ML003673316

Detailed Summary of Facts, Data and Arguments and Sworn Submission on Which Orange County Intends to Rely at Oral Argument to Demonstrate the Existence of a Genuine and Substantial Dispute of Fact with the Licensee Regarding the Proposed Ex

ML003673316
Person / Time
Site:	Harris
Issue date:	01/04/2000
From:	Curran D, Thompson G Harmon, Curran, Harmon, Curran, Spielberg & Eisenberg, LLP, Institute for Resource & Security Studies, Orange County, NC, Board of Commissioners
To:	Office of Nuclear Reactor Regulation
References
#100 21114, +adjud/rulemjr200506, -RFPFR, 99-762-02-LA, LA
Download: ML003673316 (242)

Text

January 4, 2000 UNITED STATES OF AMERICA 'C!0 J, ,, :41 NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of )

)

CAROLINA POWER & LIGHT ) Docket No. 50-400 -LA (Shearon Harris Nuclear ) ASLBP No. 99-762-02-LA Power Plant) )

DETAILED

SUMMARY

OF FACTS, DATA AND ARGUMENTS AND SWORN SUBMISSION ON WHICH ORANGE COUNTY INTENDS TO RELY AT ORAL ARGUMENT TO DEMONSTRATE THE EXISTENCE OF A GENUINE AND SUBSTANTIAL DISPUTE OF FACT WITH THE LICENSEE REGARDING THE PROPOSED EXPANSION OF SPENT FUEL STORAGE CAPACITY AT THE HARRIS NUCLEAR POWER PLANT WITH RESPECT TO CRITICALITY PREVENTION ISSUES (CONTENTION TC-2)

Submitted by:

Diane Curran HARMON, CURRAN, SPIELBERG, & EISENBERG, L.L.P 1726 M Street N.W., Suite 600 Washington, D.C. 20036 202/328-3500 Counsel for Orange County Gordon Thompson, Ph.D.

Executive Director INSTITUTE FOR RESOURCE AND SECURITY STUDIES 27 Ellsworth Avenue Cambridge, MA 02139 Expert witness for Orange County January 4, 2000 Iz 'r,)50-3

TABLE OF CONTENTS

1. IN TRO D U C TIO N .................................................................................................. 1 II. STATEM ENT OF THE CASE ........................................................................ 2 III. FACTUAL AND PROCEDURAL BACKGROUND ...................................... 4 A. History of Criticality Prevention at Nuclear Power Plants .................... 4

1. Nature of Criticality Accidents ................................................. 4

2. Regulations and agency guidance ............................................... 5

3. Evolution of Criticality Prevention in Fuel Pools ...................... 9

a. Low -density storage ........................................................ 9

b. Reliance on the neutron-absorbing properties of storage racks and the incorporation of flux traps ...................... 10

c. Incorporation of boron in the structure of storage racks .... 11

d. Ongoing administrative measures ................................. 12

e. Independent Spent Fuel Storage Installations ............... 13 B. The Harris License Amendment Application ...................................... 13 C. Orange County's Intervention in Licensing Proceeding ....................... 16 A RG UM EN T..................................................................................................................... 18 IV. THE PROPOSED LICENSE AMENDMENT FAILS TO COMPLY WITH GDC 62 BECAUSE IT IMPROPERLY RELIES ON ADMINISTRATIVE MEASURES FOR CRITICALITY PREVENTION ...................................... 18 A. The General Design Criteria Establish Minimum Design Requirements for N uclear Pow er Plants ........................................................................... 19 B. The Plain Language of GDC 62 Requires the Use of Physical Systems or Processes to Prevent Criticality, and Thereby Precludes the Use of A dm inistrative Controls ...................................................................... 20

1. The plain language of GDC 62 requires the use of physical systems or processes to prevent criticality .......................................... 20

ii

2. Physical systems and processes are distinct in nature from ongoing administrative controls ................................... 21 C. The Rulemaking History of GDC 62 Supports the Plain Language of the Regulation ...................................................................... 24

1. Pre-rulemaking documents ................................................. 24

2. Proposed GDC for criticality control ................................. 25

3. Comments on the proposed rule ........................................ 26

4. The Final Rule ..................................................................... 27 .

D. The Plain Language of GDC 62 Is Not Altered or Contradicted By Other Relevant NRC Criticality Standards .............................. 28

1. 10 C.F.R. §§ 70.24 and 50.68 ...................................... 28

2. 10 C.F.R . § 72.124 ..................................................... 33 E. The Administrative Criticality Prevention Proposed by CP&L W ould Violate GDC 62 .......................................................... 37 F. CP&L's Proposed Reliance on Administrative Criticality Prevention Measures Is Not Justified by Draft Reg. Guide 1.13 or Other NRC Staff Guidance .................................................... 38 G. Neither CP&L Nor the Staff Has Demonstrated That Public Health And Safety Will Be Adequately Protected If CP&L Relies on Ongoing Administrative Measures for Criticality Control ..... 39 H. CP&L's Criticality Accident Analysis Misapplies Applicable Staff Guidance ................................................................................... 41

1. CP&L ignores the words "at least," and evaluates only one failure instead of sets of failures ................................. 44

2. CP&L fails to determine what failures are "unlikely, independent, and concurrent...................."................... 44

3. CP&L assumes that mispositioning of fuel is an "unlikely" event when in fact it is likely .......... 45

iii

4. CP&L unreasonably assumes that a single error can lead to the mispositioning of only one fuel .......... 46 assembly.

V. CO N CLU SION ............................................................................................... 47

January 4, 2000 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of )

)

CAROLINA POWER & LIGHT ) Docket No. 50-400 -LA (Shearon Harris Nuclear ) ASLBP No. 99-762-02-LA Power Plant) )

DETAILED

SUMMARY

OF FACTS, DATA AND ARGUMENTS AND SWORN SUBMISSION ON WHICH ORANGE COUNTY INTENDS TO RELY AT ORAL ARGUMENT TO DEMONSTRATE THE EXISTENCE OF A GENUINE AND SUBSTANTIAL DISPUTE OF FACT WITH THE LICENSEE REGARDING THE PROPOSED EXPANSION OF SPENT FUEL STORAGE CAPACITY AT THE HARRIS NUCLEAR POWER PLANT WITH RESPECT TO CRITICALITY PREVENTION ISSUES (CONTENTION TC-2)

I. INTRODUCTION Pursuant to 10 C.F.R. § 2.113, Orange County hereby submits a detailed written summary and sworn submission (hereinafter "Summary") of all the facts, data, and arguments which are known to the County and on which the County proposes to rely at the January 21, 2000, oral argument. This Summary presents Orange County's legal and factual grounds for asserting that Carolina Power & Light's ("CP&L's") application to amend its Operating License by expanding the capacity of spent fuel pool storage pools at the Harris nuclear power plant fails to satisfy the criticality prevention requirements of General Design Criterion ("GDC") 62 and applicable NRC guidance, and fails to provide adequate protection of public health and safety to

2 members of the public living in the vicinity of the Harris plant.'

As required by 10 C.F.R. § 2.111 (b), the factual assertions in this Summary are submitted under the sworn declaration of Dr. Gordon Thompson, the County's expert witness regarding criticality prevention issues. A further declaration of Dr. Thompson's qualifications and experience and a description of his work on this Summary is attached as Exhibit 1.

As detailed below, this summary demonstrates that as a matter of law, CP&L's License Amendment Application must be rejected because it places impermissible reliance on administrative procedures and controls for criticality prevention, rather than relying entirely on physical systems and processes, as required by the regulations. If the Board does not find that the issue can be disposed of clearly as a matter of law, the County submits that it has submitted substantial evidence that there is a genuine and substantial factual dispute between CP&L and the County regarding whether the criticality prevention measures it has elected are acceptable under GDC 62 and applicable portions of the NRC Staff's regulatory guidance, and whether there is any basis for finding that the public health and safety can be adequately protected by CP&L's proposed criticality prevention measures.

II. STATEMENT OF THE CASE This case raises questions about the proper interpretation of GDC 62, which requires that criticality in the fuel storage and handling system of a nuclear power plant must be prevented by "physical systems and processes, preferably by use of geometrically safe configurations." This regulation clearly precludes the use of such administrative controls and procedures as control of burnup/enrichment levels and reliance on the presence of soluble boron in fuel pools. Although 1 See Letter from James Scarola, CP&L, to NRC, re: Shearon Harris Nuclear Power Plant, Docket No. 50-400/License No. NPF-63, Request for License Amendment, Spent Fuel Storage

3 the NRC Staff's current regulatory guidance countenances the use of such administrative controls, it must be disregarded in this respect because it is fundamentally inconsistent with the controlling regulation, GDC 62.

The NRC Staff's various guidance documents related to criticality prevention do contain some provisions which are consistent with GDC 62 and which provide assistance in determining whether the physical criticality prevention measures that are designed to prevent criticality in normal operation will also suffice to protect public health and safety under accident conditions.

In order to evaluate the effectiveness of criticality prevention in an accident, it is necessary to perform a criticality analysis that encompasses possible accident scenarios and evaluates the efficacy of criticality prevention measures during each scenario. A useful tool for such an analysis is the Double Contingency Principle, which is set forth in Draft Regulatory Guide 1.13, a document employed by the Staff for evaluating criticality analyses. This version of the Double Contingency Principle requires that a criticality analysis must demonstrate that criticality could not occur without at least two unlikely, independent and concurrent failures or operating limit violations. In order to make a meaningful application of the Double Contingency Principle, it is necessary to identify what are possible sets of unlikely, independent and concurrent failures or operating limit violations, and then evaluate those events in combination to determine whether the facility's criticality prevention arrangements will maintain subcriticality during each set of events. Draft Reg. Guide 1.13 also advises that in evaluating such accident scenarios, it may be assumed that initial conditions are in the normal range. However, the deterioration of those conditions in the course of each accident scenario must also be examined. In this case, CP&L has neither complied with GDC 62, nor has it made a reasonable application of the Double (December 23, 1998), (hereinafter "License Amendment Application).

4 Contingency Principle. CP&L proposes to rely for criticality prevention on the control of burnup/enrichment levels, which necessarily entails ongoing administrative procedures and controls. These procedures are not only prohibited by GDC 62, but they are inherently less reliable than physical systems and processes. CP&L has also misapplied the Double Contingency Principle, by failing to identify and evaluate the sets of unlikely, independent, and concurrent failures that could lead to a criticality accident. Instead, CP&L has addressed only one scenario in which criticality is approached: the mispositioning of a single fresh PWR fuel assembly in pool C or D.

Because it has made no attempt to identify and evaluate the sets of events that could lead to a criticality accident, CP&L has no basis for asserting that its analysis of a single event is conservative. Moreover, experience at operating nuclear power plants shows that a single error can lead to the mispositioning of multiple fuel assemblies, and that mispositioning of this kind is a likely event. Given the potential for mispositioning of multiple assemblies, CP&L's and the NRC Staff's own criticality calculations show that the spent fuel in pools C and D could become supercritical.

Accordingly, because it violates GDC 62 and misapplies the valid portions of applicable NRC Staff guidance for the conduct of criticality accident analyses, CP&L's License Amendment Application must be rejected.

III. FACTUAL AND PROCEDURAL BACKGROUND A. History of Criticality Prevention at Nuclear Power Plants

1. Nature of Criticality Accidents In operating a nuclear power plant, it is necessary to protect the facility against a

5 criticality accident. Criticality occurs when neutrons emanating from atoms of special nuclear material, as a result of fission of their nucleii, bombard other atoms and cause fission of their nucleii, setting off a chain reaction. Criticality can be prevented by providing adequate spacing of special nuclear material, and by introducing neutron-absorbing material to shield the special nuclear material and absorb the neutrons.

A nuclear fission reactor generates power because criticality is achieved under controlled conditions. At all times when fresh or spent fuel is outside a reactor, criticality must be prevented. In the case of light-water reactor fuel, a criticality accident can occur if fresh or spent fuel assemblies are brought sufficiently close together in the presence of a neutron-moderating material such as water, without the presence of sufficient neutron-absorbing material to suppress criticality. The neutron-absorbing material could be solid boron or other material incorporated into the structure of the racks where fuel assemblies are stored, or soluble boron in the water surrounding fuel assemblies.

2. Regulations and agency guidance Criticality control at nuclear power plants is governed by General Design Criterion

("GDC") 62, which requires that:

Criticality in the fuel handling and storage system shall be prevented by physical systems or processes, preferably by use of geometrically safe configurations.

10 C.F.R. Part 50, Appendix A, Criterion 62. This language clearly precludes the use of ongoing procedural or administrative measures for criticality prevention. The NRC also has regulations at 10 C.F.R. § 70.24 and § 50.68, which permit licensees to forego criticality monitors if they comply with certain measures for criticality prevention. As discussed in more detail in Section 2 For a more complete discussion of the language and history of GDC 62, see Section IV.

6 IV.D. below, these measures are consistent with GDC 62, and the Commission reaffirmed GDC 62 when it promulgated the regulations.

GDC 62 sets forth unequivocal requirements for the prevention of criticality under normal conditions. However, one can postulate accident conditions that would defeat these requirements. For example, a sufficiently severe mechanical loading could reduce the center-to center distance between fuel assemblies and thereby cause criticality, even though the configuration was geometrically safe before the loading was applied.

In 1978, the NRC Staff issued a guidance document which sought to extend the requirements of GDC 62 into the realm of accident conditions, by introducing the "Double Contingency Principle" and the concept of "realistic initial conditions."' The guidance is attached to a letter from Brian K. Grimes of the NRC Staff to "All Power Reactor Licensees,"

dated April 14, 1978 (hereinafter "Grimes Letter").4 The Grimes letter acknowledges that "[d]ue to an increased demand on storage space for spent fuel assemblies, the more recent approach is to use high density storage racks and to better utilize available storage space."' The Letter provides the following guidance for evaluation of criticality prevention under postulated accident conditions:

The double contingency principle of ANSI N 16.1-19754 shall be applied. It shall require two unlikely, independent, concurrent events to produce a criticality accident.

Realistic initial conditions (e.g., the presence of soluble boron) may be assumed for the fuel pool and fuel assemblies.6 below.

3 See Appendix A to this Summary for a further discussion of the source and development of these terms.

4 A copy of the Grimes Letter is attached as Exhibit 2.

5 Id., Enclosure 1 at 1-1.

6 Id.

7 As discussed in Appendix A, these terms are not further discussed or defined in the Grimes Letter. However, it is clear that the Grimes Letter did not allow reliance on the presence of soluble boron as a criticality prevention measure under normal conditions. Instead, the presence of soluble boron was intended to be considered solely as an initial condition in an accident scenario.

In 1981, the Staff issued a draft regulatory guide containing further guidance for the evaluation of criticality prevention measures: Draft 1, Regulatory Guide 1.13, Revision 2, "Spent Fuel Storage Facility Design Basis (December 1981) (hereinafter "Draft Reg. Guide 1.13")7. Although Draft Reg. Guide 1.13 has never been issued in final form, the Staff has applied it extensively to the review of spent fuel pool expansion applications. Like the 1978 Grimes Letter, this Draft Reg. Guide has never been approved by the Commission, but is solely a Staff guidance document.

In §§ 4.5 and 6 of Appendix A, Draft Reg. Guide 1.13 implies that credit may be taken for fuel bumup as a criticality prevention measure under normal conditions. Section 5.2 of Appendix A states that the presence of soluble boron can be regarded as a realistic initial condition under certain accident conditions, namely those associated with "Condition IV faults,"

which are not defined in the Draft Reg. Guide. As in the case of the Grimes Letter, it is clear that this Draft Reg. Guide does not allow reliance on the presence of soluble boron as a criticality prevention measure under normal conditions.8 Draft Reg. Guide 1.13 also calls for the application of the Double Contingency Principle, articulating the principle as follows:

7 A copy of Draft Reg. Guide 1.13 is attached as Exhibit 3.

8 As discussed in Attachment A to this Summary, the American Nuclear Society ("ANS") also provides guidance regarding the presence of soluble boron as an initial condition for the purposes of criticality analysis pertinent to accident conditions.

8 At all locations in the LWR spent fuel storage facility where spent fuel is handled or stored, the nuclear criticality safety analysis should demonstrate that criticality could no.

occur without at least two unlikely, independent and concurrent failures or operating limit violations.

Appendix A, § 1.4 (emphasis in original). The Draft Reg. Guide's version of the Double Contingency Principle is broadly consistent with the language of the Grimes Letter, although there are two notable differences, the first of which strengthens the standard significantly. First,

§ 1.4 specifies "at least two" criticality-inducing events, whereas the Grimes letter specifies "two" events. Second, § 1.4 refers to "failures or operating limit violations," while the Grimes Letter refers to "events."

A more recent guidance document on criticality prevention in spent fuel storage pools is a Memorandum from Laurence Kopp, NRC, to Timothy Collins, NRC, re: Guidance On The Regulatory Requirements For Criticality Analysis Of Fuel Storage At Light-Water Reactor Power Plants (August 19, 1998) (hereinafter "Kopp Memorandum").9 The Kopp Memorandum asserts the Staff s acceptance of various administrative measures for criticality prevention, such as credit for bumup and soluble boron. It also re-states, in substantially weakened form, the Double Contingency Principle:

The criticality safety analysis should consider all credible incidents and postulated accidents. However, by virtue of the double-contingency principle, two unlikely independent and concurrent incidents or postulated accidents are beyond the scope of the required analysis. The double-contingency principle means that a realistic condition may be assumed for the criticality analysis in calculating the effects of incidents or postulated accidents. For example, if soluble boron is normally present in the spent fuel pool water, the loss of soluble boron is considered as one accident condition and a second concurrent accident need not be assumed. Therefore, credit for the presence of the soluble boron may be assumed in evaluating other accident conditions."° The Kopp Memorandum thus effectively reduces the double contingency principle to a "single 9 A copy of the Kopp Memorandum is attached as Exhibit 4.

9 contingency principle.""

Thus, as the pressure has increased for higher and higher density fuel storage, the NRC Staff has increasingly relaxed the standards for criticality prevention, allowing the use of administrative measures and reducing the rigor of the accident analysis required.

3. Evolution of Criticality Prevention in Fuel Pools There is no centralized, publicly accessible database that provides detailed information about the rack configuration at each nuclear power plant spent fuel storage pool and the history of rack installation at each pool. Nevertheless, a survey of correspondence and safety reports for individual plants shows how measures for criticality prevention at nuclear power plants have evolved over time in response to increasing demand for higher and higher density spent fuel storage. This evolution has gone beyond the bounds of measures that are consistent with GDC

62. The NRC Staff has condoned violations of GDC 62 by issuing regulatory guidance that countenances these violations, and by approving many license amendment applications that permit the use of administrative measures for criticality prevention in the high-density storage of spent fuel.

a. Low-density storage When US nuclear power plants of the present generation were designed, and when many of the currently operating plants were commissioned, fuel pools were equipped with low-density fuel storage racks. The racks were designed with open frames of steel or aluminum. Center center distances between fuel assemblies were typically 10-13 inches in BWR racks and 18-22 inches in PWR racks. By using a relatively low fuel storage density -- less than 0.25 tonne U 10 Id., Attachment 4.

11 A more detailed discussion of the Kopp Memorandum appears in Appendix A to this

10 per square foot -- licensees achieved a high level of safety against criticality. The center-center distances were large enough to prevent criticality even if fresh fuel was placed in the racks and the pool was filled with unborated water. In other words, criticality prevention relied entirely on the use of a geometrically safe configuration.

As spent fuel began to accumulate at power plants, there was growing interest in achieving higher storage densities in fuel pools. This implied smaller center-center distances in the racks, resulting in a greater propensity for criticality. Beginning in the 1970s and continuing through the 1980s and 1990s, center-center distances in fuel pools were reduced in several steps.

Additional means of criticality prevention were introduced at each step."2

b. Reliance on the neutron-absorbing properties of storage racks and the incorporation of flux traps The first step toward higher density was to employ stainless steel racks with center-center distances of about 8 inches in BWR racks and 13 inches in PWR racks. Roughly speaking, this step occurred in the 1970s. The new configuration increased the fuel storage density to a level of up to 0.39 tonne U per square foot. The reduced center-center distances in this configuration yielded a greater propensity for criticality than was exhibited by the previous open-frame racks.

Nevertheless, the rack designers were able to achieve a subcritical margin of reactivity, relying in part on the absorption of slow neutrons by the stainless steel in the rack structures. This neutron absorption phenomenon was in turn assisted by the moderation of fast neutrons by water confined in passages ("flux traps") between the fuel assemblies. At this stage of evolution in fuel storage density, criticality prevention relied partly on the distance between fuel assemblies and Summary.

12 See US Department of Energy, Spent Fuel Storage Fact Book, DOE/NE-0005, April 1980; and USNRC, Draft Generic Environmental Impact Statement on Handling and Storage of Spent

11 partly on the neutron-absorbing properties of the racks.

c. Incorporation of boron in the structure of storage racks The second step toward higher density in fuel pools was to employ stainless steel racks which incorporated boron in solid form within the rack structures. Roughly speaking, this step occurred in the 1980s. Boron is an absorber of neutrons, and thereby suppresses criticality.

Thus, the incorporation of solid boron allowed center-center distances to be further reduced. A common method of incorporating solid boron is to attach Boral panels to the racks. To construct a Boral panel, boron carbide is dispersed in aluminum, and this material is fabricated into sheets which are clad with aluminum. These "panels" are then attached to the spent fuel storage racks.

Incorporation of solid boron within the rack structures allowed a subcritical margin of reactivity to be maintained while center-center distances were reduced to 6.5 inches in BWR racks and 10.5 inches in PWR racks, thereby achieving a fuel storage density up to 0.58 tonne U per square foot. In this configuration, criticality prevention relied to a lesser degree than previously on the distance between fuel assemblies and to a greater degree on the neutron absorbing properties of the racks. "3Most, perhaps all, fuel pools at US nuclear plants have been Light Water Power Reactor Fuel, NUREG-0404 (2 volumes) Appendices B and D (March 1978).

13 In pursuit of even higher storage densities in fuel pools, the nuclear industry has also studied fuel storage options involving a reduced presence of water between the fuel rods. Water moderates fast neutrons, so a reduced presence of water can yield a subcritical margin of reactivity even as the spacing between fuel assemblies or rods is reduced. One water-displacing option is to place spent fuel assemblies inside cans and to fill all empty space inside each can with small metal beads, thereby achieving a fuel storage density of 0.75 tonne U per square foot.

A second option is to compact fuel assemblies by crushing the fuel spacers until rods are nearly touching, thus achieving a fuel storage density of about 0.95 tonne U per square foot. A third option is to dismantle the fuel assemblies and store the rods in close contact with each other inside cans, thus achieving a fuel storage density of about 1.1 tonne U per square foot. None of these options has been generally adopted. See US Department of Energy, Spent Fuel Storage Fact Book, DOE/NE-0005 (April 1980).

12 equipped for some years with racks that incorporate solid boron within the rack structures, often in the form of Boral panels.

d. Ongoing administrative measures In recent years, a number of licensees have further increased the density of spent fuel pool rack storage. As the fuel is packed closer and closer together, fixed neutron-absorbing material such as Boral panels becomes less and less effective in preventing criticality. Therefore, licensees have introduced ongoing administrative procedures for criticality prevention. These measures consist of (a) relying on the presence of soluble boron into the spent fuel pool water, (b) controlling the burnup level of the fuel, and (c) controlling the age of the fuel assemblies.

Using these ongoing administrative methods, the density of storage of intact fuel assemblies in a fuel pools has been increased beyond the level that was achieved by adopting center-center distances of 6.5 inches in BWR racks and 10.5 inches in PWR racks.

These three methods exploit phenomena as follows. First, increased burnup of a fuel assembly will, over a broad range of conditions, decrease the assembly's reactivity because of the ingrowth of neutron-absorbing isotopes and the reduced enrichment in U-235 that occur with increased burnup.' 4 Second, the presence of soluble boron in the pool water will decrease reactivity because the soluble boron absorbs neutrons. Third, aging of a fuel assembly will decrease the assembly's reactivity due to the decay of Pu-241 (with a 14-year half-life) and the ingrowth of its decay product Am-241.

14 Bumup is the accumulated fission energy released by a fuel assembly. Its effects on criticality are exploited by restricting the combined burnup/enrichment parameters of fuel assemblies that are placed in the fuel storage racks. Note that in some instances, the reactivity of a fuel assembly will initially increase with bumup, then decrease with higher levels of burnup.

13

e. Independent Spent Fuel Storage Installations There is an alternative to adopting ever-higher densities of fuel storage in an existing fuel pool. That alternative is to construct an independent spent fuel storage installation ("ISFSI").

ISFSI's have been built at several US nuclear plant sites. In each case, a dry storage technology has been employed. As of September 1998, installations of this kind were licensed at 11 nuclear plant sites.'"

B. The Harris License Amendment Application There are four spent fuel storage pools at Carolina Power & Light Company's

("CP&L's") Harris nuclear power plant. Only two of the pools, designated "A" and "B," are currently in operation. At present, pool A contains 6 PWR racks with a total of 360 spaces, and 3 BWR racks with a total of 363 spaces. Pool B contains 12 PWR racks with a total of 768 spaces and 17 BWR racks with a total of 2,057 spaces. Under the present license, one additional BWR rack with a total of 121 spaces could be placed in pool B.

CP&L now seeks a license amendment to activate pools "C" and "D.', 6 The purpose of the license amendment is to allow CP&L to use the Harris facility to store spent fuel generated at CP&L's one-unit Harris PWR station, its two-unit Brunswick BWR station, and its one-unit Robinson PWR station. If granted, the license amendment would allow the placement in pool C of up to 11 PWR racks with a total of 927 spaces and 19 BWR racks with a total of 2,763 spaces; 15 See NRC Information Digest: 1998 Edition, NUREG- 1350, Volume 10, Appendix H (November 1998).

16 CP&L's proposed changes to its Technical Specifications are described in Enclosure 5 to the License Amendment Application. Enclosure 7 is a non-proprietary report entitled "Licensing Report for Expanding Storage Capacity in Harris Spent Fuel Pools 'C' and 'D' (Rev. 2). By letter dated March 17, 1999, CP&L submitted Rev. 3 to Enclosure 7, which reflects the release of some information that previously had been considered proprietary. Aside from the additional disclosures, the content of Rev. 3 is the same as Rev. 2.

14 and the placement in pool D of 12 PWR racks with a total of 1,025 spaces. CP&L envisions this placement occurring in three campaigns in pool C, followed by two campaigns in pool D.

For all four spent fuel pools at Harris, CP&L intends to ensure that Keffective will be less than or equal to 0.95 when the racks are flooded with unborated water, including an allowance for uncertainties"7 The proposed means for achieving this objective for pools C and D are different from the means for preventing criticality in pools A and B, however. For pools A and B, a subcritical margin of reactivity is now achieved during normal operation in two ways:

through the rack's neutron-absorbing properties, which are enhanced by incorporating solid boron in the rack structures; and by maintaining a nominal 10.5 inch center-center distance in the PWR racks and a nominal 6.25 inch center-center distance in the BWR racks. These conditions will continue to apply in pools A and B after pools C and D are activated.

For pools C and D, CP&L proposes to space the PWR spent fuel assemblies significantly closer together than they are placed in pools A and B. A nominal 9.017 inch center-center distance will be maintained in the PWR racks, while a nominal 6.25 inch center-center distance will be maintained in the BWR racks. The PWR rack spacing is close to the smallest distance that is physically possible for intact PWR fuel, because the PWR fuel assemblies used in the Harris reactor have a square cross-section that is 8.43 inches wide.' 8 For this configuration, the distance between the fuel assemblies and the neutron-absorbing properties of the racks, taken together, will not be sufficient to maintain the desired subcritical margin of reactivity under normal conditions, still less under accident conditions. Therefore, CP&L proposes to introduce an additional means of criticality suppression for PWR fuel in pools C and D.

17 Keffective is the neutron multiplication factor in a finite array of fuel, allowing for neutron leakage.

15 CP&L proposes to introduce new, ongoing administrative measures that would limit the combination of burnup and enrichment of the PWR spent fuel in pools C and D to an "acceptable range." The range of acceptable bumup/enrichment values is shown in Figure 5.6.1 of the proposed technical specifications, Enclosure 5 to the License Amendment Application.

According to CP&L: "The burnup criteria will be implemented by appropriate administrative procedures to ensure verified burnup as specified in the proposed Regulatory Guide 1.13, Revision 2, prior to fuel transfer into Spent Fuel Pools C or D."'9 CP&L further states that:

"Strict administrative controls will prevent an unacceptable assembly, as determined by the acceptance criteria stated in Section 4.2, from being transferred to Harris Pools C and D."2 ° According to CP&L, burnup is not a criterion of acceptability for storage of BWR fuel in pools C and D. The reactivity of an acceptable BWR fuel assembly will be limited by restricting its U-235 enrichment to 4.6 wt% and by the requirement that, for a Standard Cold Core Geometry ("SCCG") array of the fuel, Kinfinite must be less than or equal to 1.32 at all times during the life cycle of the assembly.2" CP&L calculations indicate that a BWR assembly of the type to be placed in pools C and D will, in a SCCG array, be maximally reactive (i.e., exhibit its 18 Y'ee Harris FSAR Table 1.3.1-1, Amendment No. 30.

19 License Amendment Application, Enclosure 7, Revision 3 at 4-4.

20 Id. at 4-17. CP&L's License Amendment Application does not provide details about these administrative controls. In its June 14, 1999, RAI Response (Exhibit 5), CP&L provides some information about the controls that will apply to PWR fuel from the Robinson station. See Exhibit 5. However, that information is not sufficient to support a thorough assessment of CP&L's administrative controls, including an assessment of their probability of failure.

Similarly, none of the documents provided by CP&L during the discovery phase of this proceeding provide sufficient information about relevant administrative controls to support an assessment of their efficacy. In a deposition, CP&L employee provided general information about CP&L's computer program for tracking the movement of fuel at the Harris plant, but was unfamiliar with the details of the program, such as how information used in the program is verified. See Transcript of Deposition Michael J. DeVoe, P.E. at 9-25 (October 20, 1999),

attached as Exhibit 6.

16 maximum value of Kinfinite) when its bumup is approximately 12,000 MW-days per tonne U.22 C. Orange County's Intervention in Licensing Proceeding On January 7, 1999, the NRC published a notice of opportunity for a hearing on the proposed license amendment, at 64 Fed. Reg. 2,237. Orange County filed a timely hearing request and intervention petition on February 12, 1999. On April 5, 1999, Orange County submitted contentions challenging the adequacy of the License Amendment Application. Orange County's contentions included a challenge to the adequacy of CP&L's criticality measures. The claims raised by the contention were two-fold. First, Orange County contended that CP&L's proposed reliance on Draft Regulatory Guide 1.13, which permits reliance on administrative measures for criticality prevention, was precluded by GDC 62, a duly promulgated regulation.

GDC 62 requires the use of "physical systems and processes." Second, Orange County argued that even if CP&L could rely on the regulatory guidance, it could not satisfy the "double contingency" principle set forth in the Draft Reg. Guide:

At all locations in the reactor spent fuel storage facility where spent fuel is handled or stored, the nuclear criticality safety analysis should demonstrate that criticality could not occur without at least two unlikely, independent, and concurring failures or operating limit violations.

Draft Reg. Guide 1.13 at 1.13-12 (emphasis in original). CP&L's proposed administrative controls on criticality would not satisfy this requirement because only one failure or violation, namely placement in the racks of PWR fuel not within the "acceptable range" of bumup, could cause criticality. Orange County's Supplemental Petition to Intervene at 10-13.

21 Kinfinite is the neturon multiplication factor in an infinite array of fuel.

22 See page 4-10 of Revision 3 of Enclosure 7 to license amendment application. See also letter from Donna B. Alexander, CP&L, to U.S. NRC, enclosing response to April 29, 199, Request for Additional Information (June 14, 1999) (hereinafter "June 14, 1999 RAI Response"),

attached as Exhibit 5.

17 In LBP-99-25, Memorandum and Order (Ruling on Standing and Contentions), the Licensing Board ruled that Orange County had standing, and admitted two of the County's contentions. 50 NRC 25 (1999). As admitted by the Licensing Board, Contention TC-2 (Inadequate Criticality Prevention) reads as follows:

CONTENTION: Storage of pressurized water reactor (UPWR") spent fuel in pools C and D at the Harris plant, in the manner proposed in CP&L's license amendment application, would violate Criterion 62 of the General Design Criteria ("GDC") set forth in Part 50, Appendix A. GDC 62 requires that: "Criticality in the fuel storage and handling system shall be prevented by physical systems or processes, preferably by use of geometrically safe configurations." In violation of GDC 62, CP&L proposes to prevent criticality of PWR fuel in pools C and D by employing administrative measures which limit the combination of burnup and enrichment for PWR fuel assemblies that are placed in those pools. This proposed reliance on administrative measures rather than physical systems or processes is inconsistent with GDC 62.

50 NRC at 35. In ruling on the contention, the Licensing Board used CP&L's "two-basis construct," construing the bases of the contention as follows:

a. Basis 1 -- CP&L's proposed use of credit for burnup to prevent criticality in pools C and D is unlawful because GDC 62 prohibits the use of administrative measures, and the use of credit for burnup is an administrative measure.

b. Basis 2 -- The use of credit for burnup is proscribed because Regulatory Guide 1.13 requires that criticality not occur without two independent failures, and one failure, misplacement of a fuel assembly, could cause criticality if credit for burnup is used.

The Board found that that the first basis raises "essentially a question of law," and that the second basis raises the following "question of fact":

Will a single fuel assembly misplacement, involving a fuel element of the wrong burnup or enrichment, cause criticality in the fuel pool, or would more than one such misplacement or a misplacement coupled with some other error be needed to cause such criticality?

LBP-99-25, 50 NRC at 36.23 23 As discussed below in Section IV.H and in Appendix A, the Board's summary of the

18 As required by 10 C.F.R. § 2.1111, the Board offered the parties an opportunity to invoke the hybrid hearing process outlined in Subpart K. This process establishes a 90-day discovery period, followed by the filing of a detailed written summary of all facts, data and arguments that each party intends to rely on to support the existence of a genuine and substantial dispute of fact regarding any admitted contentions. Following this filing, an oral argument is held. CP&L invoked the hybrid hearing process, and therefore this Summary is being filed herewith.

ARGUMENT IV. THE PROPOSED LICENSE AMENDMENT FAILS TO COMPLY WITH GDC 62 BECAUSE IT IMPROPERLY RELIES ON ADMINISTRATIVE MEASURES FOR CRITICALITY PREVENTION.

As demonstrated below, the proposed License Amendment Application fails to comply with GDC 62 because it improperly relies on administrative measures for criticality prevention.

In addition, the License Amendment Application is inconsistent with the valid and applicable portions of NRC Staff guidance for analysis of criticality prevention measures. Orange County submits that these issues may be decided as a matter of law, by applying GDC 62 and NRC Staff guidance to the clear and undisputed evidence regarding CP&L's proposed criticality prevention measures. If the Board decides that it is unable to rule for Orange County on these submissions, the Board should find that Orange County has raised a genuine, substantial and material factual and legal dispute with CP&L, and order that Contention TC-2 proceed to a trial pursuant to 10 C.F.R. § 2.1115.

Double Contingency Principle as found in Draft Reg. Guide 1.13 is not fully consistent with the language of the Reg. Guide itself, or with Orange County's contention. Orange County does not believe, however, that the Board intended to issue a definitive interpretation of the Draft Reg.

Guide with this admissibility ruling.

19 As discussed in more detail in Section I of Orange County's Detailed Summary and Sworn Submission of Facts, Data and Arguments, etc., With Respect to Quality Assurance Issues, the Licensing Board must allocate the burden of proof to the Applicant in considering whether the standard for going forward with an adjudicatory hearing is satisfied.

A. The General Design Criteria Establish Minimum Design Requirements for Nuclear Power Plants.

The Commission's General Design Criteria ("GDC") for Nuclear Power Plants establish the basic principles of nuclear power plant design. They constitute:

minimum requirements for the principal design criteria for water-cooled nuclear power plants similar in design and location to plants for which construction permits have been issued by the [Nuclear Regulatory] Commission.

Appendix A to 10 C.F.R. Part 50, Introduction (emphasis added). The General Design Criteria constitute basic guidance for the more detailed NRC safety regulations. They are "intended to provide engineering goals rather than precise tests or methodologies by which reactor safety

[can] be fully and satisfactorily gauged." Petitionfor Emergency and Remedial Action, CLI-78 6, 7 NRC 400, 406 (1978), quoting Nader v. Nuclear Regulatory Commission, 513 F.2d 1045 (D.C. Cir. 1975). As the Commission noted in that case, there are a "variety of methods for demonstrating compliance with GDC," including regulatory guides, standard format and content guides for license applications, the Standard Review Plan, and Branch Technical Positions. Id.

Although the Commission allows flexibility in developing methods for compliance with the general requirements of the General Design Criteria, the fundamental principles of the GDC must be adhered to in choosing those methods. Thus, for example, in Nader v. Ray, the Court of Appeals held that a set of detailed standards for prevention of a loss of coolant accident was consistent with the broad requirement of GDC 35 for a "system to provide abundant emergency

20 core cooling." 513 F.2d at 1051-53. But see Consumers Power Co. (Big Rock Point Nuclear Plant), ALAB-725, 17 NRC 562, 567 571 (1983).24 B. The Plain Language of GDC 62 Requires the Use of Physical Systems or Processes to Prevent Criticality, and Thereby Precludes the Use of Administrative Controls.

1. The plain language of GDC 62 requires the use of physical systems or processes to prevent criticality.

General Design Criterion 62 is entitled "Prevention of criticality in fuel storage and handling." GDC 62 instructs that:

Criticality in the fuel storage and handling system shall be prevented by physical systems or processes, preferably by the use of geometrically safe configurations.

The language of GDC 62 is quite clear: criticality control measures must be carried out by 24 In Consumers Power, the Appeal Board found that a remotely controlled makeup line for the spent fuel pool constituted a "physical system" for criticality control, and therefore was consistent with the requirement of GDC 62 that criticality must be maintained through "physical systems or processes." Id. at 571. In the County's view, the use of a makeup line is an impermissible administrative procedure, because it requires ongoing reliance on human action to turn on the flow of water into the makeup line. Two aspects of the Consumers Power decision give it questionable applicability to this case, however. First, the Appeal Board noted that it had been provided with "no evidence" to suggest that the make-up line was not a physical system within the "broad, general terms" of the GDC. 17 NRC at 571. Here, in contrast, Orange County has provided the Board with evidence of (a) the clear basis for distinguishing physical measures from ongoing administrative measures, and (b) the Commission's intent to preclude the use of procedural controls for criticality control. See Sections B. L.a and B. 1.b, below. Second, the circumstance addressed in the Consumers Power decision, involving the hypothetical exposure of high-reactivity (fresh or nearly-fresh) fuel to boiling water, foam or mist, is now implicitly addressed in Staff guidance which establishes a Keffective value of 0.98 for such a scenario, rather than requiring measures for maintaining Keffective below 0.95. See Kopp Memorandum at 4-5 (Exhibit 4). The Staff guidance is provided in the context of fresh fuel storage in a new fuel storage facility (vault), but logically must apply to pool storage of high reactivity fuel that could become critical in the presence of boiling water, foam or mist. Indeed, the informational Appendix A to ANSI/ANS-8-17-1984, American National Standard, Criticality Safety Criteria for the Handling, Storage and Transportation of LWR Fuel Outside Reactors (January 13, 1984), indicates that "void formation by boiling" is a normal condition for the purpose of evaluating the potential for criticality in a fuel pool. Thus, the question of whether a makeup line constitutes a physical measure for purposes of eliminating a boiling, misting or

21 physical systems or processes. The phrase "physical systems or processes" is not defined in Appendix A to Part 50, but it may be understood by reference to the example provided in GDC 62 of an acceptable physical system or process: a geometrically safe configuration. In other words, fuel storage racks must be configured in such a way as to prevent criticality, without resort to any ongoing administrative measures. Standing alone, the plain language of GDC 62 clearly dictates that CP&L must rely solely on physical measures to avoid criticality. Because CP&L intends to rely in part on ongoing administrative measures, i.e., control of bum-up and enrichment, its license amendment application must be rejected based on the plain language of GDC 62.

Moreover, in contrast to some of the other General Design Criteria, nothing about GDC 62 remains open-ended or subject to later revision. For instance, with respect to the definition of a loss of coolant accident, footnote 1 of Appendix A to Part 50 states that "[flurther details relating to the type, size, and orientation of postulated breaks in specific components of the reactor coolant pressure boundary are under development." Thus, GDC 62 is distinct from other criteria that "have not as yet been suitably defined." Nader v. NRC, 513 F.2d at 1052.

2. Physical systems and processes are distinct in nature from ongoing administrative controls In the prehearing conference, members of the Licensing Board questioned the distinction between physical systems and processes and administrative measures. Concededly, any physical measure has some administrative component, and any administrative measure has a physical component. However, there is a basic difference between the nature of physical systems and processes, on the one hand, and administrative measures, on the other hand.

foam environment in a spent fuel pool has effectively been mooted.

22 If a subcritical margin of reactivity is to be maintained in a fuel pool solely by use of a geometrically safe configuration, then administrative controls will be needed to ensure that the fuel racks provide the required configuration. That configuration must be maintained during normal operation and after specified insults, such as an earthquake or the drop of an object onto a rack. The necessary administrative controls may be stringent, but they will be applied on a one time basis. After the fuel racks are designed, fabricated and installed, ongoing administrative controls will not be required.

Similarly, if a subcritical margin of reactivity is to be maintained in a pool partly by exploiting the neutron-absorbing properties of the fuel racks, then one-time administrative controls will be needed to ensure that those properties are provided. For example, if Boral panels are attached to the racks, then one-time administrative controls will be needed to ensure that the Boral panels are properly designed, fabricated and installed. Periodic inspections may be needed to ensure that the Boral panels or other neutron-absorbing materials retain their needed properties, but these inspections will be comparatively straightforward.

By contrast, prevention of criticality by ongoing administrative controls will require continuing actions by human beings to carry out these measures, such as inputting information into a computer system, and operating and maintaining equipment. These measures must be carried out throughout the period when criticality is possible. For example, if the presence of soluble boron is to be exploited as a means of criticality suppression in a fuel pool, then administrative controls must ensure that the concentration of soluble boron in the pool water never falls below a specified level. These administrative controls must be implemented on a continuous, ongoing basis, with complete reliability. The controls must apply to an entire pool, and to canals or other pools that are interconnected with that pool.

23 Similarly, if restrictions on fuel burnup/enrichment or fuel age are to be exploited as means of criticality suppression in a rack in a fuel pool, then ongoing administrative controls must ensure that a fuel assembly is never placed in the rack unless its bumup/enrichment or age is within a specified range. Ongoing administrative controls on fuel bumup/enrichment or fuel age can be specified for an entire pool, for a particular rack, or for particular spaces within a rack.

At a number of nuclear plants, a "checkerboard" pattern of fuel placement has been specified, wherein particular spaces in the repeating checkerboard pattern have particular restrictions on fuel bumup/enrichment. These administrative controls must be effective on each occasion when a fuel assembly could be placed in the pool.

Ongoing administrative controls are inherently less reliable than physical systems and processes, because they involve the repetition of tasks numerous times, thus providing multiple and cumulative opportunities for error. They must also be implemented by human beings, and thus are prey to human error. A related factor noted by the NRC Staff in an Information Notice is the potential unfamiliarity of fuel handling personnel with procedures:

Refueling activities are safety-significant operations that are not conducted on a routine basis. In addition, fuel handling activities are often performed by contractor personnel under the supervision of licensee personnel. As a result, fuel handling personnel may not be familiar with the fuel handling equipment or may feel that their experience in fuel handling operations permits them to ignore some requirements for procedural use and adherence.

Information Notice 94-13 (February 22, 1994).25 Thus, while physical systems and processes entail some administrative controls, these are one-time controls that generally are completed before the system or process is put to use. By contrast, the use of restrictions on fuel bumup/enrichment or fuel age, or reliance on the presence 25 A copy of this Information Notice is attached to Appendix A as Exhibit A-16.

24 of soluble boron, as means of criticality suppression will require ongoing administrative controls.

This requirement can never be relaxed, and the controls must be implemented on a completely reliable basis. Over time, ongoing administrative controls of this kind will have a much higher cumulative probability of failure than one-time controls.

C. The Rulemaking History of GDC 62 Supports the Plain Language of the Regulation.

The rulemaking history of GDC 62 makes it even more clear that in promulgating GDC 62, the Commission intended to impose the fundamental requirement that criticality must be controlled by physical rather than administrative or procedural measures. Early in the rulemaking process, and in the proposed rule, the Commission considered language favoring physical systems or processes, but permitting procedural measures. In response to comments, however, the Commission removed the reference to procedural measures, and established a clear requirement that physical systems and processes must be used. In addition, while the General Design Criteria were originally proposed as guidance, they ultimately were promulgated in the form of minimum requirements.

1. Pre-rulemaking documents To Orange County's knowledge, a set of draft General Design Criteria first appeared as an attachment to an Atomic Energy Commission ("AEC")26 press release of November 22, 1965, entitled "AEC seeking public comment on proposed design criteria for nuclear power plant construction permits."27 The attachment included draft Criterion 25, which proposed the following language relating to prevention of criticality in fuel handling and storage facilities:

The fuel handling and storage facilities must be designed to prevent criticality and to 26 The Atomic Energy Commission was the predecessor agency to the NRC.

27 The Press Release and attached documents are attached as Exhibit 7.

25 maintain adequate shielding and cooling for spent fuel under all anticipated normal and abnormal conditions, and credible accident conditions. Variables upon which health and safety of the public depend must be monitored.

During the following year, the AEC continued to revise the language of the proposed GDC in response to comments made by AEC staff and by members of the Advisory Committee on Reactor Safeguards ("ACRS"). A revised draft of October 6, 1967, prepared by the AEC, contained draft Criterion 10, which stated:

Possibilities for inadvertent criticality must be prevented by engineered systems or processes to every extent practicable. Such means as geometric safe spacing limits shall be emphasized over procedural controls."8 The same language appeared again in an October 20, 1966 draft, which was attached to a letter of October 25, 1966 from J.J. DiNunno of the AEC to David Okrent of the ACRS.2 9 Another draft of a GDC for criticality prevention appears as a February 6, 1967, attachment to a letter from J. J. DiNunno of the AEC to Nunzio J Palladino of the ACRS, dated February 8, 1967.30 In this draft, the potential for criticality in fuel handling and storage facilities was addressed by Criterion 61, which stated:

Possibilities for criticality in new and spent fuel storage shall be prevented by physical systems or processes to every extent practicable. Such means as favorable geometries shall be emphasized over procedural controls.

2. Proposed GDC for criticality control On June 16, 1967, the AEC Director of Regulation proposed a set of draft GDCs to the AEC Commissioners, "for consideration by the Commission at an early date".3" The set of 28 Internal AEC memorandum from G.A. Arlotto to J.J. DiNuuno and Robert H. Bryan (October 7, 1966), and attached Revised Draft of General Design Criteria for Nuclear Power Plant Construction Permits (October 6, 1966), attached as Exhibit 8.

29 The October 25, 1966, letter and attached draft are attached to this Summary as Exhibit 9.

30 The February 8, 1967 letter and attached draft are attached to this Summary as Exhibit 10.

31 Note by the Secretary, W.B. McCool, to AEC Commissioners re: Proposed Amendment to

26 GDCs was described as a proposed amendment to 10 CFR 50. The potential for criticality in fuel handling and storage facilities was addressed by draft Criterion 66, which stated:

Criticality in new and spent fuel storage shall be prevented by physical systems or processes. Such means as geometrically safe configurations shall be emphasized over procedural controls.

Shortly thereafter, this language appeared in the Commission's notice of proposed rulemaking for the General Design Criteria, 32 Fed. Reg. 10,213 (July 11, 1967).32 Thus, throughout the early development of the GDC for criticality control, the concept of procedural controls was included in the language of the criterion.

The introduction to the General Design Criteria stated that they were "intended to be used as guidance in establishing the principal design criteria for a nuclear power plant." 32 Fed. Reg.

at 10,215.

3. Comments on the proposed rule Comments on the proposed GDC show persistent effort by the nuclear industry to influence the evolution of many of the GDCs, but comparatively little concern about the criterion that became GDC 62. The Commission did, however, receive an influential comment on criticality prevention from the Nuclear Safety Information Center, Oak Ridge National Laboratory (ORNL).33 The ORNL commented as follows:

We do not understand the implication of 'or processes' at the end of the first sentence, nor do we believe that it is practical to depend upon procedural controls to prevent accidental criticality in storage facilities of power reactors. Hence, the last sentence of this criterion should be changed to read as follows: 'Such means as geometrically safe 10 CFR 50: General Design Criteria for Nuclear Power Plant Construction Permits (June 16, 1967). The Note and relevant excerpts from Appendix B to the Note are attached as Exhibit 11.

32 A copy of the Federal Register notice is attached to this Summary as Exhibit 12.

33 ORNL's comments on the proposed rule were contained in an attachment to a letter of September 6, 1967 from William B. Cottrell of ORNL to H. L. Price of the AEC, attached as Exhibit 13.

27 configurations shall be used to insure that criticality cannot occur.'34 On July 15, 1969, the AEC prepared a set of revisions to the GDC, based on comments by the ACRS and the nuclear industry. As discussed in the accompanying cover letter, a major difference between the proposed GDC and the revised GDC was that the revised GDC

"[e]establish "minimum requirements" for water-cooled reactors, whereas the published criteria were "guidance" for all reactors." The revised GDC included GDC 62, entitled "Prevention of Criticality in Fuel Storage and Handling:"

Criticality in the fuel storage and handling system shall be prevented by physical systems or processes, preferably by use of geometrically safe configurations.

On June 4, 1970, the AEC prepared another revision to the GDC, containing the identical language of GDC 62 that had been prepared on July 15, 1969. This revision was circulated to other members of the AEC and the Atomic Industrial Forum (AIF), a nuclear industry trade organization.16 Although the AIF recommended substantial changes to other GDCs contained in the revised draft, it accepted the new draft GDC 62 without any proposed alteration.

4. The Final Rule On February 20, 1971, the AEC published the General Design Criteria in final form.37 The introduction to the GDC's now characterized them as "minimum requirements" for the design of 34 Id., Attachment containing "Specific Comments" at 11.

35 Letter from Edson G. Case, AEC, to Dr. Stephen H. Hanauer, ACRS (July 23, 1969),

enclosing General Design Criteria for Nuclear Power Units (July 15, 1969), attached as Exhibit 14.

36 See Memorandum from Edson G. Case, NRC, to Harold L. Price, et al., AEC, re: Revised General Design Criteria (October 12, 1970), and enclosed letter from Edward A. Wiggin, AIF, to Edson G. Case, NRC (October 6, 1970) Attached to the Wiggin letter is a marked-up version of the June 4, 1966, revised draft of the GDC. The Case Memorandum and enclosed documents are attached as Exhibit 15.

37 Final Rule, General Design Criteria for Nuclear Power Plants, 36 Fed. Reg. 3,255 (February 20, 1971). A copy of the Federal Register notice is attached as Exhibit 16.

28 nuclear power plants, rather than "guidance" as had been proposed. In addition, the final rule included GDC 62, which provided that:

Criticality in the fuel storage and handling system shall be prevented by physical systems or processes, preferably by use of geometrically safe configurations."

The final rule removed the language in the proposed rule that had included "procedural controls" in the set of acceptable measures for controlling criticality. Instead, "physical systems or processes" became the only acceptable means of criticality control. Moreover, geometrically safe configurations were clearly identified as the "preferred" type of physical system or process, in lieu of "emphasized" controls. It can be assumed that ORNL's comment regarding the impracticality of procedural controls had an important influence on this near-final step in the evolution of GDC 62. Thus, the rulemaking history of GDC 62 illustrates the importance placed by the Commission on physical systems and processes, in contrast to procedural controls.

D. The Plain Language of GDC 62 Is Not Altered or Contradicted By Other Relevant NRC Criticality Standards.

GDC 62's plain language, requiring the use of physical systems or processes to prevent criticality, is consistent with other relevant NRC regulations for criticality prevention that were promulgated afterwards. In particular, GDC 62 is consistent with the NRC's requirements for criticality prevention in 10 C.F.R. § 50.68 and 10 C.F.R. § 70.24. Both the language of these regulations and their regulatory history demonstrate that the Commission considers physical systems and processes to be essential to preventing criticality in the storage of spent or fresh fuel.

1. 10 C.F.R. §§ 70.24 and 50.68 Aside from GDC 62, prior to 1998 the NRC's only criticality-related regulation for operating nuclear power plants consisted of 10 C.F.R. § 70.24, which required criticality

29 monitoring for any licensee authorized to possess significant quantities of special nuclear material ("SNM"). The regulation included a provision authorizing licensees to seek an exemption where good cause was shown. 10 C.F.R. § 70.24(d).

On December 3, 1997, the NRC concurrently published in the Federal Register a proposed rule and a direct final rule, making changes to 10 C.F.R. § 70.24 and adding a new section 50.68.38 The purpose of the amended regulations was to eliminate the requirement for case-by-case exemptions from § 50.24, and establish a blanket exemption for licensees who agreed to follow a set of criticality accident prevention requirements in the new section 50.68.

The new set of rules was based on the NRC's experience that a "large number of exemption requests ha[d] been submitted by power reactor licensees and approved by the NRC based on safety assessments which concluded that the likelihood of criticality was negligible."'39 The discussion of safety in criticality control which followed this assertion made it clear that the finding of negligible risk was based in part on the assumption that during fuel storage, physical measures such as design features would be used to prevent criticality:

At a commercial nuclear power plant, the reactor core, the fresh fuel delivery area, the fresh fuel storage area, the spent fuel pool, and the transit areas among these, are areas where amounts of SNM sufficient to cause a criticality exist. In addition, SNM may be found in laboratory and storage locations of these plants, but an inadvertent criticality is not considered credible in these areas due to the amount and configuration of the SNM.

The SNM that could be assembled into a critical mass at a commercial nuclear power plant is only in the form of nuclear fuel. Nuclear power plant licensees have procedures and the plants have design features to prevent inadvertent criticality. The inadvertent criticality that 10 CFR 70.24 is intended to address could only occur during fuel-handling operations.

In contrast, at fuel fabrication facilities SNM is found and handled routinely in various configurations in addition to fuel. Although the handling of SNM at these facilities is 38 Proposed Rule, Criticality Accident Requirements, 62 Fed. Reg. 63,911; Direct Final Rule With Opportunity to Comment, Criticality Accident Requirements, 62 Fed. Reg. 63,825.

39 62 Fed. Reg. at 63,825, Col. 3.

30 controlled by procedures, the variety of forms of SNM and the frequency with which it is handled provides greater opportunity for an inadvertent criticality than at a nuclear power reactor.

At power reactor facilities with uranium fuel nominally enriched to no greater than five (5.0) percent by weight, the SNM in the fuel assemblies cannot go critical without both a critical configuration and the presence of a moderator. Further,thefresh fuel storage arrayand the spentfuel pool are in most cases designed to prevent inadvertent criticality, even in the presence of an optimal density of unboratedmoderator. Inadvertent criticality during fuel handling is precluded by limitations on the number of fuel assemblies permitted out of storage at the same time. In addition, GeneralDesign Criterion (GDC) 62 in Appendix A to 10 CFR Part50 reinforces the prevention of criticality infuel storageand handling throughphysical systems, processes,and safe geometricalconfiguration. Moreover, fuel handling at power reactor facilities occurs only under strict procedural control. Therefore, the NRC considers a fuel-handling accidental criticality at a commercial nuclear plant to be extremely unlikely. The NRC believes the criticality monitoring requirements of 10 CFR 70.24 are unnecessary as long as design and administrativecontrols are maintained. 40 Thus, in promulgating § 50.68, the Commission affirmed the language of GDC 62 which restricts criticality prevention measures to physical systems and processes.

The language of § 50.68, as it was finally promulgated, contains a list of measures for criticality prevention that can be implemented in lieu of maintaining a criticality monitoring system.4" Although these provisions contain some references to procedures and administrative measures, they do not undermine or contradict the general requirement of GDC 62 for physical criticality prevention measures. For instance, subsection (b)(1) requires that:

Plant procedures shall prohibit the handling and storage at any one time of more fuel assemblies than have been determined to be safely subcritical under the most adverse moderation conditions feasible by unborated water.

