ML003730645

Connecticut Coalition Against Millstone and Long Island Coalition Against Millstone Supplemental Response to Northeast Nuclear Energy Company'S First Request for Production

ML003730645
Person / Time
Site:	Millstone
Issue date:	05/30/2000
From:	Burton N Connecticut Coalition Against Millstone, Long Island Coalition Against Millstone
To:	Northeast Nuclear Energy Co (NNECO), Office of Nuclear Reactor Regulation
Cater C, SECY
References
+adjud/rulemjr200506, -RFPFR, 50-423-LA-3, ASLBP 00-771-01-LA, RAS 1861
Download: ML003730645 (234)

Text

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of:  : Docket No. 50-423-LA-3 Northeast Nuclear Energy Company (Millstone Nuclear Power Station,  : r.

Unit No. 3)  : ASLBP No. 00-771-01-LA CONNECTICUT COALITION AGAINST MILLSTONE AND LONG ISLAND COALITION AGAINST MILLSTONE SUPPLEMENTAL RESPONSE TO NORTHEAST NUCLEAR ENERGY COMPANY' S FIRST REQUEST FOR PRODUCTION The Connecticut Coalition Against Millstone ("CCAM") and Long Island Coalition Against Millstone ("CAM") (collectively, "Intervenors") herewith supplement their production of documents in response to the Northeast Nuclear Energy Company's First Request for Production, as follows:

Documents submitted by Orange County In the Matter of Carolina Power & Light (Shearon Harris Nuclear Power Plant), Docket No.

50-400-LA, ASLBP No. 99-762-02-LA, as follows:

(1) 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 Orevention Issues (Contention TC-2);

(2) Appendix B to the above Detailed Summary; and (3) Appendix C to the above Detailed Summary.

CONNECTICUT COALITION AGAINST MILLSTONE LONG ISLAND COALITION AGAINST MILLSTONE By:

S 0,35 By-0z

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of:  : Docket No. 50-423-LA-3 Northeast Nuclear Energy Company (Millstone Nuclear Power Station,  :

Unit No. 3)  : ASLBP No. 771-01-LA CERTIFICATE OF SERVICE I hereby certify that copies of "Connecticut Coalition Against Millstone and Long Island Coalition Against Millstone Supplemental Response to Northeast Nuclear Energy Company's First Request for Production" and the documents identified therein in the above-captioned proceeding have been served on the following by deposit in the United States Mail, first class, this 30th day of May, 2000.

David A. Repka, Esq. Charles Bechhoefer Winston & Strawn Chairman 1400 L Street NW Atomic Safety and Licensing Board Washington DC 20005 U.S. Nuclear Regulatory Commission Washington DC 20555 Office of the Secretary U.S. Nuclear Regulatory Commission Dr. Richard F. Cole Washington DC 20555 Administrative Judge Atomic Safety and Licensing Board Adjudicatory File U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Washington DC 20555-0001 Atomic Safety and Licensing Board Panel Dr. Charles N. Kelber Washington DC 20555 Administrative Judge Atomic Safety and Licensing Board Office of Commission U.S. Nuclear Regulatory Commission Appellate Adjudication Washington DC 20555-0001 U.S. Nuclear Regulatory Commission Washington DC 20555 Ann P. Hodgdon Office of General Counsel U.S. Nuclear Regulatory Commission Washington DC 20555 147 Cr s Highway Redd' Ridge CT 06876 Tel. F203-938-3952

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)

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

TABLE OF CONTENTS I. IN TR OD 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 AR GU MEN 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 dministrative 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 R egulation ..................................................................... 24

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

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

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

4. The Final R ule .................................................................... 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 SIO N ................................................................................................ 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 & I ight'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 applicaLion 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.2 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."3 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 batter utilize available storage space."5 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 I-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 burnup as a criticality preventi6n 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.' 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 not 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.12

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 fuei 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 Arn-241.

14 Burnup is the accumulated fission energy released by a fuel assembly. Its effects on criticality are exploited by restricting the combined bumup/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 burnup, 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 BWIR 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 PWRý 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." 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 burnup/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 bumup 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. 21 '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 See 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 burnup, 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 ("PWR") 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 bumup 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 cf 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 stanidard 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 burn-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 a fuel pool solely by use of a If a subcritical margin of reactivity is to be maintained in controls will be needed to ensure that the geometrically safe configuration, then administrative must be maintained during fuel racks provide the required configuration. That configuration or the drop of an object onto a normal operation and after specified insults, such as an earthquake but they will be applied on a one rack. The necessary administrative controls may be stringent, and installed, ongoing administrative time basis. After the fuel racks are designed, fabricated controls will not be required.

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

retain their needed to ensure that the Boral panels or other neutron-absorbing materials straightforward.

properties, but these inspections will be comparatively controls will require By contrast, prevention of criticality by ongoing administrative such as inputting information continuing actions by human beings to carry out these measures, These measures must be into a computer system, and operating and maintaining equipment.

For example, if the presence of carried out throughout the period when criticality is possible.

in a fuel pool, then soluble boron is to be exploited as a means of criticality suppression of soluble boron in the pool water administrative controls must ensure that the concentration must be implemented on a never falls below a specified level. These administrative controls must apply to an entire pool, continuous, ongoing basis, with complete reliability. The controls that pool.

and to canals or other pools that are interconnected with

23 Similarly, if restrictions on fuel bumup/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 burnup/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 burnup/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.

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

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

in promulgating GDC The rulemaking history of GDC 62 makes it even more clear that that criticality must be 62, the Commission intended to impose the fundamental requirement measures. Early in the controlled by physical rather than administrative or procedural language favoring rulemaking process, and in the proposed rule, the Commission considered In response to comments, physical systems or processes, but permitting procedural measures.

and established a clear however, the Commission removed the reference to procedural measures, In addition, while the General requirement that physical systems and processes must be used.

were promulgated in the Design Criteria were originally proposed as guidance, they ultimately form of minimum requirements.

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

following language relating to prevention of criticality in fuel criticality and to The fuel handling and storage facilities must be designed to prevent to the NRC.

26 The Atomic Energy Commission was the predecessor agency 7.

27 The Press Release and attached documents are attached as Exhibit

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.2" 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.V0 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"." The sLt 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 to 10 CFR 50. The potential for criticality in GDCs was described as a proposed amendment by draft Criterion 66, which stated:

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

notice of proposed Shortly thereafter, this language appeared in the Commission's Reg. 10,213 (July 11, 1967).32 Thus, rulemaking for the General Design Criteria, 32 Fed.

control, the concept of procedural throughout the early development of the GDC for criticality controls was included in the language of the criterion.

that they were "intended to be used The introduction to the General Design Criteria stated for a nuclear power plant." 32 Fed. Reg.

as guidance in establishing the principal design criteria at 10,215.

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

Laboratory (ORNL). The ORNL commented as 33 at the end of the first sentence, We do not understand the implication of 'or processes' procedural controls to prevent nor do we believe that it is practical to depend upon reactors. Hence, the last sentence of accidental criticality in storage facilities of power

'Such means as geometrically safe this criterion should be changed to read as follows:

Power Plant Construction Permits (June 16, 10 CFR 50: General Design Criteria for Nuclear B to the Note are attached as Exhibit 11.

1967). The Note and relevant excerpts from Appendix to this Summary as Exhibit 12.

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

27 34 configurations shall be used to insure that criticality cannot occur.'

the GDC, based on comments by the On July 15, 1969, the AEC prepared a set of revisions to cover letter, a major ACRS and the nuclear industry. As discussed in the accompanying was that the revised GDC difference between the proposed GDC and the revised GDC reactors, whereas the published criteria

"[e]establish "minimum requirements" for water-cooled GDC 62, entitled "Prevention of were "guidance" for all reactors. " The revised GDC included 3

Criticality in Fuel Storage and Handling:"

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

containing the identical On June 4, 1970, the AEC prepared another revision to the GDC, This revision was circulated to language of GDC 62 that had been prepared on July 15, 1969.

(AIF), a nuclear industry trade other members of the AEC and the Atomic Industrial Forum to other GDCs contained in organization." Although the AIF recommended substantial changes proposed alteration.

the revised draft, it accepted the new draft GDC 62 without any

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

(July 23, 1969),

35 Letter from Edson G. Case, AEC, to Dr. Stephen H. Hanauer, ACRS attached as Exhibit (July 15, 1969),

enclosing General Design Criteria for Nuclear Power Units 14.

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

Plants, 36 Fed. Reg. 3,255 (February 37 Final Rule, General Design Criteria for Nuclear Power as Exhibit 16.

20, 1971). A copy of the Federal Register notice is attached

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

shall be prevented by physical systems Criticality in the fuel storage and handling system safe configurations."

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

by the Commission on physical systems and processes, or Contradicted By Other D. The Plain Language of GDC 62 Is Not Altered Relevant NRC Criticality Standards.

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

systems and processes to be essential to preventing

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

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 39 safety assessments which concluded that the likelihood of criticality was negligible." 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 and the frequency with which it is controlled by procedures, the variety of forms of SNM criticality than at a nuclear power handled provides greater opportunity for an inadvertent reactor.

enriched to no greater than five At power reactor facilities with uranium fuel nominally cannot go critical without both a (5.0) percent by weight, the SNM in the fuel assemblies Further,thefresh fuel storage critical configuration and the presence of a moderator.

to prevent inadvertent criticality, array and the spentfuel pool are in most cases designed Inadvertent even in the presence of an optimal density of unborated moderator. of fuel on the number criticality during fuel handling is precluded by limitations addition, GeneralDesign assemblies permitted out of storage at the same time. In the prevention of Criterion(GDC) 62 in Appendix A to 10 CFR Part50 reinforces systems, processes, and safe criticalityin fuel storage and handlingthrough physical power reactor facilities occurs geometrical configuration. Moreover, fuel handling at considers a fuel-handling only under strict procedural control. Therefore, the NRC extremely unlikely. The NRC accidental criticality at a commercial nuclear plant to be CFR 70.24 are unnecessary as long believes the criticality monitoring requirements of 10 40 as design and administrativecontrols are maintained.

the language of GDC 62 which restricts Thus, in promulgating § 50.68, the Commission affirmed processes.

criticality prevention measures to physical systems and contains a list of measures for The language of § 50.68, as it was finally promulgated, maintaining a criticality monitoring criticality prevention that can be implemented in lieu of to procedures and administrative system.41 Although these provisions contain some references requirement of GDC 62 for physical measures, they do not undermine or contradict the general (b)(1) requires that:

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

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

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

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

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

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

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

to perform an accident analysis that Subsections (b)(2) and (b)(3) require the licensee if water accidentally enters the fresh fuel racks.

demonstrates criticality will be prevented, even that may be exempted from the accident analysis if it demonstrates one of two things:

A licensee 1998).

32 or that fresh fuel storage racks will not be flooding will be prevented by administrative measures, to prevent flooding, is in addition to the used. The first option, use of administrative measures in a place that is removed from the presence design features by which fresh fuel racks are located prevention measure, but as a of water. Thus, it cannot be viewed as a primary criticality features. If the second option is secondary measures used as a back-up to the primary design not used, i.e., that the fresh fuel is stored elected, the licensee must show that fresh fuel racks are meet the same criticality prevention in a fuel pool. If fresh fuel is stored in a pool, it must discussed below). Under these requirements as apply to spent fuel (see subsection (b)(4),

42 absence of soluble boron.

requirements, the fuel must remain subcritical, even in the or (b)(3) that is inconsistent with the Accordingly, there is nothing about subsections (b)(2) must be used to prevent criticality.

requirement of GDC 62 that physical systems and processes pools. Although this Subsection (b)(4) relates to the storage of fuel in spent fuel sense that it discusses the parameters for provision also mentions administrative measures in the water, the provision also makes it clear that taking credit for the presence of soluble boron in the administrative measures:

criticality ultimately must be prevented without resort to the spent fuel storage racks If no credit for soluble boron is taken, the k-effective of must not exceed 0.95, at a 95 loaded with fuel of the maximum fuel assembly reactivity with unborated water. If percent probability, 95 percent confidence level, if flooded spent fuel storage racks loaded credit is taken for soluble boron, the k-effective of the not exceed 0.95, at a 95 percent with fuel of the maximum fuel assembly reactivity must borated water, and the k-effective probability, 95 percent confidence level, if flooded with 95 percent confidep-e must remain below 1.0 (subcritical), at a 95 percent probability, level, if flooded with unborated water.

must be controlled (i.e.,

Thus, the basic requirement of subsection (b)(4) is that criticality the presence of soluble boron in the Keffective maintained below 1.0) without considering of fresh fuel in a pool should also 42 As discussed in note 23 above, arrangements for storage

33 43 water.

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

§ 50.68, or even mentioned.

2. 10 C.F.R. § 72.124 of criticality at The Commission has also promulgated regulations for control These regulations are inconsistent Independent Spent Fuel Storage Installations ("ISFSI's").

the use of physical systems or with GDC 62, because they do not unequivocally require standard. 10 C.F.R. § processes for criticality control, and instead apply a practicability 72.124(b) provides as follows:

of an ISFSI or MRS must be Methods-of criticality control. When practicable the design materials (poisons),

based on favorable geometry, permanently fixed neutron absorbing the design shall provide for or both. Where solid neutron absorbing materials are used, positive means to verify their continued efficacy.

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

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

ensure that the fuel remains subcritical in the presence of through (8), are not relevant to this 43 The other provisions of § 50.68, subsections (b)(5) proceeding.

34 plant or a preamble to the ISFSI regulations, an ISFSI is "not coupled to either a nuclear power 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 bumup 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, iý 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 I (January 21,

35 requirements for Monitored Retrievable promulgated in 1988, when the Commission added 4 The 1988 rulemaking fundamentally altered the Storage ("MRS") to the ISFSI regulations. "

control at ISFSI's, which had been promulgated Commission's existing regulation for criticality with the original set of ISFSI regulations in 1980.

explicitly and unequivocally required Section 72.73(b) of the original ISFSI regulations material - i.e., physical systems and the use of geometric spacing and/or fixed neutron-absorbing processes - for criticality control:

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

favorable geometry (spacing), permanently fixed are used, the design shall provide for or both. Where solid neutron absorbing materials In criticality design analyses for positive means to verify their continued efficacy.

for the neutron absorption of rack underwater storage systems, credit can be taken structures and the water within the storage unit.

of Spent Fuel in an Independent Spent Fuel Final rule, Licensing Requirements for the Storage (November 12, 1980).

Storage Installation, 45 Fed. Reg. 74,693, 74,710 amend the Part 72 regulations to On May 27, 1986, the Commission proposed to to "clarify matters that have arisen since part 72 encompass the licensing of MRS facilities and Licensing Requirements for the Independent was made effective on 11/28/80." Proposed Rule, Radioactive Waste, 51 Fed. Reg. 19,106. The Storage of Spent Nuclear Fuel and High-Level for methods of criticality control, Federal Register notice included the following provision

§ 72.93:

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

favorable geometry (spacing), permanently fixed can be taken for fixed neutron absorbing or both. In criticality design analyses, credit material present within the storage structure.

1987).

Independent Storage of Spent Nuclear Fuel and 45 Final Rule, Licensing Requirements for the 31,651 (August 19, 1988).

High-Level Radioactive Waste, 53 Fed. Reg.

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

require the use of favorable geometry and permanently governing methods for When the final rule was promulgated in 1988, the provision rule contain a mandatory requirement controlling criticality was transformed. No longer did the measures were called for only "if for favorable geometry and fixed poisons; instead, these the following "double practicable." The Commission had also added to § 72.124(a) 1986 proposed rule:

contingency" provision, not found in the 1980 rule or the systems must be designed to be Spent fuel handling, packaging, transfer, and storage accident is possible, maintained subcritical and to ensure that, before a nuclear criticality have occurred in sequential changes at least two unlikely, independent, and concurrent or 46 the conditions essential to nuclear criticality safety.

rule for this eleventh hour No justification can be found in the preamble to the final from the proposed rule. The only substitution of language that was so completely different mention of the changes is the following discussion:

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

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

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

19,124.

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

of the continued efficacy of solid neutron poisons the rule "to make it correspond to other Parts Instead, the Commission claimed to have changed review practice." The Commission did of the Commission's regulations and standard criticality with, and indeed none can be not identify what other regulations this new rule is consistent Nor did the Commission attempt to identified: this is a rationalization without substance.

justify it, or explain why the describe the alleged "standard criticality review practice,"

the change. By making such a major Commission failed to give public notice prior to making public notice or permitting public substantive change in the final rule, without first providing Act, which renders the rule comment, the Commission violated the Administrative Procedure invalid.47 by CP&L Would Violate E. The Administrative Criticality Prevention Proposed GDC 62.

the bumup/enrichment of As described above in Section III.B, CP&L proposes to restrict CP&L asserts that these PWR fuel in order to suppress criticality under normal conditions.

administrative controls" that will burnap/enrichment limits will be carried out through "strict 48 to Harris Pools C and D.

prevent an unacceptable assembly from being transferred controls to enforce This reliance on ongoing administrative procedures and F.2d 107, 135 (D.C. Cir. 1976);

47 See American Frozen FoodInstitute v. Train, 539 525, 533 (D.C. Cir.), cert. denied, 459 U.S.

Connecticut Light and Power Co. v. NRC, 673 F.2d F.2d 765, 771-72 (D.C. Cir. 1988), cert.

835 (1982); FloridaPower & Light Co. v. U.S., 846 ofAmerica v. FAA, 169 F.3d 1, 6-8 denied, 490 U.S. 1045 (1989); Air TransportAssociation (D.C. Cir. 1999).

38 and intent of GDC 62, which is to ensure that burnup/enrichment limits violates the language safe configuration of the assemblies, physicalsystems and processes, preferably geometrically relies on the presence of soluble boron to are used to control criticality. Similarly, CP&L This violates the plain meaning and intent of GDC prevent criticality under accident conditions.

of soluble boron in the spent fuel pools require 62, because the introduction and maintenance and do not constitute physical systems or ongoing administrative actions and procedures, processes.49 Criticality Prevention F. CP&L's Proposed Reliance on Administrative 1.13 or Other NRC Staff Measures Is Not Justified by Draft Reg. Guide Guidance.

CP&L and the NRC Staff argued that In opposing the admissibility of Contention TC-2, to prevent criticality is permitted by Draft its reliance on cbntrol of bumup/enrichment levels generally that "if there is conformance with Reg. Guide 1.13. The Commission has stated guides, there is likely to be compliance with the GDC." Petitionfor Emergency and regulatory As the Board has recognized, however, this Remedial Action, CLI-78-6, 7 NRC 400, 406 (1978).

regulatory guides necessarily govern." LBP-99 is "not a blanketendorsement of the notion that between a regulation and a regulatory guide, the 25, 50 NRC at 35. Where there is inconsistency 7 Rev. 3 at 4-17.

48 License Amendment Application, Enclosure the presence of soluble boron during an accident.

49 In one criticality analysis, CP&L relied on 5). In a subsequent response to the same See CP&L's June 14, 1999, RAI Response (Exhibit that if defined as Kinfinite less than 1, RAI, CP&L stated that a new criticality analysis shows in the presence of one mispositioned fresh subcriticality can be maintained in unborated water, attached Alexander to U.S. NRC (October 15, 1999),

PWR fuel assembly. Letter from Donna B.

of 400 ppm was found necessary to as Exhibit 17. However, a soluble boron concentration in Section "maintain Kinfinite less than the regultory limit of 0.95." Id. As discussed below an the consideration of mispositioning of a single fresh fuel assembly does not constitute IV.F, and to meet the regulatory limit of 0.95 for adequate criticality analysis. For this reason, reliance on the presence of soluble boron Kinfinite, it is necessary consider whether CP&L's GDC 62.

under abnormal conditions is consistent with

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 LouisianaEnergy Services (Claiborne 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 Staffs 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 of Double Contingency Principle to a Single Contingency Principle. Despite the many years 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." 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, a.5 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 by a 50 Orange County is aware of only one generic study of boron dilution, which was done the historical self-interested party, the Westinghouse Corporation, and which failed to summarize

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 proce.ses 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 bumup 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 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.

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."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 mispositionihg 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 the regulatory limit of 0.95.

D would, in the absence of soluble boron, cause Keffective to exceed 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.

CP&L has In considering the role of fuel mispositioning as a potential cause of criticality, Underlying this restricted its attention to the mispositioning of only one PWR fuel assembly.

to the mispostioning of restriction is an assumption that a single failure or violation will lead in Appendix C, only one fuel assembly. In fact, as demonstrated in Appendix B and discussed fuel assemblies.

experience show's that a single error can lead to the mispositioning of multiple control, In addition to its improper reliance on administrative measures for criticality discussed above has CP&L's misapplication of the Double Contingency Principle in the manner a reasonable yielded a criticality analysis that is non-conservative and inadequate to provide accident. Whether or assurance that public health and safety will be protected in the event of an Board as not the administrative measures chosen by CP&L are approved by the Licensing accident analysis must consistent with GDC, CP&L's methodology for performing its criticality 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 Harris must License Amendment Application for the expansion of spent fuel storage capacity at be rejected as inconsistent with GDC 62 and valid and applicable NRC Staff guidance.

a Moreover, CP&L's criticality prevention measures are demonstrably insufficient to provide 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 and should find that Orange County has raised material and substantial issues of law and fact, 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 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))'

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.

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

Appendix A Some Incidents Relevant to the Potentialfor Criticalityin 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 1 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 misp6ýitioning. 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 diaa 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, thý 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 Criticality in 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 nmis-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 Criticality in 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 I 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))15 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 i-his 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 Criticality in 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 Criticalityin 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 samplinlg 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)

14*) .

ESTIMATED BURDEN PER RES"ONSE TO COMPLY WITH TI US INFORMATIO COULECT1ON REQUEST. 00 HIS. REPORTED LESSONS LEARNED ARE INCORPORATED INTO THE LICENSING LICENSEE EVENT REPORT (LER) PROCESS AND FED BACK TO INDUSTRY FORWARD COMMEWS REOoA.WINO BURDEN ESTIMATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANC1IJ (MNBB 7714). U.S. NUCLEAR REGUIATORY COMMMSION. WASNINUTON. DC 20555001. AND TOTM PAPERWORK REDUCTION PROJECT SF*ia-NAME(I)

N.A._.*. BraiLdwood Unit t 2OKETý-NUN'X' 0500456 1PAGt of of (3)

B in Spect Fud Rocks and a aiticality -- aysis modeling 1@. 1 I Fnilre Comply With inlitp Basis Due to Degrdaion of Boamil V 1 nACnLtTYNAMcImrrY~vE FiT 22 96 96 010 02 09 II 99 Ih,iurnn,* llanl 2 o300045 1 21 H-W-ood Uoit 2

1. -- 1I rACiE

-I nrv NAME [IX).KE[ NUIIBER Bvron Unit I and 2 I I I I I I I I _____________________ I I -

I I -- I *rO 1*R ,REM[ENTS Tf OF If CFR 5: (C-e-- ow---*of t)i*1 THIS REPORT IS S JBum1TW PURSUANT I

-- 10 2.2203(aX3XI) &3"(SX2XWA) 13 IIEX) 02Qa3i SSO7~X2iv) 7.1c 20M O20.fl03(iNX4) S.1S(.N2Xv)OTE 50.4CX) .

30730(a2XvWWA) (SpedCy in S..... S0liUMYI) .5O* , m2XvWX 50j3(XVHIB) below adimaT¢ .

Xv) 5O.7T3(U2XM) S0.730(20X,) NAC Frm 366A

, -W OErAJI*," 3fWO TNIS L M 112b

%tIfI I.

MA.PI*K flnchi& Area ( o'k)

ANAML  !(815) 45872801. Etension 20 8

3. schrIwswtUn. Licensing Engineer l

• gei* *r 7lTWIm,,-AI ll*tOIT 11,311

-COMPLETEE CM= svso L5UMPON.' lOKlf TO

X Da DBoraflex B959 N Panel StUIDMIIISIO X I Nu DATE (15) f -- EXPECTED SUBMISSION DATE)

Am,1 to 140o . L.. Approxiamwy Ahml skwk . -sp, rypcwi lift_ 16) received on Analysis of Neutron Attenuation test data for Braidwood's Spent Fuel Racks The largest gap has a width of greater than 6/21/96, shows Boraflex shrinkage and gaps. that assumed panel exceeds four inches. A gap of greater than four inches in any Boraflex a in the current criticality analysis. The spent fuel storage racks are designed to maintain The cause of this event was determined to be KofE S 0.95 when flooded with unborated water.

failure of the Boraflex due to deterioration as a result of improper material selection.

(SFP) boron concentration and silica Corrective actions include controls on Spent Fuel Pool that there is concentration. The safety analysis contained in this report concludes maintain a Keff S 0.95.

reasonable assurance that the Braidwood SFP will criticality analysis CAC-96-248, "Byron 1 3/25/97, a modeling deficiency was identified in Analysis with Credit for Soluble Boron", dated and Braidwood Spent Fuel Rack Criticality on all four faces October 31, 1996. This analysis assumed Boral poison plates were located model did not reflect the actual (as of all Region 1 storage cells. The criticality which are located on the interior designed) configuration of the Boral poison plates, new Region 1 fuel storage racks but are not present on the periphery of the portions of the Subsequent to the discovery of this modeling deficiency.

Region I storage cells.

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

7- Pm M" VALN'lrL#CAR f-rELA*TO-Y COMUION APPROVED BY OMN -NO )1364104 EATWATED DURDL MR.RESPONSE TO COMPLY. Wrfh THIS IWyRMATKINae COLLECTION REOMET 500H13 REPORTD LCENSEE EVENT REPORT (LER) LESSON$ LEUIE, BKgTO P1aOCESSA.%'DFED ARE MROMMATED 1,-To THE UCESING i,.sND TY FORWARD COUMtEN-M TEXT CONT ATION REOARDG BIRDEN ESThMATE TO THE IRF3RMATW%. -ND RECOR1IMA&NAGEPL%-7 BRANCH (r.6 F33jL. %*NLCUAR "RUOLIATORY CO1MLS5ION. WASHINGTON. C 205,M,..-AND I THE PAPERWORK MEDUCTION PROJECT VAUALTY ?EAM9 (1) DOCKET MUMBER (2) LsILMuR (6) Pi (3)

SBaidwood Unit 1 05000456 96 1 010 I 02  : of 18 tif Mm jp, is rapilied. arn addiiona copies of NRC Form 366AX(17)

A. VLMTTCCZTZCWI MI*O TO ZrXVT:

Unit(s): 1 Event Date: 08/21/96 Event Time: 1224 Hours Reactor Mode(s): 1 Power Level(s): 1001 RCS [(BI Teinp./Press. NOT I tiOP Unit(s)$ 1 Event Date: 03/25/97 Event Time: 1700 Hours Reactor Modets): 1 Power Level(s): 100% RCS [AB) Temp./Press. NOT I MOP

a. DhCRIPTlIW OF KVZZT:

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

On 8/21/96, 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 (SFP) 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 1987, ComEd first identified gamma-radiation induced damage to the Boraflex polymer.

The 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 tj 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," Westinghouse, June 1994, supplemental criticality analysis. Sufficient margin exists within this supplemental criticality analysis to accommodate the anticipated levels of cracking and gapping associated with the first stage of 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 ir 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, c-e stronger the local flow dnd thus the dissolution rate increases.

The recent Blackness Testing campaigns at Byron and Braidwood indicate progress int:

the second stage of damage has occurred, and that the maximum gap width allowed in the

mm. M#.CLAB RATOST COSOG3UOW A9I3Onr[ By 0oW.%Q 315s104

[E.s'MAT*E P.L'M4 PMR RErSPONSE To COmPLY ITrH TKIS 1NF0R3AATlON COUIECflrMzgiw 500 HIS REPMRTED LICENSEE EVENT REPORT (LER) MXMA.ss *1) WCSORP.A.DA~u LLS D AC* T0L%'DrW hCDMM FMewARD UCLUsI YTOn7oTHE C1AM.,Ts TEXT CO)NTI]NATIC4 RARDIG KIM.4 hSThMTE TO THE LFoL%%T% i.D WRMco MtgdW13WT At.'CH (T4 F0 iL S %tCtl.%n Mt'LOATtCC0IS0 WAZ 0GTM.Dc 2O5-. AND THE PAPER WORK JEIRXTMO rwEcT1 FACM I AMS(I) OCK"I .'lk) 3. WMU If Braidwood Unit 1 05000456 96 010 1 02 3 Ie f is (urnVs mis reqld s addkaia copusdNRC Faorm 366AX17)

Based on the above facts, Bralioodd already has large numbers of storage locations in the second stage of degradation. The degradation mechanism associated with the second stage proceeds slowly, however it iS both difficult to predict ansd measure the extent of damage. Although Blackness Testing Is useful for measuring cracks, gaps. and rastage, It does not measure as overall reductlon in boron density. Therefore Blackness Testing provides incomplete information regarding the current state of a given storage location. An approwed methodology to measure boron spatial density does not currently exist for PURs. Therefore, the gaps recently found at Braidwood Station may not represent the full extent of Boraflex degradation.

When assessing the current state of the storage racks, the following factors, aLosn with others, are consideredt they Include the slow nature of the degradation process, the continued presence of some Boraflex, the succossful perfomance of the surveiLLance coupon program, the inclusion of Docal in the Region I rack design, and the potential for additional reactivity margins due to burn-up.

COMEd has perform calculation'to support a short term recoendation Of m0intalning greater than 2000 PI1? soluble boron in the Spent ruel Pool tocompensate for the degradation of the Boraflax. These calculations are -very.conservative. The 2000 Pip

.limit is intended to approxLmate the total reactivity suppression worth oe the installed. Boraflex in both the Region L and Region 2. fuel storage racks. Therefore,

@ven if all soraflex were to be removed from the Spent ruel Racks, the 2000 PPM volue Is adequate to maintain the Spent I~el Pool at 1 0.95 Keff.

based on the recent Blackness Test data, It cannot be stated with certainty that Technical Specification b.6.-.1 is met. This specification states, -The spent tueL storage racks are desLgned and shall be MaintaLned with a Keff S 0.95 when flooded with unborated vater ...... Therefore, the racks are in an "lndetermanate" state of operability as defined in MRC GmsorIc Letter 91-18, they have been conservetiLvily declared inoperable, and comipensatory measures that were Initiated in 1995 were verified.

ThLs event is being reported pursuant to IOCRSO.7JIa) t2) (Li} (5) any event or condition that resulted in the condition of the nuclear power plant being in a condition that was outside the design basis of the plant On 03/25/9, additional reviews by ComJd Identified a mode*Lnq defi*cLn-:y in criticality analysis CAC-96-249, 'Byron and Braidvood Spent Fuel Rack :r.iL:a&L-:y Analysis with Credit for Soluble Boron', datrd October 31, L996. This analysis was performed to support Technical Specification Amendnent No. 06 for Byron Uni -- L Z A nd Amencdment No. 78 for Braidwood Units 1 and 2. issued April 2, 1997. This ".ttPem" critLcaLit- analysis was performed due to the degradation of the Borafle Ln h*_h spent fuel racks. The deficiency is due to inadequate model.inq of rhe physi:aL :z'i:..;or of the Bora! panels within the Byron and Braidwood Reqion L Fue, St3r*ree Ra-ts.

Due to ComEd's con,:erns regarding :he industry's experiences w-ith lore!.** Ie::t-r during the mid-90's, Borel panels were placed in the tl-jx raos )f! - . e-.-r' during initial fabrication. The B~oral panels were Lrior*.'pd exist between each cell withkn a Re*ion 1 ,.k , ." / ti. .i. > 4--"

included in the assumptions of the -wod'el for the Q.qt )n r a1Ici

U -}L REGUIATOK CtNUMM AsvuasmaysiWCKAXGf'.. uaes0

-, ISTDAflD BL/D* I1E8 TO1COMWLY MTH TMH ORMWATION COLLWICK 7EST 5OHRS U3FORTED LICENSEE EVENT REPORT (LER) KMoEYD:UBMz5DCONMon" TEXT CONTINUATION EA5.DIM 5.*UMey ESIDIATE TmHE MO"m.10 V.,D RECOMM 14ANAGUt. i*MA, (T-46)k LI .*UtCLEAR IURIATMY COM3OMLWASlL TON. DC 2015%4001. &%1 THE PAPRWOMK MtlD M MOM=CT vi..a u.*u sai, E.rarld5ood Unit 1 05000456 96 oM10o 02 4 or 19 (If~~ ,rn1iu~mm, ma~dakj copus .NRC Fan 366AX17)

- The array is Infinite in lateral Ix and yl extent. There Is no interface requirements between Region I storage racks.

- Boral poison plates were on all four faces of all storage cells.

The criticality model did not reflect the actual Ia" designed) configuration of the loral plates. which are located on the Interior portions of the new Region I fuel storage cocks, but were not designed to be installed on the peripbery of the Region I storage cells. Thus, Region I periphery storage cells actually contain Doral plates on only three sidevo and the four corner cells actually Contain only two Interior boral plates.

In march 1997, the Westinghouse criticality engineer was reviewing the S3? rack perLphesal geometry in an attempt to regain storage locations that were lost due to constraints required In the 1996 analysis. Drawings of the Byron &ad Braidwood 5rP and tacks were supplied to Westinghouse at their request. On Match 20, the Westinghouse criticality engineer contacted Coud Nuclear Fuel ServiceS (NrS) math a concern that there may not be Boral on the peripheral walls of the Region 1 racks due to the girdle bar geometry. As a result of subsequent Coomrd reviews of rack drawxng and discussions with ,the vendorsi responsible for .const-tuction and seismic analysis of- the racks, Comd -"

concluded that moral poison plates were neither present on the periphery of the Reqion 1 storage cells not designed to -be in these locations.

C. Ca-M Ova"m The cause of this event was determined to be failure of the Soraflex due to deterioration as a result of Improper material selection In 1907, ColAd first identified gamma radiation-induced damage to the Doraflex polymer.

That damage pogrestoses through two stages. rirst, the Doraflex cracks and shrinks, producing crack& and gaps. Second, after the polymer has sustained significant dasmage, the Botaflex becomes brittle and is 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 ,.n vendor performed criticality analyses, and inadequate reviews of the analyses input and assumptions against manufacturing drawings during the criticality analyses revLews and verifications. The infinite array criticality analysis methodology wes not appropriate for the unique placement of Docal in the Reqgon I racks, and did not properly model the i.nterface between Region I racks.

loral panels were placed in the flux traps of the Region 1 racks durinq initial fabcication. The manufacturing drawings for the Region I racks are xnadeqcate to determine Boral panel placement, and no as-built drawings of the wqion I racks were ever generated. The original criticality analysis for the ReqLon I racks was perforvmd in 1987 by a vendor. The criticality "model", originally generated by the vendor, was transferred to Westinqhouse, and was carried forward through all subsequent :Lrticality analyses. The original analyLsi and all subsequent analyses assu.med f ar Reqion ". an infinite array and Boral poison plates on all four fa:es o*f all. -els. Cr:iza'::y analysis :AC-96-248 assumed that the boraflex poiz--" was mve Ev[ :frm -,e stror.&T.

racks, and that the Boraflex was replaced with water Thw analys.1 4'3 -rjo'j.e*J --.

Region I storage cells with dorel panels on all four f6:.s. Th.s &:i, prevL7-4s sna ysP&

did not specifically model the peripheral Reqgon I ce .i that io nz: have 3Dra. pane.1 on their exterior faces.

I

(4.95) K~lES S413a EMATED .. NMPIER RESPONSE TO COMMY WITH T1lls D1VOIJIATKON COLLECTION RVEQLsT 300 HRS R[?ORThD LESOS LEARE ADED AR ACKRCOToDliE "nATD M70TTilt LICL%SING LICENSEE EVENT REPORT (LER) POS0s PROCESS AND FED BACK TO O",MDMTY FORWAD C(M\EN-rS TEXT CONTDNUATION REGARDLN BULz uAmTE To THE RF0M,%ATIo% %D RECORMO MA.AGEMENTBRA..%*(Ti1 73)1.5 "tCLE.%

REO.LATOIR COuMMIsM.WASINGTON. DC 20!5.oo0o ,. "ND THE PAPERWORK REDITIO PROJECT FAC Y NAME (3) DOCKZT NUMBE a) LE NL*MDM* (6) F.UGE t Braidwood Unit 1 05000456 96 010 02 5 o.e of on spun is .qoin, m addiama cops of NRC FOM 366AX 17)

D. AIIU I oV SuurlTI

• Recent slackness Testing indicates that the degradation of the Braidwood Spent Fuel Packs exceeds that assumed in the criticality analysis. This could lead to a condition where the Technical Specification reactivity limits for the SrP could be exceeded.

Based on a comparison witb prior analyses by Coond NucLear FUel Services for the Byron/Breaidwood reactor care&, maintaining SrP boron concentration >2000 PPM will ensure that the requiremmts for maximm reactivity in the SrP are met, even assuming the Boraflex panels are Ineffective from a reactivity mitigation standpoint. The analysis assumed enriched fuel with no burnup (e.g. maximMm reactivity) in close proximity to other assemblies. The physical separation of assemblies in the Spent Fuel Racks is greater than the separation in the core. In addition, the spent fuel assemblies are at much lower ro'actLvity due to burnup from incore operation. For these reasons, there is reasonable assurance that the Sraidwood Spent Fuel Pool maintains a Keff.* 0.95.

After discovery of the modeling deficiency on 3/25/97, supplemental criticality, analyses for the actual Reqion I cell 4oral geosaatries waere performed and demonstrated that, with administrative controls in place regarding boron concentration and- fuel placement in Region I rack interface, acceptance criteria for spent fuel storage is met. The supplemental criticality analyses utilized the same assumptions, codes, procedures, and uncertainties used to support the 1996 criticaliLty analysis (CAC-96 248) but with Boral panels located only in the Interior cell-to-cell interfaces. The supplemental analyses modeled the fo*llowing Region I rack geometries:

I. Corner cell of rack facing two concrete walls.

2. Periphoeal cell of rack facing one concrete wall.

3. Empty row of cells facing a full row of cells across & Region I to Region 1 rack interface.

4. Checkerboard pattern of cells across a Region 1 to Region I rack interface.

Calculations were performed for the four rack geometries to verify that with a maximum nominal enrichment of U-235, that Keff is less tham 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 reactlvities were compared to the all cell Keff calculated in section 3.2.1 of CAC-96-248. The all ceaL Keff (from CAC-96-249) was verified to be greater than the reactivities cal:ulated fo: these four rack geometries. The biases and uncertainties calculated in CAC-96-249 remain valid for use with the four analyzed rack geometries.

By determining that the &l- cell Keff remains bounding, the conclusions of CAC-96-249 are applicable for the four analyzed rack geometries analyzed for the following acceptance criteria:

I. Assuminq no sol.Lble boron, the maximum nominal enrichment o! U-235 could be stored and a Keof of less than 1.0 is maintained,

2. Taking credit f'ra minimum concentration of soluble boron of 2000 ppm, a Keft of less than or equal to 0.95 is maintained, and

3. Assuming the SF? water temperature postulated aczidefnt and Zak.nq .:reiLt for a minimum :oncentration of soluble boron of 2100, ppm, a Kef o ,ss than or equal t, 0.95 is maintained.

Additional cases for the m=sloaded assembly were performed. these :ases were calculated at no soluble bearon conditions and the resultinq reaCt1VLt1* ýw~r Showy -,

rFOM (4-w) M" u.a NXUCZAR RO1VATORY CKOhMIMl APPROVED BY3 WIR~s OM wwX*

s

.%335.1" ESTMATED BLRD'EIJ PER RESPONSE TO COI4PLY WITH Tills INFORMATION COLLECTION REQLEST 50 0 HRS REPORTED LICENSEE EVENT REPORT (LER) LSON PROCESS AN*D , FED'BACK ITO D TO IDUSLTRY FORWARD CO.%EI.'T%-S TEXT CONTINUATION REGARDING BL7RDN ESTIMATE TO THE INFORMAT1O'X.. ND RECORDS MANAGEMENT BRANCH (14 F) . Us " .IC.LAX REGULATORY COMMISSION. WASHINGTON. DC 205ý55-00 1 ANE THE PAPERWORK REDUCTION PROJECT FACILITY MRAW(I) DOCKE- l4Jmc(2) LERIIM3E(6) PArG 3)

YW& UQLV4ifLA "V15Uo%

Braidwood Unit 1 05000456 96 010 02 6 of to (If mOM spa is neund, wn addiioo*ma wps ofNRC Form 366AX 17) be loss than the a11 call Keff from CAC-96-248. The dropped assembly accidents are not affected by the foral configuration.

An additional criticality analysis was performed taking credit for 2000 ppm soluble boron and no Boral 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 criticality analyses ace conservative since, in reality, an appreciable amount of Bocaflex remains in place in addition to the administrative requirement to maintain at least 2000 ppm in the SFP. It is concluded that the safety analysis impact due to the incorrect modeling of the Boral configuration is minimal.

3. COCTZV3 ACTMOS:

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

Evaluation has shown that 2000 PPM soluble Boron will compensate* for even fully deteriorated Boraflex. Therefore, Braidwood will administratively maintain >2000 PPM soluble Boron until further review of the Boraflex issue. This will be tracked by NTS item 1456-180-96-01001.

This item has been completed.

Spent ruel 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 NTS item 0456-180-96-01002.

This item has been completed.

The lonq 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 Keft.

The analysis for this amendment is in progress. Submittal to the NRC is expected in mid-1997. This will be tracked by NTS item 0456-180-96-01003.

This item has been completed.

ComEd his created a Boraflex Issue Coummttee to work with the industry to resolve this Issue.

Resolut-on: 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 NTS Item 0456-180-96-010ER.

This item has been completed.

As a result of the supplemental criticality analyses for the actual Region 1 ::el-' Bocal geometries, the following administrative control has been impleemented in 3-azizn procedures !BwAP 2364-9, paragraph C.l1. This change is tracked by NTS item 4 4)-18, 96-0105101. -No assembly may be placed in a Region 'ak r'a-e: 'oration race "

another assembly across a Region 1 rack interface."

This item has beer completed.

ETIMATED U.,DE!N PER RESMNSE TO COMPLY wIrT THIS WNORSIAT1ON COLLECTIO REQUEST. 500 HRS REPORTED TO INl[UTR'

• FOR IUANRCOO THE UCLINSZNG KrOWAR*D LICENSEE EVENT REPORT (LER) LESNS A,-ATE PlROCESS A,'D FED BACK Co.%N.*f.%'TS TEXT CONTIATION REoARDINO LUDEN ESTIMATE TO THE TNFOR.IATK.N .%_\D RECORDS MAAMIENT BRANCH (4 F3311 LUs .UCLEAR, "AREGATORtYCOMMzSSIN. WASHINTO.N.DC 20555OO0I.AND T THE PAPERWORK REDUCXTI0 PROJECT FAaCUT NAME (1) DOCK-I N1MU (C) LERnNumB(R) PAGE (3)

Braidwood Unit 1 05000456 96 010 02 -7 of to (C me mpm is rspued, un addiaima caopes of NRC Form 366AX17)

-Dated on this additional administrative control, Braidwood station repositioned fuel assemblies in the spent fuel pool. CorEd has subsequently verified that the storage

=onfiguration of fuel assemblies in Byron and Braidwood srI meet the criteria specified An CAC-96-243 and met the supplemental criticality analyses performed for the actual megion 1 cell Boral geometries.

'This it.. ham been completed.

ComEd will review the spent fuel pool criticality analysis for other Comrd facilities that may be susceptible to similar problems. These reviews will verify that the gcurrent analyses conservatively consider the potentially limiting geometries associated with peripheral cells of adjacent fuel racks, especially as it relates to the placement cf fixed poisons such as Botal or Boraflex on the outer faces or peripheral cells.

This will be tracked by NWS item 1456-180-96-0L05102.

This item has been completed.

Mrs5 will submit required reading for the entire staff to clarify the responsibilities of NFS - engineers when performing an. nacceptance review" of externally yenerated calculations: (1) verify all ComrEd 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. CornEd. This will be tracked by'NTS item-0456 3S0- 96-010S103.

"This item has been completed.

3475 will revise NFS procedures governing the review and approval of controlled work to clarify the responsibilities of NHS engineers when performing an "acceptance review" of externally generated calculations: (1) verify all ComEd 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 CouEd. This will be Tracked by NTS tite #456-180-96-0105104.

Trhis item has been completed.

A review of regulatory requirements/guidance on fuel pool rack criticality analysis w11l be performed to ensure other requirements are adequately addressed. This will be Tracked by NTS item 0456-180-96-0103105.

This Item has been completed.

Obtain As-built drawings for the fuel pool racks. This will be tracked by NTS item

• 456-180-96-0105106.

Velete corrective action:

Communications 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 Zhe 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.

"rhe Boraflex and Criticality Analysis issues have been submitted to the ComEd Pact 21 Committee for consideration of reportability under 10CFR Part 21. This review will be tracked by NTS item 1 456-180-96-0105107.

"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 4456-180-96-010SIER.

"This item has been completed. I

(4,45)l lOMI1JIL id"C NLCLEAR REGULATORY COMMIII5ON A1?ftID u BV ulaownD NOW' 31mo5S4 ESTIMATED LIDL3DN MER RLSPONSE To COMIPLY WInh This DqORMATIONCOECfl OE- .LW.. So* Has REPrTE UrS5 NS DART ED I THCs%

LICENSEE EVENT REPORT (LER)

TM CONTMINATION READN BL .EESUMri To *THE OmRu.IT% %%XD RECORDS MANAGEMENT BRANJCH (T4 F3U) V 5k. LC1LR EOLLATORY COMMIKsIN, WA2hl?%T0r.Dc 20355o00.,V.,I THE PAPEIRWORK REDLXflON PROJECT VACILIT NAME (1) DOCKE UMDER (2) LEA NIJM3E(O ~E YEJAA sEURIUAL "%EVUK)%

braidwood Vnit 1 05000456 96 1 010 1 02 Sot J 9

(it EAT EiCs repild, adftdiosi

&me copies of NRC Fan. 366AX 17)

Borafles degradation, that was bounded by the current Criticality Analysis, wasn previously identified at Braidwood Station INTS 0 456-201-95-2155). Anticipatory a ~ensasory and mitigating actions werer put in place and included: administratively maintaining the SIM boron concentration greater than 2000 ppm, maintaining the SFP temperatuare as low as possible, restricting the removal of Silica from the SF?,

mninimizin transfer of SF?. water into the RCS during refueling operations, and developing a Boraflex committee to review and approve 'long term solutions to the Doraflex degradation problem for ComrA~. The corrective actions of the previously identified 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 theo event. The absence of floral. poiaon plates on the periphery cells could have beinn'ldentified during the blackness tettlhq in. 1991. -The increase in neutron signal may hay, been attributed to degradation o.fBorafltex-'rather than lack of Boral 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 toabe accurate).

A review of industry events did not find previous occurrences of errors, in criticality analyse& due to Bortl poison sheieti, not being installed.

a. COWn mw2 FAILURE DATA:

KAXUFACI1JRZR NOMENCLATURE MODEL MFG. PART NO.

Bisco Prcoducts Boraflex Panels NA NA Inc.

EXHIBIT B-2 Braidwood Unit 1:

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

• m OM39 ULL NUCLEAR RE04LATORY COMMISSMO APPROED BY 0MB NO. 3"""41 0410 EXFUIS OWAM LrXSAD BemPut OWSONS To CcI.yV fT To*

LICENSEE EVENT REPORT (LER) L=me1DW.

sralsvc W t060004" 1OF8 huqWprphCfleap*e of Spewt Fuel Resutftg InTechrkal Specificallo Violation Due to Peisonld Enm.

_ OW-l

_ _ _ _ _ _ _ _ _ _ _ _ _

D.Lawuli 20-203M? 2n1nern (8545-80 3061 )

- NN-419101I U IEI%

I Wx" PU Kf LkU1 1 4W OP . Ul. P-Tv J ~ -- 0 During the verification of Spent F~uel. Pool storage locations, it was disccrered that an*m fuel assembly was stored in Region 2, and rot in the recuired check~rboard conticuration, based upon the burnup versus iniitiai einrichment limits specified oy Tecfl. Spec. 5.6.1.1.b.2. The cause at this event wan personnel error. The burnup versus initial. enrichment lLm~ts, which determine acceptabloe fuel. storage conti-jurations, were chainged by Technical Spe.:iticatior, Am~endment 58. A calculationl performed prior to this change to verity that the new limits were met contained an incorrect burnliD. The *:alculat ion was not

ý.r'devendentlv voxizi-iedo, so Ene- error was nct icentiftied. Immediate correcti-4e actions were to reiocate Assembly .-46V into Regicn 1.of the Spent Fuel Pool.

Additional corrective a(7tions were counseling of the inrlividlial regarding expectations And procedtire revision. This event ro-sults-i in no safety :-oncerns.

Two Prev iuus fue I mispos it 1on i nl event's wr --ý du-, r - per -,nne 1 e r r r andj proc~edurAl ano management *ief iciencies.

-~~00LR uv-OO Meu~s LL.mm APPROVED EXPIRES NO 31164104 BY WeB413WNO4

• -2 9SThA1E uOLWA" PER RESPONSE TO COPLY VWDI TMS WNMTOMMORMY1ON COLLECTION F*9NR1pWE LESSOM LEARNED RPWEaSr: §i0H "

ARE 1NCOF`GATZD LICENSEE EVENT REPORT (LER) THE UCV*00 FRI~MM P COMADWTS AM FE SUVN REGARSOM MsrM.

1AC 10111TW*'I*

TEXT CONTINUATION TE 9WOU AM RD WGO(ME DWN 6 r^ " NUCLEAR REOULATORY CSW.

DC 20W6000. AM TO THE WAinGT*W PROW10T FAMUMOK PLAN C NDI, XOg PRIOR TO EVENT:,

UN Unit 1 raidwood I EVENT DATE: 0710F96 EVENT TIME: 1045 MODE: 1 RX POWER: 100 RCS [AB) TEMPERATURE/PRESSURE: NOT/NOP S. DESCRIPTIOM OF EVENT:

There were no systems or components inoperable at the beginning of this 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 mne requirements or Technical Specification 5.6.1.1.b.2, "Fuel Storage

- Region 2". 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 investigation into this event, ComEd Nuclear Fue-1 services transmitted a 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 1 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 Specificatior, 5.6.1.1.b.2. Upon discovery, fuel assembly 546W was immediately 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 1 of the Spent Fuel Pool. S46W was relocated into Region 2 of the Spent Fuel Pool on June 16, 1992. Prior to

Yom4 Dom a.. -=hSMOAMW CSýSZU AMOED BY OiS WOM 316"1" 93-D-ED UBEll PEA IEsPONSE TO COWLY WVTM THS W4"TOWy #1FO0mION cLECT1N IWOUESl?: 5, Hft.

REORED L93ML&M 4E PE R dCON1" MW~f r~pR LICENSEE EVENT REPORT (LER) 09 W104UXMACKTO FCWAAW REGANDM* DUTA ENAO ESIV^791 J.TO TEXT CONTINUATION TM PWvo0oMW M hWU0oNr BuReA (.

K Mcomos OIAAORYSMl NWMJOIANKRrm 6r* U3.SJ.

WAN0Wn1. 20001. NATO "KEPPAPMO 002 MMMMMN P"DJ*'F Braidwood Unit 1

[Brai w T SMWMIAL L °°05000456 3 Of 6 96 I--008-- M1001

____I00 I1.

I W (it 00r Sc ii As e*red, ur ea e *WZiorJ copJ* of W frbm J6A) Ill)

D. DESCPAIPTCIN OF EVENT (continued) thjis move, procedure BwAP 2364-9, "Controlling Movements of Nuclear Fuel Into 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 S46W, based on assembly burnup, supplied by Nuclear Fuel Services.

Technical Specification Amendment 58 was incorporated on January 20, 1995, to reflect a new criticality analysis that includes fuel enrichment to 5.0 wepight percent uranium 235, andl 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 dutring A2R04: (refueling outage prior to Unit 2 Cycle 5),. Although these calculations were not required until receipt of the approved Amendment, they ware performed to verify that the new limits would be complied 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 pireviously 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.

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

!s not retained as plant documentation, and requires performance only upon movement within the Spent Fuel Pool. Since independent verificaticn is not required, the Nuclear Material Custodian was misled into thinking that independent verification was not required for the calculatiuns prior to Aimendment incorporation. Since performance is not required except prior to fuel movement in the Spent Fuel Pcxl, 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) (i) (B), any operation or condition prohibited by the plant's Technical Specifications.

kjpoV ofrOj) 11 IAJ0a'

-"--- EWIS7*OYC' W2U - 4 g'-

-TMT* NOON NMR MWONM TO CO.LY M T" hMAIMTCW IDO1ON COUICTN P=40: 9O HW.

WM" MEM EDD PCOWAYM WTO FAV47OMP LICENSEE EVENT REPORT (LER) DCucdMPFXXXWAsWo#AXToWUnWW.

NSMEMPAWo W 11W TEXT CONTINUATION Doo 02) 1" Vo

.Wv mo 11) 05000456 1 1 Mumm 4 OF6

~ridwodUnnit 1 0 0 8 --

____ ___

___ ___ ___ ___ 96j-- 00o a.,s* Ali onai 'wi*' of AC rose 3**AI tall MoSto 4"M. JA Few.

C.. S OMFl 0 VENT:

lTie cause of this event was personnel error.

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

D. SAFETY ANALYSIS:

There were no safety consequences from this event. The Spent Fuel Pool boron current concentration remained well above the value assumed for the ariticality analysis, while all fuel assemblies, adjacent or near to the umisloaded 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 had burnup greater than the analysis, and all adjacent fuel' assemblies had been minimumw burnup assumed. If the Spent Fuel Pool boron concentration at the value assumed for the current criticality analysis and -n additional ruel misloadinq had occurred, the required k-eff of 0.95 may have been exceeded.

-Now, )so SNOUdZW Q.a - MCA"t3i. ic APRVEDB NM 3164IM OM MI3WJ4

• -*2t FEXPI ESTSATED UUWE PER RESPONSE TO COMPLYWw" T45 MdOlaTORW N0WOVATION COUJCTiMON REU.EST: $S.S.

RPORTaDUISUONS LEAMEDDN AM 960R0MMTI WTO LICENSEE EVENT REPORT (LER) "MIECUS PmROCMEAMMD FWOMTO DJRIVW.

FORWAMCOMMVMNTEIGARVMBLTAN*TOml TEXT CONTINUATION T(N7fANOW1 MW WC.T S ,33).U.&MUW[REIATONY CW6 WAWTMKDC REOUCION m. 20WIM1.XNTOTE*P*PFWWW in~ZW S ill- min(2) zm u.OA -

Braidwood Unit 1 05000456 yz" I swIAL"DU6T mvWs, 5 OF 6 96j--008--1 00 (it norg ipce is requirod, ua acs t¢Jtina1 copi of Arc rtorn 366A (17 K. CORRECTIVE 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 F'uel Pool', will be revised to require independent verification of the calculations, retention as plant documentation, and performance when the burnup 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 ensure 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.

F. PREVIOUS OCCURRENCES:

LER 1-96-007 involved failure to comply with Technical specification 5.6.1.1 due to positioning fuel that did meet the burnup versus initial enrichment limits in a close-packed configuration irn-diately 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 lcaring to this event were different from those leading to th: 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 rnispositioned 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 proct. ares. A review of the corrective actions determined that they would not have prevented this event from occi~rring.

NRC FORM 38(4-95)

6 AWA

EXHIBIT B-3 Braidwood Unit 1:

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

1 4 FORM244~ ~ EGUATOY COMISIONAppROVIED my0#" pNO.3164 D ESTOMTED BUROeN MA RESPONSE TO COWY W3TH THIS MANDTORY *WMiTMO COLLIECTMO REQUEST: 50.0 HS.

REPRTD ESONS LEARNED AF19 0CORPORATED WTO ThE UC40 3 MM AC ONMW OA LICENSEE EVENT REPORT (LER)

ITME PAPENWOIW REKNJCTIO PRAOJECT 05000-458 lOF5 Draktwood Unit 1 Placement of Spent Fuel in Regards to Checked ~no DuO to Personnel Error, and Procedural and

• mar mar 4

_ _ _

I';U W I '0T0FCMi MA.NNERVU

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_____________________ r~uIM F.IM *.a5.

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7-2.22m*[aX  :)O*mmlm(4) S0.7&aW2WOMv~ h Amd I 1I-9X4)_______ . - FnloWbeowa i50.36(cX!)

i3

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• k ww'I' TkL'.N 7.inM p -

(815) 456&2801 X3051 ID[.Lawson, System Engineering (Ii s. afgWW EXPECTED SUBMISSION DATE).

JUITRACT (th1O 1400 SrAcus. f-.s that was not bounded by the existing On 6/17/96, fuel was repositioned in the Spent Fuel Pool into a configuralion Supervisor accomplishing the fuel Criticality Analysis. The mispositioning was identified by the Fuel Handling fuel in the inappropriate configuration movement and reported to the Nuclear Material Custodian (NMC) The and Independent Reviewers did not consider was immediately repositioned. Investigation concluded that the NMC the fuel moves Additiunally. this the effects on lower burnup fuel in adjacent storage locations in planning Station by the analysis vendor Causes configuration rest. iction had not been properly transmitted to Braidwood management deficiencies Corrective of the event were determined to be personnel error, and procedural and position guidance, counseling of actions taken involve preparation of a -ew procedure containing more detailed 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 atn appropniate configuration mispositioning was that the mispositioned fuel did not cause a criticality concern A previous event involvingfiuel caused by a failure to Follow procedures for fuel moves

a--

A-LARO - OxSc-M APPROVED BY OMB NO. 316" 164

- r,- 3.o V..S. -

EXPIRES 04MM ESTI, TED BURDEN PER RESPONSE TO COMPLY WTH THIS MANDATORY ONFORMATION COLLECTION RECUEST: Sc.O mS.

REPORTED LESSONS LEARNED ARE UCOCRPORPATED WTO LICENSEE EVENT REPORT (LER) THE UCENSN PROCESS AND FED AC TO 3DmJSTRY.

FOR6WARD COETS REGADING BUREN ESIMATE TO r TEXT CONTINUATION DIE W ONAEORD MENT BC 6 F3M. U.S. A*L.EAR REGULATORY p5UVN.

WA9*4GTON. DC 0546500. AND TO THE PAPERWORK REDCTION PROJECT ISUM WA (1) DoWm CIIm (21 to ~ w~

05000456 YEA S ENUAL 2 OF 6 Braidwood Unit 1 96 -- 007-- 00 o.*ifJronj)j copi5' of NRC Forw .66A) (17)

-an (I ports pAof Js tqJulrd, us.

A. PLANT CONDITIONS PRIOR TO EVENT:

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

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

Fuel moves were planned for the Spent Fuel Pool in preparationifor in which a "Blackness Testing". "Blackness Testing" consists of a technique Boreflex neutron neutron source is used to evaluate the degradation of the periodic Continued absorber material in the Spent Fuel tool storage racks. remains moderation testing is a commitment to the NRC to ensure that neutron Pool Criticality within acceptable bounds, ensuring that the Spent Fuel Nuclear Componenr Transfer Lists (NCTLs)

Analysis assumptions remain valid. this purpose on Custodian (NMC) for were prepared by the Nuclear Material NMC position.

recently assumed the 5/9/96. The NMC preparing these moves had previous NMC, on An independent review of the NCTLs was performed by the As a part of the normal review process, a second independent 5/11/96.

5.'15/96. On review was conducted by the Station Reactor Engineer (SRE), on these fuel moves, 6/17/96 at approximately 0930, during the performance of he considered the Fuel Handling Supervisor noted a fuel configuration that The suspect configuration involveýd irradiated fuel stored in to be suspect.

Requirements ter storage of fuel in Region Region 2 of the Spent Fuel Pool.

corresponding to its 2 are that either the fuel must have a specified burnup initial enrichment, or it must be stored in a "checkerboard" configuration weight percent if its initial enrichment was less than or equal to 4.2 Uranium 235. Fuel meeting the burnup-initial enrichment restriction may be stored in any configuration in Region 2. Thp suspect fuel configuration beirq involved fuel that met the burnup-initial enrich~ment restriction adja-'nt -o ýkiel "ha- did stored in a close-packed configuration immediately not mept the requirement, and was placed into the "checr.erboard'

r

• -='"" --L&*-.ssIoI APPROVD 9Y 0.B MB NO, l' 1 ,

S6i ii.OR4 s.S.

EXPRES O413WN "ESTIMATE OUURDEN PER RESPONSE TO COIPLY VMi" THS MANATORY OMIATION COLLECTION REOUESI: 50-0 MRS.

REPORTED LESSONS I*EARNED *W ECOMPORATED P0TO (LER) IFORWARD p MPEM TVE U COWMNTSr FED GMXTOWUBT~R.

REGARCM4 MAVE~N ESTMrlE TO LICENSEE EVENT REPORT TEXT CONTINUATION. no WoIATO AM ME MM*ADMET MAMM (T I M3.U.S. WX.EAR REGl"ATORY COWUSS WASMATON. DC 20M5.00t. AND TO THE PWORK REDUCTION PROJCT

-c - 2 &MDISOM tos MM 3 MaUMzv MIS (1.1 05000456 Y*" s5MLM*,f!j^L IMMV; 3 OF 6

_raidwood Unit 1 96 -- 007-- 00 a:-u .t*lrjF jnadl WjaS of MAC Form )66A) 117)

Iff More 3P.M j sO ,i.jW r,1, B. DESCRIPTION4 OF EVENT (continued) configuration. The Fuel Handling Supervisor immediately contacted the the System Engineer in cha:ge of the "Blackness Testing", who then contacted NHC. After consulting with the SRE, the NMC directed the Fuel Handling Supervisor to suspend fuel movement, and began preparing NCTL Variations further (BwAP 370-3T3) to re,,osition the suspect fuel assemblies pending The NCTL Variations were prepared by the NMC and investigation.

independently reviewed by a Qualified Nuclear Engineer (QNE) and two Senior Reactor Operators (SROs) by approximately 103C. investigation of the not suspect fuel configuration revealed that this configuration was so a specifically allowed in the Spent Fuel Pool criticality analyses, of the Problem Investigation Form was completed at 1215. Repositioning suspect fuel 'assemblies was completed before this time.

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

of this A 2opy of the an.alysis indicated there was no safety significance a minimum boron concentration in the fuel pooition.,ng other than requirrng at all times during this Spent Fuel Pool of 300 PPM, which was exceeded evrut.

(B), any Ti.> event is being reported pursuant to 10CFP50.73(a) (2) (i) operation or condition prohibited by the plant's Technical Specifications.

C. CAUSE OF EVENT:

The caus;ý. of the event .were determined to be personnel error and procedural and mari3aement deficlnnc-Les.

not identify 7he Nu':P-ir Miterial (:ustodian and one Indepenident Reviewer did plan.

the su.:nect fuel positioung during preparation of fuel movement A_thol.!. no known requirements for the olacenent of fiel at this transition

m~ssiff APRO EXPIRES BY 0MB0FONO. 3101 RAPOa.w IYIF4

• 0-921)*

-W SGQ V.S. WOW" ESTMATED IPROEN PER RESPONSE TO CWI~.Y wVrTH TMI M*NDOATORY WORIAATION ZCTION REQUEST: 50.0 MRS.

REPORTED LESSOMS LEAE FED ARE I0C0OPORATED HTO SAM TO 1DUSTRY.TO LICENSEE EVENT REPORT (LER) FORWARD COMLWTs RGARo* BURDE ESTUM THE WOFiMi AM. R*Nnos wWB.F)I DmJcH (T TEXT CONTINUATION 0 F34. U.S. CLEAR REOULATORY CO55sloN.

WASPorTON DC 20*46SM0. AND TO TE PAPERWORK REDUCTI NPWOJeCT (2) ZZ = to I (3)

DOOM Braidwood Unit 1. 05000456 YEAR SEUN PIVISION 4 OF 6 96 -- 007-- 001 iC.

Itwit pae. js zuquJro.d as.

CAUSE OF EVENT

• .WUtJionaJ capJa. of MXPCFosa 366A&

(continued):

(I7) enrichment criteria and boundary between fuel meeting the burnup-initial 2) of the Spent Fuel Pool fuel not meeting the criteria (stored in Region analysis vendor, the NMC had been transmitted to Braidwood Station by the such a questionable and the Independent Reviewer are expected to identify configuration prior to NCTL issuance.

identified the suspect fuel One Independent Reviewer of the prepared NCTLs positioning as questionable. However, the reviewer did not address the question prior to approving the NCTLs between fuel that does meet The required fuel positioning at the interface fuel that does not meet the the burnup-initial enrichment restrictiqn and not specified in any criteria in Region 2 of the Spent Fuel Pool was directing fuel Braidwood Station or Commonwealth Edison procedures movements.

than cr eqtial to 4.2 weight The requirement for positioning fuel with less enrichment percent Uranium 235 that does jnot meet the burnup-intial by the "Licensing criteria in a checkerboard configuration was transmitted Units 1 and 2",

Report On High Density Spent Fuel Racks For Braidwood Revision 0, dated August, 1988. This document addresses the assumptions interface requirements.

made for the analysis, but does not identify any against positioning The expectation to revie4 the planned fuel movements requirements intn Inclusion of all requirements was not clearly defined.

of NC-LS Lor all types of fuel movement planning, and actual preparation in sufficient fuel movement planning did not address these activit:es detail.

NQfTLs were performed The planning and independent review of the controllei using unverified and uncontrolled information.

D. SAFETY ANALYSIS:

A-.a!'y,.s by the vendor There were no safety consequences for this event.

Lndicates that the performing" the Spent Fuel Pool Criticality Analysis as long as suffi-ient mispositioned fuel did not cause a criticality conce':r.

V'R . a.

7gQ aL& .l&mT cUsszw APPROVED BY OW NO. 31ff1"

~~ EXPIRESHMIWN ESTIMATED PBJUEN PER RESPONIS TO CNIPLY WITH THS IM#MATOR' 1 TXW COLLICION REMUST: 50.AMRS.

REPORTED LESSON LEARNED AR OORPORATED SITO LICENSEE EVENT REPORT (LER) 7wuwaPOCSS IFORIWAD CO*ANTSU FM REIkROMB MTOT WIROEN O.

E7TIARTE TO TW DWOMAoM REcORM mWA4AMNT MRNC (T TEXT CONTINUATION VIU.

o FM. U.S. MICLEAR REOULATORW WAS1IGTON. DC 2U5CUOW1. A 10 THE P*APRWOR REOUCTION PROJECT 05000456 YE"A Sr.ULNLAL "nsin, 5 OF 6 Braidwood Unit 1 96 -- 007-- 00 I soV, mt c" Js requJred, u - diwitJonai copies of AC roae )66A) IU,1 S(If D. SAFETY ANALYSIS (continued):

boron existed in the Spent Fuel Pool. The required concentration for this of event is 300 PP2, Spent Fuel Pool boron concentration remained in excess If a fuel mispositioning or fuel 2300 PPM for the duration of this event.

'orop event had occurred while the fuel was mispositioned, sufficient boron a safe condition.

concentration existed to maintain the Spent Fuel Pool in E. 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 generated to require for fuel positioning.

• Procedure changes have been new procedure prior to issuing NCTLs. The new procedure execution of this an indepetdent includes a checklist, requiring both the NCTL preparer and verifier to review the proposed fuel movements for fuel positioning requirements.

The interface requirements for fuel storage that does meet the initial analysis burnup-initial enrichment requirements were received from the These requirements were reviewed against all other fuel stored in vendor.

the Spent Fuel Pool. 1,7 other instanc-s in which the requirements were not met were identified. These requirements were incorporated into Braidwood Station Procedures as BwAP 2364-3AI.

guidance The Qualification Guide will be revised to provide more specifiz regarding the necessity to review planned fuel moven.ents against positioning 7 requirements. This action will be tracked by NTS item #456-18C-96-00 01.

This event was discussed with all qualified Nuclear Engineers.

EXHIBIT B-4 Browns Ferry Unit 2:

Supplemental LER (October 9, 1980)

--.....

~iORITY AO1-a Hr.ans.O'ily Director U.S. M=3lear regulatory Ccmriiskic Offioeý of Inspecticn and hforcemnt regicri.Zl l 101. Marietta Street, Suite 3100 Atlzntal Georgia. 30303

Dear Mr. 0O'Ieilly:

TENE VALL~y AIUhjIT' - Br~wNS Fruta NL=R PIJANV UNIT 2 - DOCK=

No3. 50-260 - FACIT= CPERAT=l T.~ DPR EXPOTABL 92OIC B~a-50-260/8O37 R~EVSIW 1.

A The encl osed report is a supplemenit to my~ letter dated Septmrber 25,, 1.980, xr~ernngfuel assemblies TZ 758 and TZ 399 w~hich we~re miscriented 90 degrees. This report is submitted in acordmen with Brns Ferry unit 2 Technical Specificaticr 6.7.2.a(9).

very truly yours, J. P. CaLhoun Director of Nuclear Powr Enl osure (3) cc (Enclosure):

Director (3C)i~o Offioek of Managmi~t 4 atmand Por m~o U._S jula Catndissicri Director (40)

Offic of Inspectimr ana Eaftoement UJ.S. Nuclear Regulatory Caxzussiczi Washingtrri, DC 20555 Mr. Bill Lavallee Nuc-lear Safety Analysis Center Palo Alto, CAli1ft-mnia 94303 M!r. R. F. Sullivan, NBC Inspector, Brcwn Ferry 8An 10150

ývOt- An tunitk

ONPC FC-. SAM MkL.LJLA I UM Y IP1 UPDATE REPORT -PREVIOUS REPORT-Septeber,126;-1980 LiCENSELF EVEN R EPORTI CONTROL BLOCK: j ..... LLJ0'IPLIAIS PRINT OR TYPE ALL ACQUIRED INFOAM~ATIONW CFY* 9 LICKINS1E coDE 14 IsLICENSE PiU1mal i 26 17AICSE REPORT ~LEJ&)0 15 10 10 10 12 1 0ak7 19 1114 18 10 0i1) ifo01 01g! RI rn so 10eso 0 O OCKET NU.MOER es E5VENTODATE 14 -75 REPORT DATE Ili EVENT DESCRIPTION AND0 PROBABLE CONSEQUENCESG 1 During EOC-3 fuel shuffling operations it was noted that fuel assemblies TZ 758 and LTZ 399 were misoriented 90 . There was no previous occurrences. There was no 025 A. JaO~s"A to~ the publics See Technical- Snecification, 2.1 and 3.5.K.

1' 5.. - - - - - -

£S'STEM IlkS CAUSEfCOMP VALVE CODE CODE SL'BCOOE COMPONENT CODE sLsCOO&E SUBCoOO 10 It 12 13 11, 19 7 a.

,EUUENTIAL OCCuAI 4F NCI REPORT vSI' RE(PORT NO CODE TYPE NO LE,n no VENT YEAR r~U~L21 Z2 23 24 26 27 .18 29 30311

• PPECT SH4UTDOWN A-TACHVIANIT A\ NIP410.A onivE Comp. 0.p1tNt Art-Oi£ FUTURE ACTION OP4PLANT METHOD HOURS 023 SUIMITTED FORM lIW. $LPPLIER kI&NL-9&7'CA:IFnr TA..cEP tX 181G 0 LW(Z LZJ LO-10101041) LX-JD I-(3II IW10 18 io 1.

3% 74 37 41 42 43 444 33 ja CAUSE DESCRIPTION AND CORRECTIVE ACTIONS © in During the previous refueling out~age (BOC-3) 16 fuel assemblies were misoriented.

I1 Subsequent rework left two 7 x 7 bundles in the misoriented position._ A later Misoriented

~I review of the-BOC-3 core verification tapes confirmed there were no other I fuel assemblies. Units 1 and 3 core verification tapes will be verified.

I

r. ACILI tv VETHN r) Of

% POWER OTNEN £7ST., T DISCOVER~Y D-%CuVERY LSCRIPTION (

STA1LJS

' L.](3 1 01 01 0 ~q L (:01 Operator observation

.kCTIVITY :ONTENT LUC A~ ' OF ~EALE E G1 REI1LI IZ NSWI OF*64ELEASE Z18 AMOUNT Of AC11y'?'

NA INA 45 7 89 to II4 PFFISUNNEL EKPOSURES Nut.II #, TYPk J..:,CAIPTIU)NS

ý-,VME DESC iPtN F

d 9 '1 12 LOS! I,,IlOHAWACE TOFPACILITV "TYPE LL'CRIP t 00 WZ' I? NA PUOLICITY '-jýNL YN I',SUED II,*.;rpPhIONS I e TVA media information televhofle II~III NYAME oF Pn&PAFIEI P14ONE

Aw..Tennessee.Valley. Authority Form BF-17

"-~Browns, Ferry Nuclear. Plant BF 15.2

"- * "* " .**i

.*  :" ' 1/10/79 LER SUPPLEMENTAL INFORMATION it

'I BFRO*50- 260 i._8037 Technical Specification Involved Reported Under Technical Specification 6.7.2.a(9) 2.1 & 3.5.K Date of Occurrence 9/14/80 Time of Occurrenco 1900 Unit 2 Identification and Description'of Occurrence:

Fuel assemblies TZ 758 in c 8 re location 15-26 and TZ 399 in core location 29-28 I, were found to be rotated 90 from their correct orientation.

t Conditions -Prior to Occurrence:

unit 1 - 1055 MJe Unit 2 - refuel shutdown Unit 3 - Shutdown maintenance outage Action specified in the Technical Specification Surveillance Requirements met due to inoperable equipment. Describe.

NA Apparent Cause of Occurrence:

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

Analysis of Occurrence:

See attachment

- S Corrective Action:

have been made.

Verification and reorientation procedure for fuel loading verification documented wi~h Procedural changes include the requirements that rework will be secona party verification.

Failure Data:

KA

• Retention: ,jeriod - Lifetime; Responsibility - Administrative Supervisor ORevion:

Mccoding to Supplemental Reload Licensing document NEDO-24169A the limiting full loading error is a rotated 8 x 8 (8D274) fuel assembly and assumes a limit of 1.07.

rotation of 1800. The MHR for the limiting event is the safety Any other misorientation would result in an MCPR greater than 1.07. The "misoriented fuel assemblies were both high exposure - original 7 x 7 (7D250).

Since both aubject 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/O HCPR - 1.39 6/30/80 1{CR a 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.

4 t.

EXHIBIT B-5 Byron Station:

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

F"U.S. NUCLEAR REGULATORY COMMISSION APPOVED BY OMIN NO. 315W04104 NMC FRM~EXPIRES 04130198 ESTUMATTO AMiscI" VA(~ TO COMPLY OmTM IllSMATORT E~OWATiOM COLLECTIO11 ftEIRM3 508O Ha. RPOM tSI5 11AM. AN LICENSEE EVENIT REPORT (LER) FO=C1111T 11141111 UCWS wom IUMCKIim ra ToCIS WCOilft MMANA"W11hTIAMSN 14 F33L US. WXUCA NWIATMY C01111IS~IC WAOMO 20MMO1W. AM T11111E KC71011W PMKT MW5 PA1911111111 (See reveresl for required number of 10%OWEI~ OFm MNUUISNT AMS PAGET. WAOMITOU K 205" digitslcharscte's for each block)

DOCWE Mi CM Fam CU AW 111'i R~JT 05000454 1 OF 9 BYRON NUCLEAR POWER STATION Fujel Assemblies Located In Incorrect Region of Spent Fuel Pool EVD DAT (611 LER NWASA . REPORT DATIE 17? OTHER FAC1UTIES linLE 3 TEARI VIM m"=IN~A I 'v~wOu um DY YA "

50 Mai" DAY I 1111AII Iiel 008-00 9 DCET050 05 28 96 98 -- 06 25 FinYMA E PURSUANT TO THIE REQUIREME111TS OF 10 CFR 3: (C heck o~w or m1re111t" REPORT is SU THISTIG 150.73We12*iNe?

20.2201 W 70-.2203(a)12)ti' x 50.731alIill MOWE I3N 5O.73lrn#12l1x) 20.2203(sH3H' S0.731e)[2)iil PWR20-220318)(1? 73.71

_26.2203(*)12)(4 20.220314113114) W073(a)l21041 LEMU 11011 OTHER 20.220341 2)lW 20 2203ta1141 5O.734a)(21(iv Iv? 596@tY inAbsovcl below r0.2203101421(W __________ 50.731eII 120.22031*1(2M 50.364HOW2 - 5O7311t)vii)

CENSEE COKTACT FRTM LER 112?

X2154. 1-3-"

David D. Gotff, Systemn Engineer DESCRIG10 WTHSS REPORT 113) ____

COMPLETE ONE LN&FOR EACHCOMPONENTFAb URE CRSE WKWMA N

-MK STFC T11 rrm uxcw 0If CAUSE SYSTEM COMFUNENT WMAMSII1 oCUE wnie XECE UBISIND TEI typeiwntten he..) 116?

PETED DA) 11s)

Io~ FE ASSTRC (LUmit to 1400 spaces, i.e.. appontiry 15 anir~e-spce that fuel assemblies F37E. F4.E. and G67F were Or% 28 May. 1996, Byron Station nuclew relgineers confirmed meeting the requirements of Technical Specification (TS) reaidi i R on 2 of the SetFuel Pool ISMP without burnup requirements. nor were

.L St Th. rag qin miwwriu 2-.- The assemblieswere did not meet the minimum 32651 MWd/MTU. and 32771 5.6..2.'~ue -

eqre burnups 32851 MWd/MTU, they checkerboarded. 32838 MWd/MTU, and 32728 MWd/MTU MVN/dMTU respectively. The actual buraips were 32848 MWdIMTU.

respectively.

Thse cause of this event was cognitive personnel error. The computer spreadsheet used to verify minimum required F37E. F44E. and G67F, and the data in the spreadsheet had btarnup contained erroneous informationi fr assembliesplacement of G67F into SFP Region 2 did not have the current not been independently verified. Pers!oniiel approving revision of Burnup criteria for determiflbof of fuel assembly eligibility for placement into Region 2. Ultimately, the of TS 5.6.1.1 Amendment 68. -Fuel Storage fueal assemblies' burnups were not venifid to meet the requirements Criticality," prior to its implementation.

Region 1, as allowed by TS 5.6.1 .I.a.2, 'Fuel Storage On 29 May. 1998. the three fuel assefrties were moved into either to meet the minimum required btirnup or to Region 1.* All fuel assemblies rematinWq in Region 2 were verified be stored in a checkerboard pattern.

This event resulted in no saefty concerr . The event was bounded by both the older and the newer criticality artalyses for Region 2fuel storage. Adequate reactivity not controls were in place to ensure that the k, lr-nit of 0.95 re-quired by TS 5.6. 1. 1, Fuel Storage - Cuiticality' was challenged during this event.

or condition prohibited by the plant's TS.

This event is reportable under 10 CFR 50_73(alH2)1i)(Bl. any operation

S *30A1U.S. NUCLEAR REGULATORY COMMISSION

'* LICMSEE EVEN RNPORT (LER)

TEXT CONTINUATION DOCKET LER NUM (6 AE FACILITY NAME Q BYRON NUCLEAR POWER STATION 05000454 98 - 008 -00 TVC of more soec is rvqukivd, g W copes of NusC Fow 35641 (17)

A. PLANT CONDITIONS PRIOR TO EVENT:

Event DateTirme 05-28-96 1 1700 84F /0 psig Unit 1 Mode 5 - Cold Shutdown Rx Power Shutdown RCS (ABI Temperature/Pressure 3350F 1 321 Psig Unit 1 Mode 4 - Hot Shutdown - Rx Power Shutdown RCS[JAB) Temperature/Pressure B. DESCRIPTION OF EVENT:

a checklist "Spent Fuel Burnup Verification Checklist." is Byron Administrative Procedure (BAP) 2000-3TI, required burnup for or have not accrued the minimum used to verify, that fuel assemblies either have by linea interpolation required burnup is calculated unchackerboerded SFP Region 2 storage. The minimum Burnup as a Function of Enrichment .1 for Region Minrmum Required between values given in BAP 2000-3A1. we intended to bound TS Figure 5.8-1.

values in BAP 2000-3A1 High Density Spent Fuel Storage Racks." The For Region 2 Storage.

"Minimum Burnup Versus Initial Enrichment fue.

(engineers I and 2) completed BAP 2000-3Tn for On 10 Febuary, 1993, Byron Station nuclea engineers of 3.8, showed both assemblies with an Initial enrchment assemblies including F37E an F4E. Thechklist 2 of 32540 MWd/MTU. given by BAP wt% U-235 and a *mnimum requ"iad burnup for placement'into Region burnupsof 32848 MWd/MTU and- 32638 MWdMTU U 2000-3-A Rev 1. F37E and F44E had accrued actual wt%

was a.ppropriate for an initial enrichment of 3.8 respectively. The minimum value of 32540 MWdMTU Region 2 storage.

requirement for uncheckerlborded 235, and both assemblies met the Technical Specification part, stated that INFS) issued letter NFS:PSS:93-060 which, in On 11 February, 1993, Nuclear Fuels Services This letter showed F37E burnup requirements of TS 5.6.1.1.

fuel assemblies F37E and F44E met the minimum and 32638.4 MWd/MTU respectively.

and F44E having accumulated 32648.0 MWdIMTU locatios K moved fuel assemblies F37E and F44E into SVP On 18 August, 1993, Byron Station fuel handlers checkerboard pattern since they were not stored in a C2 and K-DS, respectively, in Region 2. The assemblies were performed in accordance in place. The moves met the minimum required burnup restrictions presently 2000-3T3 Rev 1, "PWR Station Nuclear Component Transfer List."

with page 93-104 of an approved BAP was completed prior to transfer list approval.

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

Form ( PIF)

(engineers 3 and 4) initiated Problem identification On 11 August, 1994, Byron Station engineers methods in Byron Station and NFS employed different 454-201-94-69200. This PIF documented that 2 storage. NFS used minimum burnup requirement for Region determining whether a fuel assembly meets the penalty to criticality analysis after applying a 1.03 multiplicative a polynomial fit through the points given in the interpolation burnup calculation. Byron Station used linear account for fit error and uncertainty in the assembly

25. This PIF also identified that TS Figure 5.6-1 between points which bound TS Figure 5.6-1 Amendment for the curve.

bound the criticality analysis used as the basis Amendment 25 did not. for all initial enrichments,

FORc oMA U.S. NUCLAR REGULATORY COMMSSION LICENSEE EVENT REPORT (La)

TEXT CONTINUATION FACILITY NAM I DOCKET Li NUMS II IS PAGE

- * * ~ -n

~SEMMOI. imusm IYO BYRON NUCLEAR POWER STATION 05000454 3 OF 9 go - 008 -- 00 I I I E I T1*rfit nwo, spo IPrs mxlurd, use od&Wbiiift c'opk ef N AC Fwne 36A1 tI17W B. ESCRIPTION OF EVENT (cont.)

Byron Station ard NFS continued to use different criteria for minimum requred burnup determination. The license amendment request being developed, when approved, would render the second problem moot. For the interim, engineer 3 prepared a revision request for BAP 2000-3AM to change the points used for minimum burnup determination such that both TS Figure 5.6-1 Amendment 25 n the criticality analysis would be bounded.

On 16 September, 1994, Byron Staton nuclear engineers jengineers 5 end 61 completed BAP 2000-3T1 for fuel assemblies including G87F. This checklist showed the G67F assembly with an initial enrichment of 3.809 wt% U-235 and meeting the minimum required burnup for placement into Region 2 of 32661 MWdIMTU.

G67F had accrued an actual burnup of 32728 MWd/MTU. The minimum value of 32681 MWd/MTU was conservative for an initial enrichment of 3.809 wt% U-235. Engineer 6 stated that the enrichment vaue was conservatively rounded up to 3.81 wt% U-235 when the minimum required burnup was calculated. G67F met the Technical Specification requirement for uncheckerboarded Region 2 storage.

Also on 16 September, 1994, NFS issued letter NFS:PSS:94-225 which, in part, stated that fuel assemblyv G67F did not moet the minimum burnup requirements of TS 5.6.1.1. The discrepancy between the Byron Station and NFS conclusions resulted from 'the different methods in determining eligibility of a Region 2 strage candidate. Since G87F had accrued the minimum required burnup in accordance with BAP 2000-3A1 Rev 1. It was deemed to be suitable for uncheckerboarded Region 2 storage.

On 20 October, 1994, Byron Station Onsite Review (OSR)94-078 approved a license amendment request for

. Byron Station Units 1 and 2 Technical Specifications. This amendment request later became TS Amendment

68. Ths request would, in part, revise Figurs 5.8-1 Amendment 25 to be conservativ 3% greater than the new criticality analysis. Discrete values would be provided In Figure 5.6-1 along witti s tructions that would allow lirewr interpolation between the values. In particular, the required burnup for an initial enrichment of 3.8 wt% U-235 would be increased from 32540 MWdIMTU to 32651 MWd/MTU.

The OSR 94-078 package did not document the review of incumbent fuel assemblies and their eligibility for Region 2 storage with the new minimum burnup curve. Engineer 3 and a representative from NFS porticipate in the OSR.

However, Byron Station nuclear engineers (engineers 3 rW 7) had conducted a review of the incumbent fuel assemblies over the course of several months from opproximately August to November, 1994. This review was performed by engineer 7 building a computer spreadsheet to calculate assembly eligibility, and then th*

ouput was spot checked by engineer 3 for veriflcaton. The spreadsheet required input data for initial enrichment, storage location. and actual accrued burnup, and then checked each fuel assembly against several minimum burnup criteria, including those that would become BAP 2000-3A1 Rev 2 and TS Amendment 68.

The spreadsheet calculation produced a Boolean output for each assembly. i.e., 'OK' or 7not OK' for uncheckerboarded Region 2 storage.

Initial enrichment, storage location, and actual accrued burnup data loaded into the spreadsheet for F37E.

F44E. and G67F were incorrect. This resulted in the spreadsheet producing erroneous 'OK' outputs for thos assemblies. Had correct date been loaded into the spreadsheet, the assemblies would have been properly identified as 'not OK' when compared against the minimum required burnups of SAP 2000--3A 1 and TS Amendment 68.

NRC FORM 344A U.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILT SiIiI i

NMUM 1) DOCKET L NM WhoE lM 6f)EM PAGE (3)

BYRON NUCLEAR POWER STATION 05000454 - 4 OF 9

• ~ ~ ~ 9 05044I9- 008 -0 TET M mom Np" is ,r.pad,usea a i co of NRC fwwm 36M17 B. DESCRIPTION OF EVENT (cont.)

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

On 13 December, 1994, Byron Station OSR approved revision 2 of BAP 200043A1. This revision was processed as a corrective action to PIF 454-201.94-69200, which identified that TS Figure 5.6-1 Amendment 25 did not, for all Initial enrichments, bound the criticality analysis used as the basis for the curve. The new revisk bounded both the criticality 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 MtWd/MTU to 32800 MWd/MTU. Byron Station took credit for the review performed in association with OSR 94-078 to verify compliance of the incumbent fuel assemblies. As stated before, the spreadsheet contained erroneous data for F31E, F44E, and G67F. Hence, all three ases--blies passed the review. Under SAP 2000-3A1 Rev 2.

fuel assemblies F37E, F44E, and G87F no longer met the minimum required burnup, though they all met the requirements of revision 1.

On 20 January, 1995, the Nuclear Regulatory Commission (NRC) issued Amendment cdo to Byron Station Units

• Iand.2-TS, revising Figure 5.8-1 as requested under the licensng amendment requet previously submitted.

On 23 January, 1995. Byron Station fuel hardlers moved fuel assembwy G67F into SFP location G-12 In Region 2. The assembly was not stored in a checkerboard pattern since it had been verified to meet the

-requirements of SAP 2000-3AI Rev 1. ThIs was done in accordance with page 95-5 of en approved PWR Station Nuclear Component Transfer Ust. Engineers 5 and 8 verified that BAP 2000-3T1 Rev. 1 was completed prior to transfer list approval. However, BAP 2000-3T1 Rev. 1 had been completed in Septemnber.

1994, using SAP- 2000-3AI Rev 1. BAP 2000-3A1 Rev. 2 was now 'he current revision, and assembly burnups should have benw compared to revision 2 requirements rather than the revision I requiremonts. The assembly did not meet the minimum burnup requirement of SAP 2000-3A1 Rev 2 or TS Amendment 68, though it did comply with TS Figure 5.6-1 Amendment 25.

On 25 January, 1995, Byron Station OSR 95-007 approved for use Amendment 68 and its implementation plan. The OSR 95-007 package acknowledged that TS Figure 5.6-1 was changing. The implementation plan stated that the Byron Station nuclear engineering group *will revise BAP 2000-3A1 to reflect the new burnup curve to identify nsemblies that wre acceptable to load in Region 2.' At that time, it was thought that BAP 2000-3A1 Rev 2 was more conservative then TS Figure 5.6-1 Amendment 68. Therefore, the implementation plan required no deadline for revision of SAP 2000-3A 1. The OSR package did not discuss the review that had been performed of the incumbent assemblies. Engineer 5 and the Station Reactor Engineer (SRE) participated in the OSR.

On 30 January, 1995, Byron Station OSR approved revision 3 of SAP 2000-3T2, 'NCTL Verification Checkllst." This revision provided more explicitly detailed guidance on how to perform the verification of minimum required burnups on BAP 2000-3T1 .

On & February, 1995, Byron Station OSR approved revision 2 of BAP 2000-3T1. This revision added more documentation of information so that minirr*im required burnups could be more readily end accurately determined.

FRoM U.S. NUCuLAR ROULATORY COMMSSO LICZNSEZ NVEN? REPORT (LER)

TEXT CONTINUATION FACI.ITY NAME11 ) DOCKET LEN NMUMUS11 PAGE (3)

S.... YEA lSEQUENTIAL I REVISI

BYRON NUCLEAR POWER STATION E05000454 I I NUMBEnRl ON ROF 9

- 008 - 00 F9go TEXTV mare mar eis mquWdW we .ddtabWd cope of ARC Fom 3"4t1 (1I1 B. DESCRIPTION OF EVENT (cont.)

On I March, 1995, all TS manual holders were Instructed. In a letter from the Byron Station Regulatory Assurance Department Supervisor, to Implement TS Amendments 67, 68, and S9. At this time, assemblies F37E. F44E, and GO7F, weremin Region 2 and were in violation of TS 5.6.1.1. Each had been previously approved for residence In Region 2 using a revision of BAP 2000-3A1 which reflected an earlier TS amendment.

On 17 August, 1995, Byron Station OSR approved revision 3 of SAP 2000-3A1. This revision was processed due to TS Amendment 68 changing the minimum required burnup curve. The procedure now exactly matched TS Figure 5.8-1, requiring 32651 MWd/MTU for an Initial enrichment of 3.8 wt% U-235. Again, Byron Station took credit for the review performed In association with OSR 94-078 to verify compliance of the Incumbent fuel assemblies. Two fuel assemblies were moved Into SFP Region 2 since implementation of TS Amendment 88 on 1 March, 1995. They were moved from failed fuel canisters on 1 June and 29 June. Both assemblies met the minimum burnup requirement.

On 24 May, 1996, while performinq SAP 2000-3T1 for fuel assemblies anticipated to be moved In association with upcoming spent fuel storage rack neutron attenuation testing, Byron Station nuclea engineers (engineers 7 and 9) found Indications that fuel assemblies F37E and F44E did not meet the mir*nmum burnup as required by TS 5.6.1.1 .b.2.s, "Fuel Storage - Region 2. Nor were these two assemblies stored in a checkerboard pattern as &Mowed by TS 5.6.1.1.b.2.b, Fuel Storage.- Region 2. Byron Station contacted NFS for verification of ectue burnup and minimum requred burnup and to assist the investigation Into whether tlhese fuel assemblies werefIncorrectly resid in Regio "n "2.

On 26 May. 1996, while performing SAP 2000-3T! for fuel assemblies anticipated to be moved In association with upcoming. spent fuel storage rack neutron attenuation tasting; Byron Station nuclear engineers (engineers 7 wnd 9) found Indications that fuel assembly G67F did not meat the minimum burnup as reqired by TS 5.6,1.1 ,b.2.a. Nor was this assembly stored in- checkerboard pattern as allowed by TS 5.8.1.1.b.2.b. Byron Station again contacted NFS for verification of actual burnup and minimum required burnup and to include this fuel assembly in the Investigation.

On 28 May. Byron Station nuclear engineers (engineers 7, 9 and the acting SRE) and NFS held a conference call discussing the results of the NFS Investigation into fuel assemblies F37E, F44E, and G67F. It was determined at 17:00 tht* Ai three assemblies wert in violation of TS 5.6.1. 1.b.2.

C. CAUSE OF EVENT:

The cause of F37E and F44E being Incorrectly stored in Region 2 was cognitive personnel error. The data used by the computer spreadsheet for verifying minknum required burnup was not entered correctly nor was it independently verified to be sccurate. The spreadsheet data failed to show that F37E and F44E were In SFP Region 2. Furthermore, the spreadsheet data failed to use the correct bwnup values for F37E and F44E. This resulted in assemblies F37E and F44E producing erroneous "OK" spreadsheet outputs. This faulty technical review was part of the basis for the Byron Station OSR 95-008 approval and acceptance of TS Amendment

68. The amendment was then implemented with plant conditions not conforming to the new requirements.

wORM 346A W1 U.S. MtUJC*1M REGULATORY CO S LZCZNShN EVZNT REPORT (LER)

TEXT CONTINUATION SFACMMI NAME 11 DOCKET LEM NMU , (6) PAGE 13 I LI  : H "T

j =, 1=1 BYRON NUCLEAR POWER STATION 05000454 M OF 9

'r- o08 -ooj 5 TEXT V me 'aim @d, eA/CFan 366i4 (177 C. CAUSE OF EVENT (cont.I The cause of G67F being Incorreafy stored In Region 2 was also cognitive personnel error. Personnel approving the NCTL to place G67F in SFP Region 2 failed to use the current procedure revision of BAP 2000 3A1 to verity that G67F had accrued the minimum required burnup for uncheckerboarded Region 2 storage.

The previous revision that was used did not reflect current plant conditions. This resulted in an ineligible fuel assembly being placed into Region 2.

D. SAFETY ANALYSIS:

The SFP condition throughout this event was bounded by the two criticality analyses used as the bases for TS Figure 5.6-1 prior to and after Amendment 68. All uncheckarboarded fuel assemblies, including F37E. F44E, and G87F, met the minimum bunup requirements of those analyses. However, the SFP condition failed to meet the current TS requirement, which was 3% greater than the current criticality analysis.

UFSAR section 9.1.3.2 addresses the safety evaluation far storing spent fuel in the SFP. The criticality portion Is based on the 'Byron and Braldwood Spent Fuel Rack Ciitialfty Analysis Considering Boraflex Gaps and Shrlinkea" document from Westinghouse dated June, 1994, as amended by 94CBS-G-0105 and 94C*'-G 0142. Section 6.0. Discussion of Postulateo Accidents,'addresses an abnormal condition where rescluvy would increase beyond the analyzed condition: a fuel assembly is misloaded into Region 2 which does not satisfy the requirements.

While, in the scenario considered, only one assembly is misloaded, the analysis makes several conservative assumptions:

I. All fuel assemblies contain U-235 at the nominal ervichment or its equivalent at the minimum required burnup.

2. All fuel assemblies are uwformly enriched. No credit is taken for reduced-enrichment or natural uranium axial blankets.

3. No credit is taken for -U234,U-236, or any fission product poisons. No credit is taken for any burnable absorber material which may reman in the fuel.

4. All storage locations are loaded with fuel assemblies not containing any absorption material.

5. The storage locations an infinite in lateral extent.

6. The array is moderated by pure water of 1.0 g/cc.

7. A conservative Boraflex degradation model is assumed.

8. The scenario where a frash assembly with an ennchment of 4.2 wt% is inserted into B 5x5 array of the nominal assemblies is considered.

-:C FORM 36M U.S. NUCLEAR REGULATORY COMMISSON LICERUSE EVZNT REPORT (LER)

TEXT CONTINUATION FACITY NAME 41 DOCKET LER NUMBER, I) PAGE (31 lEM 3EMM WI BYRON NUCLEAR POWER STATION 05000454 W--

! IM 7 OF 9 9I - 008 - 00 TEXT (If rWne *OecE is WAquMO, use &M~ObnoIcO'W ofMRC Form 3$6AI 1171 D- .Safety Analysis Scont.)

The maximum k, at a 95% probability with 95% confidence and Including the statistical summation of independent uncertainties is 0.9449 for Region 2 under the nominal conditions. The increase in reactivity due to the misloeded assembly is no more than 0.0438 delta k. However, only a single failure must be accounted for, so soluble boron may be credited. The reactivity from 300 ppm boron is approximaely -0.06 delta k. more then offsetting the increase from the misloading. Thus, the r,, limit of 0.95 required by TS 5.6.1.1 Is not challenged during this abnormal condition.

The situation described in this report, with three fuel assemblies misloaded rather than just one, is more conservative than the accident analysis due to the following considerations:

I. Nearly ell fuel assemblies residing in Region 2 exceed the minimum burnup requirement, making them less reactive than the reference assemblies.

2. Many fuel assemblies have reduced-enrichment or natural uranium axial blankets of six inches at both ends, reducing their reac.iviti.es.

3.. All fuel assemblles contain U-234 and U-236, and spent assemblies contain fission product poisons as wel. "These materias further reduce reactivity.

4., Not every storage location contains fuel. Locally, there awe several empty locations. Some of the fuel assemblies contain absorber material such as rod cluster control assemblies (RCCAM).

5. The SFP is finite, exhibiting nonzero neutron leakage at the boundaries.

6. The water in the SFP is normally approximately 80 degF, having a density less than 1.0 gicc. Soluble boron concentration in the SFP remained greater than 1280 ppm since January, 1995, providing at least -0.22 delta k reactivity.

7. Previous neutron attenuation testing results imply that the Boraflex in Region 2 '- as not deteriorated to the extent assumed in the ainvysis.

8. The improperly located fuel assemblies are significantly less reactive than the fresh 4.2 wt% enriched assembly assumed in the accident analysis. Fuel assemblies F37E, F44E, and G87F fell short of the required burnup by 3 MWd/MTU, 13 MWdtMTU, and 43 MWd/MTU respectively. These values are within approximately 0.1 % of the required burnup values.

The combination of the above factors ensured that the k,* limrit of 0.95 required by T1 5.6.1.1 was not challenged during this event.

U.S. NUCLEAR REGI)LAT0UY C011111111000 364A 040 LICENSZE EVENT REPORT (LER)

TEXT CONTI NUATION LMR NUMBEM6 PAO& 131 FA(IT NIF (1) DOCKET 95 00 0 8 OF 9 NUCLEAR POWER STATION 0500054 EBYRON of M7 'Form3604-1(17) tawC hfinWw Spam is reqr*4d use &ddftfnWd 00a" E. CORRECTIVE ACTIONS:

three engineers initiated PIF 454-180-96-0008, identifying On 28 May, 1996, at 17:15, Byron Station nuclear2 of the SFP. Byron Station Regulatory Assurance, fuel assemblies meppropristely residing in Region also were notified. The NRC Resident Inspector was Operations, and System Engineering management notified.

in identifying possible inadequacies and inconsistencies Concurrently, NFS initiated PIF 901-201-96-07800 The investigation results show 2 candidate fuel assemblies.

their methods of determining eligibility of Region did not contribute to the root causes of this event.

that these inadequacies and inconsistencies and G67F into 291 May, 1998, at 05:15, Byron Station fuel handlers moved fuel assemblies F37E, F44E.

On in accordance with page 96-103 of an approved PWR SFP storage locations in Region 1. This was done Station Nuclear Component Transfer Ust.

.68 assemblies residing in Region 2 using TS Amendment NFS subsequently performed a review of all fuel anod PSSCN:96-023. it consied of a listof eVery criteri. -This review was transmitted as NFS:PSS98-1421996,.and identified which assemblies had achieved March.

fuel assembly in.the Byron Station SFP as of,31 burnup for Region 2 storage., Byron Station aengineers 7 and 9 the.n verfied tatthose-.

the minimum requted There stored in Region 1 or ia checkerboard pattern.

assemblies not meeting minimum burnup were either 2. . All fuel moves i 2 oInnto Region performed since "31 were no assemblies stored inappropriately in Region accordance with

.AP 2000-3A- Rev 3.

verified in March, 1996, have had eligibility requirements in place and provides explicit guidance on the preparation and Independent SAP 2000-3T2 Rev 3 is currently was not in place at the times F37E, F44E, and GS7F were review of BAP 2000-3T1 Rev. 2. This revision presents an additional barrier to Tha guidance provided approved for u.checkerboard. d Region 2 storage. this event.

miulocating a fuel assembly that could have prevented andi provides improved documentawofl of miriiinum required burnup BAP 2000-3T1 Rev. 2 is currentlyto inorplace within Region 2. This revision was not in place at the times F37E, FEE for fuel assemblies being moved initial Region 2 storage. The improved documentation shows and GB7F were approved for uncheckerboarded an additional accrued burnup for each assembly and prdsents enrichment, minimum required burnup, and actual have prevented this event.

barrier to mislocating a fuel assembly that could to the requirements of TS Figure 5.6-1 Amendment BAP 2000-3A1 Rev. 3 is currently in place and is identical Region 2 storage eligibility. All future fuel assemblies 68 us well as the current NFS method of determining this required burnups determined in accordance with ap~proved toe Region 2 storage will have minimum 5.6-1 will have a concurrent changing TS Figure procedure or its equivalent. Any future TS Amendment new requirements. This presents an additional it reflecting the revision to BAP 2000-3A1 ansociated with have prevented this event.

barrier to mislocating a fuel assembly that could

!o this with nersons involved in the errors that contribulixii Performance expectations have been discussed everit.

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

book. NTS item 454-201-96-0008-01 tracks completion

fom 3N8" U.S. WUCEAR REGULATORY COMMSION LIC"Sii EVXNT REPORT (LiR)

TEXT CONTINUATION FACILT MAMl Mu DOCKET LER NMIM 8N PAGE 131 I I III I I IVUOWI BYRON NUCLEAR POWER STATION 05000464 so 008 - 00 Ell 9 OF 9 yETe'N ,ec i rp~d~ eedUs@ c of*M NR,*C Fm 35WA 11T3 F. RECURRING EVENTS SEARCH AND ANALYSIS:

LE5t 454:94-006, 'Fuel Assembly Located in Wrong Region of Spent Fuel Pool due to Personnel Error,"

documents a similar event. On 15 July, 1994, SED found a fuel assembly in Region 2 that neither met the minimum burnup requirements of TS Figure 6.6-1 nor was checkerboarded. The cause of this event was determined to be cognitive personnel errors. The Nuclear Materials Custodian and an independent reviewer failed to use the approved method to verify assemblies met the minimum burnup requirements for storage in Region 2.

Although the 454:94-006 event resulted in a fuel assembly incorrectly residing in SFP Region 2, the circumstances leading to this event were different from those leading to the 454-180-96-0008 event.

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)

SIGNP-TURE PAGE FOR~ LICENSE EVENT REPORT LER Numiber 454 :2~0S-pUel A@SMeablv in Wroncl t Fuel Pool due to aoai~S TitIO of Lvent:

Piep~ffle Eror occurred: 09-1 DaLto Time 46-tc 6 USES ~¶ OSR DISCIPLI~ES;REQUIRED-Acceptance by Stat ion Review:

D-a t.e Date OE -ýýSES Date crr~~**C, Date RASS Approv:ed by I A s-'ation nager a

UCENSEE EVENT REPORT UlER) lootT oIUTYNUMIBERor PAGE STM M)UCLE4R POWER TAT10U 00 5 1 tiGI 0 f7 FUEL ASSEMBLY LOCATED IN WRONG REGION OF SPENT FUEL POOL DUE TO PERSONIAEL ERROR M1WT DATE, t NUIUR PUiIT OATE ,Th,0 FkcLJtIES IVLVWD AT G~w VAR~ TEAR ýMA REu=1 ajUTHDYTEA F*CUTIEMW 7S WANU ONCE1T(9 Tl1s WREIT ISOMSTID PJSUINT TO TI REGOUMENTS OF 10 CFO IC ICK1OE 01 MONI O THE FWDOM OFERATWO MaW ,* - 421 i

]9.4O . ]2-a7y.4W , , 0.f.Tlmltrd* 71.7n ILWI-20.40tI -N57 W73"d1d L~d5 s IIw 7)MmsiI

,1 . %PC FW 51A)

II .4O5Ia*lMl~~ 5TO.3IJdII** 50 7bU~id2l]Ii*e9CIp* bI.)~

2@I4OAgOlt3 5Q.7).lt2Ni 50.7*U2NmJlI2 LICISEE CONIACT FOR T14iS LEN1 s IME ThLEPL4OE 1NUMBIR G'STAUFFER. STATIONREACTOR ENGINEER. X2249.. R CAauS SYSTEM COMPONENT

"~ ~~~~~~

MTArACTser R!PORTAIU jN ,  :

CAUSE SYSTEM IiI CoMP9WE

~

I II ~

oAIIUoACTURER 3s EWK

"

II

..IS141411 I*IS'I* Ir~COKET 14JEAC.Ci*WN ItAFiK OLWp LIN FAIIOUR( KSCI. OftlM**~

Tl145l ORT.lqlm*

SUPPLEMENTAt REP"RT t(PEC HO (XPECTEO PINTH DAY YEAR SUBMISSO 0f. yes tuppokit fXPECTE0 SUE)MSISON OKATi 170 DATE 111B1STRAC1'OLimtoi 1400 tpecn. u-.g rsnaa il. smagiwiisce ?ypewraie. looiss On July I 5, 1994. Saystem Engineering Department (SED) found fuel assembly U38J located in Region 11of the Spent Fuel Pool (SFP). The fuel assembly dio not meet the burnup requirements SPecitiea or. Technical Specifications ITS)

'Section 5. 'Design Features,' Figure 5.6-.1'Minimum Burnup Versus Initial Enrichment for Region RlStorage." The fiuclear Component Transfer Lost INCTL) incorrectly specified the placement of U38J into Region 11ar location HM5 The NCTL also did not place the assembly into. Region II in a checkerboard pattern. Administrative -ontrols require an assembly ttiat does not meet minimum burnup to be placed into Region II in a checkerboard pattern The assembly was placed into the incorrect region of the SFP on September 26. 1993 during a refueling outage or, Unir 2.

The error was discovered while preparing for the next refueing outage The assembly was moved m Region I on July 16. 1594.

This eveItinvolved no safety con-erns. The safety significance of the misplaced assembly is withir the safety analysis presented in the UFSAR. This event is reportable in accordance with 10CFR 50.731al(21(,' ) Any operatio or condiliort prohibited by the plant's Technical Specific.4tions

0 1 00 4 15 4 1 V I jiI I W I " . . . . ..

TIE avo'I ATIE ficawnin*SYsim (11151 Ca mw,od swiatelied a 00IX]

A. PLANT CONDITIONS PRIOR TO EVENT; Event Date/rime 07 I_.QL./ .

RCS [AB) Temperature/Pressure NOT/NQ P Unit 1 MODE 1 - Power gDeion Rx Power 80% in coastdo-wn Rx Power .. RCS IABI TemperaturelPressure NOT/NOP Unit 2 MODE I - PowerLOeradio B. 2QRPTION OF EVENT; Nuclear a non-licensed engineer (Engineer 1 I completed the Between mid-August. 1993 and September 10. 1993, core (Page numbers93-121 to 93-146). This 2 reactor Component Transfer Lists (NCTLs) for offloading the Unit writing of the NCTLs. he ra "-tIwo, individual was the station's Nuclear Materials Custodian or NMC. During the location HM10. This location is in U29J going to storage errors. On page 93-139, the NCTL shows fuel assembly at me time tne-i C wrote the list, did not meet the minimum a Region II rack. The burnup of the assembly. U38J on II. The NMC made a similar mistake for fuel assembly bumup requiremeht for placement into Region of 32.540 MWD/MTU.

MWD/MTU versus a required burnup page 93-143. The aictual burnup of U3BJ was 29770 a Region II rack. Both.

location is also in

.The NCTL shows-assemblt U38J going to storage location HM5. This errors were cognitive personnell errors.

Refueling Outage B2RO4,. he completed Byron Administrative After the NMC wrote the NCTL for the offload, for I of the Transfer List INCTL) Verification Checklist.' Step Procedure (SAP) BAP 2000-3T2. 'Nuclear Component to verify that.

checklist requires the preparer of the NCTL racks meet minimum burnup requirements as "Fuel assemblies entering Region II of the spent fuel configuration. Records of assemblies described in BAP 2000-3AI or are placed into a checkerboard in file 1.02.1080. which is in the NMC satellite file which meet minimum burnup requirements are kept cabinet.'

are documented on BAP 2000-3-T1. "Spent Fuel Burnup Records of assemblies that meet minimum burnup 1.02.1080. BAP 2000-3Al's title is, -Minimum Required Verification Checklist,' and are kept in file location gives Density Spent Fuel Storage Racks." This attachment Burnup as a Function of Enrichment for Region II High burnup required for storage in a Region II rack.

a listing of initial enrichment versus the minimum to of Nuclear Fuel Within a Station." requires the NMC BAP 2000-3. 'Safeguarding and Controlling Movements Pit ;SFP). The NIMC into Region II of the Spent Fuel complete BAP 2000-3-Ti for each assembly to be placed forms for assemblies placed into Region It during Outage B2R04. The started but did not complete these of this investigation.

BAP 2000-3-TI form was comoleted as part from the core. TOTE is a computer program that The -NMC used the TOTE data 'or all the assemblies discharged is accumulated burnup for each fuel assembly. The data calculates assembly burnup. TOTE data gives the total computer. The NMC used the IBM and mentally stored on the IBM mainframe and is accessible via a personal the information on BAP 2000-3.T1. Nuclear Fuel wef-t through the burnup ver ,cation. He did not complete run the code every month and after a u.nit shutdown.

Services (NFS) is responsbie 'cr running the code. They 9lR\WPF'J394

LICENSEE EVENT REPORT (LER) TEXT CONTINUATION TEAR SEQ.NUMER REVISION BMON NUCLEAR POWER STATION MXtIiT liaw"r hiury Id licntom System (EN codesm we damilld m the to es 1N

,. DESCRIPTION OF EVENT: ICont.)

Using the TOTE burnup data and the initial enrichment of each assembly, the NMC did the burnup check ussing BAP 2000-3A1. The NMC could not recall why he did not complete the forms as ;quired by procedure of why he made the error when he did the burnup checks. A review of the BAP 2000-3-Ti forms for the previous outage lFebruary 1993, BIROS) showed the NMC had completed the forms.

Discussions with the NMC identified several weaknesses in the NCTL writing process. The process is very complicated and relies heavily on the skills of the individual writing the NCTLs. The Verification ChecKlist gives criteria that the NCTLs must meet. 'However, the checklist does not describe the process on 'how to" write the NCTLs. The NMC divided the process into three major sequences: the offload, the insert shuffle, and the onload. The process as described by the NMC is given below.

First, the NMC does a comparison between the candidate loading pattern supplied by NFS. and the existinbg core loading pattern. The candidate loading pattern shows the next cycle's core loading pattern. The NMC obtained the existing loading pattern from the tagboard foi the. Unit 2 reactor core. The tagboards are located in the area where the NMC sits, The tagboards are use j to show the location of every fuel assembly and component in the SFP. the New Fuel Storage Racks, Failed Fuel Storage'Racks. and the two reactor cores. The tagboards mimic the physical layouts of each of these areas of the plant. And, the NMC keeps them up-to-date based on completed NCTLs.

Once he completed this comparison, he placed each assembly into categories., He based the categories on the insert a fuel assembly contained in the current c6cle and the insert the fuel assembly would have in the nex; cycle.

In other words, categories of assemblies are based on what they "have* and what they are "getting.' For this event, there were nine different categories. For example, assemblies that nave burnable poisons (BPs) that are getting thimble piug s (TPs) (BPs to TPsi, assemblies that have control rods (RCCAsI and are getting thimble plugs (RCCAs to TPs), and assemblies that have thimble plugs and are getting control rods (TPs to RCCAsI.

Next. the NMC arranged the categories side-by-side in the SFP such that the insert swaps can occur with the least amount of yool changes. There are five major steps to the insert shuffle.

The NfIC did this arrangement in the SFP by iteration until he obtained the most efficient layct. After the arrangement in the SFP is done, the NMC can begin writing the offload. The NMC wrote the offload such that the fuel assemblies were placed into the first open location in each of the nine categories. As he wrote the offload sequence, the NMC also ensured that each step met seven requirements and three optional items.

The N.fC went through a similar process to %..itethe insert swaps and the core onfoad sequences. In all. there v ,e eleven required checks and four desirable items for the entire refueling. During discussions, the NMC identilied an additional four criteria he met while writing the NCTLs, that were not part of BAP 2000-3T2. This brought the total number of checks the NMC met to nineteen, After the NM- wrote the three malcr sequences, they were loaded into a computer program called Shuffle Works This program .vrote the sequence on NCTL forms that the Fuel Handlers used in the field. A member o, the SED rnuclear group entered the offload and insert shuffle into the program step-by-step. This was done because Shuff' Works could not perform all of the required checks, However, the program did write the onload swduence since contained II the pool configuration after the core was offloaded and all the insert shuffl-s wvere, dime. 2' the :meal core configuration, and J) the loading sequence. Because the program had this *i- rfrriation. it. I),, leajt. \. roye tie seqiuence meeting all the aopropriaip reqriremrenls

- I I r), t.:Dc 4

LICENSEE EVENT REPORT ILER) TEXT CONTINUATION FACIJTY NAME 00CXET NUMBER LER NUMBER AGZE YER .CNUMBER IRE VISON O" NUCKL gvR EARP WU R STATiON 01 51 01 91 D 6 1 41 -- Ia1 0 a1 4 O TEXT EAMw IwIy Myikdiketwe S3"1,0m0I2, *aod~endi I DeirIgo aid0 OI1 B. DESCRIPTION OF EVENT: IContJ After the NMC wrote the NCTLs, he gave them to an independent reviewer on September 10. 1993. The independent reviewer was a non-licensed engineer (Engineer 2). Engineer 2 did not use the records of assembires that meet minