658 569 ? Light & Power Three Mile Metropolitan PWR B&W Gil 50-289 177 69(h) 6/19/74 2535 819 Island 1 Edison Company l Arkansas 1 Arkansas Power & PWR B&W Bechtel 50-313 177 69(h) 8/17/74 2568 850 Light Company Brown's Ferry 2 Tennessee Valley BWR GE O 50-260 764 185 8/28/74 3293 1065 4 Authority Peach Bottom 3 Philadelphia BWR GE Bechtel 50-278 764 185 9/1/74 3293 1065 i Electric Co. Oconee 3 Duke Power Company PWR B&W 0/Bechtel 50-287 177 69(h) 9/18/74 2568 886 Rancho Seco Sacramento Muni-PWR B&W Bechtel 50-312 177 69(h) 10/13/74 2772 913 cipal Utility District i e 4

TABLE B-1 (Cont'd) Principal Owner Reactor Nuclear Architecg) Docket No. of Fuel No. of Control Date of first Therq}) Capacity}) Electricg Plant (Utility) Type Steam Engineer Number Assemblies Assemblies Electrical Power Generation (MWt) (MWe-Net) System g Supplier,) Edwin I. Hatch 1 Georgia Power BWR GE SSI 50-321 560 137 11/11/74 2436 786 Company I9) 12/21/74 1650 530 Prairie Island 2 Northern States PWR W PS&E 50-306 121 33 Pcwer Company I9) 12/30/74 2560 845 Calvert Cliffs 1 Baltimore Gas & PWR Comb Bechtel 50-317 217 85 Electric Company ? y Notes to TABLE B-1: (a) Nuclear Steam Systems Suppliers: GE = General Electric Company; W = Westinghouse; B&W = Babcock & Wilcox; Comb = Combustion Engineering; AC = Allis Chalmers. (b) Architect Engineers: l S&W - Stone & Webster; O = Owner; S&L - Sargent & Lundy; B&R = Burns & Roe; Gil = Gilbert Associates; GHDR = Gibbs & Hill and Durham & Richardson; SSI = Southern Services, Inc.; PS&E = Pioneer Services & Engineering Company; UE&C = United Engineers & Constructors. l (c) Authorized power levels. (d)The core also includes 8 fixed Zircaloy crucifonn shim rods. I')The core also includes 16 fixed Zr-2 filler rods. (II Includes safety and regulatory rods. I9) Includes full and part length rods. (h) Includes 8 axial power shaping rods. l

TABLE B-2 TYPICAL BWR FUEL ROD DESIGN PARAMETERS (ORIGINAL CORES) Nuclear Date of First Fuel Pellet Fuel Rod Clad Clad Active Plant Steam System Electrical Diameter Diameter, Thickness Material Fuel Fuel Assembly Description Supplier Generation (Inches) OD (Inches) (mils) Length (Inches) Humboldt Bay 3 General 4/18/63 0.420 0.463 19 304 79 7 x 7 fuel rod array. Electric SS 49 fuel rods per fuel assembly. 172 assemblies in the core. Dresden 2 General 4/13/70 0.488 0.563 32 Zr-2 144 7 x 7 fuel rod array. Electric 49 fuel rods per fuel assembly. <? 724 assemblies in the core. cm Millstone 1 General 11/29/70 0.488 0.570 35.5 Zr-2 144 7 x 7 fuel rod array. Electric 49 fuel rods per fuel assembly. 580 assemblies in the core. Browns Ferry 1 General 10/15/73 0.488 0.562 32 Zr-2 144 7 x 7 fuel rod array. Electric 49 fuel rods per fuel assembly. 764 assemblies in the core. Also contains gadolinia bearing rods. Genoa Allis-4/26/68 0.350 0.396 20 348H 83 10 x 10 fuel rod array. (Also called Chalmers SS 100 fuel rods per fuel assembly. Lacrosse or 72 assemblies in the core. LACBWR) l

TABLE B-3 TYPICAL PWR FUEL R0D DESIGN PARAMETERS _(ORIGINAL CORES) Nuclear Date of First Fuel. Pellet Fuel Rod Clad Clad Active-Plant Steam System Electrical Diameter Diameter. Thickness Material Fuel Fuel Assently Description Supplier Generation (Inches) 00 (Inches) (mils) Length (Inches) i l Haddam Neck Westinghouse 8/7/67 0.3835 0.422 16.5 304 121.8 15 x 15 fuel rod array. i SS 204 fuel rods per fuel assembly. 0.3669 0.422 .24.3 Zr-4 120.0 157 assemblies per core, including 4 with Zr-4 clad. ? Surry 1 Westinghouse 7/4/72 0.3659 0.422 24.3 Zr-4 144 15 x 15 fuel rod array. e 0.3649 204 fuel rods per fuel assembly. 157 usemblies per core. l Prairie Island 1 Westinghouse 12/4/73 0.3659~ 0.422 24.3' Zr 144 '14 x 14 fuel rod array. 179 fuel rods per fuel assembly. 121 assemblies per core. t Indian Point 1 Babcock & 9/16/62 0.260 0.304-20.5 304 98.5 14 x 14 fuel rod array. (original core

• Wilcox SS 195 fuel rods per fuel assembly, supplied by (boron 120 assemblies per core.

Babcock & Wilcox) modified)

• This core contained UO -Th0 fuel material.

2 2

TABLE B-3 (Cont'd) Nuclear Date of First Fuel Pellet Fuel Rod Clad Clad Active Plant Steam System Electrical Diameter Diameter, Thickness Material Fuel Fuel Assembly Description Supplier Generation (Inches) OD (Inches) (milt) Length (Inches) 0.313 0.3415 12.0 14 x 14 fuel rod array. 304 101.5 Indian Point 1 0.280 0.3415 28.5 SS (avg.) 173 fuel rods per fuel assembly. (Core B-supplied 0.297 0.3415 20.0 120 assemblies per core. by Westinghouse) I Oconee 1 Babcock & 5/6/73 0.370 0.430 26.5 Zr-4 144 ~ 15 x 15 fuel rod array. i Wilcox 208 fuel rods per fuel assembly. m 1. o 177 assemblies per core. Palisades Combustion 12/31/71 0.359 0.4135 24 Zr-4 132 15 x 15 fuel rod array. Engineering 212 or 208 fuel rods per fuel assembly. 204 assemblies per core. Calvert Cliffs 1 Combustion 12/30/74 0.3795-0.440 26 Zr-4 136.7 14 x 14 fuel rod array. Engineering 176 or 164 fuel rods per fuel assembly. 217 assemblies per core. 1 I 4 r i r-

TABLE B-4 STANDARIZED FUEL ROD DESIGN PARAMETERS Reference Reactor Docket Fuel Pellet Fuel Rod Clad Clad Active Fuel Assembly Description Type Number Diameter Diameter Thickness Material Fuel (inches) OD (inches) (mils) Length (inches) General Electric BWR 50-447 0.416 0.493 34 Zr-2 148 8 x 8 fuel rod array, GESSAR 63 fuel rods and 1 water rod per fuel assembly. 732 fuel assemblies per core. (177 control rods.) Pestinghouse co RESAR-41 PWR 50-480 0.3225 0.374 22.5 Zr-4 164 17 x 17 fuel rod array. i 264 fuel rods per fuel assembly. 193 fuel assemblies per core. (61 full length and 8 part len th rod cluster control assemblies. Babcock & Wilcox B-SAR-241 PWR 50-481 0.324 0.379 23.5 Zr-4 143 17 x 17 fuel rod array. (this application 264 fuel rods per fuel assembly. (was withdrawn 241 fuel assemblies per core. by B&W on (76 full length control rod assemblies December 3, 1974) plus 8 axial power shaping rod assemblies). Combustion Engineering CESSAR PWR 50-470 0.325 0.382 25 Zr-4 150 16 x 16 fuel rod array. Average of 236 fuel rods per fuel assembly. 241 fuel assemblies per core. (81 full length plus 8 part length control element assemblies).

APPENDIX C - FAILURE EXPERIENCE I. TABLE C-1 This table summarizes the failure data for those commercial light water reactors which have generated electricity as of December 31, 1974 The approximate failure frequency together with the failure type, where known, are described. II. TABLE C-2 This table summarizes the categories of failure together with their associated l applicable items from Table C-1 above. C-1

. _ ~ - m - 4 i TABLE C-1 FUEL FAILURE DATA FOR C0ft9ERCIA1. LIGHT WATER REACTORS (PLANTS ARRANGED IN ALPHABETICAL ORDER) Item No. Plant Reactor Approximate Failure Frequency Failure Type Reference l Type No. Date la Arkansas 1 PWR Inspection of the 40 fuel assemblies Fabrication misaligment combined 12 2/74 containing orifice rod assemblies. with the small clearance between (0RA's) revealed that the ORA's the orifice rod and the guide tube were binding in the guide tubes of 7 nut I. D. of the fuel assemblies. This event occurred on 12/28/73, prior to initial fuel loading. l 2a Big Rock Point BWR Prior to April 1969, there were no known Failure (?) type unknown 13 Spring 4 failures and there were two suspected 1971 leakers in Type B fuel assemblies. (See item 2b below). [ 2b Big Rock Point BWR During April 1969 refueling outage, dry The observed fuel rod failures were of the same 13 Spring sipping revealed 7 leaky fuel assemblies character in all fuel types inspected and 1971 (4 type B and 3 Type E). Inspection of were limited to <20 in. of active fuel length 2 Type B and 2 Type E leaker fuel assem-in any given rod. The fuel rod failures re-blies revealed 5 and 9 failed fuel rods, sulted from heavy buildup of crud scale that i respectively. (Also noted were 8 failed caused the cladding surfaces to overheat to fuel rods in centermelt development fuel abnormally high temperatures (i.e., accelerated assembly D-50.) corrosion due to crud). I 2c Big Rock Point BWR Following Cycle 7, 100% of the core was Examination of the Type B and E leaker assemblies 6 5/72 sipped and 19 leaky fuel assemblies indicated failures are predominantly crud-related identified (5 Type B, 11 Type E, and. (i.e., accelerated corrosion due to crud). The and 3 Type EG). Type EG fuel failures gave indication of early-life hydriding. 2d Big Rock Point BWR Several fuel assemblies failed. (See Premature failure of several E fuel assemblies. 14 8/71 item 2e below). 1 7 1

TABLE C-1 (Cont'd) Item No. Plant Reactor Approximate Failure frequency Failure Type Reference No. Date Type 2e Big Rock Point BWR Following Cycle 8, 100% of the core was Examination indicated that the Type B and E 6 5/72 sipped and 17 leaky fur.1 assemblies leaker assemblies are predominantly of the identified (5 Type B, 11 Type E, and crud-related failures previously described 1 Type EG). (i.e., accelerated corrosion due to crud). The Type EG failures appeared to be divided roughly between crud-related and early-life hydride failures. 2f Big Rock Point BWR Thirty one of 84 fuel assemblies were The failed assemblies consisted of 4 types 15 8/72 found to have failed. of experimental bundles. 2g Big Rock Point BWR Cobalt target rods in 4 fuel assemblies Fuel inspection determined that several of 16 March became unlocked. the cobalt target rods had become unlocked 17 and in four fuel assemblies. The loose cobalt April rods were removed and the fuel assemblies

1973, p

recharged into outer rows in the core. resp. Analysis shows power peaking will not occur; w also change in flow distribution will not have a large effect. Unlocking resulted from insufficient force in the spring that locked the rods in position. Modification made that increases force required to unlock target rods (i.e., installed auxiliary spring which has locking force of 18 lb). 2h Big Rock Point BWR One failed fuel rod. Tie rod from "E" type fuel bundle unexpectedly 18 11/73 found on spent fuel pool floor. 21 Big Rock Point BWR Twenty three of 84 fuel assemblies. Probable cause was accelerated corrosion. 19 7/73 examined by sipping early in 1973 Evidence of internal hydriding was not observed. contained failed fuel rods. 2j Big Rock Point BWR Dry sipping of all 84 assemblies Most probable cause was accelerated 20 8/74 during lith refueling (starting 3/23/74) cladding corrosion induced by crud spalling 21 and showed 9 assemblies were leakers. and the resulting localized heating. Crud 8/74, buildup on one-cycle assemblies was minimal. resp. l

-~ .~.--. -.. ~. - .~ 2 J I l TABLE C-1 (Cont'd) Item No. Plant Reactor Approximate. Failure Frequency Failure Type. Reference. T_ype No. Da te ' 2k Big Rock Point BWR Off gas rates continued at high levels. Unreported. However, it is likely to be 20 8/74-Power derated to 63 MWe in May 1974, similar to item 2j above. After encountering other plant problems in June, decision was made to refuel once again. Dry sipping of 71 assemblies showed 15 leakers. 2t Big Rock Point BWR During nondestructive fuel examination of.- Probable cause is a pellet of higher 22 H 11/74 ' fuel which had undergone irradiation, but density which had been mislocated in a. j not yet received its final' goal exposure, rod; most likely the pellet was overlooked an anomalous pcsk in gamma activity was during cleanout of equipment during enrich-discovered. ment runs. -{ 3a Dresden 1 BWR Of 77,184 fuel segments 22 failed (<0.1%) ' Five fuel segments failed because of'. 13 ' Spring l q) Ten of these failed during fourth operating accelerated corrosion due to bowing and 5 1971 l j cycle (May 1965-February 1967). failed because of internal corrosion'due to ) end plug stringers. Twelve others (which i were operated well beyond original design - burnups) failed due to inadequate spa:e for expansion and fission gas release. 3b Dresden 1 BWR Of 400 fuel assemblies, 5 failed.. The 5 Underwater inspection of 4 of these 5 assem-13.1971 leakers in' 400 assemblies could result blies revealed no fuel rod failures. The from imperfections in 5 out of 13,000 fuel fifth assembly has one fuel rod with a cracked rods (<0.1% defects.) bottom end-plug weld. 3c Dresden 1 BWR Twenty nine fuel assemblies failed (3 type Of the fuel rod failures, approximately half 13,6 Spring III B, 19 Type III F, 7 Type V). due to brittle longitudinal cladding cracks 23

1971, caused by strain localization and half due May.1972.

to internal hydriding. and April 1972 resp. l l 4

t l TABLE C-1 (Cont'd) Item Plant Reactor Approximate Failure Frequency Failure Type Reference No. Date No. Type 3d Dresden 1 BWR Sipping results at end of Cycle 6 The 58 failed fuel rods had brittle longitu-6. May 1972 (Sept. 1969) indicated 29 leaking fuel dinal cracks characttristic of pellet-to 23 and April assemblies and at end of Cycle 7 (Sept, cladding interaction mechar, ism (longitudinal

1972, 1971) another 20 leaking fuel assemblies.

crack-strain localization failures). resp. In the 49 assemblies, 58 failed fuel rods noted (@.4% of the 14.472 fuel rods of Type III B. III F, and V reload fuel). 3e Dresden 1 BWR During Fall 1973 refueling. 46 of 464 Most probable cause was pellet-clad 24 12/74 assemblies were identified as leakers, interaction. by combination of in-core and out-of-core wet sipping. 4a Dresden 2 BWR Significant offgas release observed as Four fuel assemblies dissassembled and fuel 25 7/70 early as first week of May 1970 during rods examined. Defects observed were minor m ln operation and testing at 50% of rated and were primarily small blisters on individual power. A total of 131 fuel assemblies rods. The blisters indicate highly localized were sipped out of the core and 27 chemical reaction in the cladding; the assemblies identified as failed on basis localized points of reaction are brittle. Some of sip signals. Two other fuel assem-failed fuel was located in areas of the core blies remained out of core on basis of considerably removed from the high probability visual inspection results. (seeitem suspect areas defined by flux tilting. Cause 4dbelow) of the fuel failures has not been determined at this time, but it is most likely due to an abnormal condition introduced during fuel manufacturing (internal hydriding). The 29 defective fuel assemblies were replaced with identical assemblies that had been fabricated for Dresden 3. 4b Dresden 2 BWR Off-gas began increasing in May 1970. Investigative work in June 1070 indicated that 26 6/71 In March 1971, 215 fuel assemblies leaky fuel rods were caused by zirconium hy-removed. (See item 4e below) driding from inside of the fuel rod due to an unidentified hydrogeneous material from an l unidentified source. The 215 fuel assemblies were removed based on their confirmed leakage, other suspicious data, or statistical evaluation performed to determine potential leakers.

i i 1 l j TABLE C-1 (Cont'd) l-item Plant Reactor Approximate Failure Frequency Failure Type.

Reference:

i, No. Type No. Date-1 j 4c Dresden 2 BWR Of 724 fuel assemblies 69 identified Early-life failures caused by internal hydriding ~ 6 May and as leaker assemblies and the prospect caused by an initial hydrogen impurity inadver-23 April exists of considerable additional fail. tantly introduced during fabrication. The

1972, j

ures or incipient failures still re-specific impurity or exact means of introduction resp. L j mainingincore.(Seeitems4dand4e could not be determined. Initial fuel load was below). not vacuum outgassed. 3 J 4d Dresden 2 BWR Twenty eight leaker assemblies; of Internally initiated, localized hydriding of-27 11/73 (Cycle 1) the 28 assemblies, 19 each had at cladding caused by some unspecified hydrogenous least 1 perforated fuel rod. 4 each had impurity or impurities (trace amounts of alcohol at least I defective fuel rod, and 5 and other hydrocarbons found) most likely intro-t each had at least 1 Questionable fuel duced in the manufacture. Also 1 assembly was i l rod. It was estimated early in 1971 damaged during the outage.and was replaced. ] that 60-70 fuel assemblies were causing { the off-gas problem. t n i i a 4e Dresden 2 BWR Forty one of 668 sipped fuel assemblies Fuel rod failures caused by hydrogenous impu-27 11/73 i (Cycle 1A) identified as leakers. Thirty five of rities during manufacture. In addition to the I the 37 Dresden 2 type and 4 of 4 Dresden 41 leakers.174 other assemblies also replaced 3 type fuel assemblies each had at least in attempt to minimize further hydriding fail-1 perforated fuel rod. ures. l 4f Dresden 2 BWR Of 239 fuel assemblies sipped. 2 of the Two of 5 defective rods revealed blisters of 27 11/73 (Cycle II) 215 CY (reload) type and 1 of 7 DN (orf-type attributed to internal hydriding; other ginal core) type fuel assemblies were 3 showed nothing unusual. identified as leakers. Questionable fuel rods (7no.) and 5 defective fuel rods were replaced with other sound discharged 4 fuel rods. j 4g Dresden 2 BWR Off-gas activity during(See item 4h 1974 indicated 27 11/73 and l (CycleIII) several rod failures. 24 12/ 74 - [ ] below). resp. i ] 4h Dresden 2 BWR During Fall 1974 refueling (end of Cycle Results are preliminary at this time. Data 28 1/75 III), 615 assemblies were wet sipped out-is still being evaluated. 1 i of-core. Thirty eight defective assem-blies were detected. 'W gu--- m i

TABLE C-1 (Cont'd) 4 Item Plant Reactor Approximate Failure Frequency Failure Type Reference No. Type No. Date 5a Dresden 3 BWR As a result of quality control Deficiencies in quality control -29 4/71. audit, it was found that a small program related to fuel fabrication. number of fuel rods (5 in 10,000) contained pellets of 2.4% enrichment instead of 1.4% enrichment. 4 Sb Dresden 3 BWR As a result of fuel failures and off-gas Most probable causes are hydriding and pellet-30 4/74 limitations, first refueling moved from . clad interactions. Fifty two new assemblies Fall of 1973 to Spring 1973. One hundred and 51 reconstituted assemblies were installed. three of 724 assemblies were identified as leakers or suspect leakers. Sc Dresden 3 BWR Second refueling began 3/11/74. In-core Unreported; however, most probable causes are 31 8/74 and out-of-core sipping showed 27 deft-those of items Sb above. The 33 assemblies nite leaker assemblies plus 6 probable plus 11 high exposures assemblies (all of the c, defective assemblies. 7x7 design) were replaced by new assemblies e, of the 8x8 design, i d 5d Dresden 3 BWR On October 31, 1974, a sudden increase Most probable cause of failure is pellet-clad 32 - 1/75 in off-gas radiation occurred, indica-interaction, due to allowing rapid local ting that several fuel rods has ruptured. power changes to occur. The plant has been limited to lower power levels since 10/31/74 to reduce the off-gas rates. i 6a Duane Arnold BWR Prior to fuel loading, during fuel Poor inspection at the fabrication facility. 33 12/73 bundle inspection, it was found that one Type I bundle had a lower tie plate without the required orifice. In addition, a Type II bundle was discovered with a lower tie plate containing a Type I orifice.

.._.-___,_..___m_.. ..- ~._- -- ...-4 i s 4 l TABLE C-1 (Cont'd) l Item Plant Reactor Approximate Failure Frequency' Failure Type Reference 3 No. Type No. Date 7a Fort Calhoun 1 PWR On a fuel rod basis,' the failure rate is Unreported 34 6/74 l <0.01%. 8a Genoa BWR Fuel rods of 2 assemblies found to be bowed in 35 6/69 bottom quarter. Bowing of one may be enough to effect future behavior. Cause of bowing not known-unless locked-in tube-drawing stresses i were released by tubes standing in 540 F. water. f 8b Genoa BWR Bowed fuel rods in'13 fuel assemblies. Bowed '36 '11/69 rods on side of assenbly adjacent to fully with-l drawn control rods. 8c Genoa BWR Several fuel assemblies removed from reactor 37 -. 4/70 had fuel rods that were significantly bowed. c, Bd Genoa BWR Bowing of fuel pins first observed in May 38 1/ 71 1969. It was detemined that shroud locking. rings had been unlocked during previous opera-tion. This condition caused the fuel assentlies t to be improperly seated and produced twisting 1 and stressing of the fuel assemblies. 8e Genoa BWR Fission product leakage occurred in The fission product leakage resulted in'a 39 8/72' l i several fuel assemblies. One fuel rod stack release of I-131 in excess of technical was severed. specification limits. Inspection of one fuel assembly (No. 64) revealed a severed i fuel rod. I Bf Genoa BWR Five fuel assemblies may have cladding When failed-fuel-element-detection. system ~ 40 12/72, 2 failures. (See item 8g below) placed in service, results indicated that cladding failures may have. occurred on 5 fuel assemblies. i a l t c w w s- .,n.

TABLE C-1 (Cont'd) Item Plant Reactor Approximate Failure Frequency Failure Type Reference No. Date No. Type 8g Genoa BWR Reactor refueled twice in 1973. Visual Clad defects believed to be caused by pellet-24, 12/74, and dry sipping examinations were used clad interactions. In the April examinations, 41, 4/73, and to detect defective assemblies. Twenty 2 of the defective assemblies had severely 42 11/73, and 23 defective assemblies were ren.uved bowed rods. In November, 1 of the failed resp. in April and November 1973, respectively. assemblies had 2 badly bowed rods. There is some preliminary evidence that acce-28 1/75 lerated corrosion may have contributed to some failed fuel rods. 9a Haddam Neck PWR Two fuel assemblies. Fuel assembly difficult to latch; examination 43 5/70 showed radial vane of spider assembly, which holds absorber rods, broken from spider. A second fuel assembly was also found to have another severed vane. P 9b Haddam Neck PWR Coolant activity indicates existence of Fuel failure type unknown. 44 4/72 a few leaking fuel rods since the first reactor cycle. (See item 9c below) 9c Haddam Neck PWR During the Fall 1973 refueling, repre-Unknown. 24 12/74 sentative inspections of fuel assemblies revealed no abnorrelities. Concentration of radioactive fission products in cool-ant was indicative of a few minor defects in a few fuel rods. 10a H. B. Robinson 2 PWR Rod-control cluster failure in one fuel Vane for rod-control cluster in a fuel assembly 45 5/73

assembly, separated from the spider nut during operation.

Failure occurred in braze joint; no cause found and no other failure was found. 10b H.B. Robinson 2 PWR During the past few months, flattened fuel rods 46 10/73 have been observed in Region I fuel (unpres-surized). No collapsed cladding observed in other regions which contain pressurized fuel.

~ 4 TABLE C-1 (Cont'd) Item Plant Reactor-Approximate Failure Frequency Failure Type Reference No. Type No. Cate 10b (Cont'd) Reactor refueled in Spring, 1973, with visual 47, 8/73 and inspection of all fuel during core unload and 24 12/74 TV inspection of periphery of fuel assemblies resp. in fuel pit. Twenty four of 53 A type assem-i blies of unpressurized fuel showed one or more collapsed rods. Also in Region 1, 2 cases of severe bowing and 2 instances of failed clad-ding were noted in the unpressurized fuel. The 53 Region 1 assemblies were replaced. 10c H. B. Robinson 2 PWR One grid strap on one bundle failed. Two small sections of a fuel assembly spring 48 12/73 (See 10d below) clip grid strap made of Inconel were discovered in steam generator during a routine shutdown in November, 1973. Normal reactor coolant would readily carry the grid strap sections r, 2. into the steam generator channel head. The c) spring clip grid pieces came from a single corner area of one grid; hence, six fuel rods are partially unsupported at the one grid location. Most likely explanation is that the grid edge caught on some portion of an adjacent assembly as the affected assembly was being inserted into its core position during refueling operations. Results suggest that the grid pieces are from previously irradiated fuel. During forthcoming refueling outage, comprehensive fuel inspection to be conducted to determine location of damaged i fuel assembly and affect, if any, on sur-rounding fuel assemblies. 10d H. B. Robinson 2 PWR (Continuation of item 10c above). During refueling, starting about 5/6/74, the 49 8/74 fuel assembly was identified as No. C-08. The pieces came from its 6th grid from the bottom. No apparent damage to other assemblies. I I

TABLE C-1 (Cont'd) Item Plant Reactor Approximate Failure Frequency Failure Type Reference No. Date No. Type 10e H. B. Robinson 2 PWR During Cycle 2, number of blips per moni-Unknown. 49 8/74 tored assembly (an indication of densifi-cation) increased to about 2.0. Increased iodine ac-Ivity indicated some fuel clad failures. 10f H. B. Robinson 2 PWR During refueling starting about 5/6/74, Mechanical damage. 49 8/74 examination indicated some bent nozzle springs in three Region 4 assemblies. 11a Humboldt Bay 3 BWR Three leaky Type II fuel assemblies Failure type unknown. The 3 leaker fuel 6 5/72 detected by sipping. assemblies had exceeded their design exposure. lib Humboldt Bay 3 BWR Eleven leaker Type III fuel assemblies The failed fuel rods in the leaker fuel 6 5/72 identified. assemblies exhibit the characteristics of early-life hydride failures, n lle Humboldt Bay 3 BWR Sixteen of 86 assemblies dry sipped Unreported. 24 12/74 during Fall 1973 refueling.were identi-fied as leakers. Ild Humboldt Bay 3 BWR Refueling started in October 1974. Sixty Unreported. Elements were all in high power 50, 11/74, assemblies were selectively dry sipped. density regions. 28 and 1/75, Eleven leakers were identified. resp. 12a Indian Point 1 PWR One fuel assembly with broken top nozzle. After loading a spent fuel assembly into ship-51 8/70 ping cask and while trying to disengage the loading tool which would not release, the top nozzle was broken from the fuel assembly. Fuel rods were not damaged. Causes of nozzle and grapple failures are being investigated. 12b Indian Point 1 PWR Coolant activity has indicated one-Failure type unknown. 44 4/72 or two leaking fuel rods. (See item 12d below).

--.~ ~. -. - l I TABLE C-1 (Cont'd) Item Plant Reactor Approximate Failure Frequency Failure Type Reference No. Type No. Date 12c Indian Point 1 PWR Top nozzles on two fuel assemblies became In both cases, tack welds holding the can to 52 6/73-separated from the perforated stainless the nozzle failed during refueling and :; pent-l steel cans. fuel cask loading operations. 12d Indian Point 1 PWR Plant was not operated in 1973. Coolant 24 12/74 activity level indicated approximately 1 fuel rod failure. 13a Kewaunee PWR On 9/4/74, primary coolant activity level No apparent cause can be identified at this 53 9/74 increased suddenly. Confirmed to be a time. No indications of clad creep. leaking rod. 14a Maine Yankee PWR During receipt inspection of fuel, a Believed to be caused by excess lateral loads 54 1973 i condition of non-contact between some applied to the fuel rods during handling or Zircaloy grid spring fingers and the

shipping, fuel rods was noted.

14b Maine Yankee PWR One fuel assembly replaced because of Basket containing in-core loading detector 55 11/72 ? damaged grids. One fuel assembly had to was being removed and caught under hold-down be modified. plate of an adjacent fuel assembly, lifting it off its 4 alignment pins and damaging 2 spacer grids. The fuel assembly was replaced with a spare._ Two diagonally located support-d plate alignment pins were found to be out of.

i alignment (fabrication error); a fuel element 4

had to be modified by enlarging the pinholes before it would fit properly. j 14c Maine Yankee PWR Higher than average coolant activity Failure type not indicated yet. Reactor may 56 5/74 indicates that some fuel rods have be shut down in June 1974 (orginally scheduled j failed. (See items 14d and 14e below) for refueling next year) to correct condition. 14d Maine Yankee PWR On a fuel rod basis, the failure rate is 34 6/74 <0.1%. (See items 14b and 14e) 4 14e Maine Yankee PWR Plant shutdown on 6/28/74 (earlier than Most likely hydriding and/or pellet clad - 57, 3/74, I anticipated) due to high iodine release interactions; analysis is continuing. In 28, and rates. All assemblies sipped and 43 addition, problems were identified concerning 1/75, i leakers were identified (41, 1 and 1 in fuel pin bowing and spacer-grid damage; the . resp. j Regions B, A and C, respectively), causes for these were not reported. The i 4 bowing of fuel pins resulted in some fuel } loading problems. =

~ __ - L TABLE C-1 (Cont'd) Item Plant Reactor Approximate Failure Frequency Failure Type Reference No. Date No. Type 14f Maine Yankee PWR Licensee data indicates a factor of 10-15 Unknown at this time. Licensee is planning to 58 1/75 increase in I-131 levels in primary cool-reduce power level to 80% until scheduled re-ant system during last two months of 1974. fueling in May, 1975. Licensed power level Gross primary coolant activity has in-is presently 95%. creased from 1% to 6% of Technical Speci-fication limit. Average energy of primary l coolant sample has been drifting downward 1 which would be indicative of fuel fail-4 ures. 15a Millstone 1 BWR Off-gas trend suggests some fuel rod fail-Failure type unknown; no fuel inspection to 6 5/72 ures exist in the core (has 508 fuel date. The fuel rod failures are suspected to assemblies). (See item 15b below) be early-life hydride failures. Only a portion of the initial core fuel assemblies contain. I fuel rods which have been vacuum outgassed. rs I 15b Millstone 1 BWR Of 112 fuel assemblies discharged, 105-110 Unreported. 59 1/73 leakers determined by sipping out of core. 15c Millstone 1 BWR Plant restricted frequently to 80% power Data, and its interpretation, are.not complete 28 1/75 due to off-gas activity. Refueling at this time. started in Summer 1974; of about 460 assemblies dry sipped, approximately 25 were leakers. Some visual examinations also performed. 15d Millstone 1 BWR During refueling, while transferring an Design deficiency of the grapple. 60 9/74 unchanneled spent fuel bundle from a fuel preparation machine to a spent fuel rack in the fuel pool, the bundle fell from the main grapple to the floor of the spent fuel pool. No release of activity was- ] measured, even though the bundle was damaged 16a Monticello BWR Offgas trend suggests that some fuel rod No fuel inspection performed yet. Fuel rod 6 5/72 failures have occurred in the core (has failures probably due to early-life hydriding. 484 fuel assemblies). (See item 16b Initial core fuel loaded in Monticello was not below) vacuum outgassed during fabrication. i e

_.m i 5 0 4 i i TABLE C-1 (Cont'd)' i Item Plant Reactor Approximate Failure Frequency-Failure Type Reference f No. Type No. Date 16b Monticello BWR During the first refueling shutdown during Predominant failure mechanism in the relatively ' 61-7/74 Spring 1973, sipping identified 25 out of low exposure Cycle 1 fuel was hydriding. 484 assemblies as leakers. A total of Failed assemblies replaced by 20 Type B 7x7 j 163 fuel rods were rejected. assemblies and 5 reconstituted assemblies. Off-gas rates at end of 1973 were indicative ] of several additional failures. i 16c Monticello BWR During Cycle 2, power was administrative 1y - 61 7/74 i reduced to reduce stack off-gas activity. (See item 16d below) 16d Monticello BWR During refueling, starting about 3/15/74, Limited visual inspection indicated that pellet-61 7/74-i in-core and out-of-core wet sipping iden-clad interaction was the predominant failure. j tified 83 leaking assemblies cut of 484. mechanism. For Cycle 3, 116 8x8 assemblies

j plus 7 reconstituted assemblies were inserted in the core.

16e Monticello BWR During Cycle 3, power was administratively Not known at this time. Consideration is 61 7/74 4 limited at various levels to reduce stack being given to sipping and replacement of i off-gas activity. defective assemblies prior to the end of design i life. (AsofDecember 31, 1974, a refueling . is planned for early 1975). ] 17a Nine Mile Point BWR Above normal off-gas activity indicated Maximum reactor power will be limited until - 62 6/71 i increasing fuel rod leakage. (See item fuel is replaced. r ] 17b below) j 17b Nine mile Point BWR 100% of core (532 fuel assemblies) sipped The leaker fuel ass m blies eh-fed predominant 6 5/72 i and 38 leaky fuel assemblies identified. failure characteristics of early-life cladding i hydride attack; however,10 of the leaker assemblies had fuel rod failures attributed to fretting wear from debris trapped in spacers. Of the 38 leaker fuel assemblies, 22 were = repaired (falled rods replaced) and 14 of the 22 recharged into reactor. Additional leakers were replaced in April, 1972.' + 17c Nine Mile Point BWR During the Spring 1973 refueling,104 of Unreported. 24 12/74 532 assemblies were identified as leakers, i ,L m

TABLE C-1 (Cont'd) Item Plant Reactor Approximate Failure Frequency Failure Type Reference No. Date No. Type l I 17d Nine Mile Point BWR Refueling started 3/29/74. Wet sipping Unreported. 63 8/74 I of assemblies in core identified 28 leakers. 1Ba Oconee 1 PWR Coolant activity levels observed corre-There has been very little change in activity 64, 6/74 and spond to fission gas escape through small level as a function of time since startup. In 65 1/74, pinholes. (See item 18b below) January 1973, one-half of fuel was replaced resp. with prepressurized fuel rods. 18b Oconee 1 PWR During the Fall 1974 refueling, visual Results are preliminary at this time, but no 28 1/75 examinations and physical measurements defective assemblies were detected. Approxi-were made on a few fuel assemblies, in mately one-third of the higher burnup assemblies accordance with the Technical were discharged from the core. Difficulties Specifications. were encountered with fuel handling equipment. 19a Oyster Creek BWR During sipping operations to detect fuel Personnel and/or procedure deficiencies. The 66 5/72 leakers, 1 fuel assembly was found to have fuel assembly had been improperly loaded into been installed 90 degrees counter-clock-the core and 4 administrative checks had failed wise from its proper position. The to discover the situation. reactor had been operating 6 months in this condition. l 19b Dyster Creek BWR 100% of core (560 fuel assemblies) sipped Fuel rod failures identified predominantly 6 5/72 and 44 leaky fuel assemblies identified, had characteristics of early-life hydride I attack. Of the 44 leaker fuel assemblies, 20 l were repaired (i.e., failed rods replaced) and recharged into reactor. 19c Oyster Creek BWR 100% of core (560 fuel assemblies) sipped Relationship between fuel assembly and fuel 67 6/73 during each outage. Bundle failure rod failure frequency indicates some positive (activity release to coolant) is due to correlation in fuel rod behavior within an only a few perforated rods among the 49 assembly. Observed clustering of failures is in an assembly. Fuel rod failure rate felt due to similarity in operating environ- <0.5% even for earliest cycles. (See ment within an assembly rather than casual item 19d below) failure interaction mechanisms between rods. 19d Oyster Creek BWR In the Spring 1973 refueling, 77 of 560 Unreported 24 12/74 assemblies were identified as leakers. I 1

4 TABLE C-1 (Cont'd) Item Plant Reactor . Approximate failure Frequency Failure Type Reference No. Type No. Date 4 19e Oyster Creek BWR Refueling outage started 4/13/74. In-core Unreported. 68 8/74 I sipping procedures identified 27 leakers out of 560 assemblies. 1 20a Palisades PWR -After inspection of control blade upper Design and/or procedure - ~69 12/71 end fittings, one blade bound slightly deficiences. The 0.060 inch thick as it was being reinstalled into the core. nut capture devices bend very easily. i The binding was caused by bent guide rod Also, it is difficult to insert the { nut capture devices on two adjacent Type blade by a crane without the blade i j A fuel bundles. Further inspection catching on the edges of fuel assemblies. ( revealed 20 bent nut capture devices in t i the core, all on Type A fuel. t 20b Palisades PWR On a fuel rod basis, the failure rate is Unreported. 34 6/74 j <0.15. 5) 20c Palisades PWR During an extended shutdown period, fuel Damage caused by handling. 70 8/74 l was being stored in a spent fuel pool. os The fuel was eventually reloaded into the i reactor.' One fuel bundle was damaged and had to be replaced. 21a Pilgrim 1 BWR The first core utilized temporary Design deficiencies. 71, 11/73-poison curtains in the bypass regions 72 ar.d (zones between the channel box assemblies). 2/74, i Coolant flow through bypass flow holes resp. caused curtain vibration which resulted in damaged fuel channels. During the late December, 1973 shutdown, the channels were i i inspected.' Damage observed ranged from - i slight to through-wall wear. All damaged-channels were replaced. (See similar item 1 31c). I i 1 v ~, r n-- r ~

TABLE C-1 (Cont'd) Item Plant Reactor Approximate Failure Frequency failure Type Reference No. Type No. Date 21b Pilgrim 1 BWR During transfer, an irradiated fuel Design and/or procedure deficiencies of the 73 1/74 bundle became detached from grapple grapple. Subsequently, a switch was installed and fell about 20 feet in the spent on the grapple to indicate closure of the hook fuel pool. Grapple hook apparently by activating a light on the bridge console. 4 j was not completely latched under The immediate fix was additional ~ administrative handle of the fuel element. There was controls requiring visual monitoring of grapple l no measurable release of activity. The hook closure. nose piece and the nose piece end of the fuel channel were crushed; there were no indications of broken fuel rods. 21c Pilgrim i BWR During refueling outage, fuel sipping Unreported. The 20 7x7 design assemblies were 74 8/74 began on 1/18/74. Sixteen fuel assem-replaced by new 8x8 design assemblies. blies showed indications of cladding per-i p forations. In addition, 4 other assem-j blies were damaged. 21d Pilgrim 1 BWR From 12/17/74 through 12/31/74 station Not known at this time. 28 1/75 operation limited to about 95% of rated i power due to high airborne effluent i release rates and unexplained pertur-I bations in the Augmented Off-Gas-System. 22a Point Beach 1 PWR Low-level coolant activity observed from Failure type unknown. 44 ' 4/72 j beginning indicating one or two leaking i fuel rods. (See item 22c below) 22b Point Beach 1 PWR Seventy fuel rods in 26 unpressurized fuel Examination performed by binocular observa-75 10/72 assemblies showed indications of collapse, tion. At time of shutdown, core had 13,000 representing a collapse ratio of 3.51. effective full power hours. Prepressurized i (See item 22e below) rods exhibited no evidence of collapse. j l 22c Point Beach 1 PWR Of 105 fuel assemblies sipped, 23 were Weak relationship found between leaky fuel 76 11/72 leakers and I was suspect. assemblies and those with collapsed fuel rods. I No correlation was found between collapses and core location, burnup, or fuel-assembly in-i serts. I 1

~ TABLE C-1 (Cont'd) Item Plant Reactor Approximate Failure Frequency Failure Type Reference No. Type No. Date 22d Point Beach 1 PWR During the past few months, flattened fue' rods.,46 10/73 have been observed in Region I fuel (unpressu-rized). No collapsed cladding observed in other regions which contain pressurized fuel. 22e Point Beach 1 PWR Twenty five fuel assemblies with failed 59 1/73 rods (collapses and leaks); 6 fuel assem-blies with collapsed sections have no leaks. 22f Point Beach 1 PWR During the June 1974 startup of Cycle 3 Tentatively attributed to pellet-clad interac-4 11/74 higher than expected main coolant radio-tion in conjunction with a rapid rate of reactor activity indicated some rod defects. power increase after the refueling shutdown. 22g Point Beach 1 PWR On site examinations performed during Measurements of 3 bowed rods showed the bow to 4 11/74 late 1972 revealed some bowed fuel rods, be a few tens of mils from a true centerline. Examinations of the rods showed no signs of I abnormalities due to operation of these rods n in the bowed condition. L 23a Point Beach 2 PWR During the Fall 1974 refueling, about 48 Examinations are not complete at this time. 28 1/75 higher burnup assemblies were discharged In one of the assemblies reinserted into the from the core. The remaining assemblies core, a small chip was found aM was removed. were reinserted after visual inspection. No defective assemblies were observed. { 24a Quad-Cities 1 BWR Release rate for I-131 exceeded several Defective fuel elements will be replaced dur b3 77 7/73 times durin 24b below) g a 5-day period. (See item upcoming refueling outage. The plant was administratively limited in power 78-2/74 level at times, starting in the.last half of i 1973, to maintain stack rates at acceptable levels. i t 24b Quad-Cities 1 BWR During refueling outage starting 3/31/74, Cladding hydriding and pellet-clad interac-79 5/74 1 in-core and out-of-core sipping identi-tions. 80 and fled 29 leaker assemblies out of 724, 8/74, resp.

~ t TABLE C-1 (Cont'd) i Item Plant Reactor Approximate failure frequency failure Type Reference No. Date No. Type 25a R. E. Ginna 1 PWR Coolant activity increases observed in Leaky fuel assemblies identified by visual ex-44, 4/72 March 1970. Leaks were confined to 32 aminations and leak testing. Evaluation of 81 and fuel assemblies in Region 3. Replacing observations suggested local hydriding result-9/71 the 12 worst leaker assemblies with fresh ing from fuel-contained moisture as the likely resp. i ones reduced activity to about half the cause of the leaks; it was later confirmed that i i level prior to outage. source of leaks was moisture contained in the fuel. i 25b R. E. Ginna 1 PWR End plug separated from fuel rod. During refueling operations, one fuel element 82 10/72 would not bottom properly, protruding 1/2 in. above other core assemblies. Four days later an end plug from a Region 3 fuel assembly was retrieved from the bottom core plate. Plug to be examined to see why it separated from the fuel rod; expected reason is severe internal hydriding. i 1. un 25c R. E. Ginna 1 PWR Fuel rod end-plug recovered. No indica-About 13 days required for replacement of 48 83 2/73 tion of fuel deterioration observed after unpressurized fuel assemblies and recovery of l 48 fuel assemblies were replaced with a fuel rod end-plug from the lower core-support ,i other assemblies, plate. l 25d R. E. Ginna 1 PWR 0.4%, based on primary coolant activity Study, without visual examination, indicated 84 12/72 early in first cycle; went additional most probable cause to be internal hydriding 400 days before further defects indicated. due to moisture which was later confirmed. Prior to this, modifications in fuel produc-i tion had been introduced to eliminate this since it had been expect:1 During Spring refueling, collapsed rods observed with col-( lapsed sections ranging from 4-8 cm in length f and are the result of gradual creepdown of cladding over an unsupported length due to high differential pressure. 25e R. E. Ginna 1 PWR During Cycle 1 refueling, a large number of 85 6/73 fuel rods were observed to be in interference with the top nozzles and a few of these rods were bowed. Rod interference and bowing were due to larger-than-expected Zircaloy growth i during irradiation. l l b ~ m

y

, / ' '

i m_. TABLE C-1 (Cont'd) Item Plant Reactor Approximate Failure Frequency Failure Type Reference No. Type No. Date 25f R. E. Ginna 1 FWR Flattened fuel rods observed in Regions I II 46-10/73 and III (all unpressurized fuel) during the Cycle 1 refueling in 1972. 25g R. E. Ginna 1 PWR Some' fuel rods collapsed because of in-reactor 86 10/73i densification of fuel. Densification pheno. ~ .'e menon reported by AEC to occur at linear heat s + rates as low as 1 to 2 kW/ft (33 to 65 W/cm). '- I 25h R. F. Ginna 1 FA Fuel failures and collapsed cladding noted. 59 1/73 Final 48 non-pressurized fuel assemblies discharged from core. 251 R. E. Ginna 1 Pt During'1973, coolant activity was indi-Unknown. K 24 12/74 .ative of some fuel failures (@.05%). ?g 26a San Onofre 1 PWR Coolant activity implies existence of one Visual examination of. fuel discharged during 44 4/72 or two leaking fuel rods during second second refueling outage disclosed two damaged operating cycle. fuel rods, which corroborated radiochemistry results..No other anomalies were found. 26b San Onofre 1 PWR All fuel assemblies in the core were visually inspected during the June 1973 24 12/74-refueling. No anomalies were noted. 27a Surry 1 PWR During first half of 1974, I-131 activity Cause unknown at this time. In addition, pri

_87, 8/74 and level in primary coolant indicates about mary pressure was reduced to preclude fuel 88 3/74, 2-4 defective fuel rods (See item 27b collapse. Densification induced power spikes resp.

below) i observed in all regions of the core. The num-ber is increasing, but all spikes are rela-tively small. 27b Surry 1 PWR During the Fall 1974 refueling, visual No specific failures noted. All examinations _ 28 1/75' (binocular) inspection was performed on are not yet complete. Eighty four higher burnup all 157 assemblies. TV inspection assemblies were replaced. Number of power spikes performed on 12, 20, and 12 Region 1, 2 decreased in second half of 1974. .ar.d 3 assemblies, respectively. No de-fects were observed. - Very little crud present. Slight bowing was observed in lu

~.. TABLE C-1 (Cont'd) ~ Item Plant Reactor Approximate Failure Frequency Failure Type Reference No. Date No. Type 27b (cont'd) some assemblies. 1-131 activity levels at end of cycle 1 indicated 3-4 defective rods. 28a Surry 2 PWR During 1974, I-131 activity level in pri-Cause unknown at this time. Densification 87, 8/74 and mary coolant indicates about I defective induced power spikes observed in all regions 28 1/75, fuel rod. of the core. However, the total number has resp. not appeared to increase in the first 6 months of 1974; one additional spike observed in last half of 1974. 29a Turkey Point 3 PWR Coolant activity increased caused by Unknown. 89, 7/73 and fuel-cladding defects (failure rate very 24 12/74, below) perhaps @.01%).

small, (See item 29b resp.

n i 4 29b Turkey Point 3 PWR During the Fall 1974 refueling,157 Results incomplete. Some sipping of discharged 28 -. 1/75 assemblies were visually (binocular) assemblies may be performed later (about one-inspected. Sipping was not done due to third of core discharged). Some trouble in equipment problems. Some bowing of fuel refueling encountered due to the bowed fuel rods was observed. rods. 30a Turkey Point 4 PWR One fuel assembly dropped during initial Fuel assembly dropped 4 or 5 inches (cable 90 4/73 fuel loading. clamps did not grip cable) while being raised to the vertical position. Skeleton of fuel j assembly replaced before assembly was loaded. 31a Vermont Yankee BWR Because of excessive gaseous release Cause of activity release is believed to be 91 12/72 activity levels, power level was reduced fuel cladding perforations due to internal until mid-January 1973 shutdown. Vendor hydriding of the zirconium. Cause of hydriding indicated possibility that 183 fuel rods is excessive moisture in fuel rod as a result out of 18.302 could fail. (See item 31b of inadequate vacuum outgassing during fuel below) rod fabrication. 31b Vermont Yankee BWR Fifty four of 368 fuel assemblies iden-Cause of failure is thought to be internal 92, Apr. and l fied as ler.kers by sipping. Of the 54, hydride attack of the Zircaloy cladding. In 93 Feb. 73, 51 had perforated and/or defective fuel 8 fuel assemblies which showed no indication resp. rods (an average of 7 rods per assembly). of failure by sipping, examination revealed Three hundred seventy defective fuel that an average of 4 fuel rods per assembly rods in 51 leaking assemblies have were defective. been replaced.

TABLE C-1 (Con +.'d) Item Plant Reactor Approximate Failure Frequency Failure Type Reference No. Date No. Type 31c Vermont Yankee BWR Of 368 fuel assemblies,14 fuel assem-Hydriding noted on 14 fuel assemblies. Tem-65 1/74 blies were hydrided. Of 53 fuel channels, porary neutron-poison curtains believed to 19 had cracks, holes, and worn spots. have vibrated and rubbed against channels An additional 20 leakers were removed in because of rapid water flow. One hundred fuel the Fall of 1973. channels were replaced because of wear holes and cracks; flow holes in channels were plugged. (See similar item 21a). 31d Vermont Yankee BWR Plant was administratively limited to Problem attributed to " faulty cladding". Pro-94 8/74 lower power during 1974 due to excessive bably caused by hydriding, off-gas activity at the steam jet air ej ectors. (See item 31e below) 31e Vermont Yankee BWR During the Fall 1974 refueling, 328 See item 31d above. 28 1/75 assemblies were replaced by the new 8x8 5' design. The remaining 40 (of the Rs improved 7x7 design, including getter, etc.) were wet sipped out-of-core and no defects found. These 40 (having about 1 year of exposure) were reinserted into the core. 32a Yankee (Rowe) PWR Two assenblies removed. Two Zircaloy-clad test assemblies removed in 95 3/69 1966 because of grid and clip failures. Cor-rections were made to later test assemblies. Removal of 4 Zircaloy-clad test fuel assemblies 96 8/69 32b Yankee (Rowe) PWR propeeed because inspection indicates length change of fuel rods greater than expected. 32c Yankee (Rowe) PWR One fuel assembly damaged. Crane operator mispositioned spent fuel assem-97 9/69 bly and damaged fuel assembly and refueling equipment. 32d Yankee (Rowe) PWR Possibility of only one pinhole sized Failure (?) type unknown. Reactor coolant 44 4/72 leak. activity seems to indicate that the pinhole had either sealed itself or that no defect has existed and activity increase represented some uranium contamination on the fuel rod surface.

l l l TABLE C-1 (Cont'd) Item Plant Reactor Approximate Failure Frequency Failure Type Reference No. Type No. Date 32e Yankee (Rowe) PWR One damaged fuel assembly will not be When upper core barrel was lifted, a fuel 98 Dec. and reused; the fuel assembly was not rup-assembly stuck to it and was lifted 8 ft above 99 Nov., tured. core. While trying to reinsert fuel assembly,. 1972 it was dislodged and fell several inches to resp. top of core adjacent to its orginal position. Upper nozzle and upper fuel assembly wrapper sheet were damaged. The fuel ' assembly will 3 not be reused. The hang-up was caused by a 4 small foreign object locking the assembly in place; marks were found on the upper core-support plate. 32f Yankee (Rowe) PWR One fuel assembly slightly damaged. The fuel assembly was slightly damaged while 100 3/73 the upper core barrel was being removed. A new pressurized fuel assembly was used as the replacement. l b 32g Yankee (Rowe) PWR During refueling shutdown starting No failures were noted. The 12 assemblies 101 11/74 5/10/74, 12 Core X predetermined fuel were considered acceptable for continued assemblies were given close surveillance. second and third cycles of operation in Core Some crud, discoloration, and abrasions XI. were noted. 33a Zion 2 PWR During initial fueling, fuel assembly Personnel error. Fuel assembly may have been 102 12/73 B62P was dropped during handling. No set in a flow hole instead of a fuel aligment i visible damage, but the fuel assembly pin. was replaced by another. I

_.=-... .~ t TA8LE C-2 -FUEL FAILURE

SUMMARY

CATEGORIZATION (See Notes 1, 2 and 3 below) 1 1 Category See Item No. in Table C-1 4-Internal Continination: 2c, 2d(e) 3c, 4a(c.d),' 4b (c.e) 4f, 5b, Sc(?), (21 items) lib.14c(d.e)(?),15a(b)(?),16a(b)17a(b),19b, 24a(b),25a.25b(c)(?).25d,31a(b),31c,31d(e). Manufacturing Defects: la,' 21, 3a, 3b(?),' Sa, Sa, 9a,10a,14b. (9 items) { . Mechanical Damage: 4d,10c(d),10f,12a,12c,14b,14e,15d,17a(b), (18 items) 20c ' 21b, 21c, 30a, 31c, 32a(?), 32c, 32e, 32f. Fuel Cladding Interactions: 3c, 3d, 3e(?), Sb Sc(?), 5d, 8f(g)(?),14c(d.e)(?), (11 items) 16c(d), 22f, 24a(b). Accelerated Corrosion: 2a(b),2c'2d(e),21,2j,3a,8f(g)(?). (7 items) Fuel Rod Bowing: Ba,8b,Bc,Bd.8f(g)(?),10b,14e,22g,25e, 1 { (11 items) 27a(b),29a(b). I Cladding Collapse: 10b, 22b(e), 22d. 25d, 25f, 25g(h). I' (6 items) Other: Miscellaneous Design i Deficiences: (5 items) 29,2h(?),20a,21a',32b. Unknown or Unreported: 2f, 2k, 4f, 4g(h), 7a, 8e, 9b(c),10b, 4 (36 items) 10e, lla, lic, lid,12b(d),13a,14f, 4 15a(b),15c,16c(d),16e,17a(b),17c.. j' 17d,18a(b),19c(d),19e,20b,21c, 21d,22a(c),25h,251,26a,27a(b). r 28a,29a(b),32d. F Notes: 1. Items 14a 19a, 23a, 26b, 32g, and 33a are not included above since no failures were evidenced. 2. To avoid duplication, some items have been combined. For example, 24a(b) is considered one item for purposes of this table; 24a noted an increase in gaseous radioactive effluent while 24b denotes the results of fuel inspection for the same Core. 3. Due to lack of specific data. certain of the failures for BWR fuel attributed to hydriding (internal contamination) may have been initially caused by fuel-cladding interactions. I i i 1 C-24 l'

APPENDIX D - FUEL PERFORMANCE REPORTING REQUIREMENTS Table D-1 presents the technical specification reporting requirements pertaining to fuel performance for various reactor plants. The plants are arranged in order of date of first electrical generation, 4 I' i 1 i a t t I ( D-1 1

.. ~. -. -.. - ..~ + 1 h 4 4 -TABLE D-1 TECHNICAL SPECIFICATION REPORTING REQUIREMENTS PERTAINING' TO FUEL PERFORMANCE FOR VARIOUS REACTOR PLANTS 1 Reactor First Plant Type Electrical Reference Technical Requirement ] Generation Appendix A Section 6.2.B.1: Semiannual Operating Report; Paragraph a.(2) requires the report i Indian Point 1 PWR 9/16/62 .to License to include performance characteristics (e.g., equipment and jftsel performance). No. DPR-5 Section 6.3.A.1:. Covers Abnormal Occurrence Reports. 4 Section 6.3.A.2: Unusual Events Report; Paragraph (b) includes "any substantial variance from performance specifications contained in the technical specifications O g, or in the Safety Analysis Report". h Appendix A Section 3.5.2, added by Change No.12 ' dated 8/28/73, covers the reports required. Genoa BWR 4/26/68 to License There is no specific-reference to fuel performance. a No. DPR-45 Section 3.5.2.4 requires a monthly operations report and Paragraph (e).~ which states, " Principal maintenance performed and replacements made in the reactor and associated systems, including a report on various tests performed on components . of the reactor and associated systems " could cover fuel inspections. i. i i l 3

m TABLE D-1 (Cont'd) TECHNICAL SPECIFICATI0fl REPORTING REQUIREMENTS PERTAINING TO FUEL PERFORMANCE FOR VARIOUS REACTOR PLANTS Reactor First Plant Type Electrical Reference Technical Requirement Generation Appendix A Section 6.6.1: Semiannual Operating Report; Paragraph a.1 states the report i Oyster Creek 1 BWR 9/23/69 to License should include a summary of performance characteristics (e.g., equipment and No. DPR-16 fuel performance). Section 6.6.2: Paragraph (a) covers Abnormal Occurrence Reports. I Section 6.6.2.b: Unusual Events; Paragraph 2 requires a report for " discovery ? of any substantial variance from performance specifications contained in the technical specifications or in the Safety Analysis Report." Appendix A Section 15.6.6.A.1: Covers Abnormal Occurrances, s Point Beach 1 PWR 11/6/70 to License Section 15.6.6.A.3.b: Requires a report for "any substantial variation in an No. DPR-24 unsafe direction of a system or component involving nuclear safety from predicted performance presented in the Technical Specifications or Final Facility Descrip-l tion and Safety Analysis Report". l Section 15.6.6.B: Semiannual Operations Report; Paragraph 1.b requires the report to contain a summary of changes in perforvance characteristics (e.g., ~ i i major equipment and fuel performance). i J

TABLE D-1 (Cont'd) TECHNICAL SPECIFICATION REPORTING REQUIREMENTS PERTAINING TO FUEL PERFORMANCE FOR VARIOUS REACTOR PLANTS Reactor First Plant Type Electrical Reference Technical Requirement Generation Point Beach 1 (cont'd) Section 15.6.6.C: Special Reports; Paragraph 3 requires a " limited program of non-destructive fuel inspections...of the two lead burnup assemblies during the second and third refueling shutdowns". This extra requirement was to provide the Commission with added verification of the safety and reliability of pre-pressurized Zircaloy-clad nuclear fuel. j' Appendix A Section 6.6.1.d: Covers Abnormal Occurrences. Palisades PWR 12/31/71 to License Section 6.6.5: Semiannual Operating Report; Paragraph (a) includes "a narrative No. DPR-20 summary of operating experience and of changes in facility design, performance characteristics and operating procedures related to safety occurring during the i reporting period". Section 6.6.6: Special Reports; Paragraph 4 requires "a comprehensive report presenting the results of the initial preoperational, startup, power ascension and rated-power test programs (including operating record of in-core and instru-mentation) shall be submitted within one year of the commercial service date". l 1

TABLE D-1 (Cont'd) TECHNICAL SPECIFICATION REPORTING REQUIREMENTS PERTAINING TO FUEL PERFORMANCE FOR VARIOUS REACTOR PLANTS Reactor First Plant Type Electrical Reference Technical Requirement Generation Appendix A Section 6.6.A.2: First Year Operation Report; Paragraph (c) requires "an assess-Quad Cities 1 BWR 4/12/72 to License ment of the perfor1 nance of structures, systems and components important to safety" No. DPR-29 Section 6.6.A.3: Semiannual Operating Reports; Paragraph a.(2) states the report should include " performance characteristics (e.g., equipment and fuel l performance)." l E Section 6.6.B.1: Covers Abnormal Occurrence Reports. Section 6.6.B.2: Unusual Event Reports; Paragraph 6. covers " discovery of any substantial variance from performance specifications contained in the technical specifications or in the Safety Analysis Report." Section 6.6.B.3: Special Reports; Table 6.6.1, item (c) requires a " summary status of fuel performance after each refueling outage starting with (the) second refueling outage". Additional detail as to what to include is provided in Section 1.1, since this plant's fuel operating conditions reflected linear power generation rates and exposures higher than experienced previously in BWR plants.

TABLE D-1 (Cont'd) TEt9NICAL SPECIFICATION REPORTING REQUIREMENTS PERTAINING TO FUEL PERFORMANCE FOR VARIOUS REACTOR PLANTS ~~~ Reactor First Plant Type Electrical Reference Technical Requirement Generation Appendix A Section 6.6.A: Semiannual Operations Report; Paragraph 1.b. requires the report Surry 1 PWR 7/4/72 to License to include " performance characteristics (e.g., equipment and fuel performance)". No. DPR-32 Section 6.6.8.1: Covers Abnormal Occurrence Reports. Section 6.6.B.2: Unusual Safety Related Events; a report is required for "any )( substantial variance, in an unsafe or less conservative direction, from per-formance specifications contained in the Technical Specifications or from per-formance specifications, relevant to safety related equipment, contained in the Final Safety Analysis Report" l Section 6.6.D.b: First Year Operation Report; Paragraph (3) requires "an assess-ment of the performance of structures, systems and components important to safety" Appendix A Section 6.6.1.2: First Year Operation Report; Paragraph d. includes "an assess-Oconee 1 PWR 5/6/73 to License ment of the performance of structures, systems, and components important to No. DPR-38 safety". Section 6.6.1.3: Semiannual Operating Report; Paragraph a.(2) includes "Perfor-mance characteristics (e.g., equipment and fuel performance)".

. - - - _ - - _ - _ _ -. ~. _ _ 4 1 2 'f TABLE D-1 (Cont'd) TECHNICAL SPECIFICATION REPORTING REQUIREMENTS PERTAINING i TO FUEL PERFORMANCE FOR VARIOUS REACTOR PLANTS l Reactor First Plant Type Electrical Reference Technical Requirement Generation Oconee 1 (cont'd) Section 6.6.2.1: Covers Abnomal Occurrences or Unusual Events; for the latter, this would be "any substantial variance from performance specifications contained } in the Technical Specifications or the Safety Analysis Report". l Section 6.6.3.6: Fuel Surveillance Program Report; the program provides for examination of fuel rods at the end of the first, second,' and third cycles of { this reactor to determine if fuel rods have maintained their integrity and to l determine the extent, if any, of dimensional changes in diameter and length. l Both visual and dimensional examinations are specified. This program was to-i substantiate the fuel perfomance for this generation of PWR reactors. l Appendix A Section 5.12.1.b: First Year Operation Report; Paragraph 3 requires "an l Fort Calhoun PWR 8/25/73 to License assessment of the performance of structures, systems, and components important I No. DPR-40 to safety". Section 5.12.1.c: Semiannual Operating Report; Paragraph 1.(b) requires the report to include a summary of " performance characteristics (i.e., equipment and fuel performance)". l

TABLE D-1 (Cont'd) TECHNICAL SPECIFICATION REPORTING REQUIREMENTS PERTAINING TO FUEL PERFORMANCE FOR VARIOUS REACTOR PLANTS Reactor First Plant Type Electrical Reference Technical Requirement Generation Fort Calhoun (cont'd) Section 5.12.2.a: Covers Ab,ormal Occurrence Reports. Section 5.12.2.b: Unusual Events; Paragraph 2 requires a report for the " discover of any substantial variance from performance specifications contained in the technical specifications or in the Safety Analysis Report". Section 5.12.3: Special Reports; Paragraph (1) provides an example of a subject [ for such a report, namely, " Fuel performance following each refueling er partial refuelino" Appendix A Section 1.8: Abnormal Occurrences; similar to requirements of Reg. Guide 1.16. Rancho Seco PWR 10/13/74 to License Section 6.12.1: Semiannual Operating Reports; Paragraph A.2 requires a summary No. DPR-54 of " Performance Characterisites (e.g., equipment and fuel performance) directly related to nuclear safety". Section 6.12.2: Non-Routine Reports; Paragraph B.2 covers "any substantial variance from performance specifications contained in the Technical Specifica-tions or in the Final Safety Analysis Report".

TABLE D-1 (Cont'd) TECHNICAL SPECIFICATION REPORTING REQUIREMENTS PERTAINING TO FUEL PERFORMANCE FOR VARIOUS REACTOR PLANTS Reactor First Plant Type Electrical Reference Technical Requirement Generation i Rancho Seco (cont'd) Section 6.12.3: Special Reports; Paragraph B (First Year Operation Report) l i requires the report to cover "an assessment of the performance of structures, -- j systems and components important to safety". Appendix A Section 6.0.H.I.a(2): First Year Operation Report; Paragraph (c) includes "an Edwin I. Hatch 1 BWR 11/11/74 to License assessment of the perfomance of structures, systems, and components important ? No. DPR-57 to safety". Section 6.0.H.I.a(3): Semiannual Operating Reports; Paragraph (a)(ii) requires the report to include " performance characteristics (e.g., equipment and fuel performance)". Section 6.0.H.2.a: Covers Abnonnal Occurrence Reports. Section 6.0.H.2.b: Unusual Event Reports; a report is required for "any sub-stantial variance in an unsafe or less conservative direction from perfonnance specifications contained in these Technical Specifications or from performance specifications relevant to safety related equipment contained in the FSAR". Appendix A Section 6.7.A.2: First Year Operation Report; Paragraph'(c) requires an assess-Prairie Island 2 PWR 12/21/74 to License ment of the perfomance of structures, systems and components important' to safety. No. DPR-60

l o-

...-._..___m _m .__4 m - j. i, I TABLE D-1 (Cont'd) TECHNICAL SPECIFICATION REPORTING REQUIREMENTS PERTAINING 1 1 TO FUEL PERF0lMANCE FOR VARIOUS REACTOR PLANTS i j Reactor First 1 Plant Type Electrical Reference Technical Requirement l Generation i I j Prairie Island 2 (cont'd) Section 6.7.A.3: Semiannual Operating Reports; Paragraph a.(2) states the report 1 1 j (Sameas should include a summary of " performance characteristics (e.g., equipment and j ) Appendix A fuel performance)". i I to License Section 6.7.B.1: Covers Abnormal Occurrence reports. I No. DPR-42 Section 6.7.B.3: Special Reports; Item 6 requires a " fuel surveillance report i ? j g for Prairie within 3 months after first refueling outage of Unit 1. Surveillance will be I i Island 1)

nade of the fuel assemblies to determine effects caused by phenomena such as i

4 fuel densification and shall include visual observation of at least one high ' l; power density fuel assembly". 1 Appendix A Section 6.12.1.b: First Year Operation Report; Paragraph 3 requires "an assess-- Calvert Cliffs 1 PWR 12/30/74 to License ment of the performance of structures, systems, and components important to a No. DPR-53 safety". j Section 6.12.1.c: Semiannual Operating Reports; Paragraph 1.(b) requires that 2 these reports include " Performance characteristics (e.g., equipment and fuel J performance)". i i

f TABLE D-1 (Cont'd) TECHNICAL SPECIFICATION REPORTING REQUIREMENTS PERTAINING TO FUEL PERFONUICE FOR VARIOUS REACTOR PLANTS Reactor First Plant Type Electrical Reference Technical Requirement C::: ration Calvert C1tffs 1 Section 6.12.2: Nonroutine Reports; Paragraph (a) covers Abnormal Occurrences. (Cont'd) Paragraph (b), unusual Events, subparagraph 2 covers " discovery of any substan-tial variance from performance specifications, contained in these Technical Specifications or in the Safety Analysis Report". Section 6.12.3: Special Reports; this section requires a special report covering inspections, tests, and maintenance that ar= appropriate to assure safe opera-tion. An example of a subject for such a report is Paragraph (c) " Fuel Performance". .... _ _.....}}