This provision simply requires licensees to have a procedure which forbids them from handling or storing any fuel assemblies for which the licensees are unable to maintain 40 62 Fed. Reg. at 63,825-26. (emphasis added) 41 See Final Rule, Criticality Accident Requirements, 63 Fed. Reg. 63,127 (November 12,

31 subcriticality. It does not explicitly address whether, for the number of assemblies that are permitted to be handled or stored, criticality control must be accomplished through physical measures or may be addressed by administrative measures. However, it is noteworthy that the provision assumes that at least one administrative measure, reliance on the presence of boron in the pool water, will not be available.

Subsections (b)(2) and (b)(3) provide that:

(2) The estimated ratio of neutron production to neutron absorption and leakage (k effective) of the fresh fuel in the fresh fuel storage rack shall be calculated assuming the racks are loaded with fuel of the maximum fuel assembly reactivity and flooded with unborated water and must not exceed 0.95, at a 95 percent probability, 95 percent confidence level. This evaluation need not be performed if administrative controls and/or design features prevent such flooding or if fresh fuel storage racks are not used.

(3) If optimum moderation of fresh fuel in the fresh fuel storage racks occurs when the racks are assumed to be loaded with fuel of the maximum fuel assembly reactivity and filled with low-density hydrogenous fluid, the k-effective corresponding to this optimum moderation must not exceed 0.98, at a 95 percent probability, 95 percent confidence level. This evaluation need not be performed if administrative controls and/or design features prevent such flooding or if fresh fuel storage racks are not used.

These requirements relate to the storage of fresh fuel in fresh fuel storage racks. Fresh fuel storage racks are free-standing racks that surround the fresh fuel with air. By design, no water is present that could act as a moderator. The absence of water as a moderator is a physical system or process for criticality control, built into the design of the fresh fuel storage facility. This is consistent with GDC 62.

Subsections (b)(2) and (b)(3) require the licensee to perform an accident analysis that demonstrates criticality will be prevented, even if water accidentally enters the fresh fuel racks.

A licensee may be exempted from the accident analysis if it demonstrates one of two things: that 1998).

32 flooding will be prevented by administrative measures, or that fresh fuel storage racks will not be used. The first option, use of administrative measures to prevent flooding, is in addition to the design features by which fresh fuel racks are located in a place that is removed from the presence of water. Thus, it cannot be viewed as a primary criticality prevention measure, but as a secondary measures used as a back-up to the primary design features. If the second option is elected, the licensee must show that fresh fuel racks are not used, i.e., that the fresh fuel is stored in a fuel pool. If fresh fuel is stored in a pool, it must meet the same criticality prevention requirements as apply to spent fuel (see subsection (b)(4), discussed below). Under these requirements, the fuel must remain subcritical, even in the absence of soluble boron.42 Accordingly, there is nothing about subsections (b)(2) or (b)(3) that is inconsistent with the requirement of GDC 62 that physical systems and processes must be used to prevent criticality.

Subsection (b)(4) relates to the storage of fuel in spent fuel pools. Although this provision also mentions administrative measures in the sense that it discusses the parameters for taking credit for the presence of soluble boron in the water, the provision also makes it clear that criticality ultimately must be prevented without resort to administrative measures:

If no credit for soluble boron is taken, the k-effective of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95 percent probability, 95 percent confidence level, if flooded with unborated water. If credit is taken for soluble boron, the k-effective of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95 percent probability, 95 percent confidence level, if flooded with borated water, and the k-effective must remain below 1.0 (subcritical), at a 95 percent probability, 95 percent confidence level, if flooded with unborated water.

Thus, the basic requirement of subsection (b)(4) is that criticality must be controlled (i.e.,

Keffective maintained below 1.0) without considering the presence of soluble boron in the 42 As discussed in note 23 above, arrangements for storage of fresh fuel in a pool should also

33 water.

It should also be noted that the type of ongoing administrative measure proposed by CP&L in the instant case, i.e., control of burnup/enrichment levels in the fuel, is not condoned by

§ 50.68, or even mentioned.

2. 10 C.F.R. § 72.124 The Commission has also promulgated regulations for control of criticality at Independent Spent Fuel Storage Installations ("ISFSI's"). These regulations are inconsistent with GDC 62, because they do not unequivocally require the use of physical systems or processes for criticality control, and instead apply a practicability standard. 10 C.F.R. § 72.124(b) provides as follows:

Methods of criticalitycontrol. When practicable the design of an ISFSI or MRS must be based on favorable geometry, permanently fixed neutron absorbing materials (poisons),

or both. Where solid neutron absorbing materials are used, the design shall provide for positive means to verify their continued efficacy.

The ISFSI regulations do not apply to the instant proceeding, however. The Harris operating license amendment is being considered under Part 50 of the regulations, which govern nuclear power plant operating licenses. It is not being considered under Part 72, the ISFSI regulations.

Section 72.124(b) is also inapplicable to this case because design and operation of an ISFSI is fundamentally different than the design and operation of a nuclear power plant, such that the Commission might have grounds for establishing a more relaxed standard for criticality control at ISFSI's than for nuclear power plants. As recognized by the Commission in the ensure that the fuel remains subcritical in the presence of boiling water, foam or mist.

43 The other provisions of § 50.68, subsections (b)(5) through (8), are not relevant to this proceeding.

34 preamble to the ISFSI regulations, an ISFSI is "not coupled to either a nuclear power plant or a fuel reprocessing plant." 43 Fed. Reg. at 46,309. The Commission saw "a need for a new regulation covering the requirements for extended spent fuel storage under static storage conditions involving no operations on such materials." Id. (emphasis added). In contrast, the operations in a fuel storage building of a nuclear power plant cannot be considered "static."

Fresh fuel is constantly being brought into the fuel building and moved through the fuel transfer canals and pools into the reactor. The same equipment and personnel are used to move both fresh and spent fuel. Also, at a nuclear power plant there will be occasions when spent fuel with a reactivity nearly as high as, or even higher than, the reactivity of fresh fuel is stored in fuel pools. This could occur, for example, during a full core offload.

Thus, at an operating nuclear power plant there is the constant possibility that fresh fuel will be placed inappropriately into a spent fuel storage pool. Indeed, such mispositioning has occurred in the past. ' By requiring physical systems and processes for the control of criticality, GDC 62 ensures that criticality will be avoided, regardless of the burnup level or age of fuel that is placed in the pool. It is much less likely that fresh or highly reactive fuel would be placed in an ISFSI, and thus there may not be the same need to insist on physical measures for criticality prevention at an ISFSI.

Although the Board need not reach this far in finding that 10 C.F.R. § 72.142(b) has no precedential value in this case, it is also noteworthy that § 72.142(b) was not duly promulgated in compliance with the procedural requirements of the Administrative Procedures Act, 5 U.S.C. § 553, for public notice and opportunity to comment. The current language of § 72.124(b) was 44 See examples cited in Appendix B: Braidwood Unit 1, (July 10, 1996); Cooper Station (March 5, 1990); Crystal River Unit 3 (November 9, 1987); Oyster Creek Unit 1 (January 21,

35 promulgated in 1988, when the Commission added requirements for Monitored Retrievable Storage ("MRS") to the ISFSI regulations.45 The 1988 rulemaking fundamentally altered the Commission's existing regulation for criticality control at ISFSI's, which had been promulgated with the original set of ISFSI regulations in 1980.

Section 72.73(b) of the original ISFSI regulations explicitly and unequivocally required the use of geometric spacing and/or fixed neutron-absorbing material

- i.e., physical systems and processes - for criticality control:

Methods of criticalitycontrol. The design of an ISFSI or MRS must be based on favorable geometry (spacing), permanently fixed neutron absorbing materials (poisons),

or both. Where solid neutron absorbing materials are used, the design shall provide for positive means to verify their continued efficacy. In criticality design analyses for underwater storage systems, credit can be taken for the neutron absorption of rack structures and the water within the storage unit.

Final rule, Licensing Requirements for the Storage of Spent Fuel in an Independent Spent Fuel Storage Installation, 45 Fed. Reg. 74,693, 74,710 (November 12, 1980).

On May 27, 1986, the Commission proposed to amend the Part 72 regulations to encompass the licensing of MRS facilities and to "clarify matters that have arisen since part 72 was made effective on 11/28/80." Proposed Rule, Licensing Requirements for the Independent Storage of Spent Nuclear Fuel and High-Level Radioactive Waste, 51 Fed. Reg. 19,106. The Federal Register notice included the following provision for methods of criticality control,

§ 72.93:

Methods of criticality control. The design of an ISFSI or MRS must be based on favorable geometry (spacing), permanently fixed neutron absorbing materials (poisons),

or both. In criticality design analyses, credit can be taken for fixed neutron absorbing material present within the storage structure.

1987).

45 Final Rule, Licensing Requirements for the Independent Storage of Spent Nuclear Fuel and High-Level Radioactive Waste, 53 Fed. Reg. 31,651 (August 19, 1988).

36 51 Fed. Reg. at 19,124. These proposed changes to the 1980 criticality control regulation were minor: they added a reference to an MRS, and they took out the sentence requiring the verification of continued efficacy of fixed poisons. Significantly, the proposed rule continued to require the use of favorable geometry and permanently fixed poisons as mandatory measures.

When the final rule was promulgated in 1988, the provision governing methods for controlling criticality was transformed. No longer did the rule contain a mandatory requirement for favorable geometry and fixed poisons; instead, these measures were called for only "if practicable." The Commission had also added to § 72.124(a) the following "double contingency" provision, not found in the 1980 rule or the 1986 proposed rule:

Spent fuel handling, packaging, transfer, and storage systems must be designed to be maintained subcritical and to ensure that, before a nuclear criticality accident is possible, at least two unlikely, independent, and concurrent or sequential changes have occurred in the conditions essential to nuclear criticality safety.46 No justification can be found in the preamble to the final rule for this eleventh hour substitution of language that was so completely different from the proposed rule. The only mention of the changes is the following discussion:

Comment: A comment was received concerning the removal of the requirement for verifying continued efficacy of solid neutron poisons.

Response: Several changes have been made to the criticality section of the final rule to make it correspond to other Parts of the Commission's regulations and standard criticality review practices. Verification of solid neutron poisons has been retained. Double contingency criteria and requirements for criticality monitors have been added. It is not the intent of the revision concerning criticality monitors to require monitors in the open areas where loaded casks are positioned for storage as that system is static. Monitors are required where the systems are dynamic.

46 53 Fed. Reg. at 31,674. The 1980 rule and the proposed 1986 rule had provided that: Spent fuel handling, packaging, transfer, and storage systems must be designed to be maintained subcritical and to prevent a nuclear criticality accident. 45 Fed. Reg. at 74,710; 51 Fed. Reg. at 19,124.

37 53 Fed. Reg. at 31,656. Here, the Commission effectively admitted that the changes had nothing to do with a response to comments: the provision relating to the comment regarding verification of the continued efficacy of solid neutron poisons was not changed at all, but was "retained."

Instead, the Commission claimed to have changed the rule "to make it correspond to other Parts of the Commission's regulations and standard criticality review practice." The Commission did not identify what other regulations this new rule is consistent with, and indeed none can be identified: this is a rationalization without substance. Nor did the Commission attempt to describe the alleged "standard criticality review practice," justify it, or explain why the Commission failed to give public notice prior to making the change. By making such a major substantive change in the final rule, without first providing public notice or permitting public comment, the Commission violated the Administrative Procedure Act, which renders the rule invalid.47 E. The Administrative Criticality Prevention Proposed by CP&L Would Violate GDC 62.

As described above in Section IlI.B, CP&L proposes to restrict the bumup/enrichment of PWR fuel in order to suppress criticality under normal conditions. CP&L asserts that these burnap/enrichment limits will be carried out through "strict administrative controls" that will prevent an unacceptable assembly from being transferred to Harris Pools C and D.48 This reliance on ongoing administrative procedures and controls to enforce 47 See American Frozen FoodInstitute v. Train, 539 F.2d 107, 135 (D.C. Cir. 1976);

Connecticut Light and Power Co. v. NRC, 673 F.2d 525, 533 (D.C. Cir.), cert. denied, 459 U.S.

835 (1982); FloridaPower & Light Co. v. US., 846 F.2d 765, 771-72 (D.C. Cir. 1988), cert.

denied, 490 U.S. 1045 (1989); Air TransportAssociation ofAmerica v. FAA, 169 F.3d 1, 6-8 (D.C. Cir. 1999).

38 burnup/enrichment limits violates the language and intent of GDC 62, which is to ensure that physicalsystems andprocesses, preferably geometrically safe configuration of the assemblies, are used to control criticality. Similarly, CP&L relies on the presence of soluble boron to prevent criticality under accident conditions. This violates the plain meaning and intent of GDC 62, because the introduction and maintenance of soluble boron in the spent fuel pools require ongoing administrative actions and procedures, and do not constitute physical systems or processes.49 F. CP&L's Proposed Reliance on Administrative Criticality Prevention Measures Is Not Justified by Draft Reg. Guide 1.13 or Other NRC Staff Guidance.

In opposing the admissibility of Contention TC-2, CP&L and the NRC Staff argued that its reliance on control of bumup/enrichment levels to prevent criticality is permitted by Draft Reg. Guide 1.13. The Commission has stated generally that "if there is conformance with regulatory guides, there is likely to be compliance with the GDC." Petitionfor Emergency and Remedial Action, CLI-78-6, 7 NRC 400, 406 (1978). As the Board has recognized, however, this is "not a blanket endorsement of the notion that regulatory guides necessarily govern." LBP-99 25, 50 NRC at 35. Where there is inconsistency between a regulation and a regulatory guide, the 48 License Amendment Application, Enclosure 7 Rev. 3 at 4-17.

49 In one criticality analysis, CP&L relied on the presence of soluble boron during an accident.

See CP&L's June 14, 1999, RAI Response (Exhibit 5). In a subsequent response to the same RAI, CP&L stated that a new criticality analysis shows that if defined as Kinfinite less than 1, subcriticality can be maintained in unborated water, in the presence of one mispositioned fresh PWR fuel assembly. Letter from Donna B. Alexander to U.S. NRC (October 15, 1999), attached as Exhibit 17. However, a soluble boron concentration of 400 ppm was found necessary to "maintain Kinfinite less than the regultory limit of 0.95." Id. As discussed below in Section IV.F, the consideration of mispositioning of a single fresh fuel assembly does not constitute an adequate criticality analysis. For this reason, and to meet the regulatory limit of 0.95 for Kinfinite, it is necessary consider whether CP&L's reliance on the presence of soluble boron under abnormal conditions is consistent with GDC 62.

39 regulation is controlling. A regulation has the force of law; in comparison, a regulatory guide is a set of recommendations setting forth acceptable methods for complying with the regulation.

Such documents "are useful as guides," but "insofar as the adjudicatory process is concerned, they represent the opinions of one of the parties to that process and as such cannot be viewed as necessarily controlling." Potomac Electric Power Co. (Douglas Point Nuclear Generating Station, Units 1 and 2), LBP-76-13, 3 NRC 425, 432 (1976). See also Louisiana Energy Services (Claibome Enrichment Center), LBP-91-41, 34 NRC 332, 354 (1991). Therefore, a Reg. Guide cannot be relied on to modify or circumvent the requirements of duly promulgated regulations like the General Design Criteria.

To the extent that they permit prevention of criticality through administrative procedures and controls, Draft Reg. Guide 1.13 and the Kopp Memorandum violate the plain language and intent of GDC 62. Therefore, in this respect they must be disregarded.

G. Neither CP&L Nor the Staff Has Demonstrated That Public Health And Safety Will Be Adequately Protected If CP&L Relies on Ongoing Administrative Measures for Criticality Control.

Although the Staff's regulatory guidance is fundamentally at odds with GDC 62, the Staff's practice of permitting ongoing administrative measures for the prevention of criticality in spent fuel pools is well-entrenched. In recent years, the NRC Staff has approved many applications similar to CP&L's, setting a trend toward higher and higher density of spent fuel storage and greater and greater reliance on administrative controls to prevent criticality.

Astoundingly, the Staff has pursued this course for over two decades without conducting any safety analysis to determine whether its radical departure from the requirements of GDC 62 could be justified on safety grounds. The Staff has never done a systematic analysis of the potential for criticality accidents when reliance is placed on administrative measures instead of

40 physical measures. Although the Staff has advocated the Double Contingency Principle in evaluating criticality accidents since 1978, it has made no attempt to determine what combinations of fuel handling or pool management errors would violate the Double Contingency Principle. Instead, as discussed above and in Appendix A, it has merely watered down the Double Contingency Principle to a Single Contingency Principle. Despite the many years of accumulated licensee experience with spent and fresh fuel storage, the Staff has never attempted to conduct a systematic review of the operating experience of licensees with fuel mispositioning or fuel incidents relevant to boron dilution.5" The Staff does not even maintain a systematic data base of the experience of nuclear power plant licensees with such problems as mispositioning of fuel assemblies and soluble boron management errors.

In fact, as discussed in Appendix B, the limited information that was provided by the Staff in discovery, and that Orange County was able to find in the Public Document Room, shows that there is a significant history of incidents relevant to failure of criticality prevention in fuel pools. These incidents include mispositioning of fuel assemblies and incidents relevant to boron dilution, including one boron dilution event. Significantly, the record includes events in which a single error resulted in the mispositioning of more than one fuel assembly, such as the mispositioning of 184 fresh fuel assemblies in the Oyster Creek spent fuel pool in 1986. The record also includes incidents that are relevant to the prevention of criticality solely through the use of physical systems and processes, notably some errors in criticality analyses. These incidents raise questions about the size of the safety margin achieved when preventing criticality solely through the use of physical systems and processes, and the wisdom of cutting into that 50 Orange County is aware of only one generic study of boron dilution, which was done by a self-interested party, the Westinghouse Corporation, and which failed to summarize the historical

41 safety margin by placing reliance on less-reliable ongoing administrative measures.

As set forth in Appendix C, experience at U.S. nuclear power plants shows that fuel mispositioning, involving placement in a pool of one or more fuel assemblies with inappropriate burnup/enrichment or age, is a likely occurrence. Experience also shows that the concentration of soluble boron in a pool can fall below specified levels. Some accident sequences could yield substantial reductions in soluble boron concentration. From a qualitative perspective, it is clear that criticality scenarios which involve the failure of ongoing administrative controls have a much higher probability of occurring than criticality scenarios involving failure of physical controls. Also, Appendix C shows that significant onsite and offsite radiation exposures are potential outcomes of a criticality event in a fuel pool, including Harris pools C and D. Under the circumstances, there is no basis for concluding that the public health and safety can be protected through reliance on administrative measures for criticality prevention at the Harris nuclear power plant.

H. CP&L's Criticality Accident Analysis Misapplies Applicable Staff Guidance.

As discussed above, CP&L's criticality analysis is fundamentally deficient because CP&L relies on administrative measures for criticality prevention, in violation of GDC 62. To the extent that it condones this unlawful practice, current NRC guidance is also invalid.

In examining the lawfulness and reasonableness of CP&L's criticality prevention measures, it is necessary to go beyond a determination that physical systems and processes are required for criticality prevention. Even where such physical measures are used and are effective in preventing criticality during normal operation, it is necessary to perform an accident analysis to determine whether such measures are adequate to prevent criticality under a range of accident record of relevant events. See Appendix C.

42 conditions. For this purpose, portions of the NRC Staff's guidance for criticality control provide useful guidance that is consistent with GDC 62. In particular, the Double Contingency Principle provides a method of analysis that is useful for evaluating the potential for criticality accidents.

As set forth in Draft Reg. Guide 1.13, the Double Contingency Principle requires a nuclear criticality safety analysis to demonstrate that criticality could not occur "without at least two unlikely, independent, and concurrent failures or operating limit violations." CP&L has misapplied this guidance in four principal respects. First, CP&L ignores the words "at least,"

and evaluates only one failure instead of sets of failures; second, it fails to determine what failures are "unlikely, independent, and concurrent;" third, it assumes that mispositioning of fuel is an "unlikely" event when in fact it is likely; and fourth, it unreasonably assumes that a single error can lead to the mispositioning of only one fuel assembly.

Before addressing CP&L's misapplication of the Draft Reg. Guide in more detail, it is necessary to point out that in admitting "Basis 2" of Contention TC-2, the Board summarized the thrust of the contention in a manner that is overly narrow and inconsistent with the contention.5' The Board's summary of Basis 2 shortens Draft Reg. Guide 1.13's statement of the Double 51 The Board characterized Basis 2 as follows:

Basis 2 - The use of credit for burnup is proscribed because Regulatory Guide 1.13 requires that criticality not occur without two independent failures, and one failure, misplacement of a fuel assembly, could cause criticality if credit for bumup is used.

The Board also found that:

The second basis raises a question of fact: Will a single fuel assembly misplacement, involving a fuel element of the wrong bumup or enrichment, cause criticality in the fuel pool, or would more than one such misplacement or a misplacement coupled with some other error be needed to cause such criticality?

LBP-99-25, 50 NRC at 36.

43 Contingency Principle from "at least two independent, unlikely, and concurrent failures" to "two independent failures." The decision also contains language implying the assumption that one failure would lead to the misplacement of no more than one fuel assembly, and that the Double Contingency Principle is a single failure criterion. The Board also refers to "the required single failure criterion," when in reality the criterion is a double contingency standard.

Orange County believes that in admitting Basis 2 of Contention TC-2, the Board intended to permit the litigation of whether CP&L's criticality analysis satisfies the accident analysis criteria set forth in Draft Reg. Guide 1.13, as quoted and discussed by by Orange County at page 12-13 of its Supplemental Petition to Intervene.5 2 Orange County does not interpret the Board's summary of the contention's basis to constitute a definitive interpretation of Draft Reg. Guide 1.13, which after all is the subject of the contention. As the Board noted in admitting Basis 2, "Clearly the nature of this amendment, introducing as it does the presence of high density racks 52 The contention stated as follows:

Draft Reg. Guide 1.13 does not support the administrative measures proposed by CP&L.

Although Appendix A contains some language implying that the design of spent fuel racks against criticality can take credit for burnup (pages 1.13-13, 14, 15), other parts of the Draft Reg. Guide clearly proscribe such activity. For instance, at page 1.13-9, the Draft Reg. Guide states that:

At all locations in the LWR spent fuel storage facility where spent fuel is handled or stored, the nuclear criticality safety analysis should demonstrate that criticality could not occur without at least two unlikely, independent, and concurring failures or operating limit violations.

(emphasis in original). CP&L's proposed administrative controls on criticality would not satisfy this requirement because only one failure or violation, namely placement in the racks of PWR fuel not within the "acceptable range" of bumup, could cause criticality.

Note that "misplacement of a spent fuel assembly" is identified in the Draft Reg. Guide as one of nine "credible normal and abnormal operating occurrences."

The contention did not summarize Draft Reg. Guide 1.13 or assert that Orange County's only

44 on the site, involves a change that may call into question conformance with this aspect of the regulations." Id. at 36. In order to evaluate whether the License Amendment Application complies with this provision of Draft Reg. Guide 1.13, it is necessary to closely examine each aspect of the Double Contingency Principle as set forth in the Draft Reg. Guide, without attributing the Board's general summary of the Draft Reg. Guide as a definitive interpretation of its meaning.

CP&L's criticality accident analysis for pools C and D violates the guidance of Draft Reg. Guide 1.13 in the following respects:

1. CP&L ignores the words "at least," and evaluates only one failure instead of sets of failures.

Draft Reg. Guide 1.13 calls for the analysis of situations involving "at least" two failures or violations of operating limits. Analysis that meets this requirement must identify the sets of failures or violations that might cause criticality, and then evaluate these failures or violations in combinations of at least two, to determine which combinations will cause criticality. This process will yield an "envelop" of criticality which bounds the combinations of failures and violations that produce criticality. That envelope cannot be identified if failures or violations are evaluated one at a time. When the envelope has been identified, the Double Contingency Principle can be applied, with consideration as to whether failures or violations are unlikely, independent and concurrent. See Appendix C for a more detailed discussion.

CP&L has not gone through this process, but has only considered a single failure, limited to the mispositioning of one fresh PWR fuel assembly.

2. CP&L fails to determine what failures are "unlikely, independent, and concurrent."

concern was the misplacement of a single fuel assembly.

45 When the envelope of criticality has been determined for a particular situation, such as the storage of PWR fuel in Harris pools C and D, application of the Double Contingency Principle requires a determination, for each failure or violation represented in the envelope, as to whether that failure or violation is unlikely, and whether it is independent of and concurrent with the other failures or violations represented in the envelope. For Harris pools C and D, the most significant failures or violations will be fuel mispositioning events and boron dilution events.

CP&L has failed to determine if these events are unlikely, independent, or concurrent.

3. CP&L assumes that mispositioning of fuel is an "unlikely" event when in fact it is likely.

In considering possible criticality accidents at Harris pools C and D, CP&L assumes that the mispositioning of fuel is an unlikely event. CP&L offers no evidence to support this assumption. In fact, as shown in Appendix B and discussed in Appendix C, experience shows that fuel mispositioning is likely. Moreover, in a criticality accident involving fuel mispositioning and soluble boron dilution, these events will typically be consecutive rather than concurrent. High-reactivity fuel could be mispositioned in a fuel pool prior to or after a boron dilution event, or at both times if an event sequence involving mispositioning of multiple fuel assemblies spans a time period during which boron dilution occurs. Were CP&L to treat fuel mispositioning as a likely occurrence, then the criticality analysis would necessarily consider fuel mispositioning in combination with a complete absence of soluble boron, even employing the invalid, non-conservative version of the double Contingency Principle which is articulated in the Kopp Memorandum. Similarly, were CP&L to consider mispositioning and soluble boron dilution as consecutive occurrences, the criticality analysis would necessarily consider these occurrences in combination. Calculations by CP&L and the NRC Staff, summarized in

46 Appendix C, show that mispositioning of a single fresh PWR fuel assembly in Harris pools C or D would, in the absence of soluble boron, cause Keffective to exceed the regulatory limit of 0.95.

Mispositioning of more than one assembly could result in a supercritical configuration, potentially critical on prompt neutrons alone.

4. CP&L unreasonably assumes that a single error can lead to the mispositioning of only one fuel assembly.

In considering the role of fuel mispositioning as a potential cause of criticality, CP&L has restricted its attention to the mispositioning of only one PWR fuel assembly. Underlying this restriction is an assumption that a single failure or violation will lead to the mispostioning of only one fuel assembly. In fact, as demonstrated in Appendix B and discussed in Appendix C, experience shows that a single error can lead to the mispositioning of multiple fuel assemblies.

In addition to its improper reliance on administrative measures for criticality control, CP&L's misapplication of the Double Contingency Principle in the manner discussed above has yielded a criticality analysis that is non-conservative and inadequate to provide a reasonable assurance that public health and safety will be protected in the event of an accident. Whether or not the administrative measures chosen by CP&L are approved by the Licensing Board as consistent with GDC, CP&L's methodology for performing its criticality accident analysis must be rejected as inconsistent with valid and applicable NRC Staff guidance.

47 V. CONCLUSION For the foregoing reasons, the criticality prevention measures proposed in CP&L's License Amendment Application for the expansion of spent fuel storage capacity at Harris must be rejected as inconsistent with GDC 62 and valid and applicable NRC Staff guidance.

Moreover, CP&L's criticality prevention measures are demonstrably insufficient to provide a reasonable level of protection to public health and safety.

Orange County has demonstrated that the License Amendment Application must be rejected as a matter of law. If the Board declines to reject the application as a matter of law, it should find that Orange County has raised material and substantial issues of law and fact, and order the parties to proceed to an adjudicatory hearing on Contention TC-2.

Respectfully submitted, Diane Curran HARMON, CURRAN, SPIELBERG, & EISENBERG, L.L.P.

1726 M Street N.W., Suite 600 Washington, D.C. 20036 202/328-3500 Counsel to Orange County Gordon Thompson, Ph.D.

Executive Director INSTITUTE FOR RESOURCE AND SECURITY STUDIES 27 Ellsworth Avenue Cambridge, MA 02139 Expert witness for Orange County

48 I, Dr. Gordon Thompson, declare under penalty of perjury that the technical facts presented in the above Summary and Sworn Submission, including its appendices, are true and correct to the best of my knowledge and that all expressions of opinion regarding technical matters are based on my best professional judgment.

Gordon Thompson, Ph.D.

January 4, 2000

Appendix A The Double Contingency Principle

1. Introduction In addressing the potential for inadvertent criticality in spent fuel pools, the Nuclear Regulatory Commission (NRC) Staff and the American Nuclear Society (ANS) have employed the concept of a "double contingency principle". This appendix describes and compares the versions of this concept that have been articulated by the NRC Staff and the ANS.

2. The Grimes Letter In 1978, the NRC Staff issued guidance for spent fuel pool modifications, entitled "Review and Acceptance of Spent Fuel Storage and Handling Applications."

The guidance was attached as Enclosure No. 1 to an April 14, 1978 letter from Brian K Grimes to "All Power Reactor Licensees." This letter and its enclosures are hereafter described as the "Grimes letter". In addressing the potential for a criticality accident, the Grimes letter states:

"The double contingency principle of ANSI N 16.1-1975 shall be applied.

It shall require two unlikely, independent, concurrent events to produce a criticality accident."

Id., Enclosure 1 at page III-1.

Thus, the Grimes letter states that a criticality analysis must demonstrate that two unlikely, independent, concurrent events must occur before there is a criticality accident.

Immediately following the statement quoted above, the Grimes letter goes on to suggest that:

"Realistic initial conditions (e.g., the presence of soluble boron) may be assumed for the fuel pool and fuel assemblies."

The concept of "realistic initial conditions" is not defined in the Grimes letter, and is therefore open to interpretation. It is not plausible that the authors of the Grimes letter intended to say that soluble boron concentrations will never fall below their specified level. Instead, the Grimes Letter reasonably presumes that,

Appendix A The Double Contingency Principle PageA-2 at the outset of an accident sequence, conditions in the spent fuel pool will be in a "normal" range.

Any sequence of events that leads to a criticality accident in a fuel pool will have an end point, namely the criticality event. By projecting backward in time from the end point, one will always be able to identify an earlier point in time at which the pool's characteristics were in their normal range. For example, at this earlier point, the concentration of soluble boron in the pool water would have been as specified by licensee procedures or Tech Specs. One could reasonably describe the conditions at the earlier point in time as realistic initial conditions.

As a sequence of events unfolds toward a criticality accident, conditions will change in a manner specific to that sequence. For example, the concentration of soluble boron in the pool water might fall, and this occurrence might be preceded or followed by placement in the pool of fuel assemblies with a higher than-specified reactivity. Alternatively, an earthquake or the falling of a large object into the pool might reduce the center-center distance in the fuel racks. To apply the double contingency principle, as articulated in the Grimes letter, one must identify "events" of this kind and determine if they are "unlikely",

"independent" and "concurrent".

3. Draft Regulatory Guide 1.13 The double contingency principle was re-stated and revised in Appendix A of Proposed Revision 2 to the NRC staff's Draft Regulatory Guide 1.13, dated December 1981, titled "Spent Fuel Storage Facility Design Basis". Paragraph 1.4 of Appendix A states:

"At all locations in the LWR spent fuel storage facility where spent fuel is handled or stored, the nuclear criticality safety analysis should demonstrate that criticality could not occur without at least two unlikely, independent, and concurrent failures or operating limit violations."

This paragraph is broadly consistent with the statement of the double contingency principle in the Grimes letter, but there are two notable differences.

First, Paragraph 1.4 specifies "at least two" criticality-inducing events, whereas the Grimes letter specifies "two" events. This difference significantly strengthens the double contingency principle, as explained below. Second, Paragraph 1.4 refers to "failures or operating limit violations" whereas the Grimes letter refers to "events".

Appendix A The Double Contingency Principle Page A-3 The Draft Reg. Guide's use of the phrase "at least two" to modify the number of failures or violations that must be considered is significant, because it indicates that the drafters of the guidance were concerned about identifying potential interactions of causative events (failures or violations), beyond a single occurrence. 1 Thus, if a combination of two causative events is shown to cause criticality, and there is any possible doubt about the events being unlikely, independent and concurrent, then the Draft Reg. Guide indicates that this occurrence of criticality would be unacceptable.

Similarly, by referring to "failures or operating limit violations" rather than "events", the Draft Reg. Guide makes the double contingency principle more useful, by giving clearer guidance regarding the events that must be considered.

4. A Definition by the American Nuclear Society The ANS has provided a definition of the double contingency principle, although not specifically in the context of fuel management. This definition appears in ANS Standard ANSI/ANS-8.1-1983, "American National Standard for Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors",

approved October 7, 1983 and reaffirmed November 30, 1988. It should be noted that ANSI/ANS-8.1-1983 was endorsed by Revision 2 to the NRC staff's Regulatory Guide 3.4, "Nuclear Criticality Safety in Operations with Fissionable Materials at Fuels and Materials Facilities", dated March 1986.

ANSI/ANS-8.1-1983 defines the double contingency principle as follows:

"Process designs should, in general, incorporate sufficient factors of safety to require at least two unlikely, independent, and concurrent changes in process conditions before a criticality accident is possible."

Id. at page 3.

Note that ANSI/ANS-8.1-1983 is a revision of ANSI N16.1-1975, which is the ANSI standard that is cited in the Grimes letter.

1 Appendix C describes how a fuel pool's envelope of criticality can be determined. This envelope bounds the combinations of events that can cause criticality. Determining the envelope of criticality is a necessary precursor to applying the double contingency principle.

Appendix A The Double ContingencyPrinciple Page A-4

5. Another Statement by the American Nuclear Society A statement of the double contingency principle appears in ANS Standard ANSI/ ANS-57.2-1983, "American National Standard Design Requirements for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Plants",

approved October 7, 1983. In addressing the scope of criticality safety assessment, ANSI/ ANS-57.2-1983 states:

"At all locations where spent fuel is handled or stored, the nuclear criticality safety analysis shall demonstrate the criticality could not occur without at least two unlikely, independent and concurrent incidents or abnormal occurrences."

Id., Paragraph 6.4.2.1.4.

Similar language appears in ANS Standard ANSI/ANS-8.17-1984, "American National Standard Criticality Safety Criteria for the Handling, Storage, and Transportation of LWR Fuel Outside Reactors", approved January 13,1984, reaffirmed March 20, 1997. ANSI/ANS-8.17-1984 states:

"The fuel unit and rods should be handled, stored and transported in a manner providing a sufficient factor of safety to require at least two unlikely, independent, and concurrent changes in conditions before a criticality accident is possible."

Id., Paragraph 4.11.

In addressing the role of neutron-absorbing materials, such as boron, in preventing criticality, ANSI/ ANS-8.17-1984 states:

"Reliance may be placed on neutron-absorbing materials, such as gadolinium and boron, that are incorporated in the fuel material itself, or in structures or equipment, or in both. However, when reliance is placed on neutron-absorbing materials, control shall be exercised to maintain their continued presence with the intended distributions and concentrations. Extraordinary care should be taken with solutions of absorbers because of the difficulty of exercising such control and with fuel units containing burnable poison to identify the maximum reactivity condition to be considered."

Id., Paragraph 4.9.

Appendix A The Double Contingency Principle Page A-5 ANSI/ANS-57.2-1983 provides specific guidance regarding the assumptions about soluble boron that should be made in a criticality analysis. At Paragraph 6.4.2.2.9, ANSI/ ANS-57.2-1983 states:

"The presence of a soluble neutron absorber in the pool water shall not be considered in the evaluation of ks for PC I, II and III. In the analysis for PC IV and V faults, the initial presence of soluble neutron absorber may be assumed, if it is normally used, until addition of unborated makeup begins."

(emphasis in original)

In this context, ks is the evaluated maximum neutron multiplication factor in the fuel racks. Plant Conditions (PC) I through V are defined at pages 2-3 of ANSI/ ANS-57.2-1983. PC I events are "those events that are expected to occur regularly or frequently in the course of normal operation at the facility". PC II events are those with an estimated frequency of a least 1 per 10 reactor-years. PC III events are those with an estimated frequency of at least 1 per 100 reactor-years but less than 1 per 10 reactor-years. An example of a PC III event would be a loss of offsite power for up to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. PC IV and V events "are not expected to occur during the life of the facility, but are postulated because their consequences would include the potential for the release of significant amounts of radioactive material". Their estimated frequency is between 1 per 1 million reactor-years and 1 per 100 reactor-years. An example of a PC IV or V event would be a loss of offsite power for up to 7 days.

6. A Current Interpretation by the NRC Staff In recent years the NRC staff has articulated, and used for licensing purposes, a particular interpretation of the double contingency principle. This interpretation is set forth in a regulatory guidance document attached to an internal NRC Staff memorandum by Laurence Kopp to Timothy Collins, dated August 19, 1998 (hereafter known as the "Kopp memorandum"). The Kopp memorandum articulates the double contingency principle as follows:

"The criticality safety analysis should consider all credible incidents and postulated accidents. However, by virtue of the double-contingency principle, two unlikely independent and concurrent incidents or postulated accidents are beyond the scope of the required analysis. The double-contingency principle means that a realistic condition may be

Appendix A The Double Contingency Principle Page A-6 assumed for the criticality analysis in calculating the effects of incidents or postulated accidents. For example, if soluble boron is normally present in the spent fuel pool water, the loss of soluble boron is considered as one accident condition and a second concurrent accident need not be assumed.

Therefore, credit for the presence of the soluble boron may be assumed in evaluating other accident conditions."

Kopp memorandum at page 4.

This interpretation has been employed by the NRC staff in approving amendments to operating licenses for a number of nuclear power plants. In illustration, consider the NRC's issuance on June 29,1998 of Amendments No.

102 and No. 80, respectively, to the operating licenses for Vogtle Units 1 and 2 (Facility Operating Licenses NPF-68 and NPF-81). Those amendments allowed an increase in Vogtle Unit 1 spent fuel storage capacity from 288 to 1,476 assemblies. The NRC Staff's accompanying Safety Evaluation Report addressed criticality analysis in the context of potential accidents, and indicated that the double contingency principle can be applied in that context. The report states:

"However, for such events, the double contingency principle can be applied. This states that the assumption of two unlikely, independent, concurrent events is not required to ensure protection against a criticality accident."

Id. at page 5.

The Kopp memorandum's articulation of the double contingency principle differs significantly from the statement in the Draft Reg. Guide, because it does not require the consideration of "at least two" unlikely, independent and concurrent events." It also substitutes the word "events" for the Draft Reg.

Guide's instruction to consider "failures or operating limit violations," thereby returning to the less-useful language of the Grimes letter.

Moreover, the Kopp memorandum provides incorrect guidance regarding the need to consider reductions in the concentration of soluble boron in the pool water. In the excerpt quoted above, the Kopp memorandum states that "credit for the presence of the soluble boron may be assumed in evaluating other accident conditions". This statement is incorrect because there could be situations in which a reduced concentration of soluble boron, occurring in combination with one other failure (e.g., the mispositioning of some fuel assemblies), causes criticality without the other failure being unlikely,

Appendix A The Double Contingency Principle PageA-7 independent and concurrent. The other failure might be likely (i.e., the "unlikely" requirement is not satisfied), might share an underlying cause with the reduced concentration of soluble boron (i.e., the "independent" requirement is not satisfied), or might precede or follow the reduction in soluble boron concentration (i.e., the "concurrent" requirement is not satisfied). In any of those situations, the Kopp memorandum would provide incorrect guidance.

7. A Comparison of the Various NRC and ANS Interpretations The sources cited here show two schools of interpretation of the double contingency principle. The first school of interpretation encompasses the Grimes letter, Draft Regulatory Guide 1.13, and the relevant ANS standards. The second school of interpretation encompasses the Kopp memorandum and the current licensing practice of the NRC Staff.

The first school says that at least two abnormal events must occur before there is criticality. 2 The second school says that a criticality accident is acceptable if it follows just one abnormal event. Moreover, the Kopp memorandum incorrectly advises that the presence of soluble boron can always be assumed in evaluating the potential for another event to lead to criticality.

Overall, the second school provides a significantly weaker standard of protection against inadvertent criticality. This divergence between the two schools is much more significant than the comparatively minor divergences of interpretation that exist within the first school.

Within the first school, the most detailed guidance for application of the double contingency principle is provided by ANSI/ANS-57.2-1983. This document provides, as described above in Section 5, specific guidance about the assumptions that should be made regarding the presence of soluble boron.

The guidance in ANSI/ANS-57.2-1983 may be useful, insofar as it does not conflict with the full application of the double contingency principle, as set forth in effectively identical language in Draft Reg. Guide 1.13 and ANSI/ANS-57.2 1983. Full application of the double contingency principle requires the determination of the envelope of criticality for the fuel pool in question, and the 2 The Grimes letter takes a minority position within the first school by not requiring "at least two" abnormal events. This discrepancy could be ascribed to the relatively early date of the Grimes letter. At that time, the complexities of criticality analysis may not have been fully appreciated.

Appendix A The Double Contingency Principle Page A-8 systematic evaluation of events represented in that envelope to determine if they are unlikely, independent and concurrent.

Appendix B Some Incidents Relevant to the Potential for Criticality in Fuel Pools INTRODUCTION This appendix describes a variety of incidents at US nuclear power plants, including mispositioning of fuel assemblies in spent fuel storage racks, other fuel management errors, a soluble boron dilution event, other errors in managing soluble boron, and erroneous criticality calculations. These incidents shed light on the potential for inadvertent criticality in fuel pools.

The original source of information on the incidents described here was a set of Licensee Event Reports (LERs) supplied to Orange County by the NRC Staff during discovery in the operating license amendment proceeding regarding CP&L's proposal to increase spent fuel storage capacity at the Harris nuclear power plant.

The historical record summarized here is almost certainly incomplete, for three reasons. First, the LERs supplied by the NRC Staff were not systematically selected through a search of the full body of LERs, and the NRC Staff does not keep a database of incidents relevant to mispositioning of fuel or the dilution of soluble boron. Second, each relevant incident that has been identified by a nuclear plant licensee was not necessarily reported to the NRC by submission of an LER. Third, it is highly likely that a significant number of relevant incidents have occurred but have not been identified by the responsible licensee.

The remainder of this appendix consists of a set of incident descriptions. The descriptions are arranged by alphabetic order of the plants where the incidents occurred.

Braidwood Unit 1: August 21, 1996 and March 25,1997 (Licensee Event Report 456/96-010-02 (August 11, 1998))1 On August 21, 1996, an analysis of blackness test 2 data was received by the licensee, indicating shrinkage and gaps in the Boraflex in the spent fuel racks.

1 A copy of this LER is attached as Exhibit A-1.

Appendix A Some Incidents Relevant to the Potentialfor Criticality in Fuel Pools Page A-2 The largest gap exceeded the dimensions that had been assumed in the then current criticality analysis. This situation arose because of deterioration of the Boraflex. In response, the licensee initiated the process of requesting a license amendment to allow credit for soluble boron as a means of criticality control.

On March 25, 1997, a modelling deficiency was identified in a criticality analysis dated October 31, 1996. That analysis had incorrectly assumed that Boral poison panels are located on all four faces of all storage cells in Region 1 of the spent fuel pool. The same assumption had been carried forward through successive criticality analyses since 1987. In fact, the peripheral Region I cells do not have Boral panels on their exterior faces.

Braidwood Unit 1: July 10,1996 (Licensee Event Report 456/96-008-00 (August 5, 1996))3 During the verification of spent fuel pool storage locations, it was discovered on July 10, 1996 that one fuel assembly stored in Region 2 did not comply with a Tech Spec requirement that the assembly should be stored in a checkerboard configuration, based on its burnup level. Contrary to that requirement, the assembly was stored in a close-packed configuration.

The non-complying fuel assembly had been discharged from the reactor core on October 11, 1991 and relocated to Region 2 of the pool on June 16, 1992. Initially, its storage configuration met Tech Spec requirements for burnup. Those requirements became more stringent on January 20, 1995, at which time the assembly should have been relocated to Region 1 or to a checkerboard configuration in Region 2. Neither step was taken, because the burnup of this assembly was incorrectly entered into a spreadsheet program that was used to determine if assemblies were stored appropriately. The spreadsheet calculations were not independently verified.

2 Blackness testing is a technique in which a neutron source is used to evaluate the degradation of Boraflex neutron-absorbing material in spent fuel storage racks.

3 A copy of this LER is attached as Exhibit A-2.

Appendix A Some Incidents Relevant to the Potentialfor Criticalityin Fuel Pools Page A-3 Braidwood Unit 1: June 17,1996 (Licensee Event Report 456/96-007-00 (July 15,1996))4 On June 17,1996, while spent fuel assemblies were being repositioned in the spent fuel pool, the Fuel Handling Supervisor noted a fuel configuration in Region 2 of the pool that had a potential for criticality that was not bounded by the existing criticality analysis. This configuration had been specified by the Nuclear Material Custodian on May 9, 1996, and the configuration had then been accepted by two independent reviewers, on May 11, 1996 and May 15, 1996. The licensee attributed this incident to personnel error, and to procedural and management deficiencies.

Neither the number of assemblies involved in this incident, nor the details of the configuration, are stated in LER 456/96-007-00. The potentially critical configuration involved the interface between: (a) fuel whose burnup level allowed it to be placed at any location in Region 2; and (b) fuel whose burnup level required that it be checkerboarded. Calculations performed for the licensee indicated that criticality in this configuration would be suppressed by the presence of soluble boron in the pool water at a concentration exceeding 300 ppm.

In addition, the LER reports that a licensee review of plant records revealed one previous instance of fuel mispositioning. In that instance, fresh fuel was mispositioned in the spent fuel pool during transfer from the New Fuel Storage Vault. The cause was attributed to "personnel error due to a lack of a questioning attitude and failure to follow procedures."

Browns Ferry Unit 2: September 14,1980 (Licensee Event Report (October 9, 1980))5 During a refuelling outage, two fuel assemblies in the core were found to be rotated 90 degrees from their correct orientation. These two assemblies were among sixteen assemblies that had been loaded with an incorrect orientation during the previous refuelling outage. During that outage the incorrect orientation was detected for each of the sixteen assemblies, but was corrected for only fourteen assemblies. Thus, two assemblies remained in an incorrect orientation until the next outage.

4 A copy of this LER is attached as Exhibit A-3.

5 A copy of this LER is attached as Exhibit A-4.

Appendix A Some Incidents Relevant to the Potentialfor Criticality in Fuel Pools Page A-4 Byron Station: May 28, 1996 (Licensee Event Report 454/96-008-00 (June 25, 1996))6 On May 28, 1996, three fuel assemblies were found to be present in Region 2 of the spent fuel pool without meeting Tech Spec requirements. The assemblies did not meet the minimum burnup requirements, nor were they checkerboarded.

The required (actual) burnups (in MW-days per tonne U) were: 32,651 (32,648);

32,651 (32,638); and 32,771 (32,728). Two of the three non-complying assemblies were placed in Region 2 in August 1993, and the third assembly was placed in Region 2 in January 1995.

In the period August-November 1994, Byron Station engineers had built a computer spreadsheet to calculate assembly compliance with criteria for placement in Region 2. This spreadsheet did not detect the non-compliance of the three assemblies, because the spreadsheet was loaded with incorrect data for the assemblies' initial enrichment, storage location, and burnup.

When first placed in Region 2, each of the three assemblies was in compliance with minimum burnup requirements as then calculated. Subsequent re calculations led to increased minimum burnup requirements (operative in December 1994), which put the assemblies out of compliance. Although the degree of non-compliance was relatively small, it is significant that the non compliance arose from faulty data entry and was not detected for a long period.

Byron Station: July 15,1994 (Licensee Event Report 454/94-006-00 (August 15, 1994))7 On July 15, 1994, one fuel assembly was found to be present in Region 2 of the spent fuel pool without meeting Tech Spec requirements. The assembly did not meet the minimum burnup requirements, nor was it checkerboarded. The required (actual) burnup (in MW-days per tonne U) was: 32,540 (29,770). The non-complying assembly was placed in Region 2 in September 1993.

The Nuclear Materials Custodian (NMC) mistakenly allocated two non complying fuel assemblies for placement in Region 2. This mistake arose because inappropriate procedures were used for assembly allocation. A reviewing engineer detected the NMC's mistake for one fuel assembly but not the other.

6 A copy of this LER is attached as Exhibit A-5.

7 A copy of this LER is attached as Exhibit A-6.

Appendix A Some Incidents Relevant to the Potentialfor Criticalityin Fuel Pools Page A-5 The reviewing engineer's failure to detect both of the NMC's mistakes arose from the reviewing engineer's use of inappropriate procedures.

Catawba Unit 1: March 5, 1990 (Licensee Event Report 413/90-016-00 (April 19, 1990))8 The Boric Acid Tank (BAT) and the Refueling Water Storage Tank (FWST) were major sources of borated water at the plant. On February 5, 1990 the plant's Chemistry Department was informed by operations personnel that the BAT was the declared source of borated water. From February 5 through February 26, 1990, the Chemistry Department took samples from the BAT and the FWST, to comply with Tech Spec requirements.

During the period March 5 through March 12, 1990, the Chemistry Department failed to take a sample from the FWST as required by the Tech Specs. During that period the Chemistry Department continued to believe that the BAT was the declared source of borated water. On March 14, 1990 the Chemistry Department contacted operations personnel to confirm this belief, but was informed that the BAT had been inoperable since March 1, 1990.

The licensee attributed this incident to personnel error and deficient communication between departments.

Cooper Station: November 18, 1986 (Licensee Event Report 298/86-034-00 (December 18, 1986))9 On November 18, 1986, during a refuelling outage, it was discovered that fresh fuel with a U-235 loading in excess of the Tech Spec limit had been stored in the spent fuel pool during three cycles of plant operation. The Tech Spec limit on U 235 loading was 14.5 grams per axial centimeter.

During Cycle 7, fresh fuel with a U-235 loading slightly higher than the Tech Spec limit was stored in the spent fuel pool between February 3, 1981 and April 27, 1981. The same phenomenon occurred during Cycle 10, between July 23, 1984 and July 17, 1985. During Cycle 11, fresh fuel with a U-235 loading of 14.6 grams per axial centimeter was stored in the spent fuel pool for some period prior to the determination on November 18, 1986 that the Tech Spec limit had been violated.

8 A copy of this LER is attached as Exhibit A-7.

9 A copy of this LER is attached as Exhibit A-8.

Appendix A Some Incidents Relevant to the Potentialfor Criticalityin Fuel Pools Page A-6 The Tech Spec limit of 14.5 grams per axial centimeter on U-235 loading was introduced in June 1978 as part of Tech Spec amendments that provided for installation of high-density fuel racks in the spent fuel pool. Criticality calculations performed at that time were based on a fuel design for which the U 235 loading was 14.5 grams per axial centimeter.

Crystal River Unit 3: November 9,1987 (Licensee Event Report 302/87-026-00 (December 1, 1987))10 On November 9, 1987, the reactor vessel was completely defuelled. It was discovered that a fresh fuel assembly with a U-235 enrichment of 3.85 % had been placed in the "A" spent fuel pool. The Tech Spec limit on the enrichment of fuel in the "A" pool was 3.5%.

This event occurred because a mistaken entry was made on a Fuel/Control Assembly Move Sheet. The intention was to move an assembly from location M42 in the "B" spent fuel pool to the "A" spent fuel pool. The assembly in location M42 would have complied with the Tech Spec requirements for placement in the "A" pool. Location M43 was mistakenly entered on the Move Sheet, leading to transfer of the non-complying fresh fuel assembly from the "B" pool to the "A" pool. This transfer was detected about 80 minutes after its occurrence.

Hope Creek Station: December 12, 1995 (Licensee Event Report 354/95-042-00 (March 25, 1996))"

On December 12, 1995, during a refuelling outage, a visual inspection of the reactor core revealed that one fuel assembly was 180 degrees out of its proper orientation. The mis-oriented assembly had not been moved since its emplacement on April 3, 1994. A visual inspection of the core had been performed at the time of emplacement, using a video camera. This inspection had not detected the mis-orientation of the assembly. A previous mis-orientation at Hope Creek had been detected during post-emplacement inspection.

10 A copy of this LER is attached as Exhibit A-9.

11 A copy of this LER is attached as Exhibit A-10.

Appendix A Some Incidents Relevant to the Potentialfor Criticalityin Fuel Pools Page A-7 McGuire Unit 1: July 11, 1994 (Licensee Event Report 369/94-005-00 (August 10, 1994))12 On July 10, 1994, while the reactor was at 100% power, plant personnel began to drain the spent fuel pool transfer canal. During the drain-down, a water misting system was used to keep the walls of the transfer canal wet to minimize potential airborne contamination. This misting system added demineralized, un-borated water to the transfer canal. During the drain-down, the spent fuel pool was separated from the transfer canal by a gate. Drain-down was accomplished by lowering a submersible pump into the transfer canal. It appears that the discharge from the submersible pump was directed into the pool.

By a route not specified in LER 369/94-005-00 (but presumably via the submersible pump), approximately 28,000 gallons of demineralized, un-borated water were added to the spent fuel pool during the drain-down process. This occurred on July 10 and 11, 1994. According to measurements performed on July 12,1994, the addition of the demineralized water to the pool had lowered the soluble boron concentration in the pool from 2,105 ppm to 1,957 ppm. The Tech Specs require a boron concentration in the pool of 2,000 ppm.

The licensee attributed this incident to a variety of personnel errors and procedural deficiencies. The LER states: "Personnel interviewed did not have a good understanding of their responsibilities associated with Reactivity Management."

McGuire Unit 1: October 24,1991 (Licensee Event Report 369/91-016-00 (November 25, 1991))13 Plant personnel discovered that 11 fuel assemblies had been stored in the spent fuel pool in a manner contrary to Tech Spec requirements. These requirements stipulated that, if a checkerboard pattern was used in Region 2 for storage of fuel that would have been non-complying if not stored in a checkerboard pattern, then one row between normal storage locations and checkerboard locations would remain vacant. The requirement for a vacant row was not satisfied from March 23, 1990 through October 23, 1991. The licensee attributed this error to poorly written procedures.

12 A copy of this LER is attached as Exhibit A-11.

13 A copy of this LER is attached as Exhibit A-12.

Appendix A Some Incidents Relevant to the Potentialfor Criticalityin Fuel Pools Page A-8 It should also be noted that 9 of the 11 previously designated fuel assembly locations were changed on March 23, 1990 in order to maximize the number of open locations in anticipation of a core offload.

Millstone Unit 2: February 14,1992 (Licensee Event Report 336/92-003-01 (June 25, 1992))14 On February 14, 1992 it was discovered that a calculational error existed in the criticality analysis for the Region 1 spent fuel storage racks. The originally calculated value of Keffective was 0.922. The newly calculated value of Keffective, for the same conditions, was 0.963. This error arose from the use of two inappropriate assumptions in the earlier calculations.

Oconee Unit 1: January 8,1996 (Licensee Event Report 269/96-001-00 (February 7, 1996))Is On December 14, 1995, a fuel assembly was lifted from its location in the spent fuel pool, so that the assembly could be visually inspected. After the inspection, the assembly remained suspended from the refuelling bridge. This situation was discovered on January 8,1996 by two fuel handlers who were starting preparations for loading a dry cask some days later.

The two fuel handlers proceeded to lower the suspended assembly into the open location immediately beneath the assembly. Their intention was to allow an identification of the assembly in order to determine its correct location and to trace its previous movements. Through this action the fuel handlers returned the assembly to its location of December 14, 1995, although they did not know this prior to lowering the assembly.

The licensee reviewed previous operating experience, industry-wide and at the Oconee site, in an effort to identify related incidents. Findings from this review were summarized in Attachment A of LER 269/96-001-00, but with limited supporting detail. Some of the information in Attachment A is excerpted in the following two paragraphs.

Four related NRC Level IV Violations were recorded at Oconee in the period 1992-1995, as follows: (a) in November 1990, a fuel assembly was placed in a wrong location in the reactor core; (b) a similar event occurred in February 1993; 14 A copy of this LER is attached as Exhibit A-13.

15 A copy of this LER is attached as Exhibit A-14.

Appendix A Some Incidents Relevant to the Potentialfor Criticalityin Fuel Pools Page A-9 (c) in September 1991, a fuel assembly was placed in an incorrect location in the spent fuel pool; and (d) in August 1994, a refuelling sequence was altered without proper documentation and procedural control, and a fuel assembly was retrieved from an incorrect location in the spent fuel pool and placed in the reactor core.

Related incidents identified from industry-wide experience included: (a) several fresh fuel assemblies were received and placed in incorrect rack locations; (b) six fuel assembly mispositioning events occurred during refuelling and defuelling operations; (c) unauthorized movement of a defective, encapsulated spent fuel rod occurred; (d) four events occurred which involved inadequate oversight of refuelling operations and inadequate performance by refuelling personnel; (e) a control rod was inserted in the wrong fuel assembly; and (f) six events occurred that involved human performance deficiencies while reactor core components were being handled.

Oyster Creek Unit 1: January 21, 1987 (Licensee Event Report 219/87-006-00 February 24, 1987))16 On January 21, 1987 it was discovered that fresh fuel with an enrichment higher than the Tech Spec limit had been stored in the spent fuel pool, beginning on February 27, 1986. The Tech Spec limit on average planar enrichment was 3.01 wt% U-235.

A total of 204 fresh fuel assemblies, with an average planar enrichment of 3.19 wt% U-235, were received at the plant in 1986. The dry storage vault had a capacity for 140 assemblies. Thus, 64 fresh assemblies were initially stored in the spent fuel pool. As the refuelling outage progressed, more assemblies were taken out of the dry storage vault, channelled, and stored in the spent fuel pool.

Ultimately, 184 noncompliant fresh assemblies were stored in the spent fuel pool prior to the start of core reload in August 1986. By the time the core had been fully reloaded (on September 14, 1986), all of the fresh fuel had been removed from the spent fuel pool.

The licensee ascribed this occurrence to personnel error. Specifically, the plant's safety analysis did not take into account the possibility that fresh fuel would be stored in the spent fuel pool.

16 A copy of this LER is attached as Exhibit A-15.

Appendix A Some Incidents Relevant to the Potentialfor Criticality in Fuel Pools PageA-10 Susquehanna Unit 1: October 6, 1993 (NRC Information Notice 94-13, (February 22, 1994))17 During reactor defuelling operations, personnel performing the fuel handling activities removed an incorrect fuel assembly from a peripheral location in the reactor core. On becoming aware of this error, the personnel involved returned the assembly to its prior position in the core. That action was contrary to licensee procedures, which required that: (a) the assembly was to be placed in the spent fuel pool; and (b) fuel handling activities were to be halted until the cause of the error was determined and corrected.

Three Mile Island Unit 1: February 4, 1998 (Licensee Event Report 289/98-002 01 (April 3, 1998))18 Tech Specs at this plant require sampling of spent fuel pool water for soluble boron content, both monthly and between 24 to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after completion of each water addition. On January 23,1998, water was added to the pool between 0918 and 1705 hours0.0197 days <br />0.474 hours <br />0.00282 weeks <br />6.487525e-4 months <br />, but no sample was subsequently taken within the specified time period. A further water addition was made on January 27, 1998 between 1410 and 1817 hours0.021 days <br />0.505 hours <br />0.003 weeks <br />6.913685e-4 months <br />. The pool was then sampled at 0430 hours0.00498 days <br />0.119 hours <br />7.109788e-4 weeks <br />1.63615e-4 months <br /> on 28 January 1998 and again at 0830 hours0.00961 days <br />0.231 hours <br />0.00137 weeks <br />3.15815e-4 months <br /> on January 29, 1998. On February 4, 1998 a Staff Chemist noticed that this sampling sequence did not meet Tech Spec requirements for timely sampling after the January 23 water addition.

The licensee attributed this incident to personnel error and the absence of a warning sign that was supposed to be attached to the wall directly behind the valve used to fill the spent fuel pool. The missing sign would have reminded personnel to notify the Chemistry Department of the need for sampling.

A previous failure to perform sampling after a water addition to the pool had occurred in June 1996. In response to that failure, the licensee had modified the plant procedures. One of the modifications was to require placement of a warning sign -- the same sign that was absent in January 1998.

17 A copy of this Information Notice is attached as Exhibit A-16.

18 A copy of this LER is attached as Exhibit A-17.

Appendix A Some Incidents Relevant to the Potentialfor Criticality in Fuel Pools Page A-1I Waterford Station: February 18, 1994 (NRC Information Notice 94-13, Supplement 1 (June 28,1994))19 While the reactor was at 100% power, an unknown object was found hanging from the fuel-handling machine in the fuel-handling building. The object was subsequently identified as a capsule containing a defective fuel rod that had been removed from an irradiated fuel assembly several years earlier and then stored in a rack in the spent fuel pool.

Licensee investigations suggested that the capsule had become attached to the fuel-handling machine during unauthorized use of the machine between February 11 and February 18, 1994. The licensee speculated that one of the people assigned to prepare for a March 1994 refuelling outage had inadvertently lifted the capsule while practicing the use of the hoist. No keys or special knowledge were needed to operate the fuel-handling machine. None of the personnel questioned by the licensee admitted to unauthorized use of the machine.

This Information Notice offered some suggestions to licensees to prevent unauthorized or unintended use of fuel-handling equipment, including locking circuit breakers in a deenergized position and placing placards that warn against unauthorized use.

Various plants and incidents (NRC Information Notice 94-13 (February 22, 1994))20 Various fuel-handling incidents occurred at Vermont Yankee, Peach Bottom, Susquehanna and Nine Mile Point during the period September-November 1993.

This Information Notice drew a generic lesson as follows:

"Refueling activities are safety-significant operations that are not conducted on a routine basis. In addition, fuel handling activities are often performed by contractor personnel under the supervision of licensee personnel. As a result, fuel handling personnel may not be familiar with the fuel handling equipment or may feel that their experience in fuel handling operations permits them to ignore some requirements for procedural use and adherence."

19 A copy of this Supplement is attached as Exhibit A-18.

20 See Exhibit A-16.

EXHIBIT B- I Braidwood Unit 1:

LER 456/96-010-02 (August 11, 1998)

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Fuel Racks received on Analysis of Neutron Attenuation test data for Braidwood's Spent 6/21/96, shows Boraf lox shrinkage and gaps. The largest gap has a width of greater than four inches. A gap of greater than four inches in any Boraflex panel exceeds that assumed in the current criticality analysis. The spent fuel storage racks are designed to maintain a Keff S 0.9S when flooded vith unborated water. The cause of this event was determined to be material selection.

failure of the Boraflex duo to deterioration as a result of improper concentration and silica Corrective actions include controls on Spent Fuel Pool (SFP) boron that there is concentration. The safety analysis contained in this report concludes S 0.95.

reasonable assurance that the Braidwood SFP will maintain a Keff CAC-96-248, "Byron On 3/2S/97, a modeling deficiency was identified in criticality analysis Credit for Soluble Boron", dated and Braidwood Spent Fuel Rack Criticality Analysis with located on all four faces October 31. 1996. This analysis assumed Boral poison plates were The criticality model did not reflect the actual (as of all Region 1 storage cells.

on the interior designed) configuration of the Boral poison plates, which are located storage racks but are not present on the periphery of the portions of the new Region I fuel Region I storage cells. Subsequent to the discovery of this modeling deficiency.

analyses for the actual Region 1 cell Boral geometries were supplemental criticality

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Unlt(s): 1 Event Date: 08/21/96 Event Time: 1224 Hours Reactor Mode(&): 1 Power Level(s): 100t RCS (ABl Temp./Press. NOT I NoP Unit(a): 1 Event Date: 03/25/97 Event Time: 1700 Hours Reactor Node(s): 1 Power Level(s): 100% RCS (AD) Temp./Press. NOT / moP

31. D=CUZP W iorVVIWT:

"There were no systems or components inoperable at the beginning of this event that contributed to the severity of the event.

On /21/96, analysis results of Neutron Attenuation (Blackness) test data were received at Braidwood Station, indicating shrinkage and gaps in the Boraflex in the spent fuel racks. The largest gap has a width. greater than four inches. A gap of greater than four inches in any Boraflex panel exceeds that assumed in the current criticality analysis. An ENS phone call was made at 1349.

The Spent Fuel Pool (SrP) at Braidwood Station has fuel racks installed that utilize sheets of Boraflex for reactivity suppression. Boraflex is constructed of an organic polymer with a silica filler and neutron absorbing boron carbide interspersed within the silica filler.

In 1967, ConmEd first identified gamia-radiation induced damage to the Boraflex polymer.

'rhe damage progresses through two stages. First, the Boraflex cracks and shrinks.

producing cracks and gaps. The second phase occurs after the polymer has sustained significant damage, and consists of the Boraflex becoming brittle and susceptible tZ dissolution in the Spent Fuel Pool cooling water.

The reactivity effects associated with the first stage have been characterized in the

"'Byron and Braidwood Spent Fuel Rack Criticality Analysis Considering Boraflex Gaps and Shrinkage,m Westinghouse, June 1994, supplemental criticality analysis. Sufficient smargin exists within this supplemental criticality analysis to accommodate the anticipated levels of cracking and gapping associated with the first staqe o:

degradation.

The second stage of damage involves long-term degradation of the Boraflex. The second stage appears to commence after the Boraflex has received approximately 4E9 RADs of gamma exposure. There are a number of variables (burnup, cooling time, recent power history, etc.) that affect the exposure rate. The presence of silica in the SF?

cooling water is another indicator that storage locations have progressed into the second stage of damage. The reason for the uncertainty in the rack's condition lies in the degradation mecnanism associated with the second stage. The second stage involves slow dissolution of the Boraflex. The rate of dissolution is determined by the concentration of reactive silica in SFP solution, thermally-induced flow velocities.

and coolant temperature inside of the storage racks. The larger the panel spacing, tre stronger the local flow dnd thus the dissolution rate increases.

The recent Blackness Testinq campaigns at Byron and Braidwood indicate progress intz rhe second stage of damage has occurred, and that the maximum gap width allowed in the current criticality analysis has been exceeded.

N lC IVOMt M 5.i5 NUCLEAt lUDATOSV coSoINon AJYDovID UT 56 N4.* IIIee ETIMATED B.ID4 VEX RESPONSE TO COMPLY W1TU THIS 1INFOEMATIMONCLLWNO4 RU3ZLEST 900"3 NIPORTID AU VOIMATD W%1O THE L *.*&Ts ULo LICENSEE EVENT REPORT (E' ) U=ssoNs PRCS LERIAMNAND FID BC To L%1)tMV Po@W~nn TEXT CONTINUATION UEA auUmESYATE1 TOlM LFo(uRTlK lpO .%D

_nC05_5 M&A2*.*,*,"r UtcH ()36 I 14 FM UScLt..,.

EI'.ATORU cmoUSOt WASNtrTOK. U'IO4w..% n 3C 71E PAPERWORK RUMVflPON~CT Based on the above facts, I1CaIdmmd already has large numbers of storage locations in the second stage of degradation. The degradation mchanim associated with the second stage proceeds slowly, however It is both difficult to predict and measure the extent of damge. Although Slackness Testing Is useful tot measuring cracks, gaps, and vastage, It does not measure an overall reduction In boron density. Therefore slackness Testing provides incasplete Infocrmation regarding the current state of a given storage location. An approved methodology to measure boron spatial density does not currently exist for PM63.. Therefore, the gaps recently found at Braidwood Station my not represent the full extent of Boraflex degradation.

wen assessing the current state of the storage racks, the following factors, along with oeters, are considered: they include the slow nature of the degradation process, the continued presence of some Soraflex, the successful performance of the surveilLance coupon program, the inclusion of loCal in the Region I rack design, and the potential for adkdtional reactivity margins due to burn-up.

Coredr has performed calculations to support a short torm reccoendwatieo of maintaining greater than 2000 PIN1 soluble beron in the Spent fuel Pool to compensate for tue degradation of the soraflea. These calculations ave very. conservative. The 2000 PIp" limit is intended to approximate the total coactivity suppression worth of the installed Morailex in both the Region I and Region 2. fuel storage racks. Therefore.

even if all borafLex were to be rmoved from the Spent fuel Racks, the 2000 PP" value is adequate to maintain the Spent Iuel Pool at 1 0.95 Keff.

Based on the recent Blackness Test data, it cannot be stated with certainty that Technical Specification 5.6.1.1 La met. This specification otate. -rThe spent ifuL storage cocks are designed and shall be maintained with a Keft 1 0.9S when flooded wath unborated water,...'. Thecefore, the racks are in an 01ndetermAnate" state of operability as defined in IRC Generic Letter 91-11, they have been conservatLvvLy declared inoperable, and ompensatory measures that were initiated in 19 r#5 were verified.

This event is being reported peIcuant to 10CFrB0. 7 3(a ill (1&1(a) - any event or condition that resulted in the Condition of the nuclear power pLant Doing in 4 condition that was outside the design basis of the plant On 03/251/9', additional reviews by Coald identified a modllinLg dnetiCiMn:y in criticality analysis CAC-96-246. 'Byron and Staidwood Spent FueL Rack :rLý:.j*y.i AnalysLi With Credit for Soluble Bacon". dat-d October 31. L996. This anaLysLs was performed to sqpport Technical SpecLiicatLon Amendment No. SE for Byron Una - L a' and Amendment No. 79 for Braidwood Units I and 2. issued April ., 1997. This "*-.teem" CritLCaLit/ analysis was performed due to the degradation of the Socaflex in the spent fuel racks. The defLcLincy is due to inadequate modeling of lhe physi:a& :01CL'rat-r.1n of the Dora: panels within the Byron and Iraidwood Roqion L Fuel St*ra;e 4a-ts Due to Cotlrd5s con:erns regarding the industry's experiences with %or'a(. 1e; .1:6ta-,r 2

during the ,nAd-S('s, Boral panels were placed in the Cix* rraos if . c;.)n '.

during initial fabrication. The eral panels were ir,"'i*i ... -n- '..,

exist between each celE within 9 RegLon L ra-k. Z!ý! :4 - e..' .

included in the ass4pt ions af the n*odel for the Qenq)n it ranki I

(I4eI5J1 NR3~

ininUp MUMI ZO35UJATUSV CusssUMou Iill ArmUwwV

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-lI

" 355130,0104I

• IJIM p , ISIITUD BUlDID4 1U3 MIISII TOCOUNLY WITH THU DIOMATIONMCUIZOCNIWEMU 9 0I5S RME'I3D LICNSEE EVVENT R O T (LIR) LbONS uKMM. AN*D FqM IlTMY FOnrWAuA LIMMO AM 0WGSAUDMMTHE EDmBACK1O COMN UT TEXT CONTINUATION 3ZOARDIo BUM UD4 S TE Ta, 10S IFGbAnTIo.%k.D aCoIv ANA.ADCTMN NCU (14 )). I" %LVCL, AMUO~ATDMY cmomam wAs1L4GToK DC 3"54440Ln A%9 TM PAPERmRK REUXI-n POECT OMra

,a0d00ood Unit 1 ms kMadMjMd I ,llm 05000456 ofMdP*l~mw=3" 17) 90 0 1 A t o0 S The &rory is LnfinIte in lateral (x and yl extent. These Is no interface requ*lrmnmt* between Region I storage racks.

- SDoral poison platen vere on all four faces of all storage cells.

The criticality model did not reflect the actual (as designed) foiguration stfhe oegal plates. which age located on the interior portions of the new Megson I fuel storage rocks, but were not designed to be installed on the periphery of the Region I storage cells. Thus, Region I periphery storage cells actually contain Deral plates on only thrwe silsld and the four conor cOIls actually contain only two Interior Soeal plates.

In Norch 1997, the Westinghouse criticality engineer was coviewing the SFP rack perLpheral geometry in n attempt to regain storage locations that were lost due to constraints required In the 1996 analysis. Drawings of the Byron and Braidcood S3F and racks wer suppLled to Westinghouse at their request. an Karch 20. the Westinghouso criticality engineer contacted Coind Nuclear Fuel Services MIs) math a concern that there my not be moral on the peripheral walls of the "egIon I racks due to the girdle bar geoetry. As a result of subsequent 'Comd reviews of rack drawlags and discussions with the vendoral responsible for contr-uction and sel*emc analysis of the racks, Coiand -"

concluded that Dotal poison plates ware neither present on the peripbery of the PRAgion I storage cello nor designed to be In these Locations.

The cauen of this event was determined to be faLluCe of the boratlex due to detelioration as a result of imeropec material seLect&on.

tI 1907, CorEd first identified gaam radiation-induced damage to the Soraf lex polymer.

That damage progresses through two stages. rirst, the Soraflex cracks and shrinks.

producing cracks and gaps. Second, after the polymer has sustained significant damaqe, the Botafleo becomes brittle and ts susceptible to dissolution in the Spent ruel Pool cooling water.

The cause of the 3/25/97 event was determined to be the result of a modeling error in vendor pectormed criticality analyses, and inadequate reviews of the analyses input and assumptions against manufacturing drawings during the criticality sanlysts CevLeWs and verifications. The infinite array criticality analysis methodology wes not appropriate for the unique placement of Dotal in the Reqg-on I racks, and did not properly model the

-nterface between Region I racks.

Doral panels were placed in the flux traps of the Region I tacks during inLtial fabrication. The manufacturing drawings for the Region I racks ore inadequate to determine oreal panel placement, and no as-buLlt drawings of the Region I racks wore ever generated. The original criticality analysis for the Region L racks was performed in 1907 by a vendor. The criticaLity wmodels originally generated by the vendor, was transferred to Westinghouse, and was carried forward throuqh alL subsequent :nCtLca~ity analyses. The original analysis and all subsequent analyses assuam.d fair ;%eqon " an infinite array and Borel poison plates on all four fa:es of all *eiks. Cr-:a' 1 analysis AZC-96-240 assumed that the SOlCELex poLZ-'t was removed fr*m ".. stacar,.

racks, and that the boraf.ox was replaced with water Thir snalys., a"so nod-,ei --. q.

Region L storage cells with moraL panels on all four fa:.s. ThLs and previ-9s sna.y",.s did not specifically model the peripheral Pe0Lon I ce..% that do nz: have 3-*,ra. panf..s on their exterior faces.

I

(4953 flh~lSU3 iWl ESTIMATED 51D14ER RESPONSE TO COPLY WIMTHIruS OffORMATIOI4 COLLECTION R"MUMS 300U HIS EP03ThD LICENSEE EVENT REPORT (LER) uES UALDs W rtcLUAND F[DMBACTO AM TO"A5TUE

)LS 11 Y F 'ro FORWARD TilCoMrGE MVIMM.

TEXT CONTDM&ATION RGARnDcO BhLMnx ZUTMATE To mTE LFDLW'TK)% ,.%

R ANAGUIEWT MCN (4 U&ORD T6IOULU 5 %tWCLF.-R REGLILATORY COaSI. WARUaGTON. DC 20535.oOsm. AND" TIE PFAPEWORK REDMXN PROVIECT FACUJI NAME 4 (1) DOCKET IINIM1 a) LD MUND1I (6) PAGE ()

'n" I EQ.,U, . UI$W'b Braidwood Unit 1 05000456 96 1 010'02 1 55of II oflom spam isiuplid m a- tmd cenpm .NRC Fnm 366AMX)

D. AlUJMsMr OF SUMnr mmm Recent =Backness Testing Indicates that the degradation of the Braidwood Spent Fuel Packs exceeds that assoed in the criticality analysis. This could lead to a condition where the Technical Specification reactivity Limits for the SF? could be exceeded.

eased on a comparison with prior analyses by ConEd Nuclear PueL Services for the Iyron/araLdvocd reactor cares, maintaining SrP boron concentration 2'2000 PPM wFLj ensure that the requisemsto for maximum reactivity in the srP ace met, even assuming the Soreflex panels are ineffective from a reactivity mitigatLon standpoint. The analysis assumed enriched fuel with no burnup (e.g. maximum reactivity) In close proximity to other assemblies. The physical separation of assemblies in the Spent Fuel Packs Is greater than the searation in the core. Zn addition, the spent fuel assemblies are at much lower reoactLvity duo to burnup from incore opecaton. rotrthese reasons, there is reasonable assurance that the lraidwood Spent Fuel Fool maintains a Klff 9 0.95.

After discovery of the modeling de ficLency on 3/25/97, supplemental criticality, analyses for the actual Region I cell mBotal gosfetries were performed and demonstrated that. with administrative controls in place regarding boron concentration -and- tuell placement In Region I rack interface, acceptance criteria for spent fuel storage Is Met. The supplemental cciticalLty analyses utilized the same assumptions, codes, procedures, and uncettainties used to support the 1996 critLcaLity analysis ICAC-96 246) but with Bocal panels located only in the interlor cell-to-cell Interfaces. The supplemental analyses modeled the following Region I rack geometries:

1. Corner cell of rack facing two concrete wells.

2. Peripheral cell of rack facing one concrete wall.

3. bmpty row of cells facing a full row of cells across a Region I to Region I rack interface.

4. Checkerboard pattern of cells across a Region I to Region I rack Interface.

Calculations were performed foe the four rack goometries to verify that with a maximum nominal enrichment of U-235, that Koff io less thae 1.0. The analyses ignored the presence of Boraflex and accurately modeled Boral only on the interior rack faces.

This calculation was performed with no soluble boron assumed present in the SFP. The resuLting reactivitLes were compared to the all cell Keff calculated in section 3.2.1 of CAC-96-24S. The all ce:L Keff (from CAC-96-240) was verified to be greater than the reactivities calculated to: these four rack geometries. The biases and uncertainties calculated in CAC-96-24S remaLn valid for use with the four analyzed rack geometries.

By determining that the a&. cell Keff remains bounding, the conclusions at CAC-96-24S age applicable for the four analyzed rack geometries analyzed for the followinq acceptance criteria:

L. Assuming no so!Able boron, the maximum noninal enrLchmn*e o0 U-235 could be stored and a Ke~f of lass than 1.0 is maintained,

2. Taking credit ftr a muinimum concentration ot soluble boron of 2000 ppm. a Keff of less than or equal to 0.15 is maintained, and

3. Assuming the SF? water temperature postulated accident and :ak.nj -reiLt for a minimum :onceltcatLOL ot soluble boron of 210,3 ppm, a Ke!f! -, .ss than or equal t3 0.95 is maintained.

Additional cases for the misloaded assembly were performed. rhese cases were calculated at no soluble boron conditions and the resulting reactiVLties were shownl ýo

RM M~E~A 11.S. nUCLAARKGURATORV CONAUSrIfON A PPROV=D BY 0565 NO. 3)1"s4" ESTUMATED BLIUEN PER RESPONSK TO COaWLY 6TrH TIllS INFOIMATION COUSJACTM RE40LIW: 500 HRS REPORTED UCENSEE EVENT REPORT (LER) LUoNs FEDACKTO FocESSA.,%D m9ACK TORY

%*,*Y F01WAD THE. co.UL*,-rs COE%75 TEXCr CONTUIUATION K7DX ESTIMATE TO THE L'FOR.AT1. AND rEOAADWO RECoRD MA',U Mr MBRAWN (14 f)3) V S -'C,1AR REMULATORY cOMmSOK WASUJgTON. DC 2*@35.oO0. A.Sc THE PAPERWORK REU*XTN PECT FACUMrV NAM3(l) DOCK"T RUN=iis () LE.*.UlBA 5(6)) PACK(3) va SMNnAL I U%1*o*%

taidwcood Unit 1 addiuss d qm is reqhdt, - Wopaw (I- m e q=

05000456 oNRC Foe. 366AX 17) 96 1 010 1 02 I 6 of le be less than the a11 call Kff f from CAC-96-248. The dropped assembly accidents are not affected by the Dlotal configuration.

An additional criticality analysis was performed taking credit for 2000 ppm soluble boron and no Doral and no Boraflex present in the spent fuel racks. This analysis verified that Keff was less than or equal to 0.95 for all storage locations based on fuel assembly locations at the time of the event discovery.

The supplemental crLticality analyses are conservative since, in reality, an appreciable amount of Doraflex remains in place in addition to the administrative requirement to maintain at least 2000 ppm in the SrP. It is concluded that the safety analysis impact due to the Incorrect modeling of the Boral confLguratLon is minimal.

1. e nACTIONS:

The following are actLons being taken to either minimize the Boraflex degradation or mitigate the effects of Boraflex degradation.

Evaluation has shown that 2000 PP1 soluble Boron will compensate, for even fully deterLogated Botaflex. Therefore, Braidwood will administratively maintain >2000 PPX soluble Boron until further review of the Doraflex Issue. This will be tracked by NTS Item 1456-180-96-01001.

This item has been completed.

Spent Iuel Pool silica reduction using Reverse Osmosis will be restricted until the licensing amendment to allow for soluble boron credit is approved. This will be tracked by "TSitem 4456-180-96-01002.

This item has been completed.

The long term corrective action for this situation consists of submittal of a Licensing amendment to allow soluble boron to be credited in maintaining the pool S 0.95 Keff.

The analysis for this amendment is in progress. Submittal to the NRC is expected in mid-1997. This will be tracked by HTS Item 1456-180-96-01003.

This Item has been completed.

ComEd has created a Boraflex Issue Committee to work with the industry to resoLve this Issue.

Resolution: The revised Braidwood Criticality Analysis does not credit Boraflex. This item has been completed.

An effectiveness review will be performed for all corrective actions listed above.

This will be tracked by HTS item 0456-190-96-010ER.

This Item has been completed.

As a result of the supplemental criticalLty analyses for the actual Region 1 :eiL. BoraL geometries, the following administrative control has been impLemented kL s' a t1zn procedures fBwAP 2364-9, paragraph C.L). This change is tracked by NTS item S 4T-a.

96-0105101. -No assembly may be placed in a Region . ra-k ,ocat.on face *-'a r.: :

another assembly across a Region 1 rack interface.

This item has beer completed.

I

ESIMATED LUNL' .*ER RESPONSE TO COMPLY W,'r THIs nOORUATnIoCOECTION uREEST. o0O HIS RPORE*D LICENSEE EVENT REPORT (LER) LESON LEA1LNED AM INCORFORATEDIm70 THE uCEL.%5sNo FRocEss AND) nI TO [NDUTRY FORWARD COUMFN.S TEXT CONTNATION REGARDIO OILM mi FOILRATK)% *VD ESTIMATE TOTHE RECORDS MANAGEMENT* B34MM (14 I331 L'S .I'CLEAR REGRJLATOIRYCOMMzSS*. WASHLT. DC Z05-54001, A.%TT THE PAPERWORK ItDCfON PROJECT PAaCK= XAMEt() *NUMUIR* ()

DOCKT LEE NU,-ER () P.AGE (1) vimS 38OWNmIAL 0XVISM braediood Unit 1 05000456 96 010 02 7 of I0 CIf mrs spm is nmquin mu addidmaldms wpm NRC Foa 366AX 17)

.Msed on this additional administrative control, BrSaidwood station repositioned fuel iassemblies in the spent fuel pool. Comud has subsequently verified that the storage 4configurtioLn of fuel assemblies in Byron and Braidwood srp meet the criteria specified In cAC-96-240 and meet the supplemental criticality analyses performed for the actual Sagion I cell Sorel geometries.

Ihis item has been completed.

Comld will review the spent fuel pool criticality analysis for other Comld facilities what may be susceptible to similar problems. These reviews will verify that the current analyses conservatively consider the potentially limiting geometries associated with peripheral cells of adjacent fuel racks, especially as it relates to the placement

@of fixed poisons such as Soral or Boraflex on the outer faces or peripheral cells.

,Xhis will be tracked by iTS item 1456-100-96-010S102.

oThis item has been completed.

N1rS will submit required reading for the entire staff to clarify the responsibilities

• • f MFS engineers when performing an "acceptance review" of externally. jeneraated ealculations A(1) verify all Cornd specific inputs to. the analysis (such as physical dimensions, setpoints, and limits), and (2) verify. that., the vendor's methodologies and assumptions are valid when applied to Comud. This will be tracked by 'WI item-1456 S0-96-OlOS01.

"This item has been completed.

IdrS will revise HFs procedures governing the review and approval of controlled work to clarify the responsibilities of iFS engineers when performing an "acceptance review" of externally generated calculations: (1) verify all Comld specific inputs to the analysis (such as physical dimensions, setpoints, and limits), and (2) verify that the vendor's methodologies and assumptions are valid when applied to Comud. This will be itracked by NT3 item 1456-10-96-0103104.

"This item has been completed.

A review of regulatory requirements/quidance on fuel pool rack criticality analysis will be performed to ensure other requirements are adequately addressed. This will be

,racked by item 3TS 1456-10-96-0103105.

"This item has been completed.

Obtain As-built drawings for the fuel pool racks. This will be tracked by NT3 Item 0456-160-96-0103106.

XDelete corrective action:

Comumunications with the vendor confirmed as-built drawings were not generated, however.

,the station does maintain the design drawings for the fuel pool racks. The intent of

,the corrective action was to ensure future critically analyses correctly model the lack of boral poison plates on the periphery of the racks. The revised criticality analyses and procedures support the current configuration.

The Boratlex and Criticality Analysis issues have been submitted to the CoasEd Pact 21 Committee for consideration of reportability under iOCtR Part 2L. This review will be tracked by HTS item I 456-180-96-0105107.

"This Item has been completed.

An effectiveness review will be performed for all corrective actions initiated as a zesult of the 3/25/97 event. This will be tracked by NTS Item 4456-LSO-96-0O1SLER.

"This item has been completed. I U

(9'). N3~ VOAML -r. B u ~

ULS NSII.ARRUULATOSY

• C0M3UONM m mJm ORES NQ

-AVPROV 6lYuwvm m J108164 i*

ESTDATED BLI..N PU REFON TO COMLY WmTH THIS 91FORMATI1td COLLECTION REQUEST go00143 REM)BTIED REPORT (LER) FROC LSANDEDM AU TOII4LTU LoRW A1LXRDCo LICENSEE EVENT TE~~rCONTINcATM P= A1 J0 BA=KT1O MLTM1f. FMAXaD C0%M TE NTUATON REOAtD=J rtB.*tDE TO THE 1NFO3.4TO. 4,%D ME*STMATE micoN MANAGEMETf BRANCH c(4 F33 L S.V I CL.L.R EmJLATOQY COWSON. WAHIWOTO. DC 30554Q01. ,%X THE PAFERWOK REDUCTI*OCONECT NAM 0) SFAtCIrTNUMMNt (2)

DOCKET LiEt NUMmU(6) P.Ro0)

Braidwood Unit 1 05000456 96 *1 00  !*9ofI m

(Uafu isupmiid ad m - INRCB bm 366AX17)

F. HIMCU cczu Socafles degradation, that was bounded by the current Criticality Analysis, wan previously Identified at 59aidwood Station (NTJS 456-201-95-2155). Anticipatory omlpensatory and mitigating actions were put In place and included: administratLvely maintalning the SrF boron concentration greater than 2000 ppm, maintaining the Sip tuseritare as low as possible, restricting the removal of Silica fram the SrIP, minissizing transfer of SrP water into the RCS during refueling operations, and developLag a Boraflex comaittee to review and approve long term solutions to the Bloraflex degradation problem for Coinrd. The corrective actions of the previousLy identLfiLed p.:oblem would not have prevented continuing deterioration of the Boraflex.

Prior activities were reviewed to determine if precursor events occurred or if prior activities may have prevented the event. The absence of eoral poison plates on the perLphezy cells could have been jLdentLfled during the blackness testing in 1991. The Increase in neutron signal may have been attributed to degradation of Socralex- rather than lack of &oral poison plates. The interpretation of test results may have been skewed by the belief that Boral poison sheets were on the periphery cells (now known not to be accurate).

A review of Industry events did not find previous occurrences of errors In criticality analyses due to Borr.l poison sheets not being installed.

6. CA0UWU FVALUMS DAIS:

MANUFACTURZR NONMENCLATURE MODEL MiG. PART NO.

Bisco Products Boraflox Panels NA NA Inc.

EXHIBIT B-2 Braidwood Unit 1:

LER 456/96-008-00 (August 5, 1996)

.mc ~ ~ 5.MUC9 uou REGRMULATORV COMUIStIM AwumW iv _ONE NO""1"4 LI&NSKE EVAMT MZORT (LECR)

~ I OON"5 Ilops meop it PbeDiM of PU Full RMUIA InTedw"~a SpelICiNll VifOw IN Duto Pow"A6 &Wur MW1 vow=5 nv 19UM WU sap WGMIF0WWR0

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%M 0 14M WoM. W. Opp L-ý WOF Duringq the verification of Spent fuel Pool storage locations, it was disccvewred that ano~ fuel assembly was stored in Region 2, and rot in the required chelckrmoara conticuration, based upon the burnup versus Lnitlai'en-r-chment.

11ml~ts specified oy Tecnf. Spec. S.6.l.l.b.2. The cause of this event wa'*

personnel error. The burriup versus initial. enrichment limits, which determine acceptable, fuel, storage confijyurations, were chainged by TeochniCal Spe.:ificatiorn Am~endment 58. A calculation performed prior to this change to verify that the new limits were met contained an incorrect burnlio. The *;alc:'lat ion wa~s not

$..dependentlv vftAIifier, s0 Env error was nct Lcoentified. ImmvTediate corre-tivie actions were to relocate Assembly 546N into Regicn I of the Spent Fuel Pool.

Additional corrective &a7tions were counseling of the individual regarding expe.ctations And procedtire revision. This event ro-sulte.) in no safety :Oncerns.

Two %previ..ua fuel misp.)sItioninI evenlts wr -! dupa t.) per!-nnel errr) and proc.e~lural ano mAnagemlent *ieficiericies.

S~ ~*&-B IMIff

raw 24a V.. WaKO ouag~- nWMO By OWE N06 55IU41

£S113DM Ga "uA Wh TO cCmfv VWII IlS W"TvmiOMTMN CLW*EON FAMEST: we H" I LICZENSEE EVEIT RMPORT (LER)

TEXT CONTINUATION nM uccsm ForavIN o m I mlMe0 eOacold Man 0

IWO.ED 0MNVA6AW NM Mmm.

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oakCs C UUmIS. NO TO 7W6 0500456 2 OF 6 B1radwood Unit 1 r7 1 96 -- 008-- 0 I0 I CMW4ZTI048 MRIOR TO EVENT:

UNIT: Braidwood Unit I EVENT DATE: 07110/96 EVENT TIME: 1045 DE: 1 RX POWER: 100 RCS [A*) TEMPERATURE/PRESSURE: NOT/NOP

a. DlrRIlPTION OF EVENT:

this There vere no systems or components inoperable at the beginning of event that contributed to the severity of the event.

On May 28, 1996, nuclear engineers at Byron Station.reported that fuel assemblies were mislocated in Region 2 of the Spent Fuel Pool that did not meet ate requirements ot Technical Specificatioh 5.6.1.1.b.2, "Fuel Storage

- Region 2m. This situation had resulted from a change in Spent Fuel Pool storage requirements, caused by Amendment 58 to the Technical Specifications, approved on January 20, 1995. On 7/10/96, as a part of the a

investigation into this event, ComEd Nuclear Fuel services transmitted listing of fuel not meeting the burnup versus initial enrichment limitations to Braidwood Station. Braidwood Station personnel immediately noted that the Nuclear Fuel Services transmittal identified 84 assemblies that should be either located in Region I or in a checkerboard configuration, but only 83 assemblies were stored to meet this requirement. Upon verifying the information, Braidwood personnel identified that fuel assembly S46W was improperly loaded into a close-packed configuration in Region 2 of the Spent Fuel Pool without meeting the burnup versus initial enrichment requirements of Technical Specification 5.6.1.1.b.2. Upon discovery, fuel assembly S46T was imuediately relocated to Region 1.

Fuel Assembly S46W was discharged from the reactor core during A2R02 on October 11, 1991. In accordance with normal Braidwood Station practices, it was originally placed into Region I of the Spent Fuel Pool. S46W was relocated into Region 2 of the Spent Fuel Pool on June 16, 1992. Prior to

,~ ~ , a. m mm -maAPPRFOVED BY OWB MO 316"1" EXPW412 94FA

- rTED mKIU Pot MW.PONH TO COMPLY VAINS IimmiT' NOw ON OU.EcLlO4 MoLuS?: Si mi.

LICENSEE EVENT REPORT (L[R) msucinm ocueoonowcxTotuxw TEXT CONTINUATION namCMmNUAD _ mA,* IM,__- (f.

~UfiS0 noUUSI. ND dDO 114U PWBmm

=cumumc raidwood Unit 1 0 50004-56 SMhIAL 06 "VISION 3 0 96 --0508- 1001 h JEA an1 fit"" opm im r.,ireo go* A"Go~ne1 coie of - Db B. DCUPTIOF EVENT (continued) this move, procedure BwAP 2364-9, "Controlling Movements of Nuclear Fuelf Irhto The Spent Fuel Racks", was performed to verify that all moved assemblies met the burnup-initial enrichment criteria. At the time of the move, the Technical Specifications were met for assembly $46W, based on assembly burnup, supplied by Nuclear Fuel Services.

1995, Technical Specification Amendment 58 was incorporated on January 20, to reflect a new criticality analysis that includes fuel enrichment to 5.0 weight percent uranium 235, and to incorporate a 3 percent uncertainty to account for inaccuracies in calculation of assembly burnup.

The Nuclear Material Custodian at that time performed calculations, using the new limits, before moving fuel into Region 2-of the Spent Fuel Pool dturing A2R04 (refueling outage prior to Unit 2 Cycle 5)1. Although these calculations were hot required until receipt of the approved Amendment, thell were performed to verify that the new limits would be c*mplied with upon approval. These calculations were performed during October of 1994.

The Nuclear Material Custodian also performed calculations on all fuel assemblies in the spent Fuel Pool at that time to check whether the previously discharged fuel assemblies met the new criteria. He performed this calculation using a spreadsheet program, which was not independently verified. This spreadsheet was later transmitted to Nuclear Fuel Services as part of the investigation into the Byron event. The spreadsheet calculation failed to identify that assembly S46W did not meet the new limits because the fuel assembly burnup as pr(vided by Nuclear Fuel Services was incorrectly entered.

BwAP 2364-9, "Controlling Movements Of Nuclear Fiel Into The Spent Fuel Racks", Revision 1, does not require an independent review of calculations,

!s not retained as plant documentation, and requires performance only upon movement within the Spent Fuel Pool. Since independent verification is not required, the Nuclear Material Custodian was misled into thinking that iindependent verification was not required for the calculatiuns prior to Amendment incorporation. Since performance is not required except prior to fuel movement in the Spent Fuel Pcil, calculations were not required prior to amendment incorporation, when the burnup versus initial enrichment limits changed.

This event is being reported pursuant to 10CFR50.73(a) (2) W)(B), any operation or condi-ion prohibited by the plant's Technical Specifications.

NRC FORM 300 (4.6)

- M .APPROVED S OMfI N06 I ESTIMTD MM* PaER MIPONU TO C¢V *AIH TWS 1

PMNTMCNR OFI3OGN ¢LLECTION inQUiI: IOn.014 WF&VORMUMWULEDAMMMCMT WW4O LZCENSZE EVENT REPORT (LER) 1EucoampOc=SAWFWW=TO drnMi.

TEXT CONTINUAT ION uI WAMW.W-_%0-_--V---* AW *r.

Wt---

6r^.usmNLCWMftmbULATM rYCM'0 Unt 105000456 Braiwo~

I

Im" PW=

MWUCVM

2 soomeg" 4 OF6 Iuf MM uiow JA r*WJgg'6 9" MWJDitJO m*pies of M- Faso .0"AMla C. CAUSE 0F EVENT

The cause of this event was personnel error.

TMe Nuclear Material Custodian at the time of incorporation of Technical Specification Amendment 58 should have performed calculations to verify ccopliance with the new limits as a Controlled Analysis, with independent verification and retention for the duration of the Operating License for Braidwood Station.

D. 8AVETY ANALYSIS:

There were no safety consequences from this event. The Spent Fuel Pool boron concentration remained well above the value assumed for the current criticality analysis, while all fuel assemblies adjacent or near to the u1sloaded assembly had burnups higher than the burnup assumed in the criticality analysis. If the Spent Fuel Pool boron concentration had.been at the value assumed for the current.criticality analysis, no safety.

consequences would have occurred because the amount of fissile. fuel contained within the mispositioned fuel assembly was bounded by the existing analysis, and all adjacent-.fuel assemblies had burnup greater than the minimum burnup assumed. If the Spent Fuel Pool boron concentration had been at the value assumed for the current criticality analysis and -n additinnal fuel misloadinq had occurred, the required k-eff of 0.95 may have been exceeded.

BA.fErA IOM14-016

-_a._* C4 m__I,

-- P-PROVED BY O M NM 3W 940 6 q

w eiIATOC gsTMTWDsiUrd" TO CO.VmY nu ESPcmu IIO NV FCONM ,T Pa cCILE E I:. 0S*Hi.

10*rlNWW9AM LAWNED COMRATWWITO LICENSEE EVENT REPORT (LER) THeucmm4CUsADDs eAXocmMY.

TEXT CONTINUATION TM DrONATM AMR Miapoe MA*ia IMCkr (r nowm aous.am m q Braidwood Unit 1 05000456 i I = I ' MU 5 0F 6 96 --0O8-- 00 U fit mor.paC' J& rSqJred, up& sJcIJanal MopJe o1 NW Ave 366M t17)

K. CORECTIVE ACTIONS:

The Nuclear Material Custodian at the time of incorporation of Technical Specification Amendment 58 has been counseled regarding failure to meet expectations.

Procedure BwAP 2364-9, "Controlling Movements Of Nuclear Fuel Into The Spent

• uel Pool", will be revised to require independent verification of the calculations, retention as plant documentation, and performance when the kburnup versus initial enrichment limits are changed. This will be tracked to completion by NTS item #456-180-96-00801.

The location of all fuel assemblies in the Spent Fuel Pool will be verified by direct observation using an Underwater camera. This will be completed prior to moving any fuel presently located in the Spent Fuel Pool, unless such movement is required to ensurc safety. This will be tracked to completion by NTS item # 456-180-96-00802.

A review of the effectiveness of corrective actions taken for this event will be conducted by one year following completion. This will be tracked to completion by NTS item # 456-180-96-00803.

V. PREVIOUS OCCURRENCES:

LER 1-96-007 involved failure to comply with Technical qpecification 5.6.1.1 due to positioning fuel that did meet the burnup versus initial enrichment limits in a close-packed configuration irnadiately adjacent to fuel that did not meet the limits in a checkerboard configuration. The causes of this event were personnel error and procedural and "-anagement leficiencies.

Although the LER 1-96-007 event resulted in misiositioninq of nuclear fuel within Region 2 of the Spent Fuel Pool, the circumstances leading to this event were different from those leading to thu subject event.

Additionally, one other occurrence involving fuel mispositioning (457-200 94-016) was noted. A review of the event determined that new fuel was ruispositioned in the Spent Fuel Pool during transfer from the New Fuel Storage Vault. The cause of the event was personnel error due to a lack of a questioning attitude and failure to follow procL jres. A review of the corrective actions determined that they would not have prevented this event from occi+rring.

NRC FORM 368 (4-95)

z ll-ý

pp Iz

EXHIBIT B-3 Braidwood Unit 1:

LER 96-007-00 (July 15, 1996)

ONG NO1504104 ByROE Mu N~ ) FORM UAL.NUCLEA REGULATORY COOMMISSION 44- Eiromo amput mrapOU To COhUY MMfl TH w MATOWY WNI -- "K T "MS.

REPORT (iLZR) POWESS me E OWE ONO otiE LICENSEE EVN 05000456 1 FI firuiWOod Unit 1 Plceet of Spent Fuel InRMAI to C Due to Personne Ew.r and PrIoedura OWd N=. fril UV 17 U

II w N

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• l.*.jiir~R! p*-----,rWT', . ....

I

________IM 1~'I.LIPa

.401tqX1 ID.73(aK WTI X2 Oz________

m, TIMM II (815) 45&280 X306 D.ULwon. System Engeneerin (Kra. omplab EXPECTED SUBWSSON DATE).

not bounded by the existing On 6/17/96, fuel was repositioned in the Srent Fuel Pool into a configuration that was accomplishing the fuel Criticality Analysis. The mispositioning was identified by the Fuel Handling Supervisor inappropriate configuration movement and reported to the Nuclear Material Custodian (NMC). The fuel in the Independent Reviewers did not consider was immediately repositioned Investigation concluded that the NMC and moves Additionally, this the effects on lower burnup fuel in adjacent storage locations in planning the fuel by the analysis vendor Causes configuration rest. iction had not been properly transmitted to Braidwood Station deficiencies Corrective of the event were determined to be personnel error, and procedural and management guidance. counseling of actions taken involve preparation of a iew procedure containing more detailed position for other fuel stored in the Spent personnel involved, revising the NMC qualification guide. reviewing requirements A safety analysis determined Fuel Pool, and immediate repositioning of the fuel to at' appropriate configuration.

involvinL.fuel mis osationing was that the mispositioned fuel did not cause a criticality concern A previous event caused by a failure to follow procedures for fuel moves

- F.,m4 SS .. ij" aa inawu Lsaiz APPROMVIEX BY PWlOMB S G~~OA 316U44 NO. M ESTIMATED JRDEN PER RSPONSE TO CMPLY WITA TS HRS.

MANDATORY 4ORNATIION COUJECTION RECUES.: 5.0

-'2)

RVUCTEO LESSONS LEARED ARE 00CORfIPOMTo1o RN MC ENTSN(O C *OMMDD THE.sePoCESS WX4O WFA " ESlrI RG ANFEDcATO TE HDIAM.

LICENSEE EVENT REPORT (LER) IF TEXT CONTINUATION *W%~TONAORDS We THE Wff W c(

6 F33. US. NUCLEAR NEG"RTORY SOW C

WAHINGTON. OC ZUUSCUI. NO TO ThE1-9 RIVUCfl PGJC 05000456 lEAF SFdjENIIAL 'F 6 Braidwood Unit 1 96 -- 007-- 00 9te So psi copteq of NAr Form )66" (11)7 A. PLANT CONDITIOIS PRIOR TO EVENT:

EVENT DATE: 6/17/96 UNIT; Bra:dwood Unit 1 EVENT TIME: 1212 MODE: I RX POWER: 100 RCS [AB) TEMPERATURE/PRESSURE: NOT/HOP B. DESCRIPTION OF EVENT:

at the beginning of this There were no systems or components inoperable of the event.

event that contributed to the severity Fuel Pool in preparation for Fuel moves were planned for the Spent which a "Blackness Testing". "Blackness Testing" consists of a technique inneutron of the Boreflex neutron source is used to evaluate the degradationracks. Continued periodic absorber material in the Spent Fuel rool storage neutron moderation remains that testing is a commitment to the NRC to ensure Criticality the Spent Fuel Pool within acceptable bounds, ensuring that Analysis assumptions remain valid. Nuclear Componern.. Transfer Lists (NCTLs)

Custodian (NMC) for this purpose on were prepared by the Nuclear Material recently assumed the NMC position.

5/9/96. The NMC preparing these moves had by the previous NMC, on An independent review of the NCTLs was performed a second independent 5/11/96. As a part of the normal review process, on 5.'15/96. On Engineer (SRE),

review was conducted by the Station Reactor of these fuel moves, performance 6/17/96 at approximately 0930, during the he considered configuration that the Fuel 4andling Supervisor noted a fuel fuel stored in to be suspect. The suspect configuration involved irradiated Region 2 of the Spen' Fuel Pool. Requirements tor storage of fuel in Region burnup corresponding to its 2 are that either the fuel must have a specified configuration initial enrichment, or it must be stored in a "checkerboard" percent weight if its initial enrichment was less than or equal to 4.2 restriction may be Uranium 235. Fuel meeting the burnup-initial enrichment Thp suspect fuel configuration stored in any configuration in Region 2.

enrichrnent restriction beirq involved fuel that met the burnup-initial adja -nt 'o 'oel ha- did stored in a close-packed configuration immediale'y Into th- "chec,.erboard' not meet the requirement, and was placed

-ro 14 u.a. sUna2 DBOn.&gM C0maS8I APPOMu BY -W NM. M66~

EXPIIES MANM ESTIWE TEWRO PER RESPONSE TO COvWLYM Th MIS UMEDATORY INIPOqM1ON COU.ECTION OEUE5Tr: U.0 HRnS.

REPORTED LESSONS LEA HCORPOATE INTO NAME (LER) MeUMe N D FM BACK UBTw, TO TO ESTMWfU LICENSEE EVENT REPORT , OVA* COLSNITP NOKGAFA) 9UPAM TEXT CONTINUATION IE UM Mme F&O5IT CMUbON URANOI (r.

4 pIW. uS. UCIEM RIeaATORY WAS@TNOTN. CC 261OH*.4. NAD TO 116I IIqRWN REDUCTMON P*ROJCT 71raidwood Unit 1 05000456 ,IM se.-?Lam MV1s5-1 3 OF 6 96 -- 007-- 100 U

fttI ma.l* . e iv is rieJret.1, u.-- &WtionaJ copies of OK foro .66AP 118)

B. DESCRIPTIONI OF EVENT 1continued) configuration. The Fuel Handling Supervisor immediately contacted the System Engineer in cha:ge of the "Blackness Testing", who then contacted the 14C. After consulting with the SRE, the NMC directed the Fuel Handling Supervisor to suspend fuel movement, and began preparing NCTL Variations (BwAP 370-3T3) to reposition the suspect fuel assemblies pending further investigation. The NCTL Variations were prepared by the NMC and independently reviewed by a Qualified Nuclear Engineer (QNE) and two Senior Reactor Operators (SROs) by approximately 103C. Investigation of the suspect fuel configuration revealed that this configuration was not specifically allowed in the Spent Fuel Pool criticality analyses, so a Problem Investigation Form was completed at 1215. Repositioning of the suspect fuel assemblies was completed before this time.

The vendor responsible for the current Spent Fuel Pool Criticality Analysis was contacted to establish whether the suspect configuration-was bounded by the exis-Lilg analysis. The vendor responded that the suspect fuel configuration did not meet the initial assumptions made for the Spent Fuel Pool Criticality Analysis, and immediately began preparing an analysis of the safety imnact of the suspect configuration.

A -opy of the aaalysis indicated there was no safety significance of this fuel positiinng other than requir-ng a minimum boron concentration in the Spent Fuel Pool of 00 PPM, which was exceeded at all times during this ev =ýnt.

Ti.:. event is being reported pursuant to IOCFP50.73 (a) (2) (i) (B), any operation or condition prohibited by the plant's Technical Specifications.

C. CAUSE OF EVENT:

The caus,.; of the event were determined to be personnel error and procedural and manalement deficiencLes.

,e Nu,:'*>'r Mdterial ('usrodian and one Indeperident Reviewer did not identify the SL:nect fuel positiotinq during preparat-on of fuel movement plan.

A-tho,:,l no known requirements for the olajcenent ot fuel at this transition

lM p W Ids V.a. WMW MWM mo~szz= APPROVED BYM MUMO.141" 45-10.1 EXPIRES "MewI ESTh*TEDO SUIN PER RESPONSE TO CCPLYVUmV THIS WNIWTOTVMOrMTION COULECTIOL UMINED 3UST: 15.0 ARE PIOORPORATED 1TOMRS.

REPORMLEOUSSC*M LICENSEE EVENT REPORT (LER) IFC*P QIA*NS ME0014NM IUMMF. ESMTNTTO .

TEXT CONTINUATION T.e ,MT REOArWG UIWN*(TE'a TO

• "*. U.S. NUCLER*a GULATOWI*OMOION4 WASMMTMUDO2*C54*. NM TO iHE PDMRIVOWS riozwU m fb000p -mm-(a) 1" OFI Braidwood Unit 1 05000456 T~A SMOLVITIAL MVS~I4OF 6 96 1--007-- 1001 U RC Foam 16*44 117) i m ,It spac is ,quzpJred. a#suljotJo J, c i*Jes of C. CAXSE OF EVENT icontinued):

boundary between fuel meeting the burnup-initial enrichment criteria and fuel not meeting the criteria (stored in Region 2) of the Spent Fuel Pool had been transmitted to Braidwood Station by the analysis vendor, the NMC and the Independent Reviewer are expected to identify such a questionable configuration prior to NCTL issuance.

One Independent Reviewer of the prepared NCTLs identified the suspect fuel positioning as questionable. However, the reviewer did not address the question prior to approving the NCTLs.

The required fuel positioning at the interface between fuel that does meet the burnup-initial enrichment restriction and fuel that does not meet the criteria in Region 2 of the Spent Fuel Pool wa* not specified in any Braidwood Station or Commonwealth Edison procedures directing fuel movements.

The requirement for positioning'fuel with less than cr equal to 4.2 weight percent Uranium 235 that does not meet the burnup-in;tial enrichment criteria in a checkerboard configuration was transmitted by the 'Licensing Report On High Density Spent Fuel Racks For Braidwood Units 1 and 2",

Revision 0, dated August, 1988. Th..s document addresses the assumptions made for the analysis, but does not identify any interface requirements.

The expectation to review the planned fuel movements against positioning requirements was not clearly defined. Inclusion of all requirements into fuel movement planning, and actual preparation of NC7Ls Lor all types of fuel movement planning did not addres; these activit:es in sufficient detail.

The planning and independent review of the controllez NWTLs were performed using unverified and uncontrolled information.

D. SAFETY ANALYSIS:

There were no safety consequences for this event. A.a!yýis by the vendor performinj the Spent Fuel Pool Criticality Analysis indicates that the mispositioned fuel did not cause a crzticality conce:r. as long as suffirient

me ,V

-P 3M 9.a. ---Sa"M .- o s--

-- A EXPRS0mY

-_ROVE- -O31U-41" N*gM0A 45-921 E51M~TE P.1N p"R ISpom jo C~Y WITH TlU O~F41TCN 0ffOMM MIiM'TOW VLtSSONS MJET: US& MRS pWORED MALVNMWIgC*OW0PATWD WO0 Tie ucu5, s"mocUSI'nSP K TOMOMMW'.

LICENSEE EVENT REPORT (LER) PO NAW 0 C0 MMDMS RG , #DlM II A ESTr9Uq~ MTO TEXT CONTINUATION "g erov AnMe0Msw A, wc IT

- P~. U..MJWCLEAR OMATOMfCODWAM ss MW=PMAC~l*

'WM~ffTCH. Dc nsnMU. MOTO TME PAMRWCX m* 0iI567 nm._.a. .

=.ZV m ii i oII Braidwood Unit 1 05000456" WM E- . 5 OF 6 nU Mfg JFris Elf it~

10Qjaalri.a jCJUWeIc"p Acs.o" Of MWC Form 346MP ILI) 96 --007-- 001 D. SAFETY ANALYSIS (continued):

boron existed in the Spent Fuel Pool. The required concentration for this of event in 300 PPM, Spent Fuel Pool boron concentration remained in excess If a fuel mispositioning or fuel 2300 PPM for the duration of this event.

diop event had occurred while the fuel was mispositioned, sufficient boron in a safe condition.

concentration existed to maintain the Spent Fuel Pool R. CORRECTIVE ACTIONS:

in the Immediate corrective actions were to reposition the fuel inappropriate configuration.

regarding this failure to The NMC and Independent Reviewers were counseled meet expectations.

A new procedure, BwAP 2364-3T3, has been created to list the requirements Procedure changes have been generated to require for fuel positioning.

prior to issuing NCTLs. The new procedure execution of this-new procedure both the NCTL preparer and an indepeihdent includes a checklist, requiring verifier to review the proposed fuel movements for fuel positioning requirements.

the initial The interface requirements for fuel storage that does meet the analysis burnup-initial enrichment requirements were received from vendor. These requirements were reviewed against all other fuel stored in the Spent Fuel Pool. "'- other instanc-s in which the requiirements were not met were identified. These requirements were incorporated intz Braidwood Station Procedures as BwAP 2364-3Al.

guidance SThe Qualification Guide will be revised to provide nmore specifiz positioning regarding the necessity to review planned fuel moven.ents against requirements. This action will be tracked by NTS item #456-18r-96-00701.

This event was discussed with all qualified Nuclear Engineers.

EXHIBIT B-4 Browns Ferry Unit 2:

Supplemental LER (October 9, 1980)

Offic fI,~ ~ x Ata 9a 3ogi.9 330 Uear NM er. :adw reg~ulaoy MPM VA=3Z AIzLM33Rny - EI~*vS F~m NutzR PiAM UNIT 2 - DOCK=

No. 50-260 - F~ThITI OPER&T LIEN DPR P1POFWL OCCIM SW7O-50-260/8037 10nMWI~

Mme enclosed report is a supplmenft to myv letter dated septerber 26,, 1980,,

ancerning fuel assemblies TZ 758 and TZ 399 which were mnisoriented 90 degrees. This report is s~fdxtted in accrdanc with Brons Ferry unit 2 Technical pcfctc 6.7.2.a(9).

very truly yciurs, TEHE VALLEY AUMRIMM

3. RL Ca~houn Director of Nuclear Powr 1n3osure (3) cc (Eaclosure):

Director (3)

Office, of 14anzamint Injanation and Program CcmtroJ.

U.S.Nuclear Paulat~rljComaxissicn Director (40)

Offic of Inpcinand -iformmv U.S. Nuclear Regulatory Cczmiissc Washingtoni, DC 20555 Mr. Bill avrallee N~ucear Safety Analysis Center Pa~lo Alto, California 94303 Mr. R. F. Sullivan, NEC Inspector,, Bxvns Perry A "A*6PA Eoual Opportunity Emnployer-0w'---

INA J" UPDATE REPORT -PREVIOUS REPORT-September*26,; 1980

.. LICENSEE.L N ruif tVJ~

CONTROL $LOCK: 'I .T(PLIA14PRINT OR4 TYPE ALL REQUIRED INFORtMATION)

A _ 1 J10 2 S 0 10 1- 1021610 17M '161-1 Iq 4810 11 11 1-J L.a.LJ G W

66nc 1 pOCKET %UMS&A as 66 VE4T DATE 14 76 RIPOAT OAT& di EVENT DESCRIPTION AND PRODAILE CONSEQUENCES()

r0=2 IDuringt EOC-3 fuel shufflinn operations-it was noted that fuel assemblies TZ 758 and TZ 399 were misorlented 900 1 There was no previous occurrences. There was noj mdanaer or hazard to the oublic. See Technical Snecification 2.1 and 3.5.1K.

=07 SVSTIMh CAUSE CAUSE COMP VALVE F9 CODE CODE 'LSuOe"0 CO%4WONENY coot SLACODE SUJeCoOE I A 6 10 11 12 13 111 to 201 z.EUUENdrAL OCCu~e'ffNCE REPORT 0 ff.,

Art.O4 wuU.!1%

L11 8'~ 0 100ECT U_ 1 131.Z72..L l'4UOO%%%

6L:ý4 1 1.iWLi 7 A"\ rTjMcHUIPT N-"'D PRIME COMP

~

TakEN ACTIONa ON PLANdT MIt'eOO MOvIRl IMITTD U~3 SORM bU.SUPPIR.&L, Clf iW0~1 i0 18 1..~2 CAUSE DESCRIPTION AND CORRECTIVE ACTIONS (

IDuring the previous refueling outage__(BOC-3) 16 fuel assemblies were misoriented.

1Subsequent rework left two 7 x 7 bundles in the misoriented position.- A later review of the BOC-3 core verification tapes confirmed there were no other misoriented fuel assemblies. Units 1 and 3 core verification tapes will be verified.

r ( L ME T4 on OFA' STAUS %PWE OT04ENSTAtiS LO.SCOVERV D.CvR ZosOPý1 (D7

.0) 101 01 01"9 NA I LIA I Operator observation

'I 10 12 v¶ 4 44 46 .4411 ACrIVIlY :ONTENT?

ELI NSILU OF RELEASEl AMOUNT Of ACIIy'?' G. LUCAN )P:OFP "lEALf G-1; LZ 0L. J NA I NA F F-7

'F.PSUNNEL EXPOSURES 1L, IPT~o f0%

I.Ut.l4f P, IOGI 10100 TP,

~ NA PIp4SONP.r, I~ajiAI ES

'.JIH43I4 DISCRIP I if8 F'.

Los!; (A 004' nAMACE TO FACLMSTV )

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PUSIC TV4

ýApCJ'r FZL ISSUE(J 'WiE.'TO SCPPI.

' y 44 1 TVA media information te lephone S IO I II.

¶ f4 it

~~AM5 OIpIPAIrbR 6 ý" - ý P1tONE.-

Tennessee -Valley .Authort Form BF-17 Browns, Ferry NuclearI1a~ BF 15.2 1/10/79 7N-1 LER SUPPLEMENTAL INFORMATION BFRO.5O*/_83 Technical Specification Involved 2.1 & 3.5.K Reported Under Technical Specification 6.*7. 2. a (9)

Date of Occurrence 9/14/80 Time of Occurrenco 1900 Unit 2.

Identification and Description' of Occurrence:

Fuel assemblies TZ 758 in c8 re location 15-26 and TZ 399 in core location 29-28 were found to be rotated 90 from their correct orientation.

Conditions 'Prior to Occurrence:

Uniit 1 - 1055 l4Je Unit 2 - refuel shutdown Unit 3 - Shutdown maintenance outage Action specified in the Technical Specification Surveilance Requirements met due to inoperable eq~uipment. Describe.

Apparen CoseofcOcurrence:

The 16 misoriented fuel assemblies were loaded out of proper orientation and core verification procedures detecte~d the errors. Rework instructions failed to accomplish the required orientation of the two fuel assemblies.

Analysis of Occurrence:

See attachment Corrective Action:

Verification and reorientation procedure for fuel loading verification~ have been made.

Procedural changes include the requirements that rework will be documented with second party verification.

Failure Data:

NA

• Retention: ýriod - Lifetime; Responbibility - Administrative Supervisor

• Revis ion:

-4'

According to Supplemental Reload Licensing document NEDO-24169A the limiting full loading error is a rotated 8 x 8 (8D274) fuel assembly and assues a rotation of 1800. The M4CPR for the limiting event is the safety limit of 1.07.

Any other slsorLentation would result In an MCPR greater than 1.07. The misoriented fuel assemblies were both high exposure - original 7 x 7 (7D250).

Since both subject fuel assemblies were not of the limiting type, were sufficiently separated to prevent interaction, and since there were no significant transients during the cycle, the safety limit of 1.07 MCPR was not exceeded.

Of the two fuel assemblies, the process computer indicates that TZ 758 at location 15-26 made the closest approach to its operational limit for 7 x 7 fuel of 1.33 MCPR on the following three occasions:

7/10/79 MCPR - 1.40 1/1/80 HMR - 1.39 "6/30/80 .CPR - 1.40 All operation was within the bounds of the reload licensing submittal.

Both of the fuel assemblies are scheduled to be removed and will not be reloaded for BOC-4.

Si

EXHIBIT B-5 Byron Station:

LER 454/96-008-00 (June 25, 1996)

U.S. NUCLEAR REGULATORY COMMSM-. APPROVED BY O1011 NO. 31 50104 NlC FORM 3" EXPIRES 04V3015 s0rl1110 eVm M INWAE t COSAY Wal leM UUNAIU euamI CoMMam NUOSI: S5OL ill& MOMS tim U MW LICDISIzz KVI T RZPORT (L=) ICOMA' K1 11 .tem MOM AnM BKo a no e RCOD iuaUhl emu"a 04 Fn VAL MKOM MMIM comma=

IS** reverse for required oumber of umOM K 2m .ismus e M PFuwSU n UWM I nm011 SIS OFFIEM M1111 1116AN R .

WS5"I1Ui K JM digischarwcter Il for eac block) --

SdoUeil~ Ur n m RAWe itti 05000454 I Of 9 BYRON NUCLEAR POWER STATION Fuel Assemblies Located in incorrect Pinioan of Spent Fuel Pool

.1 IrY rAi OTWMn ACUEJTS BF; VE s VETDATE IN E SU WR DAMUf 1" T ila mini OAM V lWIN R ~ IAT ROM O I WIN I FCUYWAW6C MO-NýI 06 25 906 5 0H000 SOC-i umimW 28 96-I 96 - 008 -00 05 gg j____--___ u~vPa 20.22031u1Wtlv)

TOTWREOUUIREMITSO 10 CPA 6: lChek one or nwire 11 CVEATIG M 0" a X 50.731@a142HO 0.iv)1*

MO I)12.- 01 - 20.22034MM 1o 50 73410112100.7h15

- -20.2203410111 -20.22031aH1,1)4

. 3612-1. 5071a12)a v "L o .22031a412)4 --

-0.22f*1f CoSTACT RM-THU LE42 COM6N.73A1U D*CME U1106 412) inW below__

@OPEEO1512011 c hMhAC X(2154-w 815-234-5441 inPUD or11 3E-6M WE-ý1D. Golf, System Engineer David spices,

• N -* I 121

"-"-I-X25 "8ee1t5-234-5rnl"1r~Sn w nytr D. Gof David Enine M OIU GAd 3 7 Comm"" 1 UON* 11D

...... etldb 5.1., " ius wer 4r-u--,*1 IS1111111110 M*dRTUrspctvV.Th eebio no - -

b - . -t - n2. he DATE (I N 4,v. apeeEXPECTED SUBMISSSON OAT".!I)N tho 1400 spacese i.e.. opprveonelI I im(OA o Tho-emcoputer snee1161 h The cL that fuel assemblies F37E. th 4E. and Gs7F were Orb 28 May. 1996, Byron Station nuclear en ner s confirmed of Technical Specification ITS) retimben i on2 of the Spent Fuel POl IfS without meeting thie requirements did not meet the minimum burnuprequirm ents8.nor were 5.0.11i.b.2, uel Store-Region 2.0 The assemblies 32651 MT5.d1MTU. and 32771 32651 MWd/MTUw roved thy 2chckerbo, d, the threquared meeiumbus wr e 13264 MWd eMTUh 32638 MWd/MTUh and 32728 MWd/MTU MRNd/MTU respectively. The actual bu were Rioups respectively.

error. The computer spreadsheet used to verify minimum require Tte cause of this event was cognitivpersonnel for assemblies F37E. F44E. and GS7F, and the data in the spreadsheet had biarnup contained erroneous information G67F into SFP Region 2 did not have the current noit been independently verified. PersoiiEl approving placement of into Region 2. Ultimately, the rewision of ornup criteria for detrminmfl of uelasiseibly eligibility for placementAmendment 68. 'Fuel Storae" were not venifed to most this requirements of TS S.6.1.1 fulel assemblies' bumnups CrIticality,0 prior to its implementation.

Or- 29 May. 1996. the three fuel assegf'lies were moved into Region 1. as allowed by TS 5.6. 1.1 .a.2. 'Fuel Storage either to meet the minimum required burnup or to Region 1. All fuel assemblies remainirl in Region 2 were verified be storeld in a checkerboard pattern.

This event resulted in no safety conceriw. The event was bounded by both the older and the newer criticality ermalyses for Region 2 fuel storage. Adequate reactivity controls were in place to ensure that the k., limit of 0.95 Storage - Criticality' was not challenged during this event re~quired by TS 5.6.1.1, "Fuel or condition prohibited by the plant's TS.

This event is reportable under 10 CFR S0.73101142)1iB), any operation

U..mcM EULTR OUS N FFORM I.'. LICIM3 Z VKNT RuWORT (LaiR)

TEXT CONTINUATION OKTLtMSRPG 3 FACILITY NMEU 11 O 9 BYRON NUCLEAR POWER STATION050442 TEXTr #raw smor ais j,.q4 use se~f i -coo of f Form 3664 1171i A. PLANT CONDITIONS PRIOR TO EVENT:

Event Date/Time 05-28-96 11700 Unit 1 Mode 5 - Cold Shutdown - Rx Power Shutdown RCS (ABI Tempereture/Pressure 840F I 0 peig Unit 1 Mode 4 - Hot Shutdown Rx Power Shutdown RCS IABI Tempersture/Pressure 3350F 1 321 pslg B. DESCRIPTION OF EVENT:

is a checklist Byron Administrative Procedure IBAP) 2000-3T1, -Spent Fuel Burnup Verification Checklist, required burnup for used to verify that fuel assemblies either have or have not accrued the minimum uncheckerboarded SFP Region 2 storage. The minimum required burnup Is calculated by linear interpolation between values given in BAP 2000-3A1, -Minimum Required Burnup as a Function of Enrichment for Region I1 intended to bound TS Figure 5.6-1.

High Density Spent Fuel Storage Racks.' The values in SAP 2000-3A1 are "Minimum Burnup Versus Initial Enrichment For Region 2 Storage."

BAP 2000-3T1 for fuel On 10 February, 1993, Byron Station nuclear engineers (engineers 1 and 2) completed The thekbldist showed both assemblies with en initial evnichmem of 3.8 assemblies including F37E and F44E.

MWd/MTU. given by BAP wt% U-235 and a minimum requiued burnup for placement into Region 2 of 32540 and 32638 MWdTU "2000-3A1 Rev 1. F37E and F44E had accrued actual burnuas of 32648 MWd/MTU enrichment of 3.8 wt% U for an initial respectively. The minimum value of 32540 MWd/MTU was appropriate Region 2 storage.

for uncheckerboeided 235. and both assemblies met the Technical Specification requirement which, in port, stated that On 11 February, 1993, Nuclear Fuels Services INFS) issued letter NFS:PSS:93-060 of TS 5.6.1.1. This letter showed F37E fuel assemblies F37E and F44E met the minimum burnup requirements MWd/MTU respectively.

and F44E having accumulated 32648.0 MWd/MTU and 32638.4 and F44E into SFP locations K On 18 August, 1993, Byron Station fuel handlers moved fuel assemblies F37E were not stored in a checkerboard pattern since they C2 and K-DS, respectively, in Region 2. The assemblies in place. The moves were performed in accordance met the minimum required burnup restrictions presently Ust."

Rev 1, 'PWR Station Nuclear Component Transfer with page 93-104 of an approved SAP 2000-3T3 was completed prior to transfer list approval.

Engineers 1 and 3 verified that SAP 2000-3T1 of a license amendment Starting in the summer months of 1994. engineer 3 was assisting in the preparation Region 2 up to 5.0 wt% U-235 and was supported by a request. This request would allow storage of fuel in new criticality analysis.

Problem Identihcation Form 4 PIF)

On 11 August, 1994, Byron Station "news (engineers 3 end 4) initiated and NFS employed different methods in 454-201-94-69200. This PIF documented that Byron Station requirement for Region 2 storage. NFS used determining whether a fuel assembly meets the minimum burnup applying a 1.03 multiplicative penalty to a polynomial fit through the points given in the criticality analysis after Station used linear interpolation account for fit error and uncertainty in the assembly burnup calculation. Byron bound TS Figure 5.6-1 Amendment 25. This PIF also identified that TS Figure 5.6-1 between points which used as the basis for the curve.

Amendment 25 did not, for all initial enrichments. bound the criticality analysis WOM62MI

SFORM -" ZJ)U.S. CLME fEGULATORY COUNIsSIOR LICUNSK WEZNT REPORT MMI)

TEXT CONTINUATION 05000454 3 OF 9 BYRON NUCLEAR POWER STATION 96 -006 - 00 TGXV INmr be~,pUd.'s dEdoe'al-i0oAWC ~ 3#W 117)

. DESCRIPTION OF EVENT (cont.t criteria for minimumn required bumnup determination. The Byron Station and NFS continued to use different when approved, would render te second problem moot. Fo. the ice--e a nent request being developed, to change th points used for minimum Interim enginser 3 prepared a revision request for SAP 2000-3A1 25 and the criticality anaysis would be burnup deermnination such that both TS Figure 5.6-1 Amendment bounded.

enginer 5 and 8) completed SAP 2000-3T1 for On 16 September. 1994, Byron Station nucleaw engineers GS7F assembly with an initial enrchmnt of 3.109 fuel assembies including GS7F. This checklist showed the placement into Region 2 of 32661 MWdIMTU.

wt% U-235 and meeting the minimum required burnup for The minimum value of 32661 MWd/MTU was GO7F had accrued en actual burnup of 32728 MWd/MTU.

Engineer 6 stated that the enrichment value was conservative for an initial enrichment of 3.809 wt% U-235. bumrup was calculated. G67F met conserva ly rounded up to 3.81 wt% U-235 when the minimum required storge.

the Tecmhical Specification requirement for uncheckerboarded Region 2 which, in pert, stated that fuel assembly Also on 16 September. 1994. NFS issued letter NFS:PSS:94-225 Byron 1.1. The discrepancy between the GS7F did not meet the minimum burnup requirements of TS 5.6. in desteirnining eligibility of a Region 2 Storage Station and NIPS conclusions resultedthefrom the differeont methods 2000-3A, Rev 1. It can.didate. Since G7F- had accrued minimum required burnup in accordance with BAP 2 storage.

wa deemed to be suitable for uncheckerboorded Region Review 1OSA)94-078 approved a licensebecame amendment request for On 20 October. 1994, Byron Station Onsite, This amendment request later TS Amendment "ByronStation Units 1 and 2 Technical Specifications. be conservativ 1% gret than the "68. This request would, in part, revise Figure 5.6-1 Amendment 25 to utructions that would In FIgure 5.6-1 alng with new crMicoty analysis. Discrete values would be provided 3.8 values. In particular. the required burnup for an initial enrichment o allow kler interpolation between the to 32651 MWd/MTU.

wt% U-235 would be increased from 32540 MWd/MTU of incumbent fuel assemblies and their eligibility fo The OSR 94-078 package did not document the review 3 and a representative from NFS partcipatet Region 2 storage with the new minimumIburnup curve. Engineer in the OSM.

7) had conducted .n a review of the incumbent fuel Howeve., Byron Station nuclear engineers engineers 3 August to November. 1994. This review aissemablies over the course of several months from approxiamtely calculate assembly eligibility, and then uthe was performed by engineer 7 building a computer spreadsheet to ouput was spot checked by engineer 3 for vedflicat on. The spredsheet required input data for initial accrued burnup. and then checked each fuel assembly against several enric.m.nt. storage location. and actul SAP 2000-3AI Rev 2 and TS Amendment 68.

minimum burnup criteria, including those that would become or "not O' for for each assembly. i.e..

• The spreadsheet calculation produced a Sooleen output uncheckerboerded Region 2 storage.

burnup data loaded into the spreadsheet for F 37E.

Initial enrichment, storage location, &nd actual acaued producing erroneous 'OK outputs for those F44E, a G67F were incorrect. This resulted in the spreadsheet the assemblies would have been properlV assemblies. Had correct data been loaded into the spreadsheet, mum required burnups of BAP 2000-3AI and TS identified as not OK" when compared against the mnir Amendment 68.

NCFORM 354U.S. NUCLEAR REGULATORYCO IS8 LXZC 1USUZVE REPORT l (LZR)

TEXT CONTINUATION FACIITY M f It DOCKET LE Mumm PAGE 13 BYRON NUCLEAR POWER STATION 05000454 .1 4 OF 9 96 00 -- 00 TDC0 Mfmwe wee. /.csie4 we -doft cowimf AWCFwm 36(kIR71 B. DESCRIPTION OF EVENT Icont.)

On 26 October, 1994, PIF 454-201-94-69200 was closid with the understanding that Byron Station and NFS would continue to use different methods for determining minimum required bumup for Region 2 storage. This would serve as a diverse means to identify essembles; suitable for Region 2 storage.

On 13 December. 1994, Byron Station OSR approved revision 2 of BAP 2000-3A1. This revision was processed as a corrective action to PIF 454-201-94-69200. which identified that TS F'gure 5.6-1 Andmerw 25 did not, for al initial enrichments, bound the criticality analysis used as the basis for the curve. The new revlsk bounded both the critiality analysis and TS Figure 5.6-1 Amendment 25. Under the new revision, the minimum required burnup for an initial enrichment of 3.8 wt% U-235 was increased from 32540 MWd/MTU to 12800 MWd/MTU. Byron Station took credit for the review performed in association with OSR 94078 to verify compliance of the Incumbent fuel assemblies. As stated before, the spreadsheet contained erroneous date for P3/E. F44E, ard G67F. Hence, all three seer iblies passed the review. Under SAP 2000-3A1 Rev 2.

fuel assemblies F37E, F44E., and G67F no longer met the minimum required burnup, though they all met the requirements of revision 1.

On 20 January. 1995. the Nuclew Regulatory Commission (NRC) issued Amwedment d6 to Byron Station Units 1 end.2 TS. revising Figure 5.6-1 as requested under the licensing ameridment request previously submt.

On 23 Jauwry. 1995, Byron Station fuel handlers moved fuel assembly G67F into SFP location G-L12 in Reagion 2. The assembly was not stored in a checkerboard pattn since it had been verified to meet the

'requirements of SAP 2000-3A1 Rev 1. This was done in accordance with page 96-5 of an approved PWR Staton Nuclea Component Transfer Ust. Enginews 5 and 8 verified that SAP 2000-3T1 Rev. 1 was completed prior to trnsfer list approval. However, SAP 2000-3T1 Rev. I had been completed In Septembeir.

1994, using SAP 2000-3A I Rev 1. SAP 2000-3A1 Rev. 2 was now -he current revision, and assembly burnups should have been compared to revision 2 requirements rather then the revision I requirement. The assembly did not meet the minimum burnup requirement of SAP 2000-3A IRev 2 or TS Amendment 66, though it did comply with TS Figure 5.6-1 Amendment 25.

On 25 January, 1995, Byron Station OSA 96-007 approved for use Amendment 68 and its implmentaon plan. The OSM 95-007 package acknowledged that TS Figure 5.6-1 was changing. The implementation plan stated that the Byron Station nuclear engineering group "wiN revise SAP 2000-3A1 to reflect the new burnup curve to identify assemblies that we acceptable to load in Region 2." At tha time. it was thought that SAP 2000-3A I Rev 2 was more conservative than TS Figure 5.6-1 Amendment 68. Therefore, the implementation plan required no deadline for revision of BAP 2000-3A 1. The OSR package did not discuss the review that had beeo performed of the incumbent assemblies. Engineer S ard the Station Reactor Engineer ISRE) participated in the OSR.

On 30 January, 1995, Byron Station OSR approved revision 3 of SAP 2000-3T2. NCTn Verification Checkllst." This revision provided more explicitly detailed guidance on how to perform the verification of minimum required burnups on SAP 2000-3TI On 8 Februwy, 1995, Byron Station OSR approved revision 2 of SAP 2000-3T1. This revision added more documentation of information so that minirr,jm required burnups could be more readily and accurately determined.

U.S. NjcEA REGULýATORY COO""*"IO F F GSA MEORT (LZR) 14461 LICINSIB 3VDIT TEXT CONTINUATION DOCKET LER HUMMUR PAG FACILIT MANI II FAGG TYEASEUENiAL IREVMS 6 OF 9 BYRON NUCLEAR POWER STATION 05000454 -JSE -O and were in violation of TS 5.6.1. 1.renewed Each hed been prvously F37r. 441E. and GS7F, were in Region 2000-3A1 which an earlier TS approved for residence In Region 2 using a revision of SAP approved revisionburnup 3 of BAP 2000-3A1. o This revision wes processed On 17 August. 1996. Byon Station 0SRminimum required curve nowexactly matched due to TS Amendiment66 3265 for an inhitlenrichment of 3.8 wt% U-235. Again. Byron Station T Figure .ing association with OSR 94079 to verify compliance of the incuombnt took rdit for the roelew performed Inwere moved into SFP ion 2 since implmenton of TB Amendment fuel asse.mbies. Two f. asse.mbies were moved from feiled fuel canisters on 1 June a 29 June. Both assembles 88 on I March, 1996. They met the minimum burnup requirement.

SAP 2000-3T1 for fuel assemblies anticpated to be moved in association On 24 May. '1996. while performing rdck neutron attenuation testing. Byron Station nuclear engineers (evngier with uocrun spent fuel storage mblle F37E end F44E did not meet the mnimumx burnup as require 7 and 9) oun indications that fuelase- Nor were these two assemblies stored in a checkerboard by TS 5.6 .11.b.2.s. Ofuel Storae" - Region 2.'Storage - Region 2. Byron Station contacted NFS for pattern as allowed by TS 5.6.1.1.b.2.b a to assist the investigation into whether these verifpation of actual burnup end ree m u rnup fuel asemlies were incorrectly o In Region 2.

fuel assemblies anticipated association to be moved din byiT*

On 26 9)mnMay.. 1996. whole performing BAP 200043T1 for no... t meet. the.,,,., Station,,,nuclear r, minmmbru engineers lerifneers 7wan o7i*

spent fuel storaq rack spund aesebly neutrona on testng Byron assmbl-toe In-s-chec....kerbo*ar patern as aillowed by TB* 5o.1.b*.2.b... Byron..

and to include ti ai6..1.b2.a Ntor weemm .thi- o eiiaino culbru n **

iiu eurdbru Station.agai co wntce fuel assmbly in the investigation.

fuel-..-,,

,-a TS-ebls P31 P4" 1 andG .Itwa cal dicussing the results all of the NtS ,-vsiatonnt violatio of.. .6.1.1....

three assemblie wae In determined at 17:00 that C. *AUSE OF EENT:

nRgo 2"w" contv"esne err h aaue Tghe cause of F371 rand _F4 en norclysoe bu p entrecor rectly nr--. wa it speashe for eiyn inmmrqie

,was-not byrthe coputerll faled txonsho thtF71adP41wreI F ineedntyvrfidtoeacuae Tesrushee dtae . rc bunu ale fo P31adP4.thcIsa Rein.Futemore, the spreadsheet data failed touete 9.006 approva-,nd a accepta hncewoT Amrenmentt review was pert of the basis fr te. Byo Station........ pan conditions,, not conforming totenwreurmns SB. The inrendment was then implemented wit

U.S. NjcLAR- MEOULA75ORCcoMa55lO

-e LICIDINiz KvzMT REORT (LiR)

TEXT CONTINUATION DOCIMET LER NMIER1 AG VA-CIJI NAMfflwI 06 OF 9 05000454 BYRON NUCLEAR POWER STATION V Cmisq*euM we e*dWmMaM C OfARC Avon' 36W 1171 AEM'wpe C. CAUSE OF EVENT (cont.)

also Cognitive Personnel eror. Pe~rsone The caum of GS7F being WlfCoredV stoned in Region 2 was use the current procedur revls0or Of SAP 2000 approvinig the NCTL to place G66W In SFP Region 2 failed to for uncheckerboarded Region 2 storage.

3A- to verify that G67F had ac-.d the minimum required burnup resulted in an Ineligible fuel current Plant conditions. This The previous revision that wal usd did not reflect assembly being placed Into Regiona 2.

D. SAFETY ANALYSIS:

analyses used as the basesF4E.for TS The SFP condition throughout this event warn bounded by the two criticality fuel assemblies. Including F37E.

after Amendment 68. All unchckerboerded Figure 5.6-1 prior to and failed to of those analyse. However, the SFP condition "andG67F, met the minimum burmp requirements grater than the current criticality analysis.

meet the current TS requirement. which was 3%

section 9.1.3.2 addresses the safety evaluation for storing spent fuel in teSFP- The criticality portion UFSAR Criticality Analysis Considering Boraflex Gapsan Is based on the 'Byron and Wes. Braidwood Spent Fuel Rack1994. as amended by 94CB8-G0D-105 and 94C9-Shringe document from . .houedated June.

abnormal*condition whero reacqtlvity 0142. Section 5.0. Discussion of Postulatea Accidents, addresses en assembly is misloaded Into Region 2 which does not a fl would Increase beyond the analyzed condition:

satisfy the requirements.

is misaoaded, the analysis makes several conservative While, in the scenario considered, only one assembly assumptlons:

enrichment or its equivalent at the minimum required

1. AN fuel assemblies contain U-235 at the nominal burnup.

Uranium credit is taken for reduced-enrichment or natural

2. An fuel assemblies are unformly enriched. No axial blankets.

product poisons. No credit is taken for any burnable

3. No credit is taken for U-234. U-236. or any fission absorber material which may remain in the fuel.

material.

fuel assemblies not conltairing any absorption

4. All storage locations we loaded wth in lateral extent.

5. The storage locations ame infinite of 1.0 9/cc.

6. The array is moderated by pure water model is assumed.

7. A conservative Boraflex degradation inserted into a 5x5 array of the with an ennchment of 4.2 wt% is

8. The scenario where a fresh assembly nominal assemblies is considered.

U.S. NUCLEAR REGULATORY COMMISSION MC 3611 Was "LCex EVENT RMPORT (CLi)

TEXT CONTINUATION IDOCKET uaaIIYNAM LEN NUURPAMGE3 06000454 F-i7 O BYRON NUCLEAR POWER STATION 96-006-00

.10 FuMAV366U [1in Toff (Mmwe spae is nyqufwdo. we sddE~~Iegn' D- Safety Analysis Icont.)

The maximumn k,, at a 95% probability with 95% confidenice and Including the statistical summation otdu in reactivityu independent uncertainties is 0.9,-9 for Regon 2 under the nominal conditions. This increase de*ta I. However, only a single failure must be accounted to tde misloded assembly Is no more than 0.0438 t. more from 300 ppm boron is approximaely -0.06 delta for, so soluble boron may be credited. The reactivity of 0.95 required by TS 5.8.1.1 1 not then offsetting the increase from the misloading. Thus, the k.Ilt challenged during this abnormal condition.

is more fuel assemblies misloaded rather than just one.

The situation described in this report, with three to the following considerations:

conswvative than the accident analysis due them exceed the minimum bumup requirement. making I. Noarly all fuel assemblies residing in Region 2 less reactive than the reference essembliee.

or natural uranium axial blankets of six inches at both

2. Many fuel assemblies have reduced-enrichment ends, reducing their reactivitles.

poisons as and spent assemblies contain fission product 3.. AN fuel assemblies contain U-234 end U-236.

well. These materials further reduce reactivity.

the fuel there we several empty locations. Some of

4. Not every storage location contains fuel. Locally, (RCCAs).

rod cluster control assemblies assemblies contain absorber material such as leakage at the boundaries.

5. The SFP is finite, exhibiting nonzero neutron Soluble 80 degF, having a density less then 1.0 glcc

6. The water in the SFP is normally approximately January. 1996, providing at than 1280 ppmn since boron concentration in the SFP remained greater least -0.22 delta k reactivity.

detwiorated to imply that the Boreflex in Region 2 t as not

7. Previous neutron attenuation testing results the extent assumed in the analysis.

wt% enriched significantly less reactive th** the fresh 4.2

8. The improperly located fuel assemblies are of the Fuel assemblies F37Er. F.E. and 87F fell short assembly assumed in the accident analysis. values we and 43 MWd/MTU respectively. These required burnup by 3 MWd/MTU. 13 MWd/MTU, values.

within approximately 0.1% of the required burnup 5.6.1.1 was not that the k,., limit of 0.95 required by T; Thu combination of the above factors ensured chillenged during this event.

U.S. NUCLEAR NSWJLT0AV COMMISSION1 seWA36 040 LICUsKI IVUT RZVORT (LUr)

TEXT CONTINUATION 1W OCKIET LM - I PAGI FAC ITMW U1 05000454 98 OF 9 BYRON NUCLEAR POWER STATION W1 capeu- C FewArm 356" 1171 Mm ffiN er we" 4 ,weqm ueaa.*d, E. CORRECIVE! ACTIONS*

on 28 May. 1996, at 17:15. Byron Station nuclear2 of engine~er Initiated PIF 454-180-96-0006. identifying three fuel assemblies mappropriatily residing in Region the SFP. Byron Station Regulatory Assurance, wm also Operations, and Systam Engineering management were notified. The NRC Resident Inspector notified.

possible Inadequacies and Inconsistancies in Concrrently. NFS initiated PIF 901-201-96-0780 Identifyingfuel assemblies. The investigation results show their methods of determining eligibility of Region 2 candidate to the root causes of this event.

that he"s inadequacies and Inconsistencies did not contribute moved fuel assemblies F37E. F44E. and G67F Into On 29 May. 1996. at 05:15. Byron Station fuel handlers in accordance with page 96-103 of an approved PWR SFP storage locations in Region 1. This was done Station Nuclear Component Transfer Ust.

resWdg in Region 2 using TS Anndment 68 NFS-subsequenty performed a review of all fuel assemblies a Not of every and PSSCN:96-023. it consisted of critala. This review was transmitted as NFS:PSS-90-1421998..and ientified which assemblles had achieved March.

fuel assembly in t Byron Station SFP as of 31 yron Station engineers 7 and 9 then verified thatthose.

the minimru requited burnup for Region 2 storage, a checkerboard pattern. There stored in Region 1,or in assemblies not meeting minimum burnup were eihe 2 perflormed mince31 were no assemblies stored Inappropriately in Region 2. AN fual moves into Region 2000-3AI Rev 3.,

verified In accordance with SAP March, 1996. have had eligibility requirements explicit guidance on the preparaon and Independent SAP 2000-3T2 Rev 3 is currently in place and provides woo not in piace at the times F37E. F44E. end GS7F were rfeview of SAP 2000-3T Rev. 2. This revision barrier to approved for uncheckarboarded Region 2 storage. The guidance provided presents an additional this event.

miu tocating a fuel assembly that could have prevented and provides improved docurmentsa.onplace of mTSmum required burnup SAP 2000-3T1 Rev. 2 Is currently Inorplace within Region 2. This revision was not in atthe timnes F37E. F44E.

for luel assemnbles being moved to shows Initial Region 2 storage. The improved docuent ation ard G87F were approved for unchSckerboarded presents en additional enrichment, minimum required burnuP.that and actual accrued butup for each assembly arnd could have prevented this event.

barrier to mislocating a fuel assembly requirements of TS Figure 5.6-1 Amendment SPo2000.3Ac Rev. 3 is currently in place ad is identical to the tre fuel assemblies 8B is well a the current NFS method of determining Region 2 storage eligibility.in All accordance with this required burnups determined appoved fot Region 2 storage will have minimum Figure 5.8-1 wigl have a concurrent changing TS procedure or its equivalent. Any future TS Amendment This presents en additional the new requirements.

revision to SAP 2000-3A1 a-tsociated with it reflectingprevented this event.

barrier to mislocating a fuel assembly that could have this with oarsons involved in the errors that contribu,01

!o Performancei expectations have been discussed event.

emrphasizing of the Byron Station nuclear engineering group. required reading This LER will be discussed with all members A copy will be placed in the nuclear engineering group personnel performance expectations.

completion of this action.

book. NTS item 454-201-96-000-01 tracks

U.S. NUCM RUGULATORY COMMS*MSo "FON M 34 N-i LZCcazIBz zv UITRZORT (LIX)

TEXT CONTINUATION

-IDOcE LM PWj AG FACT I MMNAME VM SminM unn 06000454 [ 9 OF 9 BYRON NUCLEAR POWER STATION TnDET N,nore aoec. b nu*de, .a -m~n co~ f AWC F.'rm JEWt £117 F. RECURRING EVENTS SEARCH AND ANALYSIS:

of Spent Fuel Pool due to Personnel Error.

LER 4S4:94-006, wFued Assembly Located in Wrong Region the "documents a similar event. On 15 July, 1994, SED fon a fuel assembly in Region 2 that neither met this event was 5.6-1 not was checkarboarded. The cause of minimum burnup requirements of TS Figure and an indpedent reviewer Nuclear Materials Custodian determined to be cognitive personnel errors. The in to verify assemblies met the minimum bumup requirements for storage failed to use the approved method Region 2.

assembly incorrectly residing in SFP Region 2. the Although the 454:94-006 event resulted in a fuel from those lemding to the 454-180-96,-008 event.

circumstances leading to this event were different G. COMPONENT FAILURE DATA:

No components failed in association with this event.

EXHIBIT B-6 Byron Station:

LER 454/94-006-00 (August 15, 1994)

S IGNATUPE PAGE FOR LICENSE EVENT REPORT LER Num~ber Title of Lvent: Fuel Assembly in Wrnna Leatim Soent Fuel Pool due to w MLnrkm rgr-aeign  !;rmnr rual Pool due to

-Personnel Error occurred:

Date Time OSA DISCIPLINES.REQUIRED: A-%

.I1h4 Acceptance by Station Review:

OE Date SES Date RAS Date -r.NHF -;cc, Date Approved by stat ion kanager 1.19 A 1.0 , WV F

UCENSEE EVENT REPORT ("a) 0101611 5810I01 VAC.IilY"W LIT UMU POOL DUE TO PERSOWi4,L ERROR MUEL ASSEMBLY LOCATED IN WRlONG REGION OF SPENT FUEL 11I? .lCIlIlolMEl I IM in M1N DA1111 o I tm plaCT iii p INmi 1111011111 DAY TIMA M N UT-II AIM0 gAA u11"*S5W E9,...

I~

SO C f V 91W 1"REI--' X01MTOil CECE U IS TI3 T 1N Co*i*fO7JlS i~riSbOAT I TM t"LBE .

S 1.----1 um.

WWII 1="&02e.lm

, m, 2u0.131111)lleU2Nl Sw O.Rr o2_OMu01d LCOMIEI CONTACT FOR tWiS LElN TELEPHONE Rallw G. STAUFFER. STATION REACTOR ENGINEER. X2249 COMaUTi ONE UN FO cOMPONENT FAIUuE ASCIIK3 aS EACH Afs REovRT REPIIAIIREPORIVLE C.AUN SYSTEM 01111,00111 TUR 11141mnmill 0 CAUSE SYSTEM COMPONIENT MAIZWfaC TORE i

I______ ___________fl TM OATS TEAM____

SUPEETAL WUNO mai~w OE

J E EISS:NDTE) H -S DAE DI It of the Spent OnJly 5, 1994. Sy~stemn Engineering Department ISED) found fuel assembly U38J located in Regian ir. Technical Specifications ITSI Fuelool SFP). The fuel assembly dio not meet the burnup requirements speciliea for Region I Storage.' The Secion5. Design Features.' Figure 5.6- 1. 'Minimum Burnup Versus Initial Enrichment into Region 11at location HM5.

l~cerComponent Transfer Lost INCTI) incorrectly specified the placement of U38J Administrative controls require any The CTLalso did not place the assembly into Region 11in a checkerboard pattern.

pattern. The assembly

,assmblythat does not meet minimum burnup to be placed into Region 11 in a checkerboard

,vsplaced into the incorrect region of the SFP on September 26. 1993 during a refueling outage or Unit 2.

The error was discovered wvhile preparing for the next refue'ing outage The assembly vias moved m~Region I on july 16. 19!l94.

is withir the safety This evei tinvolved no safety con-erns The safety significance ot the misplaced assembly 5,O 731a)l211i, 81 Any operation analysis presented in the 'JFSAR. This event is reportable in accordance vvith 10CFR or conditiori prohibited by the plant's Technical Specifica~tions

IIF1; lo;'F :ý P394 Z

LICENSEE EVENT REPORT (LER) TEXT CONTINUATION 0OCKET NlumBl LiElMNIM PAM J"TI NAIWO TIE Sit. NUe VISI F

• IVI NUCLEi POMLI- STATONI oi0 1 5! 1 4 -1i°1, °0 ol1 1 13 9F I nu!icuwam IMD f mwdwidw a ibem ENm sa - IP VLSI 1a6y bifq Mw

. PLANT CONDITIONS PRIOR TO EVENT:

Event Date/Time 07115/94 i_0930 Unit I MODE I.j. - Powr Oger.nqima Rx Power 80% in coastdown RCS [ABI Tenmperature/Pressure NOTQIOP Unit 2 MODE _I_ - Power Oberations Rx Power 9_9% - RCS IASI TemperaturelPressure NQINQp B. DESCR PTION OF EVENT:

Between mid-August. 1993 and September 10. 1993. a non-licensed engineer (Engineer 1I completed the Nuclear Component Transfer Lists INCTLs) for offloading the Unit 2 reactor core (Page numbers93-121 to 93-146). This individual was the station's Nuclear Materials Custodian or NMC. During the writing of the NCTLs. he na4ewo This location s in errors. On page 93-139, the NCTL shows fuel assembly U29J oing to storage location HM O.

minimum a Region II rack. The bumup of the assembly, at we time tfiesNrMc wrote the list. did not meet the into Region II. The NMC made a similar mistake for fuel assembly U38J on burnup requirement for placement was 29770 MWD/MTU versus a required burnup of 32.540 MWD/MTU.

page 93-143. The actual burnup of U38J The NCTL shows asssmbty U38J going to storage location HMS. This location is also-in a Region II rack. Both errors were cognitive personniel errors.

After the NMC wrote the NCTL for the offload. for Refueling Outage B2RO4, he completed Byron Administrative Procedure (BAPl BAP 2000-3T2. 'Nuclear Component Transfer List INCTL) Verification Checklist.- Step 1 of the checklist requires the preparer of the NCTL to verify that,

"*Fuel assemblies entering Region II of the spent fuel racks meet minimum burnup requirements as described in BAP 2000-3A1 or are placed into a checkerboard configuration. Records of assemblies which meet minimum burnup requirements are kept in file 1.02.1080, which is in the NMC satellite file cabinet."

Records of assemblies that meet minimum burnup are documented on SAP 2000-3-T1. "Spent Fuel Burnup Verification Checklist.' and are kept in file location 1.02.1080. BAP 2000-3Al s title is, -Minimum Required Burnup as a Function of Enrichment for Region II High Density Spent Fuel Storage Racks." This attachment gives a listing of initial enrichment versus the minimum burnup required for storage in a Region II rack.

BAP 2000-3, 'Safeguarding and Controlling Movements of Nuclear Fuel Within a Station.' requires the NMC to complete BAP 2000-3-T1 for each assembly to be placed into Region II of the Spent Fuel Pit ;SFP). The NMC started but did not complete these forms for assemblies placed into Region II during Outage 82RO4. The BAP 2000-3-T 1 form was comoleted as part of this investigation.

The NMC used the TOTE data 'or all the assemblies discharged from the core. TOTE is a computer program that calculates assembly burnup. TOTE data gives the total accumulated burnup for each fuel assembly. The data is stored on the IBM mainframe and is accessible via a personal computer The NMC used the IBM and mentally wept thr"'ugh the burnup ver,*.cation. He did not complete the information on BAP 2000.3 TI. Nuclear Fue.

Services INFSi is respons-bie '.: running the code. They run the code every month and after a t.nit shutdown.

1....

0 1s 0 1o ' 0 14] S1 4 19 - 0 10 1 - 0 10 0 13 F 18 17 rEV.* 198 blim" iMtmtiew Solo EM3 Cd- M imaidad m 'be let a IXI B. DESCRIPTION OF EVENT: (Cont.)

assembly, the NMC did the burnup check using Using the TOTE bumup data and the initial enrichment of each the forms as ;aquired by procedure or why he BAP 20003AI. The NMC could not recall why he did not complete of the BAP 2000-3-TI forms for the previous outage made the error when he did the burnup checks. A review the forms.

(February 1993. BIR05) showed the NMC had completed NCTL writing process. The process is verv Discussions with the NMC identified several weaknesses in the writing the NCTLs. The Verification Checklist gives complicated and relies heavily on the skills of the individual does not describe the process on "how to" write the criteria that the NCTLs must meet. However, the checklist sequences: the offload. the insert shuffle, and the NCTLs. The NMC divided the process into three major below.

onload. The process as described by the NMC is given pattern supplied by NFS. and the existing core First. the NMC does a comparison between the candidate loading cycle's core loading pattern. The NMC obtained loading pattern. The candidate loading pattern shows the next core. The tagboards are located in the area the existing loading pattern from the tagboard for the Unit 2 reactor every fuel assembly and component in the where the NMC sits. The tagboards are useJ to show the location of cores. The tagboards mimic the SFP. the New Fuel Storage Racks. Failed Fuel Storage Racks. and the two reactor up-to-date based on completed physical layouts of each of these areas of the plant. And. the NMC keeps them NCTLs.

He based the categories on the Once he completed this comparison, he placed each assembly into categories.

would have in the next cycle.

insert a fuel assembly contained in the current cjcle and the insert the fuel assembly "have' and what they are *getting." For this In other words, categories of assemblies are based on what they that nave burnable poisons IBPs) that are event. there were nine different categories. For example, assemblies control rods IRCCAs) and are getting thimble plugs getting] thimble plugs (TPsI (BPs to TPs). assemblies !hat have to RCCAsI.

and are getting control rods ITPs (RCCAs to TPs), and assemblies that have thimble plugs swaps can occur with the least Next. the NMC arranged the categories side-by-side in the SFP such that the insert amount of tool changes- There are five major steps to the insert shuffle.

most efficient laylct. After the Th. NPIAC did this arrangement in the SFP by iteration until he obtained the arrangement in the SFP is done, the NMC can begin writing the offload. The NMC wrote the offload such that the nine categories. As he wrote the ottload fuel assemblies were placed into the first open location in each of the requirements and three optional items.

sequence, the NMC also ensured that each step met seven sequences. In all. there The NWC went through a similar process to ,.,ite the insert swaps and the core onload During discussions, the NMC

,A.. e eleven required checks and four desirable items for the entire refueling.

not part of BAP 2000-3T2- This identified an additional four criteria he met while writing the NCTLs, that were nineteen.

brought the total number of checks the NMC met to program called Shuffle Works.

After the NM- wrote the three malcr sequenc.s, they were lodded into a computer forms that the Fuel Handlers used in the held A member o, the SED This program .vrote the sequence on NCTL the program step by-step. This was done because Shuttle r:uclear group entered the offload and insert shuffle into However. the program did write the onload %erluence sinre it Works could not perform all of the required checks all insert vere. d,'e. 21 the :inal contained I) the pool configuration after the core was offloaded and the shljufl#'S this ,ifiornim non.i,. I)f, ifatilt. 4,rc)?e core configuration. and 3i the loading sequence. Because the program had i e sequence meeting ,ii the atipropri.tP rerwiirernents 9 9? ;P '. W P F :,'8 t ,4 4

LICENSEE EVENT REPORT (LER) TEXT CONTINUATION FImLOTi RAW OCKUT mNIS LEN NUMEN L PANE W ONIUSTATI-N CLEA RR 1f lTlT EMeW hwwy liftl,1ai$=a m3 u dmoimd a tus On n =

and. XI B. DESCRIPTION OF EVENT: ICont.)

After t NMCIwrote the N--TLs, he gave them to an independent reviewer on September 10. 1993. The independent reviewer was a non-licensed engineer IEngineer 2). Engineer 2 did not use the records of assemblies that meet minimum bunmup requirements to verify certain assemblies could be placid into Region N. He was unaware of the requirement because he failed to review SAP 2000-3 prior to performing the verifications. This was a cognitive personnel error. Instead, this individual used information from the Nuclear Fuel Services Department (NFSl. NFS sent a letter that listed assemblies by region and indicated which assemblies met the minimum burnup requirement for storage in Region II. Attached to the letter, was a printout showing the individual burnups of every assembly.

During his review. Engineer 2 found the error for fuel assembly U29J on page 93-139. but fadled to find the error fMr assembly U38Jon page 93-143. He notified the NM;C of the error for U29J and the NMC wrote a vanaton to the rd; I L. Engineer 2 did not discover the second error and stated that the cause of the error was most likely due to his performing several checks simultaneously. At the time he reviewed the NCTLs. he was performing multiple checks as he went through the NCTLs. This probably caused hin to miss the burnup check for assembly U38J.

Engineer 2 and the NMC both signed the verification checklist on September 13, 1993..

Discussions with Engineer 2 indicated that there have been errors in past NCTLs but they had been catight by the' independent reviewer. No Problem Identification Forms (PIFs) were written for these events. Although PIFs were not required for these events, opportunities to identify and correct these errors before a higher level event occurred, were missed.

Fuel Harillers placed assembly U38J into a Reg.on II rack on September 26. 1993 in accordance with the NCTL.

On July 15 1994. a non-licensed engineer lEngineer 3) discovered that fuel ýssernbly U38J was in a Region II spent Tuel rack. I ne fuel assembly had been in the Region II rack since September 26. 1993. The Fuel Handlers had placed thr. assembly in the Region II rack during the last refueling on Unit 2. The assembly did not meet the minimum burnup requirements of Technical Specification Figure 5.6-1. *Minimum Burnup versus Initial Enrichment for Region II Stordoe."

Engineer 3 discovered the error during preparations for moving fuel assemblies from Region I to Region I1for the upcoming refueling outage on Unit 1 181R06). The SED Nuclear group reviewed every fuel assembly located in Region II to ensure the assemblies either met minimum burnup or were checkerboarded. After the discovery, Fue Handlers moved fuel assembly U38J to Region I following an approved Nuclear Component Transfer List INCTLI.

The Fuel Handlers moved the assembly into Region I on July 16. 1994.

This event did rot involve any inoperable systems aid was not effected by plant operations on Unit 1 or 2. No operator actions either increased or decreased the severity of the event.

This event is reportable under 10 CFR 50.73(a12)Ii)l8I. any operation or conditton prohibited by the plant's Technical Specifications i9931RWPF .080894451

I 0 l10101441 l l l l l l 0!

111V1si -

" 1 I-,-.

• -1 "

li CmlId Syi V119 IIlS ce vwe. desdod aib14 in, aIN ri Till 1"Piuhlislsn C. CAUSE OF EVENT:

Both the NMC and the independent reviewer The primary causes of this event were cognitive personnel errors.

the minimum burnup requirements for storage in failed to use the approved method to verify assemblies meet would not guarantee this mistake would not Region II racks. It should be noted that use of the approved method enhanced. There were also several contributing recur because of a procedural weakness. The procedure wi:r beerrors.

causal factors for this event that led to the cognitive personnel preparer. The well defined and relieb he: -. or .ne skills of the The current methodolegy for writing NCTLs is not NCTL until the mcst efficient seqluence is found. This preparer goes through many manual iterations on the method is mtt conducive to minimizing human error.

well defined. Many verification steps required by The methods to be used for verification are also not may not as occurred during this event. And, some methods SAP 2000-3T2 can be done in several diffarcnt ways, performing verifications.

be as effective as others in catching errors or for verifications, the ,bility to find and correct problems By not writing PIFs for failures found during independent LER was minimized.

before they result in higher level events such as an for its intended purpose. enhancement of the Shuffle Although the Shuffle Works program is an effective program in the futu e Works program could help prevent errors of this type Refer to the Recurring Events Search and Analysis A corrective action from a previous event was ineffective.

section for an explanation.

D. SAFETY ANALYSIS:

largest reactivity increase occurs from accidentally UFSAR Section 9.1.2.3. 'Safety Evaluation.' says that "The with all other cells fully loaded. Under (his condition, the placing a new fuel assembly into a Region II rtorage cell multiplication factor would not exceed the design presence of 300 ppm soluble boron assures that the infinite

. soluble poison present 12000 ppm boron),

basisreactivity for Region II With the recommended concentration Ii were to be fully loaded with fresh fuel of 4.2%

the maximum reactivity. K, is less than 0.95 even if Region enrichment.'

in the SFP at two thousand ppm and administrativel-,

Byron Stalion noiinally maintaif.$ the boron concentration At the time it was p°,aced into tne SFP, fuel controls the concentration to gre; ter than eight hundred ppm.

per Metric. Ton-Uranium (MWD/MTIJ) and an initial assembly U38J had a burnup of "'9.770 MegaWatt-Davs the misplaced assembly and nr, safety significance enrichment of 3.802% Thereiore. the UFSAR analysis bounds II iack.

existed while the asseinbiv was in the Region I  ;',

WP -t 9

LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Lin MUM PACA

--'--X UK0C!!*l VIM IL NUM! KNION 0IWA"LlAM P0MN STATa S 0.1,4 -LD11.F11 bI'1 ou Sow END miu we dlelhml4 a *4 lot Is piii IT ill E. CORRECTIVE ACTIONS:

Corrective Actions - Leno Term were counseled.

I. The NMC and the individual that performed the independent verification the methodology to be used to write NCTLs for a

2. The SED nuclear group will write a procedure that explains on when in the process verifications will be refueling operation. In addition, this instruction will give directions 454-180-94-00600-01 tracks done and the preferred method for performing the verifications. NTS item completion of this item.

each verification on the Nuclear

3. The SED Nucleai group will determine the preferred method for performing Component Transfer List INCTLI Verification Checklist. BAP 2000-3-TI. SED will revise the checklist to:

al explicitly define the preferred method of each verification.

define these alterrate methods. These b) indicate whether alternate methods are allowed and explicitly methods will be equivalent to the preferred method, checks.

cl organize the checklist to distinguish important checks from less important method.

dl provide cautions describing the pitfalls for each items.

NTS Item 454-180-94-00600-02 tracks the completion of these program that will allow it to perform more of

4. The SED nuclear group will pursue revisions to the Shuffle Works NTS item 454 180-94-00600-03 tracks the verifications the nuclear group presently does manually.

completion of this item.

will require writing PIFs tfr errors caught during

5. Regulatory Assurance will issue PIF threshold guidelines that completion of this item.

independent reviews. NTS item 454-180-94-00600-04 tracks to include a column for recording both the

6. The SED nuclear group will revise the BAP 2000-3-Ti forvrb burnup for storage in Region II.

assembly's burnup in addition to the minimum required NTS item 4* -180-94-00600-05 tracks completion of this item.

a v:alkthrough of the entire refueling on 'paper'

7. The SED ruclear group will revise BAP 2000-3 to require of this item tagboards. NTS item 454-180-94-00600-06 tracks completion outage on Unit 1:

Interim corrective actionis for the upcoming refueling of the Niclear Group and place this a The Station Reactor Engineer will discuss this event vv;th al' members LER in the Nuclear Group Required Reading NTS item 9 454-18094 0060007 tracks this item.

II This is presently a requirement of

b. BAP 2000 3 TI will be used prior to moving any fuel into Region this action BAP 2000-3, so no NTS item is needed to track of the entire refueling procedure NTS C A -paper' tagboard will be used for a step by step walkthrough itemn 454 180-94 006 08 tracks this item

• 9,4 1p, WP F .? R0!i94 "

LICENSEE EVENT REPORT ILER) TEX F CO~NTINUATION PWCTT "IN DOCIF411ER Li WNW pm VEEr few" blomWy el EBB sddip"M myis sm dwe 0 tho W ft fill F. RECURRING EVNT EArCVAND

- ANALYSIS; A sexct on ETS found one previiaus e-. ent of a misplaced fuel assemnbly dtu to an error in an NCTL OVA 6-1 071, 'Fuel Transfer List Error. documents this event. A review of the corrective actionts for 0 eveo t olodicated that one of the corrective actions was nor inclCerrenited Corrective action to prevent rucurence.

item" 28. states.

'SAP 7000:3 wil be revised to require The use of a prorodedira checklist wh~en developing rho NCTL This lost will include.

'S The requirement to use a tag board Ciarrantty, SAP 2000 3 does not contain this reqwerernent. Discussions with the Station Reactoir Engineer IS110.

at The remit of the event. sindicAte that this c orrective ac coon requi~red a Step by -steg walk through of the entie refueling evolution on the tagboards Howeveur. Engineer 2 indicated that the intent of the- corrective action changed The intent changed to the use of a 'pape tagbnard As o~vosed to the use of the pI'vsical tagbownf This wwould eliminate possitile errors from moving chips on tlhie ohvsica taghoards it cannot be determined **%

this requirment was not imncopoated into SAP 7000) 3 A ,w~view of th~e PJTS itern written to track comilsetwo of this corrective action indicated that the PITS 'was not, %paerf- )n exactly ,vhat changes to SAP 20WG 3 were needed. The FITS item simply stated to 'dovelrop a.Prn.PdisrP i-heck,t.t n#r itvoerities h#,w to pfepare an 14CTL.

At the timel th., checklist was develnp.r1. ft failed tnoinrorporurto this 'W~ijirwmumnt PnrnO SAP 2rX)Oc JI Therefore. triis correctivii actioni was ineffective G COMPONENT FAILVRE DATA.

tha.rp VvJ% tno faslow r ')rmpnwnant rtiir~nu th*. *., Pt,,P,

EXHIBIT B-7 Catawba Unit 1:

LER 413/90-0160-00 (April 19, 1990)

I Duke Power Company IV03) ,431.- 000 Cutawha ,Vucteur Station PO Box 256 Clocer S.C 29710

-- I I DUKEPOWER "90 8 :3:r2 3W April 18, 1990 Document Control Desk Commission U. S. Nuclear Regulatory Washington, D. C. 20555

Subject:

Catawba Nuclear Station Docket No 50-413 LER 413/98i  : .

Gentlemen:

413/90-16 concerning TECHNICAL Attached is Licensee Event Report OF A MISSED REFUELING WATER SPECIFICATION VIOLATION AS A RESULT ACTION.

STORAGE TANK SAMPLE DUE TO INAPPROPRIATE no significance with respect This event was considered to be of to the health and safety of the public.

ery truly yours, Tony B. Owen Station Manager keb\LER-NRC.TBO American Nuclear Insurers Dottie Sherman, ANI Library xc: Mr. S. D. Ebneter c/o Regional Administrator, Region II The Exchange, Suite 245 U. S. Nuclear Regulator Commission 270 Farmington Avenue 101 Marietta Street, NW, Suite 2900 Farmington, CT. 06032 Atlanta, GA 30323 Mr. K. Jabbour M & M Nuclear Consultants U. S. Nuclear Regulatory Convmission 1221 Avenues of the Americas Office of Nuclear Reactor Regulation New York, NY 10020 Washington, D. C. 20555 INPO Records Center Mr. W. T. Orders Suite 1500 NRC Resident Inspector 1100 Circle 75 Parkway Catawba Nuclear Station Atlanta, GA 30339

'?(_)(4 .-"A6')274 9/)(41

"  ;ý--C'C PD)(41-

NINC Pt;.; Me 19431 U.&. NUCLEAR REGULATORY CONNSUMo~

APPROVED 0Mg too. SIE0SIm LICENSEE EVENT REPOR~T (LERI EXPIRMS 8/31/19 FACILITY NAME I1)

I I1 I

R.M. Glover, Compliance Manager JAREACODE TEPO NIR

___________ COWI.ETE ONE LIN4EFOR EACH COMPONENT FAILURE D@1CRISED 80 8 13 1813 1 11- 1 31 213 16 IN THIS REPORT IIX CAUSE SYSTEM COMPONENT MANUPAC

_________ ITUPRI REPORTABLE:CUESSE OPNN MANUFAc. EOTU TO NPROS SUPPLEMENTAL REPORT EXPECTEC (IQ1 MOT'EA YES (if v, .,ftv ow xPIcrro suaSNIWOAI DtAT NODAE45 AMSTRACT ML,"ft to F400 gum. i at s0~aHW Rwfm.*A0*--"C* tyffimr 14" 1161 During the period of February 5 through 26, 1990, samples for the Boric Acid Tan (BT) ndthe Refueling Water Storage Tank (FWST) were collected by Chemistry (CHM) to comply with Technical Specification (T/S) requirements, On February 5, CHM had been informed by Operation (OPS) personnel that the BAT was the declared borated water source. From March 11 through March 13, the FWST was not placed into recirculation and was not sampled due to the use of the Refueling Water (FW) pump for draining of the reactor cavity.

On March 14, 1990, Unit 1 was in Mode 5, Cold Shutdown. CRM contacted the Control Room Operator (CR0) to verify that the BAT was still considered the declared borated water source. CHM was informed that the BAT had been inoperable since March 1, 1990 due to 1NV236B, being tagged out for repair. Following CHM review of data, during the week of March 5 through 12, 1990, CHM missed a T/S sample of the FWST. This event was attributed to inappropriate action, due to the individuals involved not ensuring an operable borated water source. A contributing assigned to deficient communications resulting from poor group interface cause is between CHM and OPS. Corrective actions taken included CHM procedure revisions will supply actions to take when T/S samples cannot be obtained which as well as including a T/S Operability Sheet for T/S items. Also, the above mentioned CHM corrective acitons will be communicated to OPS Shift personnel.

NRC Powm

- I

(S41 ., U.& NUCLEAR REGULATORY COM OISSO LICENSEE EVENT REPORT (LER) TEXT CONTINUATION pRovEo OW NO. 3150-410 E)CPIREs: /3110 9ACILTY NAME II NUMREN (2 )

OOCCKET LEN NISIR ig PAGE VEAR SECUENTIAL REVI:S:::JAVIO6N NUM  : NaIR Catawba Nuclear Station, Unit l 050 lololo 0 41113 90 _0. 1 6i_ 0002 OFO rvc~r (N w wm to m~v4um~

-d& AM Fern ins', Im _E BACKGROUND REFUELING WATER SYSTEM The Refueling Water (EIIS:CB] (FW) System provides a large source of borated water and the necessary equipment to:

1. Supply the Emergency Core Cooling System (ECCS) and the Containment Spray [EIIS:BE] (NS) System during the injection phase following a Loss of Coolant Accident (LOCA);

2. Transfer the borated water between the Refueling Water Storage Tank (FWST) and Refueling Cavity;

3. Provide cleanup of the refueling water by routing the water through the Spent Fuel Pool Cooling [EIIS:DAI (KF) System; and,

4. Provide for various other borated water requirements and miscellaneous flowpaths.

The FWST rormal capacity of 395,000 gallons is sufficient to provide a useable volume exceeding 350,000 gallons. This capacity assures:

a. The volume of borated refueling water needed to increase the boron concentration of initially spilled water to a level that assures no return to criticality with the Reactor at Cold Shutdown and all control rods [EIIS:ROD], except the most reactive Rod Cluster Control Assembly (RCCA), inserted in the core.

b. The volume of water sufficient to refill the Reactor vessel

[EIIS:VSLI above the nozzles (EIIS:NZLI after a LOCA.

c. A sufficient volume of water when combined with ice melt and Reactor Coolant (EIIS:AB] (NC) System spill in the containment recirculation sump following a LOCA to permit the initiation of the recirculation phase.

d. A sufficient volume of water to limit the radiation dose rate at the surface of the Refueling Cavity to approximately 2.5 mrem/hr during the period when a fuel assembly is transferred over the Reactor vessel flange.

e. A sufficient volume of water to allow the station operator adequate time to complete the valve [EIIS:VI alignment required to complete the switchover from the injection mode to the containment sump recirculation mode following a LOCA.

, C PORM -U.S. GPO 198-520- 50 U0070 I- 1J

"PIC WM,39SA UASNUCLEAR REGULATORY COM~auaou UCENSEE EVENT REPORT (LER) TEXT CONTINUATION APPROVED OM NOQ 3,gO..oi FACILI HAM III DOCKIT NUMBER (2) LEM NM BE 16 PAGE 12 Catawba Nuclear Station, Unit 1 0 151o10 1e 1 41 13 016-O9il_ 0 10 0 :3[ 0 TWt fmve spewsb ,uqm4 ves alWWC fw- XM 117122 -Z "

When draining the FWST, the water is routed to the Refueling Cavity and to one of the Boron Recycle [EIIS:CA] (NB) System Recycle Holdup Tanks (RHTs).

Approximately 290,000 gallons of water is drained to the Refueling Cavity while the remainder is drained through the KF purification loop into either one of the RHTs.

The refueling water from the Refueling Cavity is routed back to the FWST by using the normal refueling drain procedure. The water in the RHT is rerouted through the recycle evaporator feed pumps [EIIS:P] into the FWST. The water is brought back into specification by adding demineralized water or boric acid from the boric acid blender.

CHEMICAL AND VOLUME CONTROL SYSTEM The Chemical and Volume Control (EIIS:CB] (NV) System is designed to provide the following services to the NC System:

1. Maintenance of programmed water level in the pressurizer.

2. Maintenance of seal-water injection flow to the NC pumps.

3. Control of water chemistry conditions, activity level, soluble chemical neutron absorber concentration and makeup.

4. Filling, draining, and pressure testing.

The water chemistry, chemical shim and makeup requirements of the NC System are such that the following functions must be provided:

1. Means of addition and removal of pH control chemicals for Startup and normal operation.

2. Control of oxygen concentration following venting and that due to radiolysis in the core region during normal operation.

3. Means of purification to remove corrosion and fission products.

4. Means of addition and removal of soluble chemical neutron absorber and makeup water at concentrations and rates.

compatible with all phases of plant operation including emergency conditions.

The function of soluble neutron absorber concentration control and makeup is provided by the Reactor Makeup Control System employing 4 wt. percent boric acid solution from the Boric Acid Tank (BAT) and Reactor makeup water from the MAC Pon"i 35 4 0 0 WI"I'..Go

U4NFCP.,, USA U.S. WACLIAN NEGULATOR'F C~ogeSN0S UCENSEE EVENT REPORT (LER) TEXT CONTINUATION Poveo DMg NO. 31 W-410s DPIEs: E/31/U FACIurv NAMI 11) DOCKET NU IN 2 Lil NUMER 0 PAGE Is YER SEQUENTIAL =EUO:

Catawba Nuclear Station, Unit I o,0 100 4113 90 011 6 00 OF 0p

. '&w W 341i 11I1

,r (FSw Mr im Reactor Makeup Water Storage Tank (RMWST). In addition, for emergency boration and makeup the capability exists to provide refueling water or 4 wt. percent boric acid from the BAT to the suction of the charging pumps.

Two boric acid tanks are provided. The combined capacity of the tanks contains sufficient boric acid to provide for refueling plus enough boric acid for one Cold Shutdown immediately following refueling with the most reactive control rod withdrawn. There is sufficient capacity with one tank one-third full, to provide Cold Shutdown for the Unit with the most reactive rod withdrawn.

Technical Specification 3.1.2.5 states that as a minimum, one of the following borated water sources shall be OPERABLE (in MODES 5 & 6):

a. A Boric Acid Storage System with:

1. A minimum borated water volume of 5100 gallons,

2. A minimum boron concentration of 7000 ppm, and

3. A minimum solution temperature of 65 degrees F.

b. The Refueling Water Storage Tank with:

1. A minimum borated water volume of 26,000 gallons,

2. A minimum boron concentration of 2000 ppm, and

3. A minimum solution temperature of 70 degrees F.

T/S Surveillance Requirement 4.1.2.5 requires that the above borated water sources shall be demonstrated OPERABLE:

a. At least once per 7 days by:

1. Verifying the boron concentration of the water,

2. Verifying the contained borated water volume, and

3. Verifying the boric acid storage tank solution temperature when it is the source of borated water.

Chemistry procedures require sampling of the FWST once per week and sampling of the BAT twice per week.

EVENT DESCRIPTION On February 5, 1990, Unit 1 was in Mode 6, Refueling. At 0630 hours0.00729 days <br />0.175 hours <br />0.00104 weeks <br />2.39715e-4 months <br />, Chemistry (CHM) Technician A recorded in the Primary CHM logbook turnover notes that the Refueling Water Storage Tank (FWST) was in the process of makeup, and sampling was required. At 1400 hours0.0162 days <br />0.389 hours <br />0.00231 weeks <br />5.327e-4 months <br />, CHM Technician B telephoned the CRO to request that the FWST be placed in recirculation. The CRO informed the technician that makeup had stopped and that Operations (OPS) was concentrating on increasing the levels in the Boric Acid Tank (BAT). Due to the T/S requirement for once per 4AC FPORM 30 -U.S. GPO, l1908-'.- .. ',J70 (9493

1 d 4u " - U. .XLEAR amGUATOnY commima" UCENSEE EVENT REPORT (LER) TEXT CONTINUATION AP"ovEo oM" NQ 3150-4v04

• PmI: /i2iU NmEw OOCKIT OoCaCuy NUMIEN (2) ,AG, (2 Y A ...... A G,UiNTI.  ::L Catawba Nuclear Station, Unit 1 o is 10a IoI 4 1113 910 1 1 611 01 0 I 0 rW~ (ifmmmWa is momt . adm&, NMECs Am. =A'&)I seven day samples to be taken on either the BAT or the FWST, if the FWST was the "declared borated water source" then the sample would need to be taken no later than February 6. The OPS Shift Supervisor informed CHM Technician B that the BAT was the borated water source.

From February 6 through 11, 1990, at 1022 hours0.0118 days <br />0.284 hours <br />0.00169 weeks <br />3.88871e-4 months <br />, Unit 1 was in Mode 6. All required FWST and BAT samples were collected and analyzed by CHM personnel.

From February 11 through 26, 1990, Unit 1 was in No Mode, Core Defueled. CHM personnel collected and analyzed all required BAT and FWST samples.

On February 24, 1990, Unit 1 was in No Mode. CHM Technician C was informed by the CRO that the 1B Residual Heat Removal [EIIS:BP] ND Pump was on and that the Reactor cavity water was being pumped back to the FWST. At approximately 1439 hours0.0167 days <br />0.4 hours <br />0.00238 weeks <br />5.475395e-4 months <br />, Diesel Generator (D/G) lB was removed from service, as a result of work list items related to the Outage.

Unit 1 entered Mode 6 on February 28, 1990. On March 1, 1990, at 0220 hours0.00255 days <br />0.0611 hours <br />3.637566e-4 weeks <br />8.371e-5 months <br />, Unit 1 remained in Mode 6. OPS issued R&R 19-2838 on 1NV236B, Boric Acid to NV Pumps Suction, for MOVATS testing and also issued R&R 10-807 on A and B Boric Acid Transfer Pumps for the 1NV236B work. This action in combination with D/G 1B being out of service necessitated the determination, by OPS that the BAT was inoperable, due to the unavailable BAT water source alignment. This change in BAT status was unknown by CHM. BAT sampling continued at the prescribed interval.

On March 4, 1990, at 0725 hours0.00839 days <br />0.201 hours <br />0.0012 weeks <br />2.758625e-4 months <br />, Unit 1 was in Mode 6. CHM Technician C contacted the Unit 1 CRO to request that the FWST be placed in recirculation for the weekly sample. CHM Technician C was told that the FW pump was currently pumping down the Reactor cavity, and OPS was not able to state when the pump would be available. The CRO would check with the Shift Supervisor about the situation. CHM Technician C called the CRO again at 0832 hours0.00963 days <br />0.231 hours <br />0.00138 weeks <br />3.16576e-4 months <br />, and there had been no determination made. At 1930 hours0.0223 days <br />0.536 hours <br />0.00319 weeks <br />7.34365e-4 months <br />, CHM Technician D discussed the FWST status with the Unit Supervisor and was advised that the draining of the cavity had to be completed to permit FWST sampling.

On March 5, 1990, Unit 1 was in Mode 6. At 0050 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br />, the weekly FWST T/S sample for boron analysis was due, but was not collected as a result of the FW pump being in service for Reactor cavity draining. The FWST was last sampled at 0050 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> on February 26.

On March 9, 1990, Unit 1 was in Mode 6, and at 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br />, CHM Technician B called the Unit Supervisor and asked about the FWST status. The Supervisor stated that the FW pump had been tagged out and that OPS was planning to clear the tagout later in the day.

,.s. CPO. i9eo-MNC FORM 366A "-U.S. CPO. 1988--

3 I1

Pd(P 683ta " U.. NUCLALR NaGULA*,oV CoM~MUIom 131 .UCENSEE EVENT REPORT (LER) TEXT CONTINUATION A,,RovEo oWS .o.3100-o0ol XPIRS: 0/31/0 FACIUrTVNAME* II DOCKETNUMOM (2) LIN MWAEE (0 PAGE (a S........ ff*tf,,1T)':IA~ '**

'WAR ~ii!:ii SuGm4Tan vmR *Mmq MUM N Catawba Nuclear Station, Unit 1 5lololo 41113 90 0o16 _Ol0 017 o0 0 CONCLUSION This Technical Specification violation is attributed to Inappropriate Action, as a result of the individuals involved not recognizing the need to ensure an operable borated water source. The Chemistry personnel, though having contacted OPS personnel on numerous occasions to place the FWST in recirculation for sampling, did not pursue a timely resolution to the problems when continuing interferences occurred. In addition, the information discussed by CHM personnel and OPS personnel, concerning the T/S samples, was not carried out by OPS personnel in a timely manner to avoid missing a T/S sample. In the past, Chemistry personnel have understood that the boron concentrations are provided to OPS to fulfill the requirements of T/S 4.1.2.5.a.1. The requirements of 4.1.2.5.a.2 & 3 are supplied to the CRO by way of the Operator Aid Computer and as required in PT/l/A/4600/02 E, F, & G, Periodic Surveillance procedures.

Therefore, Operations is responsible for the determination of OPERABILITY as stated in T/S 3.1.2.5. CHM personnel concluded that if OPS did not place the FWST in recirculation during the period of March 1 through 15, OPS must have maintained the BAT as the declared borated water source. In addition, CHM had been told by OPS personnel earlier in the outage that the BAT was the borated water source. Conmmunication between the groups is considered a contributing cause in thac it did not achieve the necessary clarity and responsiveness to avoid the -/S violation.

The inoperability of D/G 1B and the tagout of 1NV236B necessitated the inoperability of the BAT, due to loss of its boron injection flow path. This INOPERABILITY was declared based on T/S 4.1.2.1b, which requires at least once per 31 days that each valve in the flow path is in its correct position. The current Chemistry sampling schedule for FWST and for the BAT is established in CHM procedures. If this schedule is followed as stated, regardless of concerns with the "declared borated water source", the required analyses should be.

completed per T/S.

The CHM staff completed changes to Chemistry Management Procedure 3.4.17, on April 5, 1990, which state that if a system needs to be placed in recirculation to collect a T/S sample, OPS is to be informed at the time of the recirculation request, that, if the requested action is not taken by an appropriate time, a T/S violation will occur.

Chemistry Management Procedure 3.4.17 was also changed to include statements on FWST and BAT sampling enclosures which states that the inability to collect a T/S sample is considered the same as being Out-of-Spec. A T/S Operability Notification Sheet (Attachment 1 of Station Directive 3.1.15, Activities Affecting Station Operations) will be issued by Chemistry with a comment that the T/S sample is Out-of-Spec or unattainable.

As a result of this event, emphasis will be placed on ensuring clear communication, focusing on clear description of needed actions and clear understanding of the importance of such actions.

N IC FO*iN 3 -U.S. GPO. 1966-52o0-584 dOO70 C04111*

FMRCFem WAg U U MJCLLAR REGM4AYORV COMMIMGft UCENSEE EVENT REPORT (LER) TEXT CONTINUATION . ,"W,0 oeM N ,31Wbo0o0 VMPIRUU:/I31/n IACIUTY NAME IM OOCKIT NWU R W Lan NUNSN"10 PA 9 IN YUAN sUIouRTAL NtPSE

]..,.........

v... 81"UMU OWO Catawba Nuclear Station, Unit 05000 413 9 0 _.01 6_ 0 0 6O F 0 T'UT (YVme aw bqW

/P -d~Wf um F~m JWIM CHM Technician D called the CRO at 2037 hours0.0236 days <br />0.566 hours <br />0.00337 weeks <br />7.750785e-4 months <br /> on March 10, 1990, requesting a FWST sample. Unit 1 was in Mode 6. The CRO was asked to place the FWST on the recirculation pump so that the tank could be sampled in approximately 30 minutes. At 2045 hours0.0237 days <br />0.568 hours <br />0.00338 weeks <br />7.781225e-4 months <br />, the CRO called CH4 Technician D and said that the recirculation pump would not operate and asked if CHM could sample off of the FW pump. The CHH Technician explained that their sample point was on the line off of the recirculation pump. CHM Technician D completed sampling the FWST at 2130 hours0.0247 days <br />0.592 hours <br />0.00352 weeks <br />8.10465e-4 months <br />.

On March 12, 1990, Unit 1 was in Mode 6. OPS had completed the Reactor cavity draining at 0500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br />. At 0600 hours0.00694 days <br />0.167 hours <br />9.920635e-4 weeks <br />2.283e-4 months <br />, CHM Technician D inquired about the FWST sampling, and was told that the FWST was still aligned to the cavity and recirculation had not begun. Unit 1 entered Mode 5 at 1800 hours0.0208 days <br />0.5 hours <br />0.00298 weeks <br />6.849e-4 months <br />.

CHM Technician B called the CRO on March 14, 1990, with Unit 1 in Mode 5, to verify that the BAT was the declared borated water source, and that the latest FWST sample was collected and analyzed on March 10, 1990. At that time, CHM was informed of the inoperability of the BAT, due to 1NV236B being inoperable. Due to the IB D/G being out of service, INV236B did not have an alternate power source available. CHM personnel were not aware of this condition. At 0900 hours0.0104 days <br />0.25 hours <br />0.00149 weeks <br />3.4245e-4 months <br />, the CRO called CHM Technician B and stated that the FWST had been placed on the FW pump and should be ready for sampling by 1800 hours0.0208 days <br />0.5 hours <br />0.00298 weeks <br />6.849e-4 months <br />.

On March 15, 1990, Unit 1 was in Mode 5. At 0140 hours0.00162 days <br />0.0389 hours <br />2.314815e-4 weeks <br />5.327e-5 months <br />, the Unit Supervisor and CHM Technician E sampled the FWST off of a low point drain, 1FW14, Refueling Cavity to FW Pump Strainer Lo-Point Drain. This sample was taken to ensure that the FWST was sampled within the seven day time frame. At 0800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br />, CHM Technician B called the Unit Supervisor and asked about the BAT lineup and also asked if the transfer pumps were still tagged out. CHM Technician B discussed the conversation on March 14, 1990 with the CRO, stating that the FWST was the declared borated water source. CHM Technician B then asked the CRO how OPS could declare the source without sample results. The response was that the CRO was using the percent level for the FWST to consider it operable.

Following a review of the previous FWST and BAT sample results, the Primary CHM group determined that during the week of March 5 through 12, 1990, CHM personnel missed sampling the FWST on March 5, which violated T/S 4.1.2.5.A.1, sampling frequency of the borated water source.

On April 5, 1990, Unit 1 entered Mode 3, Hot Standby, at 0526 hours0.00609 days <br />0.146 hours <br />8.69709e-4 weeks <br />2.00143e-4 months <br />. Changes were approved for Chemistry Management Procedure 3.4.17 which incorporated notification to OPS of T/S required samples and the possibility of T/S violations if samples are not collected before an appropriate time. A requirement was established for use of a Technical Specification Operability Notification Sheet (TSONS) for samples that are Out-of-Spec or unattainable.

N43PO *a U.S. GPO. 900.70 U9f-cg4*I

NOC Fw. M"A U. NUCLEA m c TuLAraTty

_0 UCENSEE EVENT REPORT (LERI TEXT CONTINUATION A.Rovuo oWe NQ 3190-406 FAILCILITY MAMU (11 DOCK*r NJUSIN (2* LEN NINUR ( PAGE Catawba Nuclear Station, Unit 0 000 413 910 01 6.. 00 08 OF0 TamT tr MMM*8 aI , C 9W"W o t.31110 n I- .~ n.. 118 !2 A search of the Operating Experience Program database for the past 24 months revealed two events, LER 414/89-018 and LER 414/89-05, that involved a missed Technical Specification sample. LER 414/89-018 was concerned with a missed sample of the Cold Leg Accumulator as a result of deficient communication. This event involved insufficient, unclear information communicated during CHM shift turnover. Also, an additional root cause was improper action; with no action taken when required because of lack of attention to detail. Corrective actions included meetings with the shift technicians to emphasize the need for effective turnover information. LER 414/89-05 involved Radiation Protection (RP) and a Turbine Building sump radiation monitor (2EMF31) sample which was not collected in a timely manner due to an inadequate sampling policy. In this event, RP procedures were changed to ensure correct, timely sample collection. This event is not considered a recurring event.

CORRECTIVE ACTION SUBSEQUENT

1) Chemistry Management Procedure 3.4.17 was revised to include:

a. Steps that will ensure that, if a system/component needs to be placed in recirculation or a valve needs to be manipulated in order to collect a T/S sample, OPS personnel are to be informed at the time of the recirculation or valve manipulation request, that if the system is not put in the configuration requested by an appropriate time, then a T/S violation will occur.

b. Steps in Enclosures for Primary Chemistry sampling that direct the CHM Technicians to complete a T/S Operability Statement (TSONS) when a T/S sample is unattainable (which is considered to be the same as being Out-of-Spec). The TSONS will provide the specific information for OPS to follow-up direct actions pertaining to T/S operability.

PLANNED

1) OPS Shift personnel will be informed of the Chemistry section's April 5, 1990 procedure changes to 3.4.17.

2) Management will emphasize the accountability of all personnel to ensure clear communication and understanding of needed action and its importance. This effort will include review and (as much as practical) standardization of each group's methods and paths of communication with Operations. This effort will be discussed with Operations personnel with emphasis on their obligation to "reach into" interfacing activity areas and ensure understanding and appropriate action.

NAC PONi 306A -US GPO. ý0070 19-43)

N~~C~m~~gA UASMJLEAP REGULATORY cmeI UCENSEE EVENT REPORT (LER) TEXT CONTINUATION AIovio 0M ,OW 31_Wm-40 FACILIY NAME (11 0eT NUMAf W NI* "IIN NUMBERI Catawba Nuclear Station, Unit 1 0 I 010101 1113 I1I 6__

910 __ 010 019 0 TMcT Wnreaw mbquou 6 adOP p AW ow4W (170 0 SAFETY ANALYSIS The usable capacity of the FWST is based on the requirement for filling the refueling cavity to a depth that limits the radiation at the surface of the water to 2.5 mrem/hr during the period when a fuel assembly is transferred over the Reactor vessel flange. This function requires more water than is necessary for a post-LOCA safe shutdown.

The NV System maintains the coolant inventory in the NC System within the allowable pressurizer level range for all normal modes of operation. This sysem also contains sufficient makeup capacity to maintain the minimum required inventory in the event of minor NC leaks. Other than the centrifugal charging pumps and associated piping and valves, the NV System is not required to function during a LOCA. During a LOCA, the NV System is isolated except for the centrifugal charging pumps and the piping in the safety injection and seal injection path.

When the Reactor is subcritical, i.e., during Cold or Hot Shutdown, refueling and approach to criticality, the neutron source multiplication is continuously monitored and indicated. Any appreciable increase in the neutron source multiplication, including that caused by the maximum physical boron dilution rate, is slow enough to allow ample time to start a corrective action to prevent the core from becoming critical.

During the period from March 5 through 10, 1990, following the missed FWST boron sample analysis, the Unit was in Mode 6. The FWST was considered the declared or assured borated water souce. All parameters for tank volume, and solution temperature were maintained within required T/S limits. The boron concentration from the February 26 analysis was 2071 ppm, and the concentration from the March 10 analysis was 2148 ppm. It is considered that the concentration did not significantly decrease during this period based on the values for these two samples.

The health and safety of the public were unaffected by this incident.19-431 -U.S. GPO. 00070

EXHIBIT B-8 Cooper Station:

LER 298/86-034-00 (December 18, 1986)

U.S. NUCLEAR REGULATORY COMMISSION NIt Pwa 30"

""43 LICENSEE EVENT REPORT (LER) TEXT CONTINUATION .,"RoVEo GUS ,o J,sa-oo0 0XPIRES:83110i K Lan NU1M81 1K PAGE lS13 PACILITY gAMI IC V ilO N* I SE L  ! A,';

Cooper Nuclear Station o Io5I0oIO 21918 816]-- 0131 4--OIO 01 21OF 0n3 EV tET W ffw @"".a fWA154naw eda NRC F., MUAW 11171 While conducting an evaluation of fuel enrichment requirements to facilitate future extended cycle (18 month) operation, a review of the existing CNS Technical Speci of fications was made by the General Electric Company (GE) to determine the extent During the course of this Technical Specifica any revisions that might be required.

tion review, an apparent violation of paragraph 5.5.B was identified. Paragraph 5.5.B states that, " . . . In addition, fuel in the storage pool shall have a U-235 fuel loading of less than or equal to 14.5 grams of U-235 per axial centimeter of However, GE advised that the barrier fuel, GE Type BP8DRB283, which had assembly".

contained a U-235 loading of approximately 14.6 grams, in been supplied for Cycle 11, Hence, storage of the new fuel excess of the 14.5 grams per axial centimeter limit.

in the Spent Fuel Storage Pool constituted a violation of the Technical Specifications.

was Upon receipt of this notification from GE on November 14, 1986, an evaluation the Spent Fuel Storage Pool.

conducted of all fuel reloads that had been stored in 7

On November 18, 1986, the determination was made that the fuel supplied for Cycle Pool from February 3, 1981 to April 27, 1981 and and stored in the Spent Fuel Storage 10, which was stored in the Spent Fuel Storage Pool from the fuel supplied for Cycle than July 23, 1984 to July 17, 1985, also contained U-235 loading slightly greater At the time of these discoveries, the plant was in 14.5 grams per axial centimeter.

had commenced on a shutdown condition for a refueling/major maintenance outage which October 4, 1986.

This event is being reported in accordance with the requirements specified in of 14.5 10CFRSO.73(a)(2)(i) in that storage of fuel with a U-235 loading in excess a violation of paragraph 5.5.B of the CNS grams per axial centimeter constitutes It appears that this limitation is based upon the U-235 Technical Specifications.

the nominal fuel design parameters associated with the loading which corresponds to of the Spent fuel type considered in the safety analysis conducted to support backfit Fuel Storage Pool in 1978 with high density fuel racks.

provided Amendment 52 to the CNS Technical Specifications, dated June 12. 1Q78, which Fuel Storage Pool, was for installation of high density fuel racks in the Spent The issued by the NRC with the aforer;entioned 14.5 grams per axial centimeter limit.

to provide the technical basis for the criticality calculations which were performed These new design racks were "oased upon General Electric type 8DR283 fuel assemblies.

of 2.83 w/o and a nominal pellet density of assemblies had an average enrichment 95.07 theoretical density (TD). The 150 inch fuel assembly design includes a 6 inch section of natural uranium at its top and bottom. The central 138 inches of these luel assemblies contain an enrichment of 3.01 w/o. The 14.5 grams/centimeter value IF based on this enrichment and the nominal density of 95.07. In establishing this fuel assembly value, however, no consideration was given to deviations from nominal designed and design parameters which are within the tolerances considered in the fuel either licensed by GF. These deviations from nominal parameters may result from mranufacturing tolerances or design improvements.

the fuel supplied bv GE for Cycle 11 was nanutfactuired with an upgraded In addition, As a result, the pellet design incorperating a ,lightly higher theoretical densitt.

With respect to the fuel 14.5 grams per axial centimerer limit was exceeded.

doe to manftfactorting provided lor (Uvcles 7 and 10, the axial linmt was: exceeded toleranLtes within the approved design envelope.

h,A*

S~im**

SF 116A UA. NUCLEAR REGULATORY COMMISSION 043) LICENSEE EVENT REPORT (LER) TEXT CONTINUATION A,0AovEOoN- *nSO31,-O, EXPAIS: 8/3110

,*ACILITV NAME IlI DOCKET wUmsEn (21 LIN IU.IM@1MI& PAGE 131 ISEQUINTIAL I Cooper Nuclear Station Ols 10101 OL 21918 6 -61 01314 -010 I013 °FOI3 TXTIN# nm h *u,*ds we u Ad *&N MAW) nCfwm 7I General Electric has advised that neither the pellet design change nor the man ufacturing deviations, which are within prescribed tolerances, constitute a safety problem. Fuel enrichment had not changed, consequently fuel reactivity had not changed. Criticality calculations performed in 1978 to support issuance of Amendment 52 to the CNS Technical Specifications are still fully applicable to storage of fuel of the present design. Hence, the cause of the Technical Specification violation is attributed to the lack of consideration of allowable fuel design parameter tolerances in calculations performed to support the 14.5 grams per axial centimeter 1 mit, coupled with a failure to recognize the impact of the slightly increased pellet density on the Spent Fuel Storage Pool limits.

Corrective action to be taken will consist of a review of Spent Fuel Storage Pool design for fuel loading and performance of calculations to update storage limits which are prescribed in the CNS Technical Specifications. Ensuing changes to the Technical Specifications determined to be appropriate will be transmitted to the NRC.

ItO ,.. MA IM,2

NRC F- 4 U.S. NUCLEAR REOULATORY COMMISSION APPROVED OMB NO 31504104 LICENSEE EVENT REPORT (LER) EXPIRES 5/31,0 FACILITY NAME III DCE "UuneR I Cooper Nuclear Station 15101010121 918111OF o213 TIL,, f~Storage of Fuel in the Spent Fuel Storage Pool with U-235 Loading in Excess of Terhn a Specification Limits due to Pellet Design Changes & Manufacturer Variances EVENT DATE to1 LEM NUMERM let REPORT OAT* I17 OTHER FACILITIES INVOLVED Ill MONTH OAV YEAR YEAR . EQUENTIAL RBVWON MoNTH DAY YEAR ACILlIT, NAMES DOCKET NUMBERiSI NUMIegn WJMIj 144HDY YA AIIYIb 0151010101 1 11 111 118816 816 013041 2 8 1 0o0 00 ,

OPERATING THIS REPORT IS SUBITTED PURSUANT TO THE REOUIREMENTS OF 10 CR  : (CCc1 Ww0"b ." of, IA. I lI I moot IIll N 30.4021(bl 0406[] 21.11231.1 73 ,,IbI POWE R.40 l l ll I I' l 21.1121 l,l 13 .711)

LEVEL -

1 01 0 1 U 0 0.40 llil)lC E0.]1 l .11211.41 OTHERA. im-, - A. ,-r

__.____________.__IO.?214.1.IIIII. I -d -l r.., ARC 2'1.410B11I1111.1 BO.?11l2l1l1& l 0.31.III1 LICENSEE CONTACT FOR THIS LER 1111 NAME TELEPHONE NtUM*ER AREA COOL D. L. Reeves, Jr. 402 8215 1 -1318 11 1 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIRED IN THIlS REPORT 1131 CAUSE SYSEM COMPONENT MAP#UjFAC REPORTAILE MANUF Tun R 'TO NPROS CAUSE SYSTEM C0OMPONAENT EdNU'AC REPO" TA PLt TUnER TO NP.-US SUPPLEMENTAL REPORT EXPECTED 1141 S~EEPECTED LU IMISSION MONI" 0Av 1EAR "IS tit tn c.pier# EXPECTED STEMISSIO" IDAIN A&STRACT IL- 0"00 WM Ii.0-.OW f1ý. W t f I . V M Q- I'Vot-Iff- ... d I As a result of an Investigation performed by the General Electric Company, and further evaluation performed by CNS personnel, it was determined that new fuel stored in the Spent Fuel Storage Pool for Cycles 7, 10, and thr current cycle, Cycle 11, contained a U-235 loading in excess of that allowed by Technical Specifications, paragraph 5.5.B. At the time of this discovery, a refueling/major maintenance outage was in progress.

1he cause of this problem is twofold in that:

1) The fuel received for Cycle 11 incorporated pellets of a newer design with a nominal density slightly higher than previois designs.

2) The fuel received for Cycles 7 and 10, while manufactured within approved desirr. tolerances, included pellets of a density in excess of the nominal va l tie.

(,eneral Ilectric his advised that while the 1'-235 loading limit of 14.5 grams per axial centimeter specified hv Technical Specil cat ions wa* exceeded, the average fuel enrichment was unchanged ind, therefore, the reactivity of the fuel had not hCen increased. Hence.. the criticality-' calctiiations Pade in supor-rt ti the high densittv Iueli rak upgrade rcrnain tf ilyv app Iicable.

Corrective ;ictio,:; to hI taken will consi.st of ; review oi ,;pent Foiel St(irage PooW dc' lgn for ftiel lolding and further ClLulation0 , to updatv storlage imitý prescr ihud in time CNS lechluital Sp,,cif;,.ltions.

86l2*30442 861-18 PDR ADOCK 05000298 N4C Fn- 344 S PDR

EXHIBIT B-9 Crystal River Unit 3:

LER 302/87-026-00 (December 1, 1987)

.&IC

$4J*w0-i ~

NoJl L..I *ILL4B* mgII o mAOIt CL IUOa 1 : . 1 0l 6 ivi . 0 3 , 1101,

• LICENSEE EVENT REPORT (LERI

?.11O .1 I1 i. i 510 10 10 1JIj2I2FT

"", ,C 2,,,, l ; i, -e:_ uf .oi E ,,'nerjz F,.t,- Ilo, je ow*. -. o Fui.w' -Sse-ijy iti irr n c Lc2 f~.,04I4 LAN N.W1160 -61 1 too., 04' 1pIn~ *aC.jLgT ."VOtlvO .1.

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na C..* .a.-'..a 10 51 0 0 1, 1 8

I 10t 17 Iut]

721 0115 01' Ll 0 NIA015,0 1"817 0 0 a0 MI 6. "I4 74 I0 ISc.-

• I HOes

..... Be,-l g°I ;1.a, M .. Io

, . o omi.I) 20 !I.I,*I.,.,mIj:m-

-ci.lsll coIaT&,V? PON t.6 LIN 011 Ilft uellI

. .~~~. - aaG* I - i a .Ig L. ,4.luFFJT, NUCLEAR SAFETY SUPERvISUR 19 a7 COMIJILl Oft .ml L & aCOin AC- co-Ao..

co 0 o@a l.iLiLVm oDaK.ll1o .ft 7-0*s llsaa i

! , ,III II IC iI$ I A4T I

! I I' ~II I 11 'l:2, I I C,. .$,0, urn I ' ,- ... I.'IC-W se' J1*,Oa u' l' *l A~'+I% 0&* 'l On November 9, 1987, Crystal River Unit 3 wa shut*C dciOW irn a refuel ir* oultage.

Th~e reactor vessel was competely defueled to facilitate inspect ion of the oore f lood valves. Fuel1 andi Control Rcxd as*zlies wo~re being moved in the spent fuel pools in preparation for the core reload. At 1715, while updiating the crnntrol room, fuel location taq boardi, it was noted that a new fuel asemiy, w;th 3.851 peroent U-235 enric~hment had been placed in the "A" Spent Fuel Pool.

Te fuel rack0s in the "A" Spent Fuel Pool are I imite to storage of fuel assentblies with 3.5 percent or less UJ-235 enrici-ment. This event Wras cauLse by

' personnel error. When move sheets *ere being pr*i~ to mov'e a fuel assembly incwn location M42 in the "B" Spent Fuel Pool to the "A" Spent Fuel Pool, location M43 was iri99dvertantly written instead of The mislocated r42. fuel assetly was ercentvd trth the *lA" Spent Fuel Pool upon detection of its

islcat on. IerrepenWent review se of ove seits, prior to actual fuel movemnt, hMs bee-n irrmpementod.

Poll...

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On November 9, 1987, Crystal River Unit 3 was shut dc.n in a refuelig Mhe reactor vessel was completely defueled to facilitate inspection outage.

of the core flood valves. Fuel ard Control Rod assemblies were being moved in the spent fuel pools in preparation for the core reload. At 1715, while updating cxntrol rom fuel location tag board, it was noted that a new fuel the assembly, wth 3.851 percent U-235 enrichment had been placed in the "A" Spent Fuel Pool.

'ftis fuel racks in the "A" Spent Fuel Pool are limited to storage of fuel assemblies with 3.5 percent or less U-235 enrichment. This event w-as caLsed by a personnel error. When move sheets were being prepared to move a fuel asssemb]y t x*n location M42 in the "B" Spent Fuel Pool to the "A" Spent Fuel Pool, location M43 was inadvertantly written instead of M42. The mislocated fuel assembly was removed from the "A" Spent Fuel Pool upon detection of its mislocation. Independen t review of move sheets, prior to actual fuel moverent, hMs *een implemented.

Nd,

-I2990048 8 '1201

[I*C 11

LICENSEE EVENT REPORT ILERI TEXT CONTINUATION .. , , W 4,*,-W I.-4 3, (a~~~~ N.O at A c i G . 3 ,0 2 a 7,;-0 1216 i-':0 10 01 3 o0 1 3 PREVICa.S SIM1IAR EVERM This is the first occurrenc of this type at Crystal River Unit 3.

EXHIBIT B-10 Hope Creek Station:

LER 354/95-042-00 (March 25, 1996)

U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3160-0104 NRC FORM 366 (4-96) -

EXPIRES 04130138 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS MANDATORY INFORMATION COLLECTION REQUEST: 50.0 HR.

REPORTED LESSONS LEARNED ARE INCORPORATED INTO THE LICENSEE EVENT REPORT (LER) IUCENSING PROCESS AND FED BACK TO IDSTRY. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (-6 F33), U.S. NUCLEAR REGULATORY CONWNSSION, WASHINGTON, DC 201550001, AND TO (See reverse for required number of THE PAPERWORK REDUCTION PROJECT digits/characters for each block)

AtLITY I N33) PNGEIII 05000-354 1 OF 4 HOPE CREEK GENERATING STATION

'rn'Te (4)

Fuel Bundle Confirmed to be Misoriented during an Operating Cycle EVENT DATE (5) LER NUMBER (6) R (7) OTHER FACILUTIES INVOLVED (M)

SEQUUITI-X.

SIg. L-ACLI ITr ITam o MY TUN I NAME OOKEIrNUMB..

D" y'"

man man1 05000 ME IUMMI 1 2 95 95 04 -- 03 25 9W FACIt NAME 12 05O00 OF 1CR  :(Check one or more) (11)

O AING 5 THIS REPORT IS SUBMI-TED PURSN TMENTS 50.T3(a)(2)(vit)

MODE (9) 0 (b) 20.t203(a)(2)(v) 50,73(a)(2)(1 20.23(a)(1) 20.2203(a){3)(i) 50.73(a)(2)(ii) 50.73(a)(2)(x)

POWER

.EVEL (10) 20.73(a)(2)-) 20.2203(aX3)(1i) 50.73(a)(2)(ii) 7371 20.203(aX4) 50.73(a)(2)(iv) X OTHER 20.2203(a)(2)(ii) beiaw or 50.A6(c)a1) 50.36(c)(2) 0.73(a)(2)(v Votary 7Report

-20.03(a)(2)(-)

LICENSEE GUNTIAGCT FOR THIS LIEK (1Z!)

NRA ITELEPIONI NUMBER Include Area Co")e Jeff Keenan, Licensing 609 - 339 - 5429 O IU V~A NE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

C A:M E,f SY S TM

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• Z,  ::::::::::: C AU*S E S Y ST E M* M U UT**. )111w5,1 a109,'P~ wrME 1EXPECTEI SABMISSR MONTH S (I complete EXPECTED SUBMISSION DATE).

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On December 12, 1995, one reactor core fuel bundle was verified to be misoriented by 180 degrees. This bundle was confirmed to have been misoriented for the last cycle of operation. The event occurred during the last refueling outage (RFO5) when a refuel bridge operator failed to In correctly rotate a bundle when moving it within the reactor core.

addition, the independent verification processes failed to identify the error. There was no safety consequence to plant operation due to this event; however, to share industry information this report is being submitted voluntarily.

Causes of this event are less than adequate procedural and human factor controls being established for the core verification process. Corrective actions included revisions to procedures and additional training with personnel performing core verification activities. In addition, an to the next assessment of fuel movement practices will be completed prior refueling outage.

NRC FORM 366 (4-95)

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (496)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1) DOCKET LER NUMBER (61 PAGE (3)

HOPE CREEK GENERATING STATION 05000-354 I 1 2 OF 4 95 -- 042 -- 00 TEXT (ifmom space is rquie, use adtional copies Of NRCHFom 366A) (17)

PLANT AND SYSTEM IDENTIFICATION General Electric - Boiling Water Reactor (BWR/4)

IDENTIFICATION OF OCCURRENCE TITLE: Fuel Bundle Confirmed to be Misoriented during an Operating Cycle

'rent Date: December 12, 1995 CONDITIONS PRIOR TO OCCURRENCE Plant in OPERATIONAL CONDITION 5 (Refueling)

Reactor at 0% of Rated Power DESCRIPTION OF OCCURRENCE On December 12, 1995, while shutdown for refueling, a visual inspection of the reactor core by refueling bridge personnel revealed a fuel bundle that was apparently 180 degrees out of proper orientation. Supervision was immediately notified and the bundle was verified to be misoriented. The misoriented bundle was positioned in a North-East (NE) orientation in lieu of the proper South-West (SW) orientation. A review of core verification video tapes from previous refueling outages confirmed that the bundle was

)soriented during the last cycle of operation.

A review of records has revealed that the mispositioning occurred at 0736 hours0.00852 days <br />0.204 hours <br />0.00122 weeks <br />2.80048e-4 months <br /> on Sunday, April 3, 1994. The bundle was picked up in a NE orientation and not rotated to the SW orientation during the fuel move.

Core verification, comprising a video monitor review of the core, was performed at that time. As part of the verification, bundle orientation was reviewed by looking at four bundles at a time (a fuel cell) during a continuous scan of the core by the refueling bridge camera.

NRC FORM 366A (4-95)

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (4-95)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1) DOCKET LER NUMBER (6 PAGE (3)

HOPE CREEK GENERATING STATION 05000-354 I1i "4 3 O*

""95j 3 OF 4 95-042 -- 00 TEXIf mre space ts requmid. use addibonal copes of NRC Form 366A) (17F ANALYSIS OF OCCURRENCE Fuel assemblies are arranged in the core according to a design that meets reactivity control requirements and core operating limits. Bundle orientation is an attribute which has an effect on this design. Multiple administrative barriers are in place to decrease the probability of bundle misplacement. Bundle placements are controlled according to procedures "Conduct of Fuel Handling" (NC.NA-AP.ZZ-0049(Q)) and "Refueling Platform and

"*iel Grapple Operation" (HC.OP-SO.KE-0001(Q)). These procedures require iel moves to be independently verified by the refueling floor bridge operator, spotter and refueling Senior Reactor Operator (SRO). A channel fastener (spring clip), located on top of the fuel assembly, acts as a physical aid in ensuring proper bundle orientation. In addition, after all fuel movements are completed, a core verification is performed in accordance with procedure "Verification of Fuel Location" (HC.RE-FR.ZZ-0008(Q)). This procedure specifically requires two scans of the core, one for identification numbers and the other for proper orientation. Additionally, this procedure had incorporated the recommendations of Service Information Letter (SIL) 347 concerning misoriented fuel bundles.

Any one of the above discussed barriers alone should have prevented the event. However, the fuel was misoriented by the refueling bridge operator, not accurately verified by the other bridge operating personnel, and not accurately verified during the independent core verification.

,PARENT CAUSE OF THE OCCURRENCE The causes for the initial bundle placement and fuel bridge verification errors have been inconclusive. The long time before discovery of the event has hindered the collection of relevant personnel data surrounding the events on the bridge at the time of the error. Although unable to develop a definitive causal factor, a comprehensive corrective action is in place to critically review fuel movement practices.

The procedures for core verification have been reviewed and have been determined to be deficient in detail, scope and level of independent review.

Specifically, the procedure was less than adequate in providing sufficient detail for "independent" reviews. Scope of the procedure was less than adequate in that it emphasized serial number checking over orientation and was ambiguous regarding the secondary review being limited to serial numbers. In addition, the procedure had less than adequate consideration for human factors controls in the taping and verification review. Finally, there was an inadequate self verification process for documenting the orientation check and having review aids for the orientation check.

NRC FORM 366A (4-95)

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (4-95)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1) DOCKET LER NUMBER (6 PAGE (3)

HOPE CREEK GENERATING STATION 05000-354 i Y-- 4 OF 4 TEXT Af more space 7- reqired use addtanal copes of NRC Form 366A) (95- 042 00 1 There were less than adequate human factor controls built into the core verification process. Verifiers document the bundle number; however, for the orientation check they are reviewing the monitor passively and react only if a problem is observed. In addition, the monitor's focus tended to be only on the channel clips. A view of the complete fuel cell would allow the verifier to have multiple indicators to assess proper orientation. A strengthening of these human factors issues will further reduce the probability of a fuel bundle misorientation event.

SAFETY SIGNIFICANCE This event had no safety significance. The misoriented fuel bundle and the adjacent fuel bundles, operated within fuel design limits during the cycle of concern. A thorough analysis concluded that thermal power, shutdown margin, average linear heat generation rate, minimum critical power ratio and linear heat generation rate were all minimally affected. Technical Specification limits were maintained throughout the cycle.

PREVIOUS OCCURRENCES There have been no previous reported events involving a fuel bundle being misoriented for a cycle of operation. However, a limited number of fuel bundle seatings and one misorientation have been corrected during the core verification process in the past.

ýORRECTIVE ACTIONS

1) The procedure for "Verification of Fuel Location", HC.RE-FR.ZZ-0008(Q),

was revised prior to the current outages core verification to correct inadequacies concerning detail, scope, and self verification.

2) The event was reviewed and self verification was stressed with current fuel handlers and reactor engineers prior to recommencing fuel movement.

3) A comprehensive assessment of fuel movement practices will be performed.

The assessment will be completed prior to the next refueling outage.

NRC FORM 366A (4-95)

EXHIBIT B- 11 McGuire Unit 1:

LER 369/94-005-00 (August 10, 1994)

I V

RECF-1!'D,

"*94 AUG 29 P4:15 August 10, 1994 U.S. Nuclear Regulatory Commission Document Control Desk Washington, D.C. 20555

Subject:

McGuire ar tion Unit 1 Docke o. 50-369 Volu a y Licensee vent Report 369/94-05 Probl on Process No.: I-M9'4-0801 Gentlemen:

Attached is a voluntary Licensee Event Report 369/94-05 concerning the Boron dilution of the Unit 1 Spent Fuel Pool during drain down and decontamination of the Transfer Canal. This report is being submitted voluntarily and is not required per 10 CFR 50.73. This event is considered to be of no significance with respect to the health and safety of the public.

Very truly yours, T.CK M cM eekin RJD/bcb Attachment xc: Mr. S.D. Ebneter INPO Records Center Administrator, Region II Suite 1500 U.S. Nuclear Regulatory Commission 1100 Circle 75 Parkway 101 Marietta St., NW, Suite 2900 Atlanta, GA 30339 Atlanta, GA 30323 Mr. Victor Nerses Mr. George Maxwell U.S. Nuclear Regulatory Commission NRC Resident Inspector Office of Nuclear Reactor Regulation McGuire Nuclear Station Washington, D.C. 20555 9408180023 940810 PDR ADOCK 05000369 5 PDR

bxc: B.L. Walsh (EClIC)

P.R. Herran (MG01VP)

R.C. Norcutt (MG01WC)

K.L. Crane (MG01RC)

B.F. Caldwell (MG01VP)

R.N. Casler (EC05N)

S.G. Benesole (ONS)

G.H. Savage (EC06E)

G.B. Swindlehurst (EC11-0842)

M.S. Tuckman (EC07H)

R.F. Cole (EC05N)

D.B. Cook (ECl3A)

G.A. Copp (EC050)

Tim Becker (PB02L)

J.I. Glenn (MG02ME)

P.M. Abraham (EC08I)

Zach Taylor (CNS)

L.V. Wilkie (CN03SR)

D.P. Kimball (ON05SR)

NSRB Support Staff (EC 12-A)

(,C 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED 3h3Jý 0104 STATD BURDEN~ PERQ j~E WIT*HTHA.A LICENSEE EVENT REPORT (LER) E , Sp FACILITY NAME(1) Ri (2) PAGE(3)

M*Guire Nuclear Station, Unit 1 05000 369 1 OF 7 TIrl-(4) Boron Dilution of the Unit 1 Spent Fuel Pool During Drain Down and Decontamination of the Transfer Canal.

EVENT DAT 5) LER NUMBER( 6) REOTDATE (7) OTHER FACILITIES INVOLVED( S HONlS DAY YEAR YEAR SEQUENTIAL REVISION MONTH DAY YEAR FACILITY NAMES OCKETNUMBERS)

NUMBER INUMBER 05000 07 1i 94 94 05 0 08 10 94 05000 ONW:RATI THIS RP IS SUBITTED PURSUANT TO REOUIREMENTS OF 10CPR f Check one or more of the followinal 111 HMD:E(9) i1 20.402(b) 20.405(c) 50.73(a)(2)(iv) 73.71(b) lOLM 100% 20.405(a)(1)(i) 50.36(c)(1) 50.73(a)(2)(v)

I 73.71(co LEV:*(1o) 20.405(a)(1)(ii) 50.36(c)(2) 50.73(a) (2)(yii) X OTHER ecif n 20.405(a)(1)(111) 50.73(a)(2)(i) 50.73(a)(2)(viii) (A) Abtcact jilow nor. in r aNd &eA

)

20.405(a)(1)(iv) 50.73(a)(2)(11) 50.73(a) (2)(viii)(B) 20.405(a)(1)(v) 50.73(a)(2)(111) 50.73(a) (2)(x)

-l= W M Pul TO' r12 1

-U FRicky Zýý .ý I I J. Deese, Manager, McGuire Safety Review Group ARM CODE 704 i 875-4065 II CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE CAUSE SYSTEM CcMPONENT MANUWACTURER REPORTABLE TO NPRDS TO NPRDS sPLENTL RE9POM EEC'D 14) EXPECTED MONTH DAY YEAR SU*MISSION TIM (if esa, complete EXPECTED SUBMISSION DATE) X NO DATE(15)

AEBSIACT (Limit to 1400 spaces, i.e. approximataly fifteen single-space typewrltten lines (16)

This report is being submitted voluntarily to provide information and lessons learned re--yrding a Reactivity Management Event. On July 10, 1994, with Unit 1 operating in Mode 1

( Operation) at 100 percent power, Mechanical Maintenance personnel began the drain down of cne Unit 1 Spent Fuel Pool Transfer Canal. During the drain down, a demineralized water misting system was used to keep the pool walls wet to minimize potential airborne contamination. Approximately 28,000 gallons of demineralized water was added to the pool during the decontamination process. The addition of the demineralized water lowered the Boron concentration from 2105 parts per million (ppm) to 1957ppm. The Technical Specification requires a Boron concentration >/= 2000ppm. The Action Statement to suspend fuel movement while the Boron concentration is less than 2000ppm was not violated. Boric Acid was added to the pool to bring the Boron concentration above 2000ppm. This event has been assigned a cause of improper Managerial Methods. Corrective actions include heightening the awareness of site personnel to Reactivity Management concerns, evaluation of work processes/controls, rewrite of the procedure used, incorporation of work involving complex ev-olutions and multiple interfaces into the Risk Assessment Process.

NRC Form 366

LZRVOFM 3-56A U.S. NUCLEAR REGULATORY COMMISSION APPROVED By 0MB NO. 3150-0104 EXPIRES 5/31/95 LICENSEE EVENT REPORT ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWAWD (LER) TEXT CONTINUATION COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (M!NBS 7714), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE MrAN nFT AN RHnCET . WASHTNGTON OC 2050n-FACILITY NAME(1) DOCKET NUMBER(2) LER NUMBER (6) PAGE(3)

YEAR SEQUENTIAL REVISION NUMBER NUMBER McGuire Nuclear Station, Unit 1 05000 369 94 05 0 2 OF 7 This is a voluntary LER.

E-VALUATION:

Background

171*ve [EIIS:ISVI 1KF-122, Fuel Transfer Tube Isolation, is located in the Spent Fuel Pool

'I Transfer Canal and is used to isolate the SFP from the Refueling Cavity in the z-,actor Building. During normal operation, a blank flange is installed on the Reactor Building side of the Fuel Transfer Tube and valve 1KF-122 is open. This allows SFP water to enter the Fuel Transfer Tube supplying a source of borated water to the Standby Makeup Pump. This pump is part of the Standby Shutdown System (SSS) and provides water to the Reactor Coolant (NC) system [EIIS:AB] and the NC pump [EIIS:P] seals if normal sources are Lost. The SSS is required to be operable during Modes 1 (Power Operation), 2 (Startup),

and 3 (Hot Standby). Technical Specification 3.9.12a requires the Boron concentration in the SFP to be maintained at >/= 2000 parts per million (ppm). The associated action mtatement requires that all fuel movement be suspended if the Boron concentration is found to be below 2000ppm.

w'-suription of Event

ý.is report is being submitted voluntarily to provide information and lessons learned regarding a Reactivity Managenent Event. On July 5, 1994, with Unit 1 operating in Mode 1 (Power Operation) at 100 percent power, Mechanical Maintenance personnel performed preliminary work in preparation for the drain down of the Fuel Transfer Canal (FTC). The w*ork included the installation of approximately 26 feet of 3/4 inch PVC pipe along both sides of the FTC. Approximately 1/16 inch holes had been drilled in the pipe at 3 to 5 inch intervals. The pipe was capped at one end and connected to a standard 3/4 inch hose on the other end. The hose was connected to a demineralized water line, but not charged.

The purpose of the PVC pipe was to provide a mist of water to the walls of the FTC while the canal was being drained. This would ensure that the walls stayed wet to minimize potential airborne contamination.

On July 10, 1994, at approximately 0030, Mechanical Maintenance personnel prepared to drain down the FTC to allow the Fuel Transfer Tube Isolation valve, 1KF-122 to be replaced. Prior to beginning work, the team held a pre-job briefing and contacted Operations personnel to obtain approval to begin work.

LrUaMRM 366A U.S. NUCLEAR REGULATORY CO'*KISSIO APPROVED BY CMB NO. 3150-0104 EXPIRES 5/31/95 "LICENSEE EVENT REPORT ESTIMATED BURDE PER RESPONSE TO COMPLY WITH THIS INRXRMATION COLLECTION REQUEST: 50.0 HRS.

(LER) FORWARD TEXT CONTINUATION COENTS REGARDING BURDEN ESTIMATE To THE IrNFOIION AND RECORDS KMIAGEIMEtr BRANCH CMNBB 7714),* U.S. NUCLEAR REGULATORY CCOMMSSION, WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE Qr MAN A (r F MF T AN DRBUD GE T W. AIS.N GTO N fn r 11nuA A FPkCILITY NAME(j) DOCKET NUMBER(2) LE ?IUMBER(6) PAGE(3)

YEAR SEQUENTIAL REVISION NUMBER NUMBER Mc~uire Nuclear Station, Unit 1 05000 369 94 05 0 3 OF 7 Uhe Mechanical Maintenance Team installed the Weir Gate and inflated the seals per Operations Procedure OP/0/A/6550/14, Draining and Filling of Spent Fuel Pool Transfer Canal and Cask Area. Operations personnel tagged the valve supplying the air to the seals in the open position. The Maintenance Team then lowered a submersible pump into the FTC and contacted the supervisor of a multi-skilled shift work team (SPOC) responsible for draining the FTC. A SPOC Team member was assigned to monitor the drain down process and

,re the pump when the canal was empty. The Maintenance Team started the pump and

...ed on the mister system to keep the FTC walls wet.

The Maintenance Team instructed the SPOC Team member to monitor SFP level, Weir Gate seal pressure, and pump operation. The SPOC Team member was also asked to check the Weir Gate eaals for leaks and ensure that the FTC walls stayed wet to minimize potential airborne contamination. During the day shift on July 10, 1994, Operations Control Room personnel went to the SFP Building and observed the drain down/mister operation. The Control Room Staff discussed the effects of the mister system on Boron concentration in the SFP. They referred to the SFP makeup procedure and decided that the system would not add more demineralized water to the pool than was allowed by the makeup procedure.

At approximately 2045, the drain down was complete and the pump was secured. To ensure that the FTC walls stayed wet, the mister system was allowed to continue to run. No ific instructions had been given to the SPOC team about turning it off.

On July 11, 1994, the Maintenance Team pumped the water that was added to the FTC by the zuister system out of the FTC so the Mechanical Maintenance team could begin work on valve 1KF-122. They also throttled the mister system back to reduce the amount of water being added to the FTC. Radiation Protection personnel had taken radiation level readings and believed the risk of airborne contamination had been reduced.

On July 12, 1994, Radiation Protection personnel contacted Chemistry personnel and informed them about the demineralized water that had been added to the pool. There was a concern about the amount of water that had been added by the mister system and its effect on the Boron concentration in the pool. Chemistry personnel completed sampling of the pool at 1100 and determined the Boron concentration to be 1957ppm.

Enough Boric Acid was added to the pool, to raise the concentration above the Technical Specification requirement of >/= 2000ppm.

I 2Ij? RM .16,6A U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 EXPIRES 5/31/95 LICENSEE EVENT REPORT ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD (LER) TEXT CONTINUATION COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (MNBB 7714), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE U__ jqApAL;FI*j.r ANnM Nu25 X';Q,.g rr WARoTN tr 20501 II FACILITY NAME(1) DOCKET NUMBER(2) LER NUMBER(6) PAGE(3)

YEAR SEQUENTIAL REVISION NUMBER NUMBER McGvire Nuclear Station, Unit 1 05000 369 94 05 0 4 OF 7 cmnclusion This event is assigned a cause of improper Managerial Methods. The following is a list of e xamples/contributing factors.

I' The personnel responsible for execution support for the Maintenance Team allowed the ing system that had been used in the past to be altered without reviewing impact on c.....ineralized water flow and thus Boron concentration.

2 ) The turnover of the job between the Maintenance Team and the SPOC Team was not adequate. The Maintenance Team was familiar with the procedure and was aware of the note in the procedure that stated, "The continuous use of misting hoses will add a substantial amnount of water which when pumped over can cause pool dilution". They did not inform the SPOC of the note and the need to be concerned about how much water was added.

3) Operations personnel questioned the addition of demineralized water to the pool, but did not verify Boron concentration of the pool or ensure that adequate controls were in place to prevent over dilution.

i' The part of the job associated with drain down of the FTC was not discussed or planned in detail. Since the drain down was being performed by an existing procedure and had L-,r performed before without incident, no one saw a need to review the process. The plan for the modification should have included all aspects of the job, including drain down and decontamination of the FTC.

5) Personnel involved with the actual drain down did not see the note in the procedure concerning the potential for diluting the pool and did not recognize that the mister system could significantly affect the Boron concentration of the pool. Personnel interviewed did not have a good understanding of their responsibilities associated with Reactivity Management (Nuclear System Directive 304).

6) The incorrect tags were hung on the air supply valves for the Weir Seals.

OP/0/A/6550/14 specifies red tags (Employee Safety) to be hung on the valves. Operations personnel hung white tags (Equipment Safety) on the valves. The procedure was not followed as required.

LERVOI 3FwA U.S. NUCLEAR REGULM'ORY CO4MISSION APPROVED BY OMB NO. 3150-0104 EXPIRES 5/31/95 LICENSEE EVENT REPORT ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 RRS. FORWARD (LER) TEXT CONTINUATION COMMENTS REGARDING BURDEN ESTIMATE To THE INFORMATION AND RECORDS MANAGEMENT BRANCH (MNBB 7714), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE FACILITY NAME( DOCKET NUMBER(2) LER NUKBER(6) PAGE(3)

YEAR SEQUENTIAL REVISION NUMBER NUMBER McGuire Nuclear Station, Unit 1 05000 369 94 05 0 5 or 7

7) The SPOC team was not qualified to the procedure and had not run the procedure previously. This situation requires that the Supervisor or qualified individual give c:lose direction to the employees involved to ensure adequate completion of the task assigned.

FI The decision, on July 11, to pump the additional water out of the FTC, without rmining the full impact was in error. Emphasis was on the work schedule and desire to

..-.urn the SSS to operation as soon as possible. The Job Sponsor, Radiation Protection

'echnician, Mechanical Maintenance Valve Supervisor, Work Window Manager, Maintenance Team

." Members, and the Maintenance Team Support Technician, reviewed the situation; however, the auiount of demineralized water in the FTC was unknown. The possibility that this amount of water could lower the Boron concentration of the SFP below 2000ppm was not considered.

Corrective actions to prevent recurrence include heightening the awareness of site personnel to Reactivity Management concerns, evaluation of work processes/controls, rewrite of procedure OP/O/A/6550/14 to better clarify the concern for ensuring the misting system does not add enough water to effect SFP Boron concentration, and incorporation of work invoiving complex evolutions and multiple interfaces into the Risk Assessment Process.

view of the Problem Investigation Process data bases for the past 24 months revealed event related to Reactivity Management. Therefore, this event is not considered to be recurring.

This event is not Nuclear Plant Reliability Program (NPRDS) reportable.

There were no radiation overexposures, or uncontrolled releases of radioactive material resulting from this event.

CORRECTIVE ACTIONS:

Imediate: 1) Chemistry personnel added approximately 1000Kg of Boric Acid to the pool.

2) Mechanical Maintenance personnel isolated the Mister system and only used it intermittently to wet the walls.

LEIFO:)W 366A U.S. NUCLEAR REGULATORY COMISSION APPROVED BY IMB NO. 3150 0104 EXPIRES 5/31/95 LICENSER EVENT REPORT ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD (LER) TEXT CONTINUATION COMMIENTS REGARDING BURDEN ESTIMATE TO THE INFOR4ATION AND RECORDS MAZNAGEMENT BRANCH (MNBN 7714), U.S. NUCLEAR REGULATORY CaOMISSION, WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104). OFFICE (I.._____

[;........

__ _____ ... ..... ....... I1Vi..

. .. p~l FAC.ILITY NAE, (l) DOCKET NUMBER(2) LER NUMBER(6) PAGE(3)

YEAR SEQUENTIAL REVISION NUMBER NUMBER McGuire Nuclear Station, Unit 1 05000 369 94 05 0 6 or 7 Suibsequent: Site Management has clarified that the Nuclear Engineering Group is responsible for work associated with the Spent Fuel Pool until improved processes/controls are in place.

v* 'ned: 1) Nuclear Engineering personnel will identify and implement a method to heighten the awareness by appropriate site personnel to Reactivity Management concerns.

2) Nuclear Engineering will evaluate work associated with the Spent Fuel Pool and recommend improved processes/controls to ensure concerns such as Foreign Material Exclusion, Dilution, Fuel integrity etc. are properly addressed.

3) Maintenance Procedure Group will coordinate with Operations and Nuclear Engineering to rewrite OP/0/A/6550/14 to specifically address the decontamination activities.

4) Superintendent of Mechanical Maintenance will ensure that the Risk Assessment process includes a review of work involving complex evolutions and multiple interfaces, not covered by existing processes, to determine if Project Managers are needed.

5) Safety Assurance personnel will lead a review of the Work Control process using the problems identified in this event as examples of specific areas to address.

SAFETY ANALYSIS:

This event had no safety significance and is being provided voluntarily to provide information and lessons learned regarding a Reactivity Management event. The Spent Fuel,--

Pool is designed to contain borated water at >/= 2000ppm Boron. However, the Licensing Basis for the plant does not take any credit for dissolved Boron in the pool for normal operation. The borated water in the pool serves two purposes. One purpose is to provide an additional margin of reactivity control above that which is required by the Final

U.S. NUCLEAR REGULAXMORY COMMISSION APPROVED BY OMB NO. 3150-0104 i y '.sp4,0 . ;6 6A EXPIRES 5/31/95 LICENSEE EVENT REPORT ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD (LER) TEXT CONTINUATION COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (MNBB 7714), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE FACILITY NAME (1) DOCKET NUMBER(2) LER NUMBER(6) PAGE(3)

YEAR SEQUENTIAL REVISION NUMBER NUMBER McG-uire Nuclear Station, Unit 1 05000 369 94 05 0 7 or 7 Safety Analysis Report. It also serves as a source of borated water for the Standby Makeup pump.

The Standby Makeup pump was removed from service to allow draining of the FTC. Therefore, tbe possibility of the diluted water being pumped into the NC System was eliminated.

A', -, the effect on reactivity control within the pool was minimal. Boron concentration

)nly two and one half percent below the Technical Specification limit. The Licensing L--.., for the plant takes no credit for dissolved Boron in the pool under normal conditions. The fuel storage racks provide all of the negative reactivity required to keep K(eff) below .95.

The Technical Specification Action Statement requires that all fuel movement be suspended, if the Boron concentration in the pool drops below 2000ppm. No nuclear fuel was moved; therefore, at no time during this event was the Technical Specification Action Statement violated.

At no time were the health and safety of the public or plant personnel affected by this event.

EXHIBIT B-12 McGuire Unit 1:

LER 369/91-0160-00 (November 25, 1991)

LICENSEE EVENT REPORT (LER)

DOCKET NUIDDER(2) PAGR3 3 FAC.LIzY Amkm(1) lMdcGuire Vuclear Station, Unit 1 05000 369 1 OF 5 VITLE(4) Qualified Fuel Assemblies Were Stored Improperly In The Unit 1 Spent Fuel Pool Due to A Defective Procedure.

REPORT DATEt7) OT'HR FACILITIES INVOLVED( I EVENT DATE 5 LER NUMBERI6)

FACILITY NAMES DOCKET NUMBI(s _

REVISION MWDAY YEAR

""oT DAY YEA Y SEUENTIAL N/A 05000 NUMBER NUMBER 10 24 91 91 16 0 11 25 91 05000 OPERATING NM THIS REPORT Is sUmurrm P AT REQUIREMENTS 0OF OONF (Check one or more of the fallowno)( 11) 20.405(c) 5o.73(a)(2)(iv) 73.71(b) 140oE(9) I 20.402(b) 50.36(c)(1) 5o.73(a)(2)(v) 73.72(c)

EKE 0% 20.405(a)(1)(i) 20.405(a)(1)(ii) 50.36(c)(2) 50.73(a)(2)(vii)

E.EL(10) (Specify in 50.73(e)(2)(viii)(A) Abstract blow 20.405(a)(1)(iii) X 50.73(a)(2)(1) 20.405(a)(l)(iv) 50.73( )(2)(11)) 50.73(a)(2)(viii)(B)

I ~ ~n msm in Tort) 50.73(a)(2)(111) S0.73(a)(2)(2C)_________

20.405(a)(1)(v)

,cry L. Pedersen, I--704Supervisor, Safety Review Group j AREA CODE 1875--4487 COMLETE ONE LINS FOR EACH PNN FAILURE DESCRTBED IN THIS REP03rT 1j)

CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE C-AUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE NPRDS TO NPRDS _ _T_

SUPPLEMENTAL REPORT EXPECTED 14) EXPECTED MONTH DAY YEAR SUBMISSION X No DATE(15)

YES (It yes, complete EXPECTED SUBMISSION DATE)

ABSTRACT (Limit to 1400 spaces, i.e. approximately fifteen single-space typewritten lines (16)

Reactor Unit personnel While reviewing Technical Specification Section 3.9.12, McGuire 1 Spent Fuel Pool in a identified 11 fuel assemblies that had been stored in the Unit 3.9.12. This Limiting manner contrary to the requirements of Technical Specification in Region 2 of the Spent Condition For Operation requires, in part, that fuel stored pattern is employed for Fuel Pool shall undergo 16 days of decay, and if a checkerboard and checkerboard storage unqualified fuel, one row between normal storage locations locations will be vacant. The vacant row provision of the specification was not At the time of discovery at satisfied from March 23, 1990 through October 23, 1991.

in Mode 1 (Power 0900 on October 24, 1991, Unit 1 was defueled, and Unit 2 was a cause of Defective Operation) at 100 percent power. This event has been assigned positions to Procedure. The fuel assemblies in question were immediately moved to establish the required vacant row.

LICENSEE EVENT REPORT (LER) TEXT CONTINUATION DOCKET NUMBER(2) LER NUMBER(6) PAGE(3)

FACILITY NAKE(1)

YEAR SEQUENTIAL REVISION NUMBER NUMBER McGuire Nuclear Station, Unit 1 05000 369 91 16 0 2 OF 5 EVALUATION:

Background

racks The Unit I Spent Fuel Pool (SFP) is composed of two regions of high density storage

[EIIS:RK]. Region 1, which contains 286 locations, has a high density fuel assembly spacing of 10.4 inches on center. This spacing is obtained by using a neutron absorbing material. Region 1 is reserved for temporary core off loading of spent fuel assemblies.

of 9.125 Region 2, which contains 1177 locations, has a high density fuel assembly spacing nches on center. Region 2 provides normal storage for irradiated fuel assemblies.

spent fuel, in Technical Specification (TS) 3.9.12 states that unrestricted storage of the acceptable Region 2, shall be limited to fuel assemblies of a specified burnup within Region 2 Storage.

range of TS Table 3.9-1, Minimum Burnup Versus Initial Enrichment for specified in TS Additionally, the TS requires that fuel not meeting the burnup criteria on each side of the Table 3.9-1 must be stored in a checkerboard fashion (empty locations storage spent fuel assembly) with an open row between the checkerboard and normal locations if stored in Region 2.

racks and fuel Free standing fuel assembly inserts, dummy assemblies, fuel storage Internal assemblies are transferred within the same unit using procedure OP/O/A/6550/11, Transfer of Fuel Assemblies. Steps 3.1 through 3.6 of the procedure detail the process storage employed by the Reactor Unit (RU) Engineers in determining the fuel assembly Enclosures 4.1, Internal Transfer Data Sheet and 4.4, Verification of locations.

semblies to be placed in Region 2, document the assembles initial and final locations, transfer dates, and required reviews and approvals.

Description of Event as directed by step On March 13, 1990, RU Engineer A completed Enclosures 4.1 and 4.4 3.1.1 of procedure OP/O/A/6550/1l. RU Engineer A forwarded the enclosures to RU Engineer B for review and approval.

final fuel On March 23, 1990, nine of the eleven previously designated and approved Operations Fuel assembly locations were changed by RU Engineer A at the request of the for the next core Handling Supervisor to maximize available storage cells in preparation off load scheduled during Unit 1 End of Cycle (EOC) 7. Procedure OP/O/A/6550/ll does not 4.4 when final specifically address the necessity of generating a new Enclosure 4.1 or qualified fuel locations are revised. Consequently, the locations for 9 of the 11 deleted by line assemblies originally recorded on Enclosure 4.1 on March 13, 1990 were through and the new locations were entered on the enclosure. Enclosure 4.1 was forwarded

Y.Tr'wII'*R RV'*dRNI' RRPORT (LERRI TEXT CONTINUATION T T-M-r-KE EVENT REPORT (LERI TEXT CONTINUATION I

FACILITY NAME(1) DOCKET NUMBER(2) LER NUME(b) ra YEAR SEUENTIAL REVISION NUMBER NUMBER Unit 1 05000 369 91 16 0 3 OF 5 MKcGuire Nuclear Station, to to the Maintenance Fuel Handling crew who transferred the assemblies in question assemblies remained locations specified by RU Engineer A. The records indicate that the 1991.

in these locations until the event discovery date on October 24, Conclusion to a Technical Deficency This event has been assigned a cause of Defective Procedure due is obscure. The because the procedural guidance provided by procedure OP/O/A/6550/11 phrasing of the procedure, procedure attempts to convey the intent of TS 3.9.12, but the completing the procedure in a direction

-3pecially Enclosure 4.4, leads the individual For example, Enclosure 4.4 aat does not comply with the full requirements of TS 3.9.12.

of the Spent Fuel Pool are states: "Verify all fuel assemblies to be placed in Region 2 4.5 (see Step 2.3) by within the limits of Technical Specification 3.9.12 and Enclosure This leads one to believe that checking the assemblies' design and burnup documentation".

and Enclosure 4.5 requirements by checking the design and burnup documentation, the TS will be satisfied. This is not the case. Also, although the "checkerboard pattern" is row requirement is contained referred to in the procedure, the only reference to the open (3.9.12.b(3)) pertaining to the in the section of the TS Limiting Condition for Operation storage of unqualified fuel. The storage of unqualified fuel is governed by the checkerboard array and requirements of procedure OP/O/A/6550/11 and TS 3.9.12, i.e.

open row physical barriers. These requirements would prevent the violation of the provision with unqualified fuel. The mis-storage of qualified fuel assemblies would be the most probable method of violating the open row. Therefore, to enhance clarity and the open row requirement

-ccuracy, procedure OP/O/A/6550/11 and TS 3.9.12 should address fuel. Additionally, d its association with the storage of qualified versus unqualified This requires the the TS requirements are not fully included in procedure OP/O/A/6550/l1.

to either stop work on the individual performing the procedure and the procedure reviewer memory to verify that all TS procedure to retrieve the information from TS or to rely on requirements have been satisfied. This is an undesirable situation since the procedure necessary to successfully should be a "stand alone" tool and contain all information complete the task.

This event is not Nuclear Plant Reliability Data System reportable.

for 24 months prior to this event A review of the Operating Experience Program Database that were assigned a cause of identified three LERs, 369/90-14, 369/90-10, and 369/90-33 Defective Procedure due to a Technical Deficiency. None of these LERs involve the same equipment or groups, therefore, this event is not recurring.

or uncontrolled releases of There were no personnel injuries, radiation overexposures, radioactive material as a result of this event.

LICENSEE EVENT REPORT (LER) TEXT CONTINUATION LER NUMBER(6) PAGE(3)

DOCKET NUMBER(2)

FACILITY NAMEE(1)

YEAR SEQUENTIAL REVISION NUMBER NUMBER 05000 369 91 16 0 4 Or 5

?{cGuire Nuclear Station, Unit 1 CORRECTIVE ACTIONS:

the Immoediate: 1) RU personnel determined the cell locations necessary to re-establish vacant row.

in

2) Maintenance Fuel Handling personnel moved the fuel assemblies question to new cell locations determined by the RU personnel.

Planned: 1) Procedure OP/O/A/6550/1I, Internal Transfer of Fuel Assemblies, will be revised by RU personnel to address all TS 3.9.12 requirements, to require the specifically maintenance of the vacant row provision, and 4.4 as necessary completion of additional copies of Enclosures 4.1 and to document changes in fuel assembly locations.

procedures

2) RU personnel will review and revise as necessary other have adequately involving fuel movement to ensure that the procedures addressed all acceptance criteria.

associated with

3) RU Training personnel will initiate additional training Reactivity Management.

SAFETY ANALYSIS:

be stored in a checkerboard configuration TS 3/4.9.12.b (3) requires unqualified fuel to row in the Spent Fuel Storage Pool. In the event checkerboard storage is used, one storage locations is to be kept vacant.

between normal storage locations and checkerboard have evaluated the impact on criticality General office Nuclear Engineering (NE) personnel Using the Keno Va module in the safety caused by the noncompliant fuel pool geometry.

determined that the loss of the SCALE III system of computer (EIIS:CPU] codes, NE has regions does not increase the vacant row between the checkerboarded and normal storage the licensing basis. Therefore, Spent Fuel Pool K eff beyond the value reported in placing the assemblies in the vacant Reactivity Management has not been jeopardized by which was not considered in the row. Additionally, the Boron concentration in the pool,

LICENSEE EVENT REPORT (LER) TEXT CONTINUATION FACILITY NAM(1) DOCKET NUMBER(2) LER NUMBER(6) PAG(3)

YEAR SEQUENTIAL REVISION NUMBER NUMBER MlcGuire Nuclear Station, Unit 1 05000 369 91 16 0 5 OF 5 above analysis, has been maintained at >/= 2000ppm and contributes an extra margin of safety. Therefore, unexpected criticality resulting from the mispositioned fuel assemblies is not a concern.

This event did not affect the health and safety of the public.

EXHIBIT B-13 Millstone Unit 2:

LER 336/92-003-01 (June 25, 1992)

U, S. NUCLEAR REGULATORY COMMISSION APPROVED OMB NO. 3150-0104 NRC Form 366 EXPIRES: 4/30/92 Estimated burden per response to comply with this information collection request: 50.0 hrs Forward comments regarding burden estimate to the Records LICENSEE EVENT REPORT (LER) and Reports Management Branch jp-530). U.S. Nuclear Regulatory Commission, Washington. DC 20555. and to the Paperwork Reduction Project (3150-0104). Office of Management and Budget, Washington. DC 20503.

DOCKET NUMBER (2) I 2AGE-31 FACILTY NAME 11)

Millstone Nuclear Power Station Unit 2 0 51 01 01 0 13 13 61 o 014 TITLE (4)

Spent Fuel Pool Criticality Analysis Error LER NUMBER (61 REPORT DATE (7)1 OTHER FACILITIES INVOLVED 18)

EVENT DATE (5) v _____ MONTH DAY YEAR FACILITY NAMES MONTý- DAY YEAR YEAR 0010 0001501201 OF 10 CFR §: (Check one or more of the following)111)

OPERATING THIS REPORT IS BEING SUBMITTED PURSUANT TO THE REQUIREMENTS MODE 19) - 20.402(b) 20.4021c) S0.73(a)(2)(iv) 73.71(b) s.73(a) (2) Iv) 73.71 (c)

TERSlotyn POWER LEVELI 20.405(a) (1) {i) 50.036 (c) (1) 0 (a)12)Ivii) OTHER Specify in (10) 01 3 10 (1) 20.4051a1(11(ii) L 50.361c112)

• S0.73.

j30Abstract below and in 20.405 (a)(1)I(ii [1] 50.73 (a)()2(i( I 50.73(a)(2)(viiil (A) Text, NRC Form 366A) 20 405fa)ll(iv) X 50. 73[a (2) I1)} 50.731a) 12)1 viiiB)

20. 40 aSO iv( 73a 121 (m 50.73(a) . ] fii)

LICENSEE CONTACT FOR THIS LER 112) itAREA COD* TELEPHONE NUMBER 4E Robert A. Borchert, Unit 2 Reactor Engineer, Ext. 4418 2 0 3 41 41 71- 11 71 9f 1 THIS REPORT (13)

COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN

-AUFC 4MANUFAC-COMPONENT TURER "O

- AaZ=:=*/II

.rBL CAUSE SYSTEI COMPONENT COMPONENT TURER MANUFAC- TO MaIP IIEUAI iiiiiiii)ii!:*i~i CAUSE SYSTE x IB [ [R K C1 4 1 9 10 N  ::!iiii.::*i!i!i:;**~~i:i*ii EXPECTED MONT DAY YEAR SUPPLEMENTAL REPORT EXPECTED 04' SUBMISSION complete EXPECTED SUBMISSION DATE) X NO DATE (15) I I YES Ill yes ABSTRACT [Limit to 1400 spaces. i.e.. approximately fifteen singe-soace typewritten lines) (16) On February 14, 1992, at 1415 hours, with the plant in Mode 1 at 30% power, Northeast Nuclear Energy Company (,NECO) was notified by ABB-Combustion Engineering (ABB-CE) that a calculational error existed was in the criticality analhsis for the Region 1 spent fuel storage racks. NNECO determined that this condition as a condition outside of the design basis of the plant. An immediate report was made to the NRC, reportable and the existing reactivity condition of the spent fuel pool was verified to be in compliance with the plant Technical Specifications. racks for The oriinal effective multiplication factor (Keff) calculated by ABB-CE for the Region 1 fuel storage dimensions, nominal spent fuel pool temperature and 4.5 weight percent enriched fuel assemblies was nominal 0.04 delta 0.9224 (without uncertainties). The discovered error results in an underprediction of approximately Keff. Revised calculations by ABB-CE indicate that Keff is actually 0.963 for the same condiuons. An investigauon by ABB-CE has traced the error to two approximations used in their calculation. to the Criticality analyses to support spent fuel storage rack design changes are complete. and proposed changes by on April 16, 1992. These changes were approved plant Technical Specifications were submitted to the NRC the NRC on June 4. 1992.

NRC Form 366A U.S. NUCLEAR REGULATORY COMMISSION APPROVED OMB NO. 3150-0104 C6_89) EXPIRES 4,30/92 Estimated burden per response to comoly with this Intormation collection request: 50.0 hrs. Forward LICENSEE EVENT REPORT (LER) comments regarding burden estimate to the Records TEXT CONTINUATION and Reports Management Branch Ip:-530). U.S. Nuclear Regulatory Commission. Washington. DC 20555. and to the Paperwork Reduction Project (3150-0104). Otlice 01 Management and Budget. Washington. DC 20503 DOCKET NUMBER (2) LER NUMBER (6) PAGE (31 FACILITY NAME (1) YEAR  ; INLVju Millstone Nuclear Power Station Unit 2 Ol5100101313169 2 0101311 011 02 F 04 TEXT (It more space is required. use additional NRC Form 366A s) (17) Description of Event On February 10, 1992, at approximately 1130 hours, Northeast Utilities (NU) was notified by an independent contractor that a higpher than expected effective multiplication factor (Keff) was calculated for the Region I fuel storage racks. On February 11, 1992, NU notified ABB-Combustion Engineering (ABB-CE) of the potential error in the spent fuel pool criticality analysis. On February 14, 1992. at 1415 hours, with the plant in Mode 1 at 30% power, Northeast Nuclear Energy Company (NNECO) was notified by ABB-CE that a calculational error existed in the criticality analysis for the Region 1 spent fuel storage racks. The Millstone 2 spent fuel storage racks were modified in May 1986, and consist of two regions: (a) Region 1 is designed to store up to 384 fuel assemblies with an initial enrichment of up to 4.5 weight percent C-235. Region 1 was designed to allow fuel assembly storage in every location. The Region I storage racks contain a neutron poison material (Boroflex), and have a nominal center-to-center pitch of 9.8 inches. (b) Region 2 is designed to store up to 728 fuel assemblies which have sustained at least 85% of their design burnup. Fuel assemblies are stored in a three-out-of-four array, with blocking devices installed to prevent inadvertent placement of a fuel assembly in the fourth location. The Region 2 storage racks have a nominal center-to-center pitch of 9 inches. The original effective multiplication factor (Keff) calculated by ABB-CE for the Region 1 fuel storage racks for nominal dimensions, nominal spent fuel pool temperature and 4.5 w/o enriched fuel assemblies is 0.9224 (without uncertainties). The discovered error results in an underprediction of approximately 0.04 delta Keff. Revised calculations by ABB-CE indicate that Keff is actually 0.963 for the same conditions. Evaluations by ABB-CE have confirmed that the Region 2 fuel storage racks are not affected by the error. NNECO determined that this condition was reportable as a condition outside of the design basis of the plant. An immediate report was made to the NRC, and the existing reactivity condition of the spent fuel pool was verified to be in compliance with the plant Technical Specifications. All fuel movement in the spent fuel pool had previously been restricted due to the observed degradation of the neutron poison material in the Region I fuel storage racks. No automatic or manual safety systems were required to respond to this event. II. Cause of Event An investigation by ABB-CE has traced the error to two approximations used in their calculation. First. ABB-CE used an incorrect treatment of the self-shielding effect in Boraflex for the epithermal energy group. This resulted in an overestimation of the neutron absorption in Region I and thus a lower calculated Keff. Second, ABB-CE used a geometric buckling term corresponding to a sparsely populated and unpoisoned array as an approximation of buckling in the poisoned configuration. This approximation also contributed to a lower calculated Keff in Region 1. Ill. Analysis of Event This event is being reported in accordance with 10CFR50.73(a)(2)(ii)(B), which requires the reporting of any event or condition that results in the nuclear power plant being in a condition outside the design basis of the plant.

  -1     -_   -

U.S. NUCLEAR REGULATORY COMMISSION APPROVED OMB NO. 3150-0104 NRC Form 366A EXPIRES 4130/92 Estimated Ourden per response to comply with this LICENSEE EVENT LICNSE REPORT RPOR EEN (LER) (ER collection request: comments regarding ourden 50.0 hrs. Forward estimate to the Records TEXT CONTINUATION and Reports Management Branch (p-530). U.S. Nuclear Regulatory Commission. Washington. DC 20555. and to the Paperwork Reduction Project (3150-0104). Office of Management and Budget. Washington. DC 20503. LER NUMBER (61 PAGE (3) FACILITY NAME (1) DOCKET NUMBER (2) YEAR N1M Millstone Nuclear Power Station Unit 2 l l_10 1316 9122l 013 OF 041 TEXT (If more space is required. use additional NRC Form 366A s) (17) The safety consequence of this event is a potential uncontrolled criticality event in the spent fuel pool. Upon consideration of the following factors, a significant margin to a critical condition was always maintained and, therefore, the safety consequences of this event were minimal: (a) The boron concentrauon of the spent fuel pool is procedurally controlled at greater than 1720 ppm, and is typically maintained at greater than 2000 ppm. (b) All new fuel assemblies previously stored in the Region 1 fuel storage racks had been arranged in a 2 out of 4 checkerboard array. (c) The maximum initial enrichment of any fuel assemblies previously stored in the Region 1 fuel storage racks was less than 4 weight percent U-235, which is less than the design enrichment of 4.5 weight percent U-235. (d) All discharged fuel assemblies previously stored in the Region 1 fuel storage racks have sustained at least one cycle of burnup. IV. Corrective Action Criticality analyses to support spent fuel storage rack design changes are complete. and proposed changes to the plant Technical Specifications were submitted to the NRC on April 16, 1992. These changes were approved by the NRC on June 4, 1992. These changes split Region 1 into 2 regions. Region A and Region B. Reeion A can store up to 224 fuel assemblies, which will be qualified for storage by verification of adequate average assembly burnup versus fuel assembly initial enrichment (reactivity credit for burnup). Region B can store up to 120 fuel assemblies with an initial enrichment of up to 4.5 weight percent U-235 and other assemblies which do not satisfy the burnup versus initial enrichment requirements of either Region A or Region C (formerly Region 2). Fuel assemblies -ill be stored in a 3 out of 4 array in Region B. with blocking devices installed to prevent inadvertent placement or storage of a fuel assembly in the fourth location. Region C is the ne- designation for the existing Region 2 storage racks. This alphabetic storage rack designation is a human factors consideration, designed to minimize the probability of a fuel assembly movement error and to provide a historical distinction between the various fuel pool configuration records. The attached fiure shows the new arrangement of the spent fuel pool. V. Additional Information There were no failed components during this event. Similar LERs: 77-23, 80-05, 83-07, 85-01, 86-10 and 91-10 Spent Fuel Storage Racks Manufacturer: Combustion Engineering Model: Hi-Cap Spent Fuel Storage Module Ells Code: DB-RK-C490

NRC Form 366A U.S. NUCLEAR REGULATORY COMMISSION APPROVED OMB NO 3150-0104 (6_89) lEXPIRES: 4/30/92 Estimated burden per resoonse to comply with this LICENSEE EVENT REPORT (LER) information collection request: 50.0 Mrs. Forward comments regarding burden estimate to the Records TEXT CONTINUATION and Reports Management Branch (p--530). U.S. Nuclear Regulatory CommissKin. Washington. DC 20555. and to the Paperwork Reduction Project (3150-0104). Office of Management and Budget. Washington. DC 20503. DOCKET NUMBER (2) LER NUMBER f6) PAGE (3) FACILITY NAME (1) Millstone Nuclear Power Station " I Unit 2 01O l 31316 12 01013 t 014 OF 014 TEXT (I1 more space is requare, use additional NRC Form 366A'S) (17) C14 H z F-. z z 0C 05 IXI 0Q

                                                             .....,   ~~    -.. .... .T-                                                                .i A:0 I

NRC Form 366

EXHIBIT B-14 Oconee Unit 1: LER 269/96-001-00 (February 7, 1996)

FNAC FORM 366 U.S. NUCLEAR REGULATORY COMMISSON APOVEJD BY OMI NO. 315*104 4 VA EXIIIES 041301 I JNPV PSi N IaiN It ant Pno WMATM* mW*oni" 3UlOMis LICFr'SEE EVENT RT;PORT (LER) .uMofTM I N PC1AN WO* SL mXMh KI*UAIMW UOmMOt 6119u061 K 2m0010. A#10uOiI Pm u MUCTM rmW1914 OAuOF UNINTm aM 110!. a r .verse for sqwred nume of s siL K 20M d0LchWasctesr for each block) r&cUJTv %AIR (T 001P . t VA PAP11111in MO ta 05000 269 1 OF: IL Oconee Nuclear Station, Unit-One "TmK" Hispositioned Fuel Assembly Due To Inadequate SeIt Checking and Management-Direction tVI AI lot L.KMVA00V 16WUKI UATIL in3 "5 'PA ALUMS EIVW.VW M5 mom0 Unit TWO 05000 270 01108 96 96 01 00 02 07 96 Oconee, Unit Three 06000 231 MOOIRT 1 RIFU.om I A Ia Zo*ZL0.WALvW C.HI 9: lB) r etv11*r

    U                          20.2203faHli                    20.22031aN3HIO m                 50 73lag12H* (A)                     SO.73184r LEVE 61,100toouWZ~*                                              a 20.22033114211n)                20.220.1aN4O                     SO. 7 3412*210v                     OTHER 20.]E203K*21I*I)~~~~a                    5.>1)                              n NSM Fore 3Se o34.2 I,. V. Wilkie,             Safety Review Manager                                             (803)           885-3518 LAM bImW            W     mmo1ý            amalI                               STWN          umsm K    toTWsae.ie.ap-aoT5srowgimpc
                          -,WUAVIAE    1W   NMI      9XEI ED 1141 tyevue h                3 M.JU                               DW      W ilt yes, awst  -d  EXiECTEO SUNSIN p   OATEI.

1I X

                                                                                        -u On December 14, 199S. with all                      three Oconee units at 100                     Pull power,.4 fuel handling team performing a fuel assembly (EA) inspection in the Unit 1&2 spent fuel pool (SFPI inadvertently left                                the PA unattended and suspended inside the SFP mast.                        It was discovered o.a January S, 1996. by fuel handling personnel during check outs for planned fuel movements.                                                  The PA was reinserted into the                   SIP     rack.        The  primary     safety       significance         of the event was the potential uncovering of the FA during a postulated event requiring actuation of the Reactor Coolant Make-up function of the SLandby Shutdown Facility              (SSF)     which uses the SFP as a water source.                              An engineering analysis concluded that the E-tel cladding would not be breached during an SSF event with this FA in the miast.                                  Therefore, IOCFR00 limits would not have been exceeded and the Final Safety Analysis Report (FSAR) analysis consequences would have bounded the event.                                      However, having an unattended FA in the mast is outside the intent of Technical Specif cation 1.8 on fuel handling and 3.18 oa the SSF.                                The root causes are inadequate self checking and lack of management expectations for formality and procedure use in fuel handling.                         Corrective actions include policy and procedure changes.

                                                                                        %Ms MXcLEA MOJIATORY COMMMUU FCM366A                           Licm          HN az            RMRT (LM23 TEXT CONTINUATION

s. PAGE.

J FACLIY MAMM I11 _-! DOCK" sm 05000 RM 2~OF 17 ocjonee N=lear Station gMit One 269 96 01 00 UET ,08u ,gee am

                         .e         w    of    A     .su     I117A I

a mgv,,s1n In addition to a spent rw1 Pool (SIP) (3Z1S:NDJ where spent fuel is stored in racks submerged under borated water, Oconee Nuclear Station has an interim Spent fuel Storage Facility on site. There spent fuel is stored In dry containers, thus the term dry cask storage" is used. Fuel hadling activities at Oconee are performed by members of a dedicated fuel hndling maintenance crew. The fuel handling suprvisor it a previously licensed Senim Reactor Operator. The crew,' work activities are primarily fuel handling activities and plant crane ([11S:3J3 maintenance. A significamt portion of the fuel handling crew,* scheduled dry work involve* shuffliug spent fuel assemblies in the SIP and support of cask storage activities. The manamum crew number for operating the refueling bridge ([1ISsFI in the SFi is one bridge operator and one spotter. Fuel Handlers ae qualified to ruel. Handling activities per Rmployee Training Qualification Standards, OP/O/A 1506/01* (iuel 6 Component Kandling) is the -0OV. TO" procedure for using the fuel handling bidge. it ito an 0.nformation Use* procedure which has no sign-off sr, is performed from memory.. and.,by management policy, is not repaired to be at the job location. mormalLy, OP/O/A/1503/09 iDocumentation of Fuel Assemblies I/or Component Shuffle Within a SP Pool) ti the -WHOM3 TO, procedure used to make miscellaneous fuel movem*nts. An enclosure. initiated by Reactor "Engineering. designates the fuel assemblies and/or control components to be moved, the starting locations, and the ending locations. The fueL handlers sign off each move as it :s made. Technical Specification 3 1 provides required prerequisites for fueL handling In the SFP. Ore requirement Ls that the SIP filtered ventLlation system [KIIS:VFI must be :perable. or fuel handling must be suspended. The SYP fiLtered ventilation system is considered inoperable whenever the fuel receivLng bay door ti open. The Standby Shurdown FIaci3L.ty (SSF) r[uES:Vri is designed to maIntaLn the plant in a safe shutdown :ondition for a 72 hour period in the event of an Appendix R fire, a turbizo building flood. a security event, a station blackout when the turbine driven emergency feedwater [eIrS:aAI pump (91IS:P] ts inoperable, =r a tornado which renders the %utxiary serviLce

         ,ater   and emergency feester          systems inoperable.               The  SFr Reactor        -ocant IRC) makeup pump CEIZS:CDr takes water                trom   the    SFP  inventory      in  order   to make-;- to the Reactor Coc~ant System                PCS1    (111     ABI  through   the      re-actor coolart pump seals.        In a'Jdation. SFP -3ooiLn              may also be lost dur-ng an

K FO6M l U.S. NU-MEAR RMaEGUTORY COWN-* SM*- '4. LICENSEE EVENT RBPO T (LER) TEXT CONTINUATION FACRJTV NAME III OCKET .-- uLIpN I"3 PAGE 06000 low 3F1

   .Oconee suclear Station, Unit One                        269          96     01        0 T~~~~lT~_     - is         cs vo    Ofce  WM 36" 1iiM NWsnw the lose MSY event such that boil-off of SF? water will also contribute to of SIp inventory. The design basis of the SSF system will allow depletion of the SIP inventory to within one foot from the top of the SF? racks assuming nO action to refill the SFP. Tochnical Specification 3.16.4 requires the 8SF EC Makeup System be operable for each unit when the RCS is at or above 2SOO.

During Unit _ COCIS (End of Cycle 16) refueling outage, which started on Nov. 2. 199S and concluded Dec 10. 1995. a fuel assembly (FAI was observed to have four intermediate spacer grids damaged. As part of the root cause eval,*ation, Reactor Engineer A desilred to perform a visual inspection of VA JOSTO (IA-I). the fuel assembly which -had been. adjacent to -the damagied assembly in .hereactor core for the fuel ycle .. o December 14. 199S. at about 0900 hours. Reactor Enineer A contacted the. Fuel Handling supervisor for support in inspeCt~ing' IA-S: Theý roquesttask was initially denied, due to workload. Subsequently, one of the planned, was deferred several hours and the Fuel. Handling Supervisor contacted Reactor Engineer A to schedule the inspection for after lunch. Around 1300 hours, two Fuel Handlers and Reactor Engineer A entered Unit 162 Spent Fuel Pool (SF?) to inspect FA-S. A pre-job briefing was perfor wtween Reactor Engineer A and Fuel Handler A but it covered only the basic* of what needed to be done. Reactor Engineer A had no procedure or -ovement enclosure for this evolution, and, since the inspection did not involve leaving an PA in a now SFP location. Reactor Engineer A felt that he did not need one. Fuel Handler A thought Reactor Engineer A had a procedure since he had called the control room to verLfy prerequisites listed in the normal fuel handling procedures. Reactor Engineer K stated that he called the control room out of habit. However. Reactor Engineer A stated that he did not inform the control room operator that fuel handling activities were about to take place. Fuel Handler A operated the Unit 162 SFP bridge by memory, which is the normal practice. Fuel Handler A stated that he felt comfortable dcing fuel handling steps by memory. Fuel Handler B acted as a runner for the ]ob Reactor Engineer A acted as a spotter, operated the video equipment. directed Fuel Handler A to SFP rack location Kt40. and directed -ast opecation (up/down) while video taping was in progress. DucLnq thLs

U.S. WUUJCLEARREUAOMY COWUMO -FORM 36" WENUT REORT (LEE)

     *g~a                       LICISZE TEXT CONTINUATION                                   O                  . C
                                                                    *       'mom LIWM                         OF If 4Paola0 S5oo-FACKMT NAPA 11DO1E I

i9 01 00 feet east to improve the evolution, the mast and FA-8 were moved several 1uel Handler A to available lighting. Also,-Reactor Engineer A requested back while PA-S extended below the rotate the fuel mast 90 degrees and end fitting, FA-S mast. After some scratches were noted oR A-B's Looverthe storage rack. was returned to its proper location and lowered into another FA selected For comparison, Reactor Engineer A decided to look at Fuel Handler A at random from the same cycle. Reactor Engineer A directed and directed mast to SVP rack location L44 to pickup TA IJO637 (FA-7) observing similar After operation (up/down) while the PA was video taped. he had seen enough. scratches on VA-7, Reactor Engineer A stated that A specifically At this point neither Reactor Engineer A nor Fuel Handler stated a need to lower VA-7 prior to proceeding. to-not .ea~ve OP/O/A/IS06/01, Limit and Precaution 2.27 ,directs personnelto irradiated portable underwater lights and cameras-in -close proximity ngineerA. begaA fuel assemblies when not being used. Theefore, Reactor. the pole .and cable. to raise the video camera. Due ,to the need to wipe down requires two people. attached to the camera as it is raised., this task personnel to However, GP/O/A/SO6/OI, Limit and Precaution 2.22 directs when a Bridge is turn off the Bridge hydraulic pump to prevent overheating In this rtee idle for IS minutes or greater and the hoist is not engaged.learned that _be the hoist was engaged. but during the investigation it was to turn off the Fuel Handling Supervisor has issued standing directions pump is off, most of punp even if the hoist is engaged- When the hydraulic to a default the control panel indications are either de-energized or a0 state. hydraulic in accordance with these instructions. Fuel Handler A stopped the A with pump. left the control console, and assisted Reactor Engineer camera was pulLing up and wiping down the video equipment. Once the control console and de-energized secured, Fuel Handler A returned to the believed that Fuel Handler A stated that he the bridge. During interviews. look at the he had lowered the FA back into the fuel rack and did not control console indications to confirm this. At 1342 hours. Fuel Handlers A and B exited the UJnit 142 SFP with Reactor Engineer A. This left FA-7 suspended and unattended Lt. the mast. the next several No fuel handling taaks in the Unit L1Z SFP occurred over weeks.

U.S. NUCLEAR REGULATORY COMUMMISS1 PRC FOAM 365A LICENSEE EVENT REPORT (LER) TEXT CONTINUATION DOCREr La NUM IE PAGE 13e FACLI NMA (1) F 17

                                                                        -05000 50pm avo 269                96         01          00 Oconeo       Nuclear Stationg Unit One OFwowsearm",                       W      3oWOFAm             1171 I

Fuel Handlers. A and C on January 6, 1996, at approximately 1030 hours, preparation for loading a dry cask later entered the Unit 1&2 SFV to start When Fuel Handler A energized the bridge and started the in the week. the control console indications and realized hydraulic pump, he observed mast. Fuel Handlers A and C initially assumed that that a FA was in the in the mast recently by other members of the crew. the FA bad been left to lower the FA in the open rack at "Fuel Handlers A and C made the decision to determine chs-M i^ order location L44 to allow an identificaLio-aof-movemmnt to determine who wibre it shoula ce area to trace the last Known was responsible. rack, Fuel Handler C while Fuel Handler A lowered FA-7 into the storage of the discovery and called the Fuel Handling Supervisor and informed him in the empty rack at L44. Fuel that Fuel Handler A had lowered the FA as NJOE7 at Unit 1&2 SFP rack locatLon Handlers A and C identified the FA L44. called the Rotating equipment. At 1130 hours, the Fuel Handling Supervisor to report the event,. It was verified that "Manager and Reactor Engineering FA-7 was the-last FA moved in Unit 112 SFP. event. The video tape At 1230 hours. a meeting was held to discuss the reviewed to see if the tape had shown the FA being put from 12/14/35 was The personnel present concluded that FA-7 had been back down in the pool. 1/8/56. All three Oconee units were in the fuel mast from 12/14/95 until at 100 % full power throughout this period. depletion of the SIP The design basis of the SS system will allow inventory to within one foot from the top of the SFP racks assuming no the SFP. A concern was raised that FA-? could have been action to refill to clad failure uncovered by an SSF event, with the potential for heating products- However, no analysis existed w.uith resultant release of fission occur or if the severity of the releases to determine if clad failure would

                                                --  r   the   FSAR      analysis           o.:   IOCFRio0.          Thus   there w.ould exceed limits from e the SSF   t     iht   have    been        unable        to    perform       its    intended wea a concern that Therefore. one function and would need to be considered past inoperable.

action item from the 1230 meeting was to start an operability evaluation temperatures and potential which would include calculation of expected clad releases. as the Station Manager) The Maintenance Superintendent 1who was acting Manager's staff meeting at 1330 discussed the event during the Satilon was at the meeting and assumed "he hours. The Operations Superintendent control room knew of the event.

U.S. NUCLEAR REGULATORY COMMssM of FORMA 3GM LICENSEE EVENT REPORT (LER) TEXT CONTINUATION DOCKET tM NUMBE WG PAGE in FACILTY NAME III 6OF. 17 0500 269 96 01 00 Oconee Nuclear Station, Unit One iscrQKuwd "S idtf Capes of AM FoWm 3654 117n { mwo 0f no the At 1500 hours, af tea: the staff meeting. the Maintenance Superintendent, inform the Rotating Equipment Manager and Fuel Handling Supervisor went to ONS NRC Resident Inspectors of the event. After briefing the senior resident, the Maintenance Superintendent. Rotating Equipment Manager. and Fuel Handling Supervisor discussed the until procedures situation and decided not to continue with fuel handling were revised to prevent this event from reoccurring. At about 1800 hours, the Senior VP of Nuclear Generation and the Site VP discussed the event and decided to zni iate a Significant Event investigation Team (SEIT). of the discovery Throughout this period, the control room was not informed On 1/9/96, at about 0630 hours, an NRC resident of the FA in the mast. log entry for the event. This was asked control room operators about the time Operations shift had heard about the event. the first At,0900 hours, this event was discussed in the daily site direction -' past meeting. Site management present discussed Issucs related to The information available at that time was operability and reportability. insufficient to reach a conclusion. Notes At 14L4 hours, a log entry was made in Unit I Log about the event. (RO). Control Room Senior Reactor Operator were added on Reactor Operator (SRO). and Unit Shift Supervisor'a turnover sheets not to move fuel in l&2 and/or 3 SiP until' after the SEIT investigation was completed. issues Discussions of operability and reportability issues continued. iTS) and FSAR discussed included compliance with Technical Specifications apply releases. TS that potentially analyses of fuel damage and resultant the Spent Fuel Pool, and in this case are 3.8. Fuel Movement and Storage in 3.18, Standby Shutdown Facility. TS 3.3 was initially not considere.; to apply, based on an interpretation that FA-7 was not moving while left in the mast. By that interpretation. fuel handling was not in progress and, therefore, the TS was not exceeded. at or TS 3.18.4 requires the SSF RC Makeup System be operable for each unit above 2S0"F in the RCS. During an SSF event the SSF RC makeup pump takes suction from the SFP and can allow depletion of the SFP inventory such that FA-7 would be uncovered. Preliminary engineering calculations indicated possible heating to clad failure with resultant release of fission products. This could result in dose consequences beyond th,- licensing bas is i I

SFORM 3646A US. NUCLEAR REGULATORY CO*-MIS--M

r LICVNSKE EVENT REPORT (LER)

TEXT CONTINUATION FArDUA_11) OOCKET .-_ (a PGR 05000 .- P*: =ý 70F 17 O-FJm 17 FAWT MW Ii 0 Oconee Nuclear Station, Unit One 269 96 01 00 T3WT07mWN aeWc a mrq,'e" uaft~ reis cW 01N 1MC Fvw366AU 1171 At 1700 hours, a decision was made to make a I hour NRC mergency NotLfication System call, based on management's conclusion that these consequences represented an u.ianalyzed condition that could significantly compromise plant safety. The notification was made at 1755 hours. On 1/10/96, the SSIT arrived on site and began an investigation. on 1/12/96, the SKIT presented -its preliminary findings in a formal exit with site management and the Senior VP of Nuclear Generation. one concern raised by the SKIT was the interpretation that leaving a FA, in the mast met the requirement to suspend fuel handling. A survey of industry practices revealed that all of the other sites contacted defined fuel handling to include any time an assembly was supported by the fuel handling bridge or crane. These other sites interpreted Psuspension of fuel movement" to mean that fuel movement should be continued until any VA in a raised position could be moved to a safe location and lowered. Aqpplying this more conservative interpretation, of "fuel handling" resulted in the conclusion that TS 3.B should be applied the entire time FA-7. was 'in the fuel mast. Since the fuel -receiving,bay- door was opened at various times during the period, making. the filtered ventilation system inoperable. the new interpretation would mean that the intent of TS 3.1.12 was not met. The operability calculations and analysis were completed and the results this are discussed in more detail in the "Safety Analysis" section of report. The analysis showed that FA-7 would not be damaged and would not result in off site releases exceeding ioCFRiO0 limits. However. another FA with a higher decay heat potentially could. Therefore. management concluded that the condition of a FA being located within the SFP mast during an SS? event is not in compliance with the intent of TS 3.18. Therefore, in addition to being reportable as an unanalyzed condition that could significantly compromise plant safety, this event would also be reportable as a condition outside the intent of Technical Specifications. In response to the SKIT preliminary concerns. "Short Term" actions were initiated to enhance programs, policies, and procedures to address the SKIT recommendations and observations. These 4ere primarily aimed at those items needed to resume limited fuel shuffles in preparation for dry cask storage and new fuel receipt prior to a refueling outage on Unit :. currently scheduled for late March, !996. Jn Feb. 1. !9-6. the SKIT issued its final report. The root causes identified are the same as the ro-t causes listed below.

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WM~ FORM 36A u.S. NUCLEAR R1EGULATORY COMUASIOU 1 IM L1CWZN Z EVENIT REPORT LER) TEXT CONTINUATION FAC"*IjTY NAME 11) DOCKET LER MUUU ISN PAGE-13 05000 NM OF 17 Oconee Niuclear Station, Unit One 269 96 01 00 -UXT 0 jmion Wa isMq*wu @m W

• oacmas m of NRC F&M 366AI 117[

The root causes of this event are related to inadequate barriers intended to minimize the potential for this type of error. Two root causes for the event have been determined: The first root cause of this event is the failure of Fuel Handler A to self-check his actions. This was a skill based error resulting from a momentary aemory lapse while performing routine actions using an Information Use procedure. The second root cause to this event is the lack of management expectations for formality in all aspects of the fuel handling process. The lack of formality was exhibited in the following actions, which were in accordance with management's expectations at the time for this type of work in the spent fuel pool, leading up to the leaving of the FA in the mast: L. The failure to write a.id process a work request for the -onduct of this activity.

2. The perception that no task specific procedure was required to conduct this activity.

3. 0P/0/A/1506/01 (Fuel & Component Handling) was being performed from memory because it was An Information Use procedure and was not required to be at the ]ub location. Performing procedutes from memory will increase the risk of human error. Requirements of OP/O/A/1506/0l were not met in that:

a) The Control Room was not specifically notified that fuel handling was in progress in the Spent Fuel Pool fSFP). b) Fuel Handler A ritated the mast 90 degrees and back at -he request of Reactor Engineer A. This was perfor-ed while the FA was not "full up" in the mast. c) Steps to lower a FA and disconnect trom the fuel grapple are included in the pro'edure but the omission of those steps resulted in FA-' being left susperded inside the fuel mast. of maWe W" $a9

364" U.S. NUCLERREGULATOY COUMISS9 NRC EFOR0M LICENSE* EWENT REPORT (LER) TEXT CONTINUATION OOCEETLF MUIN WUK PAGE Ma FACIr NMUI I II 05000 9 OF 17 oc-_-ee Nuclear Station, Unit. One 269 96 01 00 ofW MMTS @'5w S W1 WVAw at9G' CVPWf P*CFaWw36GQ IM) d) rLmit and Precaution 2.22 directs that *When a Bridge is idle for IS minutes or greater and the hoist is not engaged, turn off the Bridge hydraulic pump to prevent overbeating." This condition vas not met when Fuel Handler A secured the hydraulic pump because the hoist was engaged. Due to workarounds with the hydraulic pump and instruction from the Fuel Handling Supervisor, this had become a common fuel handling practice. inadequacy of OP/O/A/1506/01 (Fuel a Component Handling) in that it 4. did not provide steps for the fuel handler to verify that the fuel bridge mast was empty prior to shutting down the bridge.

5. The failure to provide an adequate pre-job briefing for the evolution.

The pre-job briefing did not address roles and responsibilities of the individuals involved. During most of the activities, Fuel This Handler A was acting under the direction of Reactor Engineer A. potentially led to An expectation on the pa*t of Fuel Handler A for Reactor Engineer A &o instruct him to lower .the FA. Reactor Engineer A feli it was not his responsibility to ensdre that FA-7 was lowered back into the SFP racks. Past industry and site experience was reviewed to determine if this event it was concluded that industry operating experience has not is recurring. been used effectively at Oconee to prevent fuel handling events. SER 91

15. as an example, identified fuel misvOsitionina events that occurred wicnLn tnh inouscry due in part Lo inaumquaCe inaependentt verification and self-verification techniques. Oconee reviewed the SER. revisec refueling procedures. enhanced methods of fuel har.dlers communication, and evaluated training in response to this SE&. However, these corrective actions were ineffective in preventing four fuel mispositionina events that occurred in 1992 through 1994.

An operating experience review was performed using the Oconee Problem investigation Process (PIP) data base in the area of fuel handling activities to look for similar events vith root causes similar to this event-. Attachment A to this report summarizes past fuel handling events and the related NRC violations. The first root cause (self-verification as it relates to fuel handling work practices) has contributed to four events resultinq in three NRC violations aL*conee during the period of 1992 through 1995.

U.S. NUCLEAP qREGUATORY COMMOS1ON OC FOM 364A LICENSEE EVENT REPORT (LER) FACILITY

                           *.I    hAMI I

TEXT CONTINUATION DOCKET 05000 jU 196 tat NI A~ 01 l to1 1" 0010 PAhaE In 9 OF 17 269 1 OcneNuclear Statioat Unit. One The seeadt root ciuse ilag% gg Lpagement expectations for formality process) has also contribut~ed to in all aspects of the fuel handling PIP 1-094-0707 and the fuel handling events at Cconee (particularly 1994). associated NRC violation of August 2. event is recurring with respect to Therefore, it is concluded that this handling events both root causes. The repetitive nature of these fuel lessons learned from previous events demonstrate the lack of full use of corrective acticns. for and application of too narrow a scope injuries or cer exposures. There were no radioactzve releases. personnel associated with this event. or NPRDS reportable eqpipment problems CnRRE-CITMZ ACTIONS [mmeediate into a Spent Fuel Pool (SFP) I.. Fuel HandLors lowered the fuel assembly storage rack location. fuel 'handling activities Mechanical Maintenance management suspended M. pending procedure changes. Subsequent and this event was analyzed I. Engineering calculations -.ere performed design basis releases. with respect to the potential for exceeding Ianned A for all f..eL movements. Step by step proce,-r-s will. he required assure ti-.at the fuel -as A procedure chec'C-.it wil: be provided to at the conclusion of fuel handlin. is returned to a proper end state activities in the

             .        Fornalized pre-::b brief ings for all fuel related SFP will be ..pl-.e-e.ted.

will beta-Een in 3c-ordance

4. Appropriate peri-:-ne- corrective actserms with Duke Power :i:7: s.

                                                ;-x.i     ' Aud-t   SITA) wC.1 be cer         :r.ed      t    pr--. ie
           *      . A Self 1nit-ated                                                                            -t*ork ano -'r.er SFP                         tci.-"z.:-eS t broader re-te--     :r :e.-.andli.q crocesses-
         ,tS an~A I"

U.S. NUCLEAR REGULATORY COMMISSI-MMR FOAM 36A LICENSEE EVENT REPORT (LER) TEXT CONTINUATION DOCKMET LEN KAMM let- PAGE I FACLITY NAMA I!) 05000 -1In '- 11 OFmo7 269 96 01 00 Oconee Nuclear Stationt Unit One TEXT IN mj weM aAn quwed. uses 01W C0002 of1 NRC Fot,,m 366& 1 M7 planned corrective actions I through S are concidered Commitments to the be NRC NRC. They are the only items included in this report intended to Commitments. SAFETY ANAMY5I fuel The consequences of the failure of a fuel assembly (FA) in the spent are analyzed in the Final Safety Analysis Report (MSAR), Section pool (SFP) 15.11.2.1, "Single Fuel Assembly Handling Accidents*. The FSAR accident scenario is a radioactive release from all 200 fuel rods. This accident is under at least 9 feet of water for iodine retention. The assumed to occur dose calculation with the FSAR initial condition assumptions of release rem inventories and conditions yields a dose of .66 rem whole body and 174 thyroid at the size boundary. During an event requiring the Standby Shutdown Facility (SSF) Reactor by the Coolant IRC) makeup pump., A NJO6E7 (FA-7) would have been uncovered of the SFP. A heat up calculation of air~cooling of decreasing inventory FA has been performed using the actual decay time after shutdown the assuming only radial free convection and radiation. Results indicate a maximum cladding wall temperature at the top of the FA of 1022 degrees F. Potential damage mechanisms and the applicable limiting temperatures are: cladding creep out (ballooning) and rupture 1150 deg F. accelerated oxidation 1600 deg F. metal water reaction 2200 deg F. enhanced fission gas release from the 2450 deg F. U02 pellet matrix zircaloy melting 3400 deg F. and no This calculation shows that cladding integrity would be maintained Therefore, the existing analysis in effluent radiation release occurs. Section 15.11.2.1 is still bounding. An estimation was also performed for the most limiting decay heat load possible. In this case a high powered assembly, only 72 hours after subcriticality. was assumed in the mast and rooled by air and radiation ieqrees F This analysis determined a maximum cladding temperature cf 2000 In this scenario, damage to the cladding would occur, and there would be no iodine retention in water. so the release of radiation from the assembly would be significant

NCFORM 36" U.S. NUJCLEAR REGULAT*AY COtr'S-- LICENSEE EVZNT REPORT (LER) TEXT CONTINUATION FACIUT ,&.u 0) DOCK" mUMe I16 PAME 93 o0e00 !0 ,w"I 12 OF 17 269 96 01 00 oconee Nuclear Station, Unit One irXT ~ t/fmore ace s.v use ac""i& copo, of1 ARC Fm 366M 1171 from Depletion of the SFP inventory removes the majority of the shielding direct radiation shine from the spent the spent fuel assemblies such that fuel will become significant. ESvever, the SF valls provide lateral shielding so the direct radiation shine is primarily in a vertical direction. Since the top of FA-7 was approximately 9 feet below the SFP grade, this will only add a small amount of additional direct radiation to either the on-site or off-site dose rate. having FA Since the SFP inventory must be eventually replenished remotely, to the 7 in the fuel mast does not impose any additional restrictions operability of the SSF RC makeup system. During the time period of interest, no spent fuel was moved in the SFP. spent FA Since the fuel mast provides a positive mechanical lock for the no additional potential for a and the SFP bridge is seismically designed. fuel handling accident existed using the updated Oconee PRA model. the annual frequency .of an event relying on the standby shutdown facility for-core damage mitigation *s 3.3 B-04. For the 25 day period FA-7 was ir the fuel mast, the probability becomes 2.3 R-0S. Furthermore, typical PRA calcutlations utilize a 24 hour to mitigate the accident. In tkis minimum time for the system relied upon of 36-40 hours would have been available before case a time in the range the SFP inventory is depleted to a level exposing a portion of the FA. In conclusion, during the period from Dec 14, 1995 to Jan. 8, 1996. when FA-7 was suspended in the fuel mast. FA-7 was in a static, stable position that the probability of fuel damage by another mechanism fcollision. such dropped object. seismic event, etc.' was remote. No SSF event occurred during this period. FA-7 was not damaged and did not release any radioactive materials to the public. In the unlikely happenstance that a SSF event actually did occur, an extensive period of up to 36 to 40 hours to woild have been available for compensatory actions to oe taken prior uncovering FA-7. Additionally, calculations show that FA-7 woi.ld have been adequately air cooled and no damage would be expected. Therefore. the health and safety of the public was not affected by this event

 -           a.

U.S. NUCEA F&GULAOW CUS' L -IPMC FORM 3NA I LXCENSEE EVENT MPORT (LER) TEXT CONTINUATION DOCKET L NUMUM IS PAGE 0 FACITY NAM i 05000 v1It3 OF ]17 Ocone Nuclear Stations, Unit One 269 96 01 00i TI ,wr-W mNJ sofs resav. u"e A-a ii

0. Lsr u" "-.- .. .

ATTrIACOD4 A 02n&fMAM mEZ w m Qcg=3'~ Z21 Pacement EventsR L-092-0723 Wroo _fumel ansemhly (A) was placed into Unit 1 Reactor Core ouring refueling activities as a result of inadequate self check and independent verification. Changes to the refueling procedure were implemented as corrective actions to prevent recurrence 1-092-0724 Wront FTA was placed into Unit .1 Reactor Core during, refueling activities as a result of inadequate sestcheck and: independent verification.% Chany.ei to the refueling 'procedure were implemented at corrective actions to prevent recurrence 1-094-0 7 07 Refuelina saemusnCe was altered at the requept of reantor engineers to observe nuciear insttrumentation response without proper documentation and procedural control. This was a non conservative decision made by the SRO in charge of fuel handling. Reactor Engineer, and the Fuel Handling Supervisor. Corrective actions to prevent recurrence involved a change in to prohibit sequence deviations without the refueling procedure the use of a procedure change or test procedure. refueling 1-094-0714 A .rone F. was placed into "ait L Reactor Core during ac.Lvities as a result of ina-tquate self-check and independent verification. Corrective acttons to prevent recurrence involved changes to procedures and methods of independent verification. 6 PIP . PROBLEM INVESTIGATION PROCE5S

poftC FORM 36UU S NUOLAR REOULATORY COMMMSOU WME L RIMSE EVDT REE PORT (LER) TEXT CONTINUATION DOCET LO MIMERIN PAGE 9X FACLMT NMWI 'I) 06000 VJW14 OF 17 Oconee Nuclear Station, Unit One 269 96 01, 00 U" "OJ of PAWCF f*l--3*--4 1i TF-xT (0 Wm PW is'emW" S za ga*+/-fl~v,+/--Qafeen 2-092-0024 A VA and control rod was damaged while the FA was being positioned for repair. The pLocedure was not reviewed prior to Corrective the move and the control rod and VA were damaged. procedure changes and actions to prevent recurrence involved pedestal modifications. 3-092-0470 sent spider assemblies causes delay in removal of burnable poison rods from two fuel asmemblies. Zt could not be determined whether the damage occurred as a result cf previous fuel handling activities by Dike or by the fuel -endor. sizing Corrective actions involved a*anufacturting a component Assurance during the component template to be used by Quality assem'llies. inspection performed upon unloading of the now fuel fromoits 2-093-0431 An intermediate grid strap becm torn and separated This type.of damage is caused PA during refueling* operation. of adjacent assemblies snagveach other when the grid straps during fuel movements made in the core. Corrective actions to prevent recurrence involved prevent changes togrid the refueling procedure to Ph strap damage. to provide new guidance of the 3-094-0204 A dusmy control rod assembly located in the deep end transporting the core fuel transfer canal was struck while This was a result of Inadequate self-check support assembly. of clearances. Crane control and water clartty problems contributed to the problem. Transport had to te halted to transfer perform inspections of the core support assembly, the canal liner plate. and the fuel storage racks. Corrective

o actions to prevent recurrence involved procedure changes incorporate preventive measures.

(SFP) 1-095- 1429 During reactor defueling activities, Spent Fuel Pool I bridge hoist and grapple operation was hampered several times due to unexpected interference with consolidated fuel canisters. This interference problems in disengaging rom fuel assemblies. Corrective actions involved moving the Ls consolidated fuel canisters to an area of the SFP that outside of the off-load area. damage on

             "*-095- 1462    FA NJO776 was found to have significant str-c-:jral four consecutive intermediate                   spacer   grids     :n  zhe   souqthwst cot. er. No fuel rod damage was found or ,uspected.

9 ismu LZCDISIZ VNIT RUPORT (LUR) TEXT CONTINUATION FAC~LTY NMUM "II DOCKET i E ~l~mIS)PASS 05= MAD1 I OF 17 Ocne uclear Station, Unit One _269 9 1 0

                                     -  -       A ~ R         2" AI k SW *.; i,.* R--            u ..             W  ,I","C....      .  ,

imum&wwinU azn I sT4 airig set up of the R&W fuel econstitution, eitator part. seared/fell Eom the elevator Into the cask area. The elevator part apparently shearetd whn It contacted the cask area wall. 3levator design deficiency and workes- attention to detail contributed to tbis event.

        * -MIq)   o00S5    several new fuel assemblies were rece&ved and placed In atore ceiLls tnat Were not in accordance with pieedure.                             Root causes were galluro to ColLow procedure and -satention to detail.

The fuel components were not adversely affected. l- I0)-_Q4i4 A control rod-was tiserted in tba wrens MA Corrective act& involvea procedure changes and persemi training to prevent recurrefnce. 2-N93-0676 A contractor personnel failed to follow procedural requirement& for handling fuel rods during reconslitution activitioes, whac resulted in severely bel*. fuel rod aW mueeqaent cbsltenge to the fuel cladding integrity. This resulted in a, WW- level IV violation 1PIP 2-wJ93-017*' or the failure of contractor personnel to follow proced;;al requiremets. l-1490-0002 A Sequence within the A rInsert shuffle procedure wae performed incorrectly resulting In the MqAnoSItian of S thmMbLe PIi LM the SFF. The verification process Ldentmliea ann correc-ed this dtscrepancy. NO Correc0tve actions to prevent recurrence were Identified. Sui 91 - L5 This report describes SIX kndult*rV CUS! Misvo66Ltronn@ P-.-entz during refue..ng and deueling activities as a result at inadequacies Ln procedures. independent verifiLction, and trmLnLng. Oconee's revoew ot this event resulted *n chanqes t 3 rCeia. Lnq procedure chanqes and methods of cnmmunLeatLon SEN 94-4 ThiS report lem:ribee six specific Industry events ,hniat.in.. hu*anr performance de¢fICLncLes wriLe handi'nq roector -ore

                             .ompo*nents thNat resulted Ln iAcuaL FI or .hther :or* -o-miponen 4amgqe. Jlmaqe to rfUT-LLTnq                   lquLpm*nt. Andoor rncroa&.,

pontPML f11r tamaqe "3 EuPL rr 'I'r .7.orM r..mt '.nP.

I MpC FORM 366A V.& MXIcEAR REGULATORMOUO LICENSRS WENT MpoRT (LER) TEXT CONTINUATION FACLffi NAME III V- SCOT NuM to 0OI UDUUW I ~ W 11 OCoiee Nuclear Stat !ot Unit One 26J 96 91 08~ TEXT #-M" 0 --iee, oWrN bco ca* of A w 3"4) 117 Su Oconee incorporated these industry ts aCMd lessons learned into the operations fuel handling moisoa plan 11 94-13 This report describes potential problems resulting Erom inadequate oversight of refueling operation, and inadequate performance on the part of refueling personnel based cm four industry events. Oconees -- view of this report resulted in no rec.mende actions based an actions taken with SZR 94-4. 13 94-03. Sup. I This report describe* an industry event involvtnM unauthorized movement of a defective spent fuel rod. Oconaees review of this report re"Ited..An no recommended actions. SC IainJ IV VLAIALL~m INUISMAC 2L. LnAL One example of a failure to adequately implement a refueling procedure that resulted in a FA being placed in the lw locat ion is thm core. Root causes were operator error and poor vLSIDLLaty in t?,e SaP. corrective actions to prevent recurrence involved "ounseling the bridge operator. I LYL azCJ itaLi"LL IAEDWnbas 12. 11011 ine example of failure to adequately implement a refueLnfq procedure rhat resulted in a Ph being placed in the.vronL__inmn fuaa -ieatio._ Ro'.t causes were &nsuflicient a:tentton to detaiL. insufticient procedure detail and communicetion errors. CorreOtive ee iLon to prevent recurrence invoLved procedural -hangea and fueL handLinq training W~ LaYAL Lit uQALaLM A13ZIJALY zz

. 0 0MU.. 3RiULATOAY NUcLEAr COMMS SLICENSE EVMT REPORT (LER) TEXT CONTINUATION DOCiOT 9 0 SER LEMI*. .SPAME NAMEion,

                     -ACIUoe                                                                       17 OF 17 06000            5UNU 269        96     01        @0 oconee nuclear Statione Unit One
     - WTuaCM5IOM"aD9U's eFfiCF'W 3fE4 IM)

IV remalted in two fuel assemblies being placed tie core. Root cause was inadequate by Pspw 1-092-0723 and 1-092-0724) self procedures that TWo examples of failure to implement refuelingin the Wrm

                                                                -checking.

actions to prevent recurrence involved procedural changes. fteatiw Corrective (Covered in f-in imiliarlaftIJaIkAWa= Z.. 191t xJ LI(ve pofuelina sequence was altered to observe nuclear instrumentation control. This response without proper aoc*.umentation and procvdural personnel. was perftomed at the request of Reactor Engineering ICovered by PIP 1-094-0707) m* Laul LI *LnaLA KL~b CL3LL ERmulL IAUA3L L- *U2A A FA retrieved from the wroam sp-a fuel j.ocaL= and ple-ed &n the reactor care. Root causes were inadequate, setf-check ,an independekt versticatiol. This was the fourth occurrence of failure to identify and adequately verify FA Locations. Corrective actions to prevent recurrence involved procedural changes and personnel training. (Covered by PP .1-094-0714 And PIP 1-094-0707)

EXHIBIT B- 15 Oyster Creek Unit 1: LER 219/87-006-00 (February 24, 1987)

9

                                                                                                                                                          *l NUCLEAR mSIuP*TOY            OM of

-WM n Sl,413l IugqOv@D OM0 MoO)I.-*t UCENSEE EVENT REPORT (LER) n," 1301 Oster Creek, Unit 11 O A 0 121 11 9, IOF1013

• nne RUIN niU TOSPECIFICATION TECHNICAL PERSoNNFL FRRo VIOLATION CAUSED BY IMPROPER STORAGE OF HIGHER ENRICHMENT

                                       -.                                                                             1MG. racaum       NEY@VEOCI moO      SMg4 l NONTH      DAY      YS"f               *.                       ftIAR0     9.0"T"     DAY    VIAPIAUYW~

1 1211 817817-olol- o1ool2 214 E47 OlIsla 01 IrNS TwPoar asuoNT1D us PqWAOWo T"E Sn g S1WS c11m1: w m o we A-mm- w "m1 t> [0" WUbv1 73,"" 4 mjsw [],6W7IbN3 DINSR 6 sow fbm -* A&~ Tbal. C palm

                                                                         .lCPNN      CO[TACT PON  W   INd I121 06*AWfe 1AREA  COOlD E TILIPPW      NUMBER Hari S. Sharma,                    Core Engineer                                                                   61019             917,1         -1416 1 318 OWLEITII CN     LING PON SAC       OWMENMT PAILuMI DIMMEO     DM TWO nIE*P? 913W CAUS     $TIIA                            .M      EPOATAASLPEC                                                     MA&NUPAC.            EP9101TAO o"OD Tu'IA                             CAUoE  SYST0.                    TUEI "OS.OBit?          TO 0POIO

_ I I I I I I:I  : !!ii]i:;:i SUWiMIkNTAL SIPOAT EIXPICTrD i41 Mo10rY DAY Y3AR ERPIlCTUBD

                                                                                                                                   ., . ,, , ,.o ft k   'VIS of,.
 ""PITACT towof    EXWC'r#o f&Lv, PaowMWMI&   -m' SUSIVIgwOO     DArt,
                                            .tor--ws ftf... ,-W   mow fv~ff h b PM.Os mfa 117 Oyster Creek Technical Specification 5.3.1(C) specifies that the fuel stored in the fuel pool storage racks shall not exceed a maximum average planar enrichment of 3.01 wt% U-235.                                      Contrary to the above, reload fuel bundles supplied by General Electric Company (GE) having an average planar enrichment of 3.19% U-235 were temporarily stored in the fuel pool during the llR outage in 1986. The cause of the event is attributed to personnel error in not performing a thorough safety analysis for storage of the new fuel and in not recognizing a conflict with the Technical Specifications prior to fuel storage in the spent fuel pool.

Corrective actions will consist of revising the refueling procedures, revising the Technical Specifications to raise the enrichment limitations on stored fuel, and reviewing the occurrence with engineering personnel. PD8/OJO3'A ,,,1

MWC PWmme& aU uS Jima4U.A 0rLVS CMUIMSM W UCENSEE EVENT REPORT (LeRl TEXT CONTINUATION ,PPOMu, owre No 11 -.104, ACNJrW -G- III DOCME mqa Mm,Lr LirnMIJ0014 0E PRW UB GAO~ Oyster Creek, Unit 1 Isil10l101.21,19 817 -- 010 6 - 000o2 o 0oOl3 DATE OF DISCOVERY The violation was discovered on January 21, 1987 during a subsequent review of the Oyster Creek Technical Specifications for potential changes related to the new fuel design. IDENTIFICATION OF OCCURRENCE Fuel with an average planar enrichment of 3.19 wt, U-235 was stored In the spent fuel pool beginning February 27, 1986. Technical Specification 5.3.1(C) states that the fuel to be stored in the spent fuel storage facility shall not exceed maximum average planar enrichment of 3.01 wt, U-235. This event is reportable under 10CFR5O.73 (a)(2)(i)B. CONDITIONS PRIOR TO DISCOVERY At the time of occurrence, the plant was operating In a coastdown mode in preparation for the 11R outage. At the time of discovery, the plant was at approximately 201 power starting up for Cycle 11 operation. All the fuel bundles which exceeded the Technical Specification enrichment limitations for storage in the fuel pool had been removed from the spent fuel pool and loaded in the core. DESCRIPTION OF OCCURRENCE A total of 204 GE P8DRB299 fuel bundles, with an average planar enrichment of 3.19 wt% U-235 and a bundle average enrichment of 2.99 wt% U-235. were received in 1986. At the time of fuel receipt, the dry storage vault had a capacity for 140 bundles. Initially. 64 of the new bundles were temporarily stored in the spent fuel pool. As the outage progressed, more bundles were taken out of the dry storage vault, channelled and stored in the spent fuel racks. Ultimately. 184 reload assemblies were subsequently stored in the spent fuel pool prior to the start of core reload in August 1986. At the end of core reload (September 14, 1986). all the P8DRB299 fuel In the spent fuel pool had been transferrei to the core. APPARENT CAUSE OF OCCURRENCE The cause of this occurrence Is attributed to personnel error. The safety analysis which was prepared was orienten toward the safe operation of the plant using the higher enrichment fuel during the next cycle. It did not take into account that the new fuel could conceivably be stored in the spent fuel pool (only dry storage was considered). Had this possibility been envisioned, the need for a Technical Specification change would have been rec-ynized.

u' WJCLIAn "OUL&1T*Y CONO OWC pWM XSA UCENSEE EVUfT REPORT ILERl TEXT CONTINUATION 04F, ,,evIo 0 40 "2-41o4 1 EX*il$ S31*S

&COUTV~ .&NM III                                 000111T oft~al"   In          I    LIMN~                              t Oyster Creek, Unitt                        o IsleoooIZ l1             9 817    -OIO 16I--0 P 0 !3oPO                3 A contributing factor in this event is that procedural controls were inadequate. The refueling procedure (205.0) contains a precaution regarding the Technical Specification restriction on fuel bundle enrichment, however, the refueling procedures do not require verifications to ensure compliance with the enrichment restriction associated with fuel stored in the spent fuel pool.                             Had such a verification bees performed,          the      fuel   would  not  have  been    stored    in   the spent fuel pool.

ANALYSIS OF OCCURRENCE AND SAFETY ASSESSMENT The Initial criticality analysis of the High Density Poison Racks (HDPR) assumed a uniformly enriched lattice of 3.01 wt% U-235. The analysis also did not take credit for burnble poisons as allowed In Regulatory Guide 1.13. The lattice K-infinity for the analysis was determined to be 1.33 which resulted In a HDPR cell K-effective of less than 0.95. The P8DRB299 bundles in the spent fuel pool have a maximm cold, uncontrolled K-infinity of 1.22 as determined by the fuel vendor. Therefore, it is expected that the cell K-effective did not reach or exceed 0.95. rowever, a re-evaluation of the HDPR criticality analysis is currently being performed taking credit for burnable poisons. The results of this analysis will be submitted in a supplement report. Corrective Actions Currently, there are no fuel bundles with an average planar enrichment of greater than 3.01 wt% IJ-235 in the spent fuel pool. Corrective actions will consist of the following:

1. Fuel movement procedures will have appropriate controls added that ensure Technical Specification compliance in this area.

2. Based upon the results of the HDPR re-evaluation, a Technical Specification change request will be submitted to allow fuel bundles with higher average planar enrichments to be stored In the spent fuel pool.

3. This event will be reviewed with the engineering personnel involved stressing the requirements to consider all licensing basis docuents and associated restrictions when performlig safety reviews.

SIMILAR OCCURRENCES None (0288A)

6 6

• Nuclear Post Office Box 388 Route 9 South Forked River. New Jesey 08731-0388 609 971-4000 Writer's Direct Dial Number:

February 24, 1987 U.S. Nuclear Regulatory Comission Document Control Desk Washington, DC 20555

Subject:

Oyster Creek Nuclear Generating Station Docket No. 50-219 Licensee Event Report This letter forwards one (1) copy of Licensee Event Report (LER) No. 87-006. Very truly yours, Peter V-F dIe Vice President and Director Gyster Creek PBF:KB:dam(0288A) Enclosures cc: Dr. Thomas E. Hurley, Administrator Region I U.S. Nuclear Regulatory Commission 631 Park Avenue King of Prussia, PA 19406 Mr. Jack N. Donohew, Jr. U.S. Nuclear Regulatory Commission 7920 Norfolk Avenue, Phillips Bldg. Bethesda, MD 20014 Mail Stop No. 314 NRC Resident Inspector Oyster Creek Nuclear Generating Station Forked River, NJ 08731  ; -. 2

                                                                                          ,[,

• I GPU Nuciew CorPoratiom Nuclear Post Office Box 388 Route 9 South Forked River. New Jersey 08731-0388 609 971-4000 Writer's Direct Dial Number:

February 24, 1987 U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555

Subject:

Oyster Creek Nuclear Generating Station Docket Ne. 50-219 Licenso-, Event Report This letter forwards one (1) copy of Licensee Event Report (LER) No, 87-006. Very truly yours. Peter-3-lde Vice President and Director Oyster Creek PBF: KS: dam(0288A) Enclosures cc: Dr. Thomas E. Murley, Administrator Region I U.S. Nuclear Regulatory Commission 631 Park Avenue King of Prussia, PA 19406 Mr. Jack N. Donohew, Jr. U.S. Nuclear Regulatory Coumission 7920 Norfolk Avenue, Phillips Bldg. Bethesda, MD 20014 Mail Stop No. 314 NRC Resident Inspector Oyster Creek Iuclear Generating Station Forked River, NJ 08731 U"'

EXHIBIT B-16 NRC Information Notice 94-13 (February 22, 1994)

UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION WASHINGTON, D.C. 20555 February 22, 1994 NRC IIOFMATION NOTICE 94-13: UNANTICIPATED AND UNINTENDED MOVEMENT OF FUEL ASSEMBLIES AND OTHER COMPONENTS DUE TO IMPROPER OPERATION OF REFUELING EQUIPMENT All holders of operating licenses or construction permits for nuclear power reactors. The U.S. Nuclear Regulatory Commission (NRC) is Issuing this information notice to alert addressees to potential problems resulting from inadequate oversight of refueling operations and inadequate performance on the part of refueling personnel. It is expected that recipients will review the information for applicability to their facilities and consider actions, as appropriate, to avoid similar problems. However, suggestions contained in this information notice are not NRC requirements; therefore, no specific action or written response is required. VescriDtion of Circumstances Vermont Yankee Events The Vermont Yankee facility was in a refueling outage with fuel movement in progress on September 3, 1993, when an irradiated fuel assembly became detached from the grapple after being lifted out of its position in the reactor core. The assembly fell approximately 2.4 m [8 ft] back into its orijinal location in the reactor core. The licensee suspended fuel handling and investigated the event. The licensee determined that the grapple had not properly engageJ the lifting bail on the fuel assembly and that the personnel performing the fuel handling activities had failed to verify proper grapple engagement. After completing the investigation and taking corrective actions, the licensee resumed fuel handling activities on September 7. 1993. On September 9, 1993, a fuel assembly that was being moved to a fuel sipping can was inadvertently lowered, instead of raised, striking another core component. The potentially damaqed fuel assembly was then moved to the fuel sipping can and the licensee again suspended fuel handling activities. The NRC dispatched an augmented inspection team (AIT) on September 9, 1393. to investigate the fuel handling incidents. The Aif documented its f'niings in NRC Inspecizon Report 50-27!,'93-91. issued October 21, 1993. The All con:ludad that mistakes madý b* re10

                                                               ,    ,   Zrornel 9407 15018?      DF~~              i.~

I February 22: 1994

                            'Papg2                                            of 5 Ih~ i~lIate ts. of Do      evetts1.' hidi

• eknse I Mt thrbe c Is fo. ?t. lseuim

                        .0t1ag a         effec~.t-) _jettrs                wee t are       :

c~tu VCa steps' most instaw~s. Ifftic'ly. the petrsonnel?

  "ioftotim. t* flowl b       tog ac, ities war* not            of the- requirement to PtfVisally verify gr appllosuro         INe ng agqitng and I1 ti9 fuel atssehis.

gAnam t dAt not commurni cate 0e T1he Alf fouwd -that~ tations and provide proper over ight of fue1 hand)lI 4 actilties. Peach Bottom Events With Unit 3 shut dlown for refuellng on September 23i M99,a fuel assembly coul d not be fully inserted into i4ts spent fuel rack. cell. It was thought that the fuel assembly had swelled due to irradlatiim in the core, and the fuel assembly was successfully placed in a different-cell. It was further postulated that there might be sine debris in the call. and that the cell should be checked at som future, date. On Septembesh. 24, 1993 another fuel assembly became stuck In its spenlt fuel rack cell. The licensee evaluated the material condition of the fuel assembly, calculated an allowable lifting force, and conferred with the fuel vendor. The licensee Inc-,-eased the load limit of the refueling hoist and the fuel assembly was freed from the rack with no damage to the fuel assembly. Subsequent exainations revealed that sections of local power range monitor instrument strings that had previously beer cut up were in the bottoms of three cells in the rack, including the two cells with which difficulties were experienced. The licensee believes that the debris may have fallen into the cells during a fuel pool cleanup effort coujucted during the previous summer. The licensee is currently investigating why the debris was in the spent fuel pool and why the refueling personnel did not ensure that the spent fuel rack cells did not contain any debris prior to inserting the fuel assemblies. 1pssuehanna Events The Susquehanna Steam Electric Station Unit I was shut down with defueling in progress on October 6, 1993, when the personnel performing the fuel handling activities removed an incorrect fuel assembly from a peripheral location in

   'he core. The personnel involved realized they had removed the.wrong asserr~ly and they inappropriately dEcided to return the assembly to its prior positirin in the core. The appropriate action, per licensee procedures, iioulj have t'aen to place the bundle in the spent fuel pool ard secure fue! ha'~n       !      act wit e until the cause of -the error was deta-r.'cA andc~re.

                                                       *-                           P:03
                                                                                     ."   i .of 5 19     .

rZ, .1 loweri A*9 a -.. sely Into-the during r !o fl be2 IS of the I Ssmecti the blad guide hit the side Of t *ctor vesse! becane it wos not i1*1 r high e gh to clear the vessel. T- lieise I IseenM refulag ties, revised thossociated Irocedift, and Imspcted tho =st. The reload was resumed after surveillamces os the fuel had)11, equipment

                 ',Successfully conducted. On October M1 IM93, while attempt Og to grapple fuel assembly In the fuel pool, the'personnel perforfmin the fuel
      *-  M hauing        activities heard two loud bangs and observed bubbles to the pool for of the mast t 10 seconds. Subsequent inspection revealed that one section 0S by the from Unit 2 was bent. The licensee believes that the                 mast was  weakened during the October 27        event.

I- -w impot with the reactor vessel that occurred Onk October 29, 1993. the NRC dispatched an AIT to the site to review the Y teveits. The AIT documented its findings in Inspection Report W387/93-80. Issued on December 21, 1993. The AIT concluded that facility management did not maintain proper oversight of refuel floor activities and that inadequate with the fuel V corrective actions were Implemented in the past for problems fuel handling handling equipment. The AIT also concluded that the licensee procedures were adequate for the proper completion of the fuel handling activities, although certain improvements could be made to increase the awareness of the operators concerning potential problems. Nine Mile Point Event Nine Mile Point Unit 2 was shut down with refueling in progress on November 1. 1993. when a blade guide was moved from the core into the spent fuel pool. The contractor refueling or erator disengaged the grapple and observed the correct light indication in the bridge. There was no procedural requirement to visually verify disengagement or for the Senior Peactor Operator Limited to Fuel Handling (LSRO) or the spotter to verify disengagement. The refueling operator noticed increased drag after the refutling bridge crane had been moved approximately 23 cm (9 in) toward the next location. At that time. licensee personnel determined that the blade guide was still engaged on the grapple. The bridge was returned ti its previous position, the blade guide was lowered and di;enqaqe. (positive verification was obtained this tim*e), and the operator pioceeded to "de the While lowerir3 t f,' as. bPy next component, which was a fuel assembly.

ling f dperat aftor releasb ach Co"Ont. ea PM wxojW "tattio of lowr. I ha Whim hiWdli Issl)fI e hie. te be mwifted l ratsing M nd rMtailg Ias te - he 44snot verify disengagement afte releasingi the blade ftilo, the refueling operator did mOt netify the LSM ef the einatiipetd epuipmt response (reaining connected to the blade #614t "bile the th brip).Also contrtbutinwg to the event was the fact that te'wS~um was t refueling bridge trolley beaing about which he

                          )ervnga comsmSd, rather than the hasdling of the blade uitde. Licensee review dotetfoled tOtt management expectations regarding the supervision of refueling activitites had not been clearly expressed to the LSROs.

efveliqg activities are safety-significant operations that are not conducted on a rotine basis. In addition, fuel handling activities are often performed by contractor personnel under the supervision of licensee personnel. As a resolt, fuel hatnli wg personnel may not be familiar with the fuel handling equipment Or SAy feel that their experience in fuel handling operations pmrits them to ignore some requirements for procedural use and adterence. Either of these situations could require increased management attention and overs'ght by the licensee to ensure proper and safe performance of fuel handling activities. Appendix B to Part 50 of Title 10 of the Code of Federal Regulattons to control (10 CFR 50) requires licensees to have appropriate procedures activities affecting quality (such as the actions to be taken durtqg operatton of refueliln* equipmenet). and that the procedures are u$ed and foltow*d. ln requires licensees to implement a training pro<;rim *r

.dd'tiln,        CFR 50.120 those various categories of nuclear power plant personnel to ensure that knowledge,       skills,        and   abilities    to    perform       their personnel have the necessary assigned jobs competently.          This rhle      applies        to   the  personnel     (including contractors) who operate er supervise the operation of the refueTinq equipment.      The cases discussed in this notice include situalio-s in wf',Ch toe
                                                                                          "-*'i        "i,,     ",

licen;ees failed to conduct appropriate training in tI~e u-P equipment, particularly w0K respect to, Jes*q' , odmf'cat,, , ,' 41-.- -4 *. controls for the fuel 'rast v pno-,~;~ " Toa, 1

IN 94-13 February 2Z, 1994 Page 5 of 5 handling personnel Involved In certain instance: wore variously not aware that managutent expected them to identify deviations from expected results, cease operations when an unexpected ur abnormal condition is encountered, and notify operations and/or plant management of unexpected or abnormal conditions. This Information notice requires no specific action or written response. If you have any questions about the information in this notice, please contact one of the technical contacts listed below, or the appropriate Office of INuclear Reactor Regulation (NRR) project manager. Brian K. Grimes. Director Division of Gperating Reactor Support Office of Nuclear Reactor Regulation Technical contacts: P. L. Eng. NPR E. M. Kelly, RI (301) 504-1837 (215) 337-S183 J. R. White. RI L. E. Nicholson. Rt (215) 337-5114 (2*5) 337-5128 AttAchment: List of Recer~tly Iss.aed NPC "r'or-rl.ý,-r Notic -s

Attachment IN 94-13 February 22, 1994 Page I of 1 LIST OF RECENTLY ISSUED NRC INFORMATION NOTICES uate uate ur 51e infoozation Issuance Issued to Notice No. Subject Insights Gained from 02/09/94 All holders of OLs or CPs W4-12 for nuclear power reactors. Resolving Generic Issue 57: Effects of Fire Protection System Actuation on Safety Related Equipment Turbine Overspeed and 02/08/94 All holders of OLs or CPs 94-11 for nuclear power reactors. Reactor Cooldown durinq Shutdown Evolution 02!04/94 All holders of OLs or CPs 94-10 Failure of Motor-Operated for nuclear power reactors. Valve Electric Power Train due to Sheared or Dislodged Motor Pinion Gear Key 02/031"94 All U.S. Nuclear Regulatory 94-09 Release of Patients with Commission medical Residual Radioactivity licensees. from Medical Treatment and Control of Areas due to Presence of Patients Con taining Radioactivity Following Implementation of Revised 10 CFR Part 20

                                                  -1*01 /94     All holders of OLS or CPs 94-08            Potential for Surveil                         for nuclear power reactors.

lance Testing to Fail to Detect an Inoperable Main Steam Isolation Valve 01/31,l94 All holders of OLs or CPs 93-26. Grease Solidification for nuclear power reactors. Supp. l Causes Molded-Case Circuit Breaker Failure to Close 01/28 ;94 All byproduct material and 94-07 Solubility Criteria for fuel cycle licensees with Liquid Effluent Releases the exception of licensees to Sanitary Sewerage Under authormze' solely for the Revised 10 CFR Dart 20 1 se3 .PJ 'c~rce.% 01 - Operat:r9 fiL1-CeC CP = ConstruCtor Perrit

EXHIBIT B-18 NRC Information Notice 94-13, Supplement 1 (June 28, 1994) 18

UNITED STATES NUCLEAR REGULATORY COMIISSION OFFICE OF NUCLEAR REACTOR REGULATION WASHINGTON, D.C. 20555 June 28, 1994 NRC INFORMATION NOTICE 94-13, SUPPLEMENT 1: UNANTICIPATED AND UNINTENDED MOVEflENT OF FUEL ASSEMBLIES AND OTHEk COMPONENTS DUE TO IMPROPER OPER*ATION OF REFUELING EQUIPMENT nuclear power All holders of operating licenses or construction permits for reactors. information The U.S. Nuclear Regulatory Zommission (NRC) is issaing this to an event involving unauthorized notice supplement to alert addressees movement of a defective spent fuel rod. It is expected that recipients will review the information for applicability to their facilities and consider actions, as appropriate, to avoid similar problems- However, suggestions notice are not NRC requirements; therefore, no contained in this information specific action or written response is required. Bac kground Unintended The NRC issued Information Notice (IN) 94-13, "Upanticipated and Operation of Movement of Fuel Assemblies and Other Components Due to Improper result from Refueling Equipment,* to alert addressees to problems that could performance on the inadequate oversight of refueling operations and inadequate IN 94-13 described various refueling events that part of refueling personnel. and Nine Mile Point. occurred at Vermont Yankee, Peach B',ttom, Susquehanna, operation These events demonstrate the importance of proper controls over, and Waterford Steam of, refueling equipment during use. A recent event at the fuel damage or Electric Station (Waterford) demonstrates the potential for that is not personnel hazards which could result from fuel-handling equipment properly stored and not secured from unauthorized use. Oescription of Circumstances power On February 18, 1994, the Waterford plant was operating at 100-percent an unknown o,'ect hanging from the when a senior reactor operator found fuel-handling machine in the fuel-handling building. Health physics found technicians measured radiation levels in the spent fuel pool area and Licensee personnel itmotely secured the object with vise them to be normal. to .7 Sv/hr grips and determined that underwater radiation levels were .2 (2O to 70 R/hr) at 15 centimeters (6 inches) from the object. A Combustion the object as a fuel rod encapsulation tube. Engineering employee identified No visual damage was apparent on the tube. The licensee posted a security guard in the spent fuel pool area and reported the event to the NRC. 9406210075 PDý U~z~ XýC_ W0'.c 9Lq.OI)3 4i~f/ q1z

IN 94-13, Supplement 1 June 28, 1994 Page 2 of 3 that the tube The licensee reviewed fuel storage records and determined from an irradiated fuel removed contained a defective fuel rod that had been the tube had been placed in a assembly several years earlier' At that time, center guide tube in a grid cage stored in the spent fuel racks. The licensee reviewed computer access records for the fuel-handling area and interviewed may have had access to the relevant personnel about the event. Personnel who the event. The fuel-handling machine completed questionnaires regarding used the fuel-handling had licensee determined that the refueling director and had parked the machine the day, before the object was discovered the fuel rod encapsulation fuel-handling machine at a location directly over the hoist and was not sure tube. However, the refueling director had not used was hanging from the that he would have noticed if the encapsulation tube records indicated that hoist at the time he used the machine. Surveillance attached to the fuel-handling the fuel rod encapsulation tube must have become tool sometime between February 11 and 18, 1994. tube showed that the Design drawings of the cap of the fuel rod encapsulation inner diameter of the end of outer diameter of the cap was about equal to the bound in the become the fuel-handling tool. Apparently, the cap had the top of the spent fuel fuel-handling tool when the hoist was lowered to removed from the rack and, when the hoist was raised, the tube was completely grid cage. operations for previous Although contractors had performed the fuel-handling to perform the fuel refueling outages, Waterford personnel were scheduled speculated that handling for the March 1994 refueling outage. The licensee for the March outage one of the people assigned to fuel-handling activities tube while practicing the use may have inadvertently lifted the encapsulation physics staff before of the hoist. Personnel were required to notify health records showed that accessing the refueling machine; however, health physics No keys or special no one had made such a notification during this time. machine. knowledge was needed to access the controls of the fuel-handling by closing two electrical breakers and Electrical power could be obtained The licensee questioned pushing one switch that were located on the machine. use of the several employees, but io one admitted to unauthorized fuel-handling machine. the computer that As an interim corrective action, the licensee deenergizedin a locked power controls the fuel-handling machine by opening a breaker a means to prevent the control center. The licensee planned to (1) develop lifted by the fuel rod encapsulation tube from being inadvertently procedure warning fuel-handling tool, (2) add a precaution to the operating storage location, and operators not to lower the fuel-handling tool over the training. (3) add hoist manipulations to the lesson plans for proficiency Discussion fuel and core Procedures governing the use of equipment for handling of that operation components may not prevent unauthorized or unintended energize the equipment. Precautions such as locking out breakers that areas in highly visible fuel-handling equipment and the placement of placards is forbidden Pquivrpnt declaring that unaiithorized oppration of fuel-handling

IN 94-13, Supplement 1 June 28, 1994 Page 3 of 3 may help ensure that the equipment is not used without proper authorization. accidental I Additionally, storing the fuel-handling machine in an area where movement of the hoist or grapple will not impact stored fuel or other

• 1 damage. Management attention and oversight of the operation movement or components may contribute to the prevention of Inadvertent fuel fuel and core of and core coMponent handling equipment Is Important to ensure that fuel and that plant components are protected from damage or unauthorized movement personnel are protected from unnecessary exposure to radiation.

If ihis information notice requires no specific action or written response.contact you have any questions about the information in this notice, please Nuclear the technical contact listed below or the appropriate Office of P:actor Regulation (NRR) project manager. Brian K. Grime.s, Director Division of Operating Reactor Support Office of Nuclear Reactor Regulation Technical contact: Dale A. Powers, RIV (817) 860-8195

Attachment:

List of Recently Issued NRC Information Notices

Attachment IN 94-13, Supp. I June 28, 1994 Page 1 of I LIST OF RECENTLY ISSUED NRC INFORMATION NOTICES Date of .'Information Issuance Issued to Notice No. Subject 06/21/94 All U.S. Nuclear Regulatory

94-47 Accuracy of Information Commission Material Provided to NRC during Licensees.

the Licensing Process 06/20/94 All holders of OLs or CPs 94-46 NonConservative Reactor for nuclear power reactors. Coolant System Leakage Calculation 06/17/94 All holders of OLs or CPs 1J4-45 Potential Common-Mode for nuclear power reactors. Failure Mechanism for Large Vertical Pumps All holders of OLs or CPs Main Steam Isolation 06/16/94 for nuclear power reactors. 94-44 Valve Failure to Close on Demand because of Inadequate Maintenance and Testing 06/10/94 All holders of OLs or CPs 94-43 Determination of Primary for pressurized water to-Secondary Steam reactors. Generator Leak Rate 06/07/94 All holders of OLs or CPs 94-42 Cracking in the Lower for boiling-water reactors Region of the Core (BWRs). Shroud in Boiling-Water Reactors 06/07194 All holders of OLs or CPs 94-41 Problems with General for nuclear power reactors. Electric Type CR124 Overload Relay Ambient Compensation 05/26/94 All holders of OLs or CPs 94-40 Failure of a Rod Control for pressurized-water Cluster Assembly to Fully reactors (PWRs). Insert Following a Reactor Trip at Braidwood Unit 2 05/31/94 All U.S. Nuclear Regulatory 94-39 Identified Problems in Commission Teletherapy Gamma Stereotactic Medical Licensees. Radiosurgery OL = Operating I icense CP = Construction PC m"t

EXHIBIT B- 17 Three Mile Island Unit 1: LER 289/98-002-01 (April 3, 1998)

GPU Nuclear, Inc. Route 441 South NUCLEAR '2 ' Post Office Box 480 Middletown, PA 17057-0480 Tel 717-944-7621 April 03, 1998 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555

Subject:

Three Mile Island Nuclear Station, Unit 1 (TMI-I) Operating License No. DPR-50 Docket No. 50-289 Lice-en No. 98-002, Revision I On February 4. 1998, GPU Nuclear determined that the Spent Fuel Pool was not sampled in accordance with the requirements of the Technical Specifications Surveillance Requirement (SR) specified in Table 4.1-3, item 4, which requires sampling monthly and after each makeup. A review of work activities determined that no sample was taken following a water addition on January 23, 1998. This condition was found to be reportable in accordance with 10 CFR 50.73(a)(2Xi)(B) as a condition prohibited by Technical Specifications. A subsequent analysis determined that the filling activity could not have diluted the boron concentration significantly. This condition was reported to the NRC by letter dated March 3, 1998. Attached is Revision I of LER 98-002, which provides additional information that addresses the following items: the reason for this event, the extent of the problem associated with the missing operator aid, the assessment of the safety consequences and implications of the event, and the corrective action section. The event did not affect the health and safety of the public. Please contact Adam Miller, TMI Licensing at (717) 948-8128 if you have any questions regarding this matter. Sincerely, Langenbacý ý Watfs. Vice President and Director, TNM AWM" 2I, cc: TMvfl Senior Resident Inspector Administrator, Region I TMI-1 Senior Project Manager File 98048 9804130278 980403 PDR ADOCK 05000289 S PDR

muC TOMi 366&- 1.S. U ?JOMUSSICU SJC.ZAR A3NLA1 (4,-95) LICENSEE EVENT REPORT (LER) TEXT CONTINUATION rALruZTr WMh (1) J DOC1W M (2) YEAR I LZR wulA (6) SEQUENTIALuBR REVISIONuBR PAM (31 Three Mile Island, Unit 1 05000289 98 -- 002 -- 1 5 OF 6 T (It more space is required, use additional copies of NRC Form 366A) (17) appropriate because no major replenishment of pool water is expected to take place over a short period of time." This bases appears to be consistent with the TMI-1 TS bases. The TMI-1 staff is evaluating if a request to revise the current surveillance requirement is appropriate. VII. Corrective Actions: A. Corrective Actions Taken:

1. A new Operator Aid has been posted at the valve that is used tp fill the Spent Fuel Pool (SFP) from the Reclaimed Water System. In order to ensure the Shift Supervisor tracks the need for the water sample, the new Operator Aid has been modified to add a step to require the individual doing the fill to notify the Shift Supervisor to track this item on the S/S Turnover until the SFP sample is taken and analyzed within the designated time period.

2. The Primary Auxiliary Operator Turnover Checklist has been revised to include a requirement to notify the Chemistry Department of sample requirements if a water addition to the Spent Fuel Pool has either been initiated or completed durifig the shift. The contents of this checklist are discussed at the crew briefing and the checklists of all the Operators are compiled and reviewed by control room supervision.

B. Action Planned to Prevent Recurrence:

1. This revised LER will be reviewed by all of the appropriate personnel in the Operations and Chemistry Departments. The review will be documented and the documentation maintained by the Operations Department Administrator. This action will be completed within 60 days of the issuance of this revised LER.

2. To determine the extent of the problem associated with missing Operator Aids, a spot check of Operator Aids will be performed. Each Shift Supervisor will select five (5) of the Operating Procedures for which he is the owner. This selection will only include procedures that contain Operator Aids. All of the Operator Aids contained in these 25 Operating Procedures (i.e. 5 crews at 5 procedures per crew) will be physicaly verified to insure that they are properly posted, not broken, legible, and accurate. This verification will be completed and an assessment of the verification performed by the Lead Operations Engineer prior to April 30, 1998. If the assessment reveals that the Operator Aids are in poor condition, a 100% verification will be performed for the remaining Operator Aids.

NRC FORM 366A (4-95)

3E66J U. 3. MUCL l 3zGUKLao, X CM SSIC "4-5) LICENSEE EVENT REPORT (LER) TEXT CONTINUATION rJ-n"cm a (1) DOO LUR UW (6) (3) YEAR SEQUENTIAL V N NUMBER NUMBER Three Mile Island, Unit 1 05000289 98 -- 002 -- 6 OF 6 TZ=T (If more space is required, use additional copies of NRC Form 366A) (17)

3. All of the Operating Procedures which contain Operator Aids will be reviewed to insure that the Operator Aids do not contain direction or guidance which would be the sole source of information provided during a task performance to comply with Technical Specification requirements. This review will be completed prior to April 30, 1998.

4. All licensed personnel will be given training on an overview of the contents of Technical Specifications section 4 and specific training on the sampling requirements of Table 4.1-3. This training will be completed by 12/31/98.

The Energy Industry Identification System (EIIS), System Identification (SI) and Component Function Identification (CFI) Codes are included in brackets, [SI/CFI] where applicable, as required by 10 CFR 50.73 (b)(2)(ii)(F). NRC FORM 366A (4-95)

VrW IM 366 U.S. MCi*rZAR RZWJLAIOl CO SSION APlRVW By am no. 31.50-0104 (4-95)

• winRS 04/30/96 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS MANDATORY INFORMATION COLLECTION REQUEST: 50.0 HRS.

REPORTED LESSONS LEARNED ARE INCORPORATED INTO THE LICENSERE VENT REPORT (LIM) LICENSING PROCESS AND FED BACK TO INDUSTRY. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION (See reverse for required number of AND RECORDS MANAGEMENT BRANCH (T-6 F33), U.S. NUCLEAR digit3/character3 for each block) REGULATORY COI4MISSION, dTHE PAPERWORK REDUCTIONWASHINGTON, DC 20555-0001, AND TO PROJECT (3150-0104), OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503. Three Mmil Ilnd Three Mile Island, Unit I Uni J' (2) 05000289 1A OF6 1OF6 T:LEZ, (46) MISSED SPENT FUEL POOL SAMPLE FOLLOWING A WATER ADDITION DUE TO UNFAMILIARITY

• .XiTH S~AMP[INC7 REOUIREMENTS AND A MISSING OPERATOR All)

MONTH 2 EVE" DAY 04 (5)

                          )

YEAR 98 YEAR 98 I I 002 SEQUENTIAL UBE[NUMBER (6) m REVISION 1 REPORT MONTH 04 DATE DAY 03 (7) YEAR 98 FACILITY NAME F 0T[] 7 MOVED (9) DOCKET NUMBER I NMBR. N. BRFACILITY NAME 1 00CKET NUMBER Ol*URAJM TKI(S RIEiPORT 1 S 13U PURSUA* T TO THE ACQUIR11-MUTS Or 10 CR m*: (Check one or morel (11) N 20.2201(b) 20.2203(a) (2) (v) X 50.73(a) (2) (I) 7 50.73 a( 7 (viii) 20.2203(a) I1I 20.2203(a) (3) W') 50. 3(a) (2) [ii) 50. 3(a) (21 (xl 7 20.2203(a) (3) (ii) 50. 3(a) (2) (iii) 73.71 100 20.22031a) 121 (i) 7 20.2203(a) (2) (ii) 20.2203(a) (4) 50. 3(a) 12) (iv) OTHER 7 20.2203(a} (2) (ii)7 50.36(c) (1) 50. 3(a) (2) (v) Specify in Abstract below or in NRC Form 20.2203(a) (2) (iv) 50.36(C) (2) j50.73(a) (2) (viil 3-L=*=CONITACT MR TX13 LEE Ila) NAME HE NUMBER (Include Area Code) Adam Miller, TMI Licensing Engineer (717) 948-8128

                                                            *        ~~FR                         B*.3[UZDMM*JiRr                1   TllI     EXOR,as  m     131 CAUSE              SYST          COMPONENT         MANUFATURE            R                                CUE              SYTE                COMPONENT         MANUFACTURER            REPORTABLE TO NPRDS                                                                                                       TO NPRDS
                                   .UP93              I           MR-          (14)                            -..                                                     NTH           DAY            YE 3                                                                                                  qIllldt      ew* lN (If yes,
       'T
        -J' complete EXPECTED SUBMISSION (Limit to 1400 spaces, (Litit    to 1400 spaces, i.e.,

i.e., DATE). approximately 1) ",ingle-spaed

                                                                                             'IT approximately 15 single-spaced typewritten a lines) typewritten lines)

(16) (16) I I February 4, 1998, GPU Nuclear determined that the Spent Fuel Pool was not sampled in accordance with the requirements of the Technical Specifications Surveillance Requiremcnt (SR) specified in Table 4.1-3, item 4, which requires sampling monthly and after each makeup. A review of work activities determined that no sample was taken following- a water addition on January 23, 1998. This condition was found to be reportable in accordance with 10 CFR 50.73(a)(2)(i)(B) as a condition prohibited by Technical Specifications (TS). An analysis determined that the filling activity could not have diluted the boron concentration significantly. Contributing factors for this event were Control Room supervision unfamiliarity with the sampling requirements and a missing sign by the fill valve, which serves as an Operator aid. The missing sign has been I replaced and the Primary Auxiliary Operator (AO) Turnover Checklist has been revised to include a requirement to notify the Chemistry Department of sample requirements if a water addition to the Spent Fuel Pool has either been initiated or completed during the shift. Additionally, licensed personnel will be given training on Technical Specification section 4 requirements. There were no adverse safety consequences from this event, and the event did not affect the health and safety of the public. 9804130281 980403 PDR ADOCK 05000289 R *Prp

snZIow ma.C F0 (4 -95) 36"k u~s MYNLZLR RBM28 LICENSEE EVENT REPORT (LER) TEXT CONTINUATION ACL NAM* (1) DO,.*IM LZ Mum (6) Fpa, (3) 05000289 Y EA NUMBER NUMBER Three Mile Island, Unit 1 98 -- 002-- 1 2 OF 6 (It more space is required, use additional copies of NRC Form 366A) (L0) TC I. Plant Operating Conditions before Event: TMiI-I was operating at 100% steady state power prior to and during the event described in this LER. II. Status of Structures, Components, or Systems that were Inoperable at the Start of the Event and that Contributed to the Event: None. LI. Event

Description:

The TMI-I Technical specifications Table 4.1-3.4 requires that the Spent Fuel Pool Water [DA]* be sampled monthly and after each makeup. Operating Procedure 1104-6 "Spent Fuel Cooling System" requires notification of the Chemistry Department at the completion of a Spent Fuel Pool water addition that a sample must be taken between 24 to 48 hours after the addition was completed. Contrary to thcse requirements, a water addition was made to the Spent Fuel Pool on 01/23/98 (From 0918 to 1705) without the required follow-up water sample. Another addition was made to the Spent Fuel Pool on 01/27/98 (From 1410 to 1817). The Spent Fuel Pool was then sampled on 01/28/98 at 0430 and again on 01/29/98 at 0830. These samples exceeded the 48-hour sample requirement for the addition that was performed on 01/23/98. During a routine review of work activities by the Chemistry Department, a Staff Chemist noticed that samples of the Spent Fuel Pool had been obtained on 01/28/98 and 01/29/98. He recognized that these samples were taken to comply with the requirement to sample the Spent Fuel Pool after each addition. The Staff Chemist identified a possible lack of formal tracking for samples after filling the pool to the Manager, Radwaste and Chemistry. The manager investigated the scope of the potential problem by reviewing two months of spent fuel pool boron concentration data and the computerized Control Room Logs. lie found that there was no spent fuel pool boron data following an addition to the pool on 01/23/98. The manager submitted a CAP (Corrective Action Process) Form (T1998-0066) to document the missed sample. NRC FORM 366A (4-95ý

V. S. WO.KAaL 3ZOUL110tr CC!aSz (4-95) LICENSER EVENT REPORT (LER) TEXT CONTINUATION _-DO_- LZR OU-UM (6) PAM (3) jNUMBER r.*X*Tr HA (1) blo 12) REVISION YEAR SQUENTAL I NUMBER Three Mile Island, Unit 1 05000289 98 -- 002 -- 1 3 OF 6 more space is required, use adcitional copies of NRC Form 366A) (11) T*X= (If IV. Identification of Root Cause The Primary AO was notified at the shift turnover meeting of the intent to fill the Spent Fuel Pool. This task is coordinated with processing (i.e. purifying) water with the ECOLOCHEM system on the secondary plant. As water is processed on the secondary plant it is transferred to the Reclaimed Water Storage Tank on the primary plant. This tank is then used as the source tank to fill the Spent Fuel Pool. Towards the end of the shift the AO was notified of the intent to shutdown the ECOLOCHEM system which would in turn require securing the filling of the Reclaimed Water Storage Tank and the Spent Fuel Pool. When the Primary AO terminated the filling of the Spent Fuel Pool he made an entry in his logbook and also notified the Control Room. He did not convey any sample requirement information to the Control Room. Contrary to the requirements of the Operating Procedure the Operations Department did not notify the Chemistry Department of the need to sample the Spent Fuel Pool and the Shift Supervisor did not track the need for the water sample on the Shift Supervisor's Turnover. The Shift Supervisor was aware that the Spent Fuel Pool fill had been performed, but he was unaware of the Technical Specification section 4 sampling requirements. The task of filling the Spent Fuel Pool is considered a routine evolution that does not require the operator to actually use a copy of the Operating Procedure to perform the evolution. For this reason an Operator Aid is affixed to the wall directly behind the valve used to fill the Spent Fuel Pool in order to remind the Operator of the notification and sampling requirements. However, the Operator Aid was missing from the wall on 01/23/98 when the Spent Fuel Pool was filled. Therefore there was no Operator Aid available to serve as a reminder to the Operator to notify Chemistry and the Shift Supervisor at the completion of the fill process. There has been one previous occurrence, June 13, 1996, where a water addition was made to the Spent Fuel Pool without the required follow-up sample being performed. (This is the first that resulted in an LER due to the recent change at TMI- I concerning the reportability of a missed Tech Spec Surveillance). As a result of the previous occurrence, the procedure guidance contained in Operating Procedure 1104-6 "Spent Fuel Cooling System" was enhanced to require notification of the Chemistry Department of the required sample and to track the need for a sample on the Shift Supervisor's Turnover until the sample is taken and that analyzed. As part of the procedure enhancement, an enclosure to the procedure was added to indicate an Operator Aid is posted at the Reclaimed Water supply valve. Factors which contributed to the failure to obtain the Tech Spec required sample of the Spent Fuel Pool following a water addition are:

                  "* Pertinent information not transmitted
                  "* Required procedure/document not followed
                  "* Installed Operator Aid not provided (i.e. missing)

NRC FORM 366A (4-95)

U.S. UMULEAZ EZO.a, C.. USSIW NBC toM 366" 495) LICENSEE EVENT REPORT (LER) TEXT CONTINUATION rACIMx Ru (1) DOOK.. LM Ul. u (6) R-AM (3) NUMZR (2) YJ"] SEQUENTIAL REVISONt' UHBE ThreeMileIsland,_Unit_1 _050009 98-- 002NUMBER NUMBER Three Mfile Island, Unit 1 05000289 98 -- 002 11 4 OF6 IS (17) more space is required, use additionaJ copies of NRC Form 366A) As stated these are all contributing factors. We incorporate a work ethic that relies on defense in depth to guard against errors. The last barrier in this defense is ideally the Control Room licensed supervision. In this case we inappropriately relied on an Auxiliary Operator with the use of an Operator Aid to provide the last barrier. The Control Room licensed supervision missed the opportunity to be the last barrier due to not being familiar with Technical Specification section 4, Table 4.1-3 sampling requirements. V. Automatic or Manually Initiated Safety System Responses: No safety system responses occurred or were required to occur. VI. Assessment of the Safety Consequences and Implications of the Event: The failure to obtain a sample of the spent fuel pool following its fill on January 23, 1998 had no adverse safety consequences. Section 5.4.1 of TNfl-I Technical Specifications states "When fuel is being moved in or over the Spent Fuel Storage Pool "A" and fuel is being stored in the pool, a boron concentration of at least 600 ppmb must be maintained to meet the NRC maximum allowable reactivity value under the postulated accident condition." It also states that "When fuel is being moved in or over the Spent Fuel Storage Pool "B" and fuel is being stored in the pool, a boron concentration of at least 600 ppmb must be maintained to meet the NRC maximum allowable reactivity value under the postulated accident condition. The bases of section 4.1 of the Technical Specifications states "The 600 ppmb limit in Item 4, Table 4.1-3 is used to meet the requirements of Section 5.4. Under other circumstances the minimum acceptable boron concentration would have been zero ppmb." No movement of fuel was conducted between the time the spent fuel pool was filled on 1/23/98 and the next sample was taken on 1/28/98. If fuel movements had been planned, boron samples would have been taken in accordance with procedure 1505-1. The Technical Specifications Bases clearly indicate that no minimum boron concentration is needed in the spent fuel pool for safe plant operation, except during fuel movements. Because the boron concentration of the spent fuel pool is typically above 2500 ppmb (2897 ppmb following the fill) no normal filling operation (outside of filling because of a major leak in the pool, which was not on-going) could dilute the boron concentration significantly below its initial value. During the review of this event, it was determined that the TMI-1 Technical Specification Surveillance requirement for Spent Fuel Pool water sampling is different than the Standard Technical Specification (STS) requirement, which is to verify boron concentration every 7 days. The STS bases for this surveillance frequency states: "the 7 day frequency is NRC FORM 366A (4-951

Appendix C Assessing the Probability and Consequences of Criticality Events in Fuel Pools

1. Introduction This appendix provides technical background on the potential for inadvertent criticality in a fuel pool. Specifically, this appendix describes the steps that must be taken to assess the probability and consequences of a criticality event, and sets forth some interim findings about Harris pools C and D. These findings are necessarily of an interim nature, because Orange County has not identified any systematic assessment of the probability and consequences of a pool criticality.

Neither the NRC Staff nor the nuclear industry has attempted such an assessment or compiled the record of experience and other factual data that would support an assessment. The probability of a criticality event is discussed here in terms of six steps. First, the various types of criticality scenario are identified. Second, the probability of these scenarios is explored from a qualitative perspective. Third, the process of determining the envelope of criticality in a pool is described. Fourth, the potential for fuel mispositioning is outlined, drawing upon actual experience. Fifth, the potential for a reduced concentration of soluble boron is outlined, again drawing upon experience. Sixth, available criticality calculations for PWR fuel in Harris pools C and D are summarized, thereby showing the broad outlines of the envelope of criticality for these pools. Then, the nature and consequences of a criticality event are discussed. Finally, some conclusions are presented.

2. Probability of a Criticality Event 2.1 Overview Analytic techniques are available for assessing both the probability and consequences of a criticality event in a fuel pool. For example, relevant techniques have been employed for probabilistic risk assessments (PRAs) at nuclear power plants. However, Orange County has not identified any attempt, either by the NRC Staff, the nuclear industry or any other body, to conduct a systematic assessment of the probability and consequences of a pool criticality.

Moreover, there has been no systematic effort by the NRC Staff or the nuclear industry to compile the factual data that would be needed to support such an

Appendix C Assessing the Probabilityand Consequences of CriticalityEvents in Fuel Pools Page C-2 assessment. The relevant data would be drawn from actual operating experience at nuclear facilities. In the absence of a systematic investigation, one can make only qualitative statements about the probability of a criticality event in a fuel pool, drawing from publicly available information. 2.2 Types of Criticality Scenario This discussion focusses on the potential for a criticality event under abnormal conditions. Thus, for the purposes of this discussion, we ignore the possibility that a criticality event will occur in a fuel pool under normal conditions. In other words, if the pool contains as-specified fuel in as-specified fuel storage racks, and other parameters such as water temperature and soluble boron concentration are within their specified range, then we assume that a subcritical margin of reactivity will exist. Nevertheless, criticality could occur under normal conditions if there is a major error in the calculations that are performed to support the design and installation of the fuel storage racks. Appendix B shows that errors have occurred in calculations of this kind. For example, at Braidwood Unit 1, an incorrect assumption about the location of Boral panels was carried forward through successive calculations from 1987 to 1997. Also, at Millstone Unit 2, new calculations showed a Keffective of 0.963 whereas previous calculations, which had employed two inappropriate assumptions, showed a Keffective of 0.922. That is a substantial error, in a non-conservative direction. The potential for errors of this type is smallest when the rack design relies solely on geometry (the center-center distance between fuel assemblies) to prevent criticality. Under abnormal conditions, a variety of scenarios could lead to inadvertent criticality in a fuel pool. The number of potential scenarios is greater when a greater number of means are used to suppress criticality. If the prevention of criticality in the pool under normal conditions relies entirely on the use of geometrically safe racks, then three types of scenario could lead to criticality under abnormal conditions. First, an earthquake, drop of a heavy object into the pool or other mechanical insult might alter the rack geometry sufficiently to cause criticality. Second, fuel assemblies that are more reactive (e.g., with a higher-than-specified enrichment in U-235) than the specified limit for fresh fuel entering this facility might be placed in the racks. Third, fuel

Appendix C Assessing the Probabilityand Consequences of Criticality Events in Fuel Pools Page C-3 assemblies might be placed inside or outside a rack in a manner that does not conform to the intended geometry of fuel placement. If the prevention of criticality under normal conditions relies not only on rack geometry but also on the neutron-absorbing properties of the racks, then the three types of scenario outlined above could lead to criticality. In addition, criticality might arise if neutron-absorbing material is displaced from its intended position (e.g., if Boral panels become detached from the racks). If the prevention of criticality under normal conditions relies not only on rack geometry and the neutron-absorbing properties of the racks, but also on restricted fuel burnup/ enrichment or age, or on the presence of soluble boron, then criticality could arise through one of the scenarios outlined above or through additional scenarios. These additional scenarios would involve mispositioning of fuel assemblies, a reduction in the concentration of soluble boron in the pool water, or a combination of these occurrences. In this context, "mispositioning" would involve the placement in a rack of one or more fuel assemblies whose burnup/ enrichment or age is not within the specified range. In scenarios that combine fuel mispositioning with a reduced concentration of soluble boron, the mispositioning could either precede or follow the reduction in boron concentration. 2.3 Scenario Probability from a Qualitative Perspective Some of the criticality scenarios outlined in Section 2.2 would involve significant mechanical insult (e.g., an earthquake that disrupts the geometry of a rack) or mechanical failure (e.g., the detachment of Boral panels from racks). If the pool and the racks are designed, built and operated to prevailing standards, these scenarios will have a relatively low probability. Another type of criticality scenario involves the placement of fuel assemblies inside or outside a rack in a manner that does not conform to the intended geometry of fuel placement. For example, a fuel assembly might be dropped and come to rest in a horizontal position across the top of a rack, or in a vertical position between racks. The possible configurations of this kind are limited by the arrangement of the racks and the practice of moving fuel assemblies one at a time. Thus, this type of criticality scenario will also have a relatively low probability.

Appendix C Assessing the Probabilityand Consequencesof CriticalityEvents in Fuel Pools Page C-4 The remaining types of criticality scenario involve failures of administrative controls. One scenario involves the placement in a rack of fuel that is more reactive (e.g., with a higher enrichment in U-235) than the level specified for fresh fuel entering this facility. Facility licensees, and their contractors and vendors, seek to prevent such an event by employing administrative controls of a "one-time" variety. For example, the level of U-235 enrichment of a fresh fuel assembly will be verified at several points in the manufacturing process. Occurrence of a criticality would be attributable to failure of the one-time administrative controls either during fuel fabrication or fuel delivery. This type of criticality scenario will have a relatively low probability, because one-time administrative controls have a relatively low likelihood of failure. In other criticality scenarios that involve failures of administrative controls, the failed controls will generally be of the "ongoing" variety. In particular, if restrictions on fuel burnup/ enrichment or age, or the presence of soluble boron, are exploited as means of criticality suppression under normal conditions, the implementation of those means will rely upon ongoing administrative controls. Failure of those administrative controls could lead to criticality scenarios that involve the placement in a rack of fuel assemblies with inappropriate burnup/ enrichment or age, a reduction in the concentration of soluble boron in the pool water, or a combination of these occurrences. Over time, ongoing administrative controls will have a much higher cumulative probability of failure than one-time controls. Thus, criticality scenarios that involve fuel mispositioning (the placement in a rack of fuel assemblies with inappropriate burnup/ enrichment or age), a reduction in the concentration of soluble boron in the pool water, or a combination of these occurrences, will have a much higher probability than other criticality scenarios. In illustration, Orange County concludes from the historical record presented in Appendix B that fuel mispositioning is a likely event. 2.4 Determining the Envelope of Criticality in a Pool An important step in understanding the potential for criticality in a pool is to determine the range of conditions in which criticality will occur. The boundary of this range constitutes the envelope of criticality in the pool. A determination of the envelope is a necessary precursor to a systematic assessment of the probability of a criticality event, and must also precede an application of the Double Contingency Principle (as described in Draft Reg. Guide 1.13).

Appendix C Assessing the Probabilityand Consequences of CriticalityEvents in Fuel Pools Page C-5 To illustrate the concept of an envelope of criticality, consider the set of criticality scenarios that involve fuel mispositioning (the placement in a rack of fuel assemblies with inappropriate burnup/ enrichment or age), a reduction in the concentration of soluble boron in the pool water, or a combination of these occurrences. In order to determine the envelope of criticality for these scenarios, one would begin by specifying a particular pool and rack configuration, and the most reactive fuel assembly that could be placed in the pool (this may be a fresh fuel assembly). Next, one would identify the possible range of fuel mispositioning events. Then, one would determine the combinations of fuel mispositioning events and soluble boron concentrations that will yield a Keffective of exactly I (or, if a factor of safety is used, some lesser value of Keffective such as 0.95). The set of these combinations would be the envelope of criticality in the pool, for these scenarios. Discovery in this case suggests that no entity in the United States has undertaken the calculations necessary to determine the envelope of criticality in a fuel pool. During depositions of NRC Staff witness Dr Laurence Kopp and CP&L witness Dr Stanley Turner, Orange County's attorney asked these witnesses how they would determine the envelope of criticality in a fuel pool, as defined above. Both witnesses' responses indicated that neither the NRC Staff, CP&L nor CP&L's contractor Holtec has given significant attention to developing a thorough understanding of the potential for criticality scenarios of the type discussed here. 2.5 The Potential for Mispositioning of Fuel Appendix B reviews the record of fuel mispositioning at US nuclear power plants, drawing from documents that are currently available to Orange County. These documents almost certainly do not reveal the full historical record of relevant events, for reasons that are explained in Appendix B. Nevertheless, Appendix B shows that fuel mispositioning, involving placement in a fuel pool of one or more fuel assemblies with inappropriate burnup/ enrichment or age, is a likely occurrence. Most of the relevant events described in Appendix B directly involved the mispositioning of one or more fuel assemblies in a fuel pool. The other relevant events involved fuel handling errors that affected a reactor core, or fuel handling errors that occurred in a fuel pool but did not directly lead to a mispositioning of fuel. These other events are relevant because they show that ongoing administrative controls related to fuel handling and management are likely to

Appendix C Assessing the Probabilityand Consequences of CriticalityEvents in Fuel Pools Page C-6 fail. This information supports our finding that fuel mispositioning in a pool is a likely occurrence. The fuel mispositioning events described in Appendix B included events where more than one fuel assembly was mispositioned. Notably, at Oyster Creek, up to 184 fresh fuel assemblies were inappropriately stored in the spent fuel pool. Oyster Creek's safety analysis had not considered the possibility that fresh fuel would be stored in the pool. Some of the mispositioning events described in Appendix B involved only one fuel assembly but could have involved multiple assemblies, because these events were attributable to failures in administrative controls that governed many assemblies. 2.6 The Potential for a Reduced Concentration of Soluble Boron The concentration of soluble boron in the water in a fuel pool will be reduced if water with a lower concentration of soluble boron is added. At a typical PWR nuclear plant, the additional water could come from a variety of unborated water sources that interface with the fuel pool, including: the component cooling water system (which removes heat from the fuel pool heat exchangers); the demineralizer system (which is used to sluice and refill the demineralizer); the reactor makeup system (which provides makeup for evaporation losses in the fuel pool); the fire protection system; and the service water system.' In addition, where several fuel pools are interconnected but are separated by removable gates, as are the four pools at the Harris plant, water from one pool could mix with water from another pool if a gate is removed. If one pool has a lower concentration of soluble boron, the mixing process will reduce the concentration in the other pool. A similar effect could occur if a pool enters into communication with a fuel transfer canal or the reactor refuelling cavity. Other soluble boron dilution scenarios can be postulated or have occurred. In illustration, in July 1994 the soluble boron concentration in the McGuire Unit 1 pool was inadvertently reduced from 2,105 ppm to 1,957 ppm (a 7 percent reduction). This event is summarized in Appendix B. Unborated water that was used to decontaminate a drained fuel transfer canal was transferred by a submersible pump to the fuel pool. 1 Westinghouse Electric Corp, "Westinghouse Owners Group Evaluation of the Potential for Diluting PWR Spent Fuel Pools", WCAP-14181, July 1995, page 2-7.

Appendix C Assessing the Probabilityand Consequences of CriticalityEvents in Fuel Pools Page C-7 A study by the Westinghouse Corporation sought to estimate the probability of soluble boron dilution at PWR plants. 2 This study examined a generic, "composite" plant. It sought to estimate the probability of diluting the soluble boron concentration in the fuel pool from 2,200 ppm to 1,380 ppm (a 37 percent reduction), yielding a probability estimate of 3.8x10- 7 per reactor-year. The study did not summarize the historical record of relevant events, such as the July 1994 event at McGuire Unit 1. Nor did this study examine mixing among pools, transfer canals and the refuelling cavity in situations when these volumes have previously been separated by gates. In addition, this study was performed by an interested party (Westinghouse). According to the NRC Staff's expert, Dr. Laurence Kopp, the report was never reviewed by the NRC Staff, because the Staff considered that a generic study would not be very valuable in light of the great variation among nuclear plants with respect to such factors as the volume of water that can be inserted into a pool for dilution, the mode of inserting it, and the capacity of the pools. 3 Thus, the study's estimate of the probability of soluble boron dilution should be viewed as a lower bound, and not as a reliable estimate. 2.7 Criticality Calculations for Harris Pools C and D In its application for a license amendment to activate pools C and D at Harris, CP&L provided the results of some calculations related to criticality. 4 These results were not sufficient to support an assessment of the probability or consequences of a criticality event in pool C or pool D. However, additional calculations have subsequently been performed by CP&L and the NRC Staff, and these show the broad outline of the envelope of criticality for pools C and D, for scenarios involving fuel mispositioning and the dilution of soluble boron. The NRC Staff submitted a request for additional information (RAI) to CP&L on April 29, 1999. Question 1 of that RAI requested an analysis of a fuel mispositioning event in which one fresh PWR assembly is inappropriately placed in pool C or pool D at Harris. This placement would violate the burnup/enrichment restrictions which are specified in Figure 5.6.1 of the proposed new Harris Tech Specs. 2 WCAP-1418, Westinghouse Owners Group, Evaluation of the Potential for Diluting PWR Spent Fuel Pools (July 1995). 3 Deposition of Dr. Laurence I. Kopp, Tr. at 36-39. A copy of the relevant pages of Dr. Kopp's deposition is attached as Exhibit C-1. 4 See Revision 3, Enclosure 7 to CP&L's license amendment application.

Appendix C Assessing the Probabilityand Consequences of CriticalityEvents in Fuel Pools Page C-8 In its response of June 14,1999 to the RAI, CP&L asserted that a souble boron concentration of 400 ppm would be sufficient to maintain Keffective less than 0.95 if this mispositioning event occurred. No supporting calculations were provided. The results of some additional calculations relevant to the RAI were provided by CP&L in a letter of October 15, 1999, to which was attached a letter of October 11, 1999 from Holtec. These results were supported by a proprietary Holtec document which provided some details about the calculations. The proprietary document is not cited here. For the mispositioning event postulated in the April 29, 1999 RAI, CP&L's additional calculations showed that Kinfinite would be 0.9916 (with a 95%/95% confidence level) in the absence of soluble boron. 5 These calculations assumed the placement of one fresh PWR fuel assembly (enriched 5 wt% in U-235) surrounded by PWR fuel of the maximum reactivity permitted by Figure 5.6.1 of the proposed new Tech Specs. CP&L also calculated that the maximum Kinfinite would be 0.9352 if the soluble boron concentration were 400 ppm. Further calculations showed a maximum Kinfinite of 0.8671 (0.7783) for a soluble boron concentration of 1,000 (2,000) ppm. In a variant of its calculation that assumed an absence of soluble boron, CP&L assumed that the one fresh PWR assembly is placed in a PWR cell adjacent to the BWR storage racks. Assuming that this assembly is surrounded by PWR and BWR fuel of the maximum permitted reactivity, CP&L calculated that Kinfinite would be 0.9932 (with a 95%/95% confidence level). Some related calculations were performed by the NRC Staff, and were reported in an internal NRC Staff memorandum of November 5, 1999 from Tony Ulses to Ralph Caruso. 6 This document is hereafter described as the "Ulses Memorandum". The calculations assumed a fuel mispositioning event in which 5 A fuel pool can contain a relatively large array of fuel. Thus, the difference between Keffective and Kinfinite will be relatively small for many pool situations. As a result, the approach to criticality in a fuel pool is often discussed in terms of the value of Kinfinite. The discussion in this appendix largely follows that practice. 6 A copy of this document is provided herewith as Exhibit C-2.

Appendix C Assessing the Probabilityand Consequences of CriticalityEvents in Fuel Pools Page C-9 an entire PWR rack of the type proposed for Harris pools C and D is loaded with fresh PWR fuel assemblies enriched 5 wt% in U-235. The SCALE modular code system was used by the NRC Staff for these calculations, and the Ulses Memorandum compared the results of the SCALE calculations with the results of CP&L calculations. The Ulses Memorandum reported its results in terms of a neutron multiplication factor (designated hereafter as K), without discriminating between Kinfinite and Keffective. Assuming an absence of soluble boron, the SCALE calculations yielded a K of 1.19378. For the same problem, using the CASMO (MCNP) code, CP&L calculations were said by the Ulses memorandum to yield a K of 1.2076 (1.2056). These CP&L results appear to be the results presented for PWR racks in Table 4.5.1 of Revision 3 of Enclosure 7 to CP&L's license amendment application. In that table, the CASMO result is said to be Kinfinite, whereas the MCNP result is said to be Keffective. The MCNP result makes some relatively small allowances for uncertainty, bias and temperature variation. The Ulses memorandum also provided the results of calculations for a problem in which a PWR rack in Harris pool C or D is loaded with PWR fuel burned to 41,700 MW-days per tonne U, without the presence of any soluble boron. SCALE calculations yielded a K of 0.8940, while CASMO calculations by CP&L were said to yield a K of 0.9126. This CASMO result appears to be the result presented in Table 4.2.1 of Revision 3 of Enclosure 7 to CP&L's license amendment application. In that table, a Kinfinite of 0.9126 is reported as a CASMO result before allowances are made for uncertainities and the effect of axial burnup distribution. The above-presented results may be summarized in simple terms. Assuming an absence of soluble boron, consider three cases. First, a rack filled with well burned (42,000 MW-days per tonne U) PWR fuel will be clearly subcritical, with a Kinfinite of about 0.9. Second, a rack filled with PWR fuel of the highest permissible reactivity, plus one fresh PWR assembly, will be close to criticality, with a Kinfinite of about 0.99. Third, a rack filled with fresh PWR fuel will be clearly supercritical, with a Kinfinite of about 1.2. Now consider the presence of soluble boron in various concentrations, assuming a rack in which one fresh PWR fuel assembly is surrounded by PWR fuel of the highest permissible reactivity. A soluble boron concentration of 400 ppm will yield a Kinfinite of about 0.94, while a concentration of 1,000 ppm will yield a

Appendix C Assessing the Probabilityand Consequences of CriticalityEvents in Fuel Pools Page C-10 Kinfinite of about 0.87 and a concentration of 2,000 ppm will yield a Kinfinite of about 0.78. If these results are accepted, it follows that the envelope of criticality for PWR fuel in Harris pool C or D, for scenarios involving fuel mispositioning and soluble boron dilution, will involve the placement in a pool of two or more fuel assemblies with a reactivity that exceeds the permissible level. Also, it appears that the presence of soluble boron at a concentration of 2,000 ppm will preserve a subcritical margin of reactivity even if the racks are filled with fresh fuel. Thus, the envelope of criticality will be a set of circumstances which combine the mispositioning of two or more fuel assemblies with the presence of soluble boron in concentrations between zero and some level less than 2,000 ppm.

3. Nature and Consequences of a Criticality Event The major determinant of the consequences of a criticality event will be the cumulative energy release during the event. In turn, the cumulative energy release will be determined by several factors, including the rapidity with which a critical configuration is assembled, and the manner in which the system responds when fission energy is released.

Consider scenarios in which criticality occurs in Harris pool C or D as a result of the mispositioning of PWR fuel, combined with a reduced concentration of soluble boron. In such a scenario, the threshold of criticality could be crossed in either of two ways. First, the threshold could be crossed while a fuel assembly with greater-than-specified reactivity is being placed in a rack that is already close to criticality because of previous fuel mispositioning combined with a previously reduced concentration of soluble boron. Second, the threshold could be crossed while soluble boron concentration is declining in a pool that is already close to criticality because of previous fuel mispositioning. In both cases, the threshold of criticality would be crossed relatively slowly. However, the above-summarized calculations by CP&L and the NRC staff show that the final configuration could be critical on prompt neutrons alone. For example, CP&L finds that an almost-critical configuration exists (Kinfinite is 0.99) if one fresh PWR fuel assembly is present in a rack and soluble boron is absent. The completed placement of additional fresh assemblies in nearby locations could yield a Keffective of, for example, 1.01. That configuration would be critical on prompt neutrons alone, because the delayed neutron fraction for U-

Appendix C Assessing the Probabilityand Consequences of Criticality Events in Fuel Pools Page C-11 235 fission is 0.0065. The process of assembling such a configuration is discussed in later paragraphs of this Section. In a situation of prompt-neutron criticality, the rate of fission would rise rapidly. The time between each generation of fission in a chain reaction could be about 10-4 seconds, in which case 1,000 generations of fission would occur in 0.1 seconds and 5,000 generations would occur in 0.5 seconds. If a Keffective of 1.01 were achieved for prompt neutrons alone (i.e., a Keffective of 1.0165 for all neutrons), then one fission in the first generation would lead to 2.1x10 4 fissions at 0.1 seconds (during the 1,000th generation) and 4.0x10 21 fissions at 0.5 seconds (during the 5,000th generation). Since one fission of U-235 releases about 200 MeV (3.2x10-11 Joules) of energy, the 5,000th generation of fission would release about 130 billion Joules of energy. This energy release would occur over a period of about 10-4 seconds, and would involve the burning of about 1.6 grams of U 235. For comparison, note that fission in a typical commercial nuclear reactor with a thermal power capacity of 3,000 MW will release, when the reactor is at full power, 3 billion Joules of energy per second. 7 Clearly, a fuel pool criticality event of this kind would be self-limiting, and would not proceed to the point where 130 billion Joules of energy is released in one generation of fission. The reactivity coefficients of this system are negative. Notably, a substantial energy release would lead to local boiling of the pool water, which would reduce reactivity. A cyclic process might occur, involving repeated episodes of local boiling. If initiated, such a cycle could continue until terminated by depletion of fissile material in the fuel, evaporation of water, or the addition of soluble boron to the pool. Although a criticality event would be self-limiting, the energy release could be sufficient to damage the fuel. If damaged, the fuel could release radioactive material into the atmosphere of the pool building and from there to the external environment. Also, personnel in the pool building could be exposed to direct gamma and neutron radiation released during fission. Let us turn again to the initial phase of the criticality, which was briefly addressed in earlier paragraphs in this Section. For the scenarios assumed here, the threshold of criticality would be crossed relatively slowly, either during placement of a fuel assembly or during a decline in the concentration of soluble 7 For background on this paragraph and the preceding paragraph, see: Anthony V Nero, "A Guidebook to Nuclear Reactors", University of California Press, 1979.

Appendix C Assessing the Probabilityand Consequences of CriticalityEvents in Fuel Pools Page C-12 boron. An interval of time, lasting from seconds to minutes or longer, would occur between the crossing of the threshold and the attainment of the maximally reactive configuration. During that time interval, the reactivity of the system would initially rise but would then be constrained by feedback mechanisms. A cyclic process might occur, in which reactivity repeatedly rises and falls, with a continuing rise in the peak reactivity until the maximally reactive configuration is reached. An alternative possibility is that the criticality event might self terminate because the initial energy release destroys the critical configuration. For example, local boiling in a rack cell might expel a fuel assembly that is being lowered into the cell, thereby terminating the event. The entire process of a hypothesized criticality event could be systematically analyzed, using known techniques such as those employed by PRA practitioners. No such analysis has been performed to date, so there is no analytic basis to estimate the potential radioactive release to the environment or the radiation dose within the pool building. Our scoping calculations show, however, that substantial reserves of energy are available for release during a criticality event. Thus, significant onsite and offsite radiation exposures are potential outcomes of a criticality event.

4. Conclusions Criticality could occur in a fuel pool through various types of scenario. If criticality prevention relies solely on rack geometry and the presence of solid boron, some scenarios would involve the failure of administrative controls, but these controls would be of the one-time variety.

The exploitation of fuel burnup/ enrichment or age, or the presence of soluble boron, as additional means of criticality control introduces additional criticality scenarios. These additional scenarios involve fuel mispositioning or soluble boron dilution, or combinations of these occurrences. Fuel mispositioning or the dilution of soluble boron will occur as a result of the failure of ongoing administrative controls. The probability and consequences of a criticality event in a fuel pool could be systematically investigated, but this has not been done. From a qualitative perspective, it is clear that the scenarios which involve the failure of ongoing administrative controls have a much higher probability than the other scenarios.

Appendix C Assessing the Probabilityand Consequencesof Criticality Events in Fuel Pools Page C-13 Experience at US nuclear plants shows that fuel mispositioning, involving placement in a pool of one or more fuel assemblies with inappropriate burnup/enrichment or age, is a likely occurrence. Up to 184 fresh fuel assemblies have been inappropriately placed in a pool. Experience also shows that the concentration of soluble boron in a pool can fall below specified levels. A variety of scenarios could yield substantial reductions in soluble boron concentration. Calculations performed by CP&L and the NRC staff for Harris pools C and D show that supercritical configurations could occur if two or more fuel assemblies are mispositioned and the concentration of soluble boron is reduced. Some of these configurations would be critical for prompt neutrons alone, leading to the rapid release of potentially large amounts of energy. Significant onsite and offsite radiation exposures are potential outcomes of a criticality event.

EXHIBIT C- I Transcript of Deposition of Dr. Laurence I. Kopp, Pages 35-40 (November 4, 1999)

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of CAROLINA POWER & LIGHT Docket No. 50-400-LA COMPANY ASLBP No. 99-762-02-LA (Shearon Harris Nuclear Power Plant) DEPOSITION OF: LAURENCE I- KOPP, Ph.D. DATE: November 4, 1999 Commencing at 2:15 p.m. PLACE: Goya Conference Room Four Points Sheraton Hotel 37611 U.S. Highway 19 North Palm Harbor, Florida 34684 REPORTED BY: Dale E. DeFranco, RPR Notary Public State of Florida at large ORIGINAL

1 APPEARANCES: DIANE CURRAN, ESQUIRE Harmon, Curran, Spielberg & Eisenberg 1726 M Street Northwest Suite 600 Washington, D.C. 20036 SUSAN L. UTTAL, ESQUIRE Office of the General Counsel U.S Nuclear Regulatory Commission Washington, D.C. 20555 WILLIAM R. HOLLAWAY, Ph.D. ShawPittman 2300 N Street Northwest Washington, D.C. 20037 STANLEY E. TURNER, Ph.D. Holtec International Palm Harbor, Florida GORDON THOMPSON, Ph.D. Institute for Resource and Security Studies 27 Ellsworth Avenue Cambridge, Massachusetts 02139

35 1 that's why we decided this week to actually do a 2 calculation and see if would be true for Shearon Harris. 3 And we found we are subcritical for the entire rack. 4 Q. Okay. Under what circumstances, if any, and 5 under what regulatory requirements, if any, does the NRC 6 require the reporting of errors in controlling boron 7 concentration in the water of fuel storage pools? 8 A. I'm not sure if there would be any requirements 9 for reporting that. If the boron concentration were a 10 minimum boron concentration were in tec specs and if that 11 were violated during the surveillance interval, there 12 would be a certain amount of time where one could 13 reborate and get back up to the required minimum level. 14 And that would not be really I guess reportable unless 15 one did not borate in time. There's a certain interval 16 where you come back within regulations. 17 A. I see. And if you correct it with appropriate 18 intervals it's not a reportable event; is that what 19 you're saying? 20 A. Right. 21 Q. Okay. To the extent that boron dilution events 22 are reported to the NRC, does the NRC keep any 23 centralized record of boron dilution events that you 24 know? 25 A. It would be the same as the LER's for fuel

36 1 misplacements. There would be the LER's as far as I 2 know. We don't compile them but they're available. 3 Q. Has the NRC performed or obtained any analysis 4 or evaluation of nuclear power plant operator's 5 experience with controlling boron concentrations in fuel 6 storage pools? 7 A. Not that I know of. 8 MS. CURRAN: I'd like to ask the court reporter 9 to mark as Exhibit 10 an October 25th, 1996 letter 10 from Timothy E. Collins, Acting Chief, Reactor Ii System Branch, Division of System Safety and 12 Analysis, NRC, to Mr. Tom Green, Chairman 13 Westinghouse Owner's Group.

Subject:

Acceptance 14 for Referencing of Licensing Topical Report 15 WCAP-14416-P, Westinghouse Special Fuel Rack 16 Criticality Analysis Methodology. 17 Attached to this cover letter is a Safety 18 Evaluation by the Office of Nuclear Reactor 19 Regulation relating to Topical Report WCAP-14416-P. 20 (Whereupon, Exhibit Number 10 was 21 marked for identification.) 22 Q. Dr. Kopp, are you familiar with this document? 23 A. Yes, I am. 24 Q. If you would turn to page 10 -- actually page 25 10 is a continuation of a discussion that starts on page

37 1 8, Section 3.7 entitled Soluble Boron Credit Methodology; 2 isn't that correct? 3 A. Yes. 4 Q. If you look at the second full paragraph on 5 page 10 of the SER, I'd like to ask you about a sentence 6 that reads: "However, a boron dilution analysis will be 7 performed for each plant requesting soluble boron credit 8 to ensure that sufficient time is available to detect and 9 mitigate the dilution before the 0.95 k effective design 10 basis is exceeded and submitted to the NRC for review." 11 In parentheses, "Ref, dot, 29." 12 Can you explain to me what is meant by this 13 sentence and the reference to Ref 29? 14 A. Yes. This is the new methodology that I spoke 15 of earlier. This is one of the reasons for updating the 16 Grimes letter. This is a recent approval we gave for 17 crediting partial soluble boron in spent fuel pools. And 18 since we are allowing, not for Shearon Harris, but for 19 some reactors, credit for soluble boron under normal 20 conditions to meet .95, this would now require a new 21 accident to be evaluated which would be the boron 22 dilution event. 23 For other plants, such as Shearon Harris, which 24 do not take credit for soluble boron during normal 25 conditions, the fact that they calculate the five percent

38 2 1 subcriticality margin in pure water takes care of the 2 boron dilution event, that is complete dilution. 3 For these newer plants that want to take credit 4 for the new methodology. They still must show they are 5 subcritical with no boron, k effective is less than one, 6 but to meet the k arc criteria, k effective less than or 7 equal to .95, they can take credit for a certain amount 8 of soluble boron. So because of that we require them now 9 to do a boron solution analysis to show that they would 10 get them below .95 dilution event. 11 Q. Okay. But Reference 29 in parentheses, when I 12 turn to the back of this SER, Reference 29 is "Cassidy, 13 B., et. al., Westinghouse Owners Group Evaluation of the 14 Potential for Diluting PWR Spent Fuel Pools, WCAP-14181, 15 July 1995." 16 How does that Reference 29 relate to what we 17 were just reading on page 10? 18 A. That was a companion to this Westinghouse 19 report which requested credit for partial boron. In 20 order to prove that methodology I said they have to do a 21 boron dilution event analysis. And this other report 22 that you referenced shows how to do an analysis of a 23 boron dilution event in the PWR. 24 Q. So the reason for the mention of Reference 29 25 is that this is a way for licensees to do the boron

39 2 1 dilution analysis and that, that will meet NRC approval? 2 A. When they want credit for this methodology, 3 partial boron credit, yes. 4 Q. And has the NRC approved Reference 29 for that 5 purpose? 6 A. No. The approval of a boron dilution event we 7 decide is done on a case by case basis because the plans 8 vary so much. The amount of, the volume of water that 9 can be inserted into a pool for dilution varies from 10 plant to plant through the mode of inserting it, the 11 capacity of the pools vary. We decided a generic 12 dilution event would not be worth anything or worth much, 13 so we decided to, the people that wanted to accept this 14 methodology for partial boron credit would have to do a 15 plan specific for boron dilution analysis for their 16 specific spent fuel pool. That's why that boron dilution 17 event was never approved or accepted. It was a generic 1B type of topical report. 19 Q. Okay. 20 Q. Has the NRC performed or obtained any analysis 21 of the probability and/or consequences of potential 22 accidents resulting from improper boron concentration in 23 fuel storage pool water? 24 A. Only the analysis that shows that the zero PPM 25 of boron when there's still a five-percent subcritical

CERTIFICATE OF DEPONENT I, LAURENCE I. KOPP, do hereby certify that I have read the foregoing transcript of my deposition testimony and, with the exception of additions and corrections, if any, hereto, find it to be a true and accurate transcription thereof. DATE Sworn and subscribed to before me, this the __ day of - , 19 __ NOTARY PUBLIC IN AND FOR My commission expires:

ERRATA SHEET PLEASE ATTACH TO THE DEPOSITION. IN THE CASE OF: CASE # : q c---o,-z02. LA Please read the transcript of your deposition and make note of any errors in the transcription on this page. Do NOT mark on the transcript itself. Please sign and date the transcript on PAGE Please return both Errata Sheet and transcript to: PAGE LINE ERROR/AMIENDMENT REASON FOR CHANGE

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EXHIBIT C-2 Memorandum from Tony P. Ulses, NRC, To Ralph Caruso, NRC re: Completion of Criticality Assessment of Misloading Error in Harris C and D Spent fuel Pool (November 5, 1999)

November 5, 1999 MEMORANDUM TO: Ralph Caruso, Chief BWR Reactor Systems and Nuclear Performance Section Reactor Systems Branch Division of Systems Safety and Analysis FROM: Tony P. Ulses, Nuclear Engineer /s/ BWR Reactor Systems and Nuclear Performance Section Reactor Systems Branch Division of Systems Safety and Analysis

SUBJECT:

COMPLETION OF CRITICALITY ASSESSMENT OF MISLOADING ERROR IN HARRIS C AND D SPENT FUEL POOL I have completed the analysis evaluating the potential for criticality from a misloading error if Shearon Harris begins to use high density storage racks in the currently inactive C and D spent fuel pools. The analysis discussed in the enclosed report assumes a worst case misloading error in which the entire rack is misloaded with fresh 5 w/o enriched Westinghouse 15x15 fuel which has been previously determined to be the most reactive PWR fuel type which could be loaded into the Harris pools. This analysis demonstrates that the multiplication factor will remain less than one (i.e. subcritical) for this postulated worst case scenario. The calculated eigenvalues are taken at upper 95/95 level and a manufacturing uncertainty of 1 percent has been added to the predicted value.

Enclosures:

As stated DISTRIBUTION: File Center SRXB R/F GHolahan JWermiel RCaruso Aulses JStaundenmeier RLandry UShoop FEltawila DEbert DCartson SRB TU' TO; RC$ 0~ 11/CU/9 N/ DOCUMENT NAME: GSRXB\HARRISCRIT.WPD

Evaluation of Postulated Worst Case Misloading Error for Harris C and D Spent Fuel Pools Tony P. Ulses November 2, 1999

1 Introduction Carolina Power and Light (CP&L), the operator of the Shearon Harris nuclear power plant, requested a license amendment to activate the two unused spent fuel pools at the Harris site. The proposal is to use a "high density" storage configuration which requires the use of burnup credit racks. In the context of this report burnup credit racks refer to storage racks which require that the fuel has reached a pre-specified minimum burnup before it can be safely stored. The need for this bumup requirement is dictated by the fact that the inter-assembly spacing is reduced to achieve the desired "high density" configuration. Whenever one relies on a physical process such as burnup one needs to assess the impact of an assembly being inserted into the rack that has not reached the minimum acceptable burnup. Therefore, criticality analyses have been performed to assess the effect of an assembly misloading error in the Harris "C" or "D" spent fuel pool. In this analyses it was assumed that the entire rack was misloaded with UO fuel enriched to 5 w/o U2 35 which is the highest enrichment allowed at commercial power plant's in the US. This would be the worst possible configuration. 2 Definition of Problem In this analyses we will assess the impact of a worst case misloading accident by predicting the multiplication factor of the system. To this end, we will perform three base analyses and one sensitivity calculation. Two of the base analyses are intended to assess the staffs criticality calculations against the licensee calculations and the final analyses will assess the worst case misloading accident. The two comparative calculations are important because they will allow an assessment of the licensee method's and will serve to strengthen the staff's position with respect to these methods. A brief description of the problems will follow: Typical Parameters Fuel type: Westinghouse 15x15 Assembly Enriched to 5 w/o U235 Rack type: Holtec High Density Boundary Conditions: Reflective in x, y, and z

1. of Histories: 1000 groups of 3000 particles for a total of 3 million histories Problem 1 This problem is extracted from reference 1. The rack should be assumed to be loaded with fresh fuel without soluble boron. All dimensions should be nominal.

Problem 2 This problem is the licensing basis for the storage racks. The rack should be loaded with fuel burned to 41.7 Mw]KgU. The depletion is to be performed assuming three cycles of operation with an average boron concentration of 900 ppm, a specific power of 42 kW/KgU, nominal fuel and clad temperature and slightly higher than expected moderator temperature. The criticality analyses should assume no soluble boron is present and credit will be taken for actinides and fission products. All dimensions should be nominal.

Problem 3 This problem assesses the effect of the worst case misloading accident. The rack should be loaded with fresh fuel and one should assume that the soluble boron is present. All dimensions should be nominal. 3 Description of Methods The SCALE (ref. 2) system was chosen for both the criticality analyses and the bumup calculations. The SCALE system has been extensively assessed and validated for these types of calculations (refs. 3 - 5). The SAS2H sequence was used for the depletion calculations and the CSAS6 sequence was used for the criticality calculations. Both of these sequences use BONAMI and NITAWL-II to process cross sections into a problem specific AMPX working library. SAS2H uses XSDRN and ORIGEN to deplete the fuel and CSAS6 uses KENO-VI for criticality calculations. Both the 44 group and the 238 group ENDF/B-V based AMPX libraries were used in the criticality analyses and the 44 group AMPX library was used for depletion. 4 Presentation and Discussion of Results The results for problems 1 and 2 are presented in table 1. For comparative purposes, we have included the results from the licensee's contractor (ref. 1). This comparison reveals that the licensee method seems to predict slightly higher mulitplication factors (as much as 2% overall). However, given the differences in the methods the staff considers this to be excellent agreement and this gives us a great deal of confidence in the methods being used by both the staff and the licensee. Table 1 Comparison of Results for Problem I and Problem 2 CASMO MCNP SCALE' Problem 1 1.2076 1.2056 1.19378 Problem 2 0.9126 N/A 0.8940

'The SCALE results are the staff calculation.

The multiplication factor predicted for problem 3 is 0.978 at the upper 95/95 interval using the 44 group library and 0.979 using the 238 group library. The 238 group library was also used for this problem to ensure that collapsing spectrum used to generate the 44 group library from the 238 group library did not introduce any significant bias into the results. This demonstrates that even assuming the worst case misloading error (i.e. misloading an entire rack with fresh fuel) the rack will remain subcritical when one considers the soluble boron which will be present in the pool. In order to assess the adequacy of multiplication factors predicted using Monte Carlo methods it is prudent to consider, in addition to the number of histories tracked, how well the spatial and energy domains of the problem were sampled. To this end, we have attached the spectrum

output for the global unit from KENO-VI in Appendix A and prepared several spectral plots. The information from the major edit indicates that all of the parts of the problem have been sampled. Note that the flux for region I in the global unit is zero because region 1 represents the hole containing the fuel which was inserted into the global unit. The flux should be zero in the global unit for this region. The spectral plots are presented as Figures 1 and 2. The error bars represent one standard deviation and were extracted from the major edit (see Appendix A). From these plots we can ascertain that there are no unexpected trends in the results. For example, figure 1 shows a characteristic light water moderated reactor spectrum, but the thermal peak is smaller than it would be in the reactor. This reduction is caused by the additional absorption in the rack poison. Furthermore, we can see that we had complete coverage of the energy domain and that the sampling was significant enough to reduce the standard deviation to acceptable values. 5 Conclusions Analyses have been performed to assess the effect of the worst case misloading scenario in the Harris "C" and "D" spent fuel pool. This analysis demonstrates that the maximum possible multiplication factor in the "C" and "D" spent fuel pools is 0.98 assuming that one credits the soluble boron present in the pool coolant. It should be noted that this analysis does not consider manufacturing tolerances, but the multiplication factor bias from manufacturing uncertainties is typically not larger than 1%. The staff has also been able to confirm that the methods used by the licensee contractor yield results that are consistent with the staff's results. 6 References I. "Licensing Report for Expanding Storage Capacity in Harris Spent Fuel Pools C and D," HI-971760, Holtec International, May 26, 1998. (Holtec International Proprietary)

2. "SCALE 4.3, A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation," NUREG/CR-0200, Oak Ridge National Laboratory, 1995.

3. M.D. DeHart and S.M. Bowman, "Validation of the SCALE Broad Structure 44-Group ENDF/B-V Cross Section Library for use in Criticality Safety Analysis," NUREG/CR 6102, Oak Ridge National Laboratory, 1994.

4. O.W. Hermann, et. al., "Validation of the SCALE System for PWR Spent Fuel Isotopic Composition Analyses," ORNL/TM-12667, Oak Ridge National Laboratory, March 1995.

5. W. C. Jordan, et. al., "Validation of KENOV.a, Comparison with Critical Experiments" ORNLJCSD/TM-238, Oak Ridge National Laboratory, 1986.

Spectrum of W 15x 15 Fuel in Poisoned Rack KENO- VI Results 4.0 Error bars represent +/- I sigma 3.0 uncertainty N

          " 2.0 1.0 Energy (eV)

Figure 1 KENO Predicted Spectrum for W 15x 15 Fuel Assembly

Spectrum in Outer Boral Sheeting KENO VI Results 5.0 4.0 Error bars represent +I- 1 sigma uncertainty N3.0 2.0 1.0 0.0 104 6 10 10 1021 106 t0 Energy (eV) Figure 2 KENO Predicted Spectrum in Outer Boral Sheeting

Appendix A Excerpt from KENO-VI Major Edit

1 keno-vi input for storage cell calc. for holtec rack w/ Ofluxes for global unit 15x15 w region 1 region 2 region 3 region Ogroup flux percent 4 region 5 region 6 flux percent flux percent flux percent flux deviation deviation deviation p ercent flux percent deviation de viation 1 0. OOOE+00 0.00 1. 376E-04 5.36 8.973E-06 18.11 deviation 2 1.932E-05 19.59 1.823E-05 18.58 2.150E-05 12.41

0. OOOE+00 0.00 4. 190E-04 3.46 5. 856E-05 8.68 5.653E-05 8.31 4. 404E-05 9.44 4.438E-05 3 0. OOOE+00 0 .00 1. 267E-03 1.92 1.656E-04 5.23 8.41 4 1.544E-04 5.92 1.291E-04 5.35 1.475E-04 4.97

0. OOOE+00 0.00 4.204E-03 1.14 5. 437E-04 3.46 5. 072E-04 3.55 4.983E-04 3.51 4.957E-04 5 0. OOOE+00 0.00 2. 834E-03 1.37 3. 175E-04 3.97 2.91 3.393E-04 3.74 3.345E-04 4.01 3.336E-04 6 0. OOOE+00 0.00 8 . 974E-04 2.41 9. 913E-05 5.97 3.86

1. 171E-04 7.88 1. 041E-04 6.83 7 0. OOOE+00 0.00 3. 574E-03 1.33 4.377E-04 1.040E-04 6.43 3.59 4. 251E-04 3.72 3. 972E-04 4.34 4.402E-04 8 0.OOOE+00 0 .00 4. 386E-03 1.18 5. 304E-04 3.18 3.67 5.120E-04 3.19 4. 895E-04 9 0. OOOE+00 0.00 6. 307E-03 1.08 7. 926E-04 3.15 3.44 5.272E-04 3.33

7. 232E-04 3.15 6.767E-04 3.19 7.520E-04 10 0. OOOE+00 0 .00 1.103E-02 0.78 1. 355E-03 2.44 2.96

1. 291E-03 2.43 1.246E-03 2.54 1.271E-03 11 0. OOOE+00 0.00 1. 178E-02 0.74 1. 464E-03 2.26 1. 336E-03 2.40 12 0. OOOE+00 2.36 1. 340E-03 2.54 1.391E-03 2.39 0.00 7. 178E-03 0.92 8. 611E-04 3 .00 8 . 099E-04 2.97 7.276E-04 13 0.OOOE+00 0.00 1. 595E-03 1.71 3.03 7.950E-04 2.93

2. 171E-04 5.22 1. 834E-04 5.38 1.810E-04 14 0. OOOE+00 0.00 7. 130E-03 0.92 5.70 1.772E-04 5.20

8. 294E-04 2.75 7. 465E-04 2 .97 7.295E-04 15 0. OOOE+00 0.00 6.261E-03 3.31 8.029E-04 2.87 0.92 7.122E-04 2.72 6.722E-04 2.81 6.210E-04 2.98 16 0.OOOE+00 0.00 5. 505E-03 0.92 6.581E-04 2.97

5. 951E-04 2 .66 5.580E-04 2.90 5.222E-04 2.90 17 0.OOOE+00 0.00 3.273E-03 1.15 5.567E-04 2.73 3.484E-04 3.02 3 .119E-04 3.32 2. 897E-04 3.31 18 0. OOOE+00 0.00 2.444E-03 3.040E-04 3.06 1.36 2. 262E-04 3.42 2. 102E-04 3.45 2.065E-04 3.42 19 0.OOOE+00 0.00 4.374E-04 2.197E-04 3.25 2.80 3 .954E-05 7.46 3.452E-05 6.89 3.002E-05 7.71 3.751E-05 20 0. OOOE+00 0.00 5.568E-04 2.75 4.471E-05 8.08 6.49 4.308E-05 6.56 3.613E-05 6.99 4.229E-05 6.87 21 0 OOOE+00 0.00 4.168E-04 2.97 3.427E-05 6.93 2. 959E-05 22 7.91 3.034E-05 7.48 3.174E-05 7.49 0 OOOE+00 0.00 7.767E-04 2.22 5. 679E-05 5.29 6.221E-05 5.63 5.160E-05 5.67 5.866E-05 5.56 23 0. OOOE+00 0.00 8.810E-04 2.08 6. 311E-05 4.74 6. 017E-05 24 4.94 5.510E-05 4.97 6.113E-05 4.85

0. OOOE+00 0.00 9.433E-04 1.98 5. 403E-05 5.21 5.279E-05 25 5.27 5.165E-05 5.02 5.716E-05 4.83 0 OOOE+00 0.00 7.081E-04 2.24 4.488E-05 5.09 3. 932E-05 26 5.31 3.755E-05 5.45 3.815E-05 5.11

0. OOOE+00 0.00 6.778E-04 2.21 3.444E-05 5.51 3. 357E-05 27 0.OOOE+00 5.09 2.928E-05 5.60 3.339E-05 5.59 0.00 8.796E-05 4.92 4.091E-06 13.38 5. 436E-06 12.26 4. 366E-06 28 0. OOOE+00 0.00 15.39 4.106E-06 14.27 9.516E-05 5.12 6. 096E-06 12.92 4. 348E-06 14.18 3.879E-06 14.43 29 0. OOOE+00 0.00 1. 080E-04 4.50 4.893E-06 13.21

5. 983E-06 13.01 5.313E-06 15.04 5.081E-06 12.86 6.356E-06 11.50 30 0. OOE+00 0.00 2.454E-04 3.50 1.201E-05 8.66 1.019E-05 31 0. OOOE+00 11.42 1. 107E-05 8.96 1.019E-05 8.98 0.00 1.288E-04 3.78 5. 914E-06 11.15 6. 818E-06 13.22 5.718E-06 32 0. OOOE+00 0.00 12.02 5.699E-06 12.93

1. 513E-04 3.91 6.783E-06 10.57 6. 677E-06 10.90 6. 587E-06 33 0. OOOE+00 0.00 1.739E-04 11.27 7.080E-06 9.90 3 .52 6. 496E-06 9.67 6. 806E-06 9.95 7.721E-06 10.56 6.509E-06 34 0. OOOE+00 0.00 4.281E-04 10.88 2.47 1.835E-05 6.12 1.563E-05 6.45 1.648E-05 6.63 1.735E-05 35 0. OOOE+00 0.00 6.08

6. 916E-04 1 .90 2.472E-05 5.58 2 395E-05 5.19 2. 208E-05 5.36 2.412E-05 5.32 36 0. OOE+00 0.00 6. 888E-04 1.78 2 .515E-05 4.84 2.490E-05 5.24 2. 035E-05 6.30 2.428E-05 4.73 37 0. OOOE+00 0.00 5.795E-04 1.93 1. 896E-05 5.63 1.813E-05 38 5.20 1.732E-05 5.52 1.871E-05 5.04

0. OOOE+00 0.00 3.240E-04 2.18 1. 036E-05 7.29 8. 607E-06 7.85 39 0. OOOE+00 1.OOIE-05 7.77 9.824E-06 8.23 0.00 3 .261E-04 2.37 9.701E-06 8.07 8.411E-06 7.75 7.653E-06 7.96 40 0. OOOE+00 0.00 9.549E-06 7.42 1.468E-04 3.28 3. 917E-06 10.32 4.053E-06 10.54 3.024E-06 12.74 3.141E-06 41 0. OOOE+00 0.00 3. 566E-04 2.29 11.64

1. 058E-05 6.57 9.020E-06 7.10 8.8 58E-06 7.00 9.430E-06 6.96 42 0. OOOE+00 0.00 3.604E-05 5.88 1. 009E-06 27.70 8.304E-07 19.11 43 8.564E-07 25.72 8.251E-07 21.12

0. OOOE+00 0.00 3.968E-05 5.42 9. 087E-07 18.02 1 .018E-06 16.41 44 8.949E-07 30.82 6,629E-07 20.54

0. OOOE+00 0.00 6.744E-06 11.51 2. 102E-07 37.98 1. 685E-07 40.02 1.729E-08 70.77 2.139E-07 37.00

LIST OF EXHIBITS TO ORANGE COUNTY'S

SUMMARY

AND SWORN SUBMISSION REGARDING CONTENTION TC-2

1. Declaration of Dr. Gordon Thompson in Support of Orange County's Summary and Sworn Statement Regarding Contention TC-2 (January 4, 2000)

2. Letter from Brian K. Grimes of the NRC Staff to All Power Reactor Licensees (April 14, 1978)

3. Draft 1, Regulatory Guide 1.13, Revision 2, "Spent Fuel Storage Facility Design Basis (December 1981)

4. Memorandum from Laurence Kopp, NRC, to Timothy Collins, NRC, re: Guidance On The Regulatory Requirements For Criticality Analysis Of Fuel Storage At Light Water Reactor Power Plants (August 19, 1998)

5. Letter from Donna B. Alexander, CP&L, to U.S. NRC, enclosing response to April 29, 199, RAI (June 14, 1999)

6. Transcript of Deposition of Michael J. DeVoe, P.E. (October 20, 1999)

7. AEC Press Release entitled "AEC seeking public comment on proposed design criteria for nuclear power plant construction permits" (November 22, 1965)

8. Internal AEC memorandum from G.A. Arlotto to J.J. DiNunno and Robert H. Bryan (October 7, 1966), and attached Revised Draft of General Design Criteria for Nuclear Power Plant Construction Permits (October 6, 1966) (relevant excerpts)

9. Letter from J J DiNunno, AEC, to David Okrent, ACRS (October 25, 1966), and attached October 20, 1966 draft of General Design Criteria (relevant excerpts)

10. Letter from J. J. DiNunno, AEC, to Nunzio J. Palladino, ACRS (February 8, 1967),

and attached draft of General Design Criteria (relevant excerpts)

11. Note by the Secretary, W.B. McCool, to AEC Commissioners re: Proposed Amendment to 10 CFR 50: General Design Criteria for Nuclear Power Plant Construction Permits (June 16, 1967) (relevant excerpts)

12. Notice of proposed rulemaking for General Design Criteria, 32 Fed. Reg. 10,213 (July 11, 1967)

13. Letter from William B. Cottrell, ORNL, to H. L. Price, AEC (September 6, 1967) ane enclosed ORNL comments on proposed GDC.

2

14. Letter from Edson G. Case, AEC, to Dr. Stephen H. Hanauer, ACRS (July 23, 1969),

enclosing General Design Criteria for Nuclear Power Units (July 15, 1969) (relevant excerpts)

15. Memorandum from Edson G. Case, NRC, to Harold L. Price, et al., AEC, re:

Revised General Design Criteria (October 12, 1970), and enclosed letter from Edward A. Wiggin, AIF, to Edson G. Case, NRC (October 6, 1970)

16. Final Rule, General Design Criteria for Nuclear Power Plants, 36 Fed. Reg. 3,255 (February 20, 1971)

17. Letter from Donna B. Alexander, CP&L, to U.S. NRC (October 15, 199), enclosing letter from Scott H. Pellet, Holtec International, to Steven Edwards, CP&L (October 11, 1999)

CONTENTION TC-2: EXHIBIT 1 Declaration of Dr. Gordon Thompson in Support of Orange County's Summary and Sworn Statement Regarding Contention TC-2 (January 4, 2000)}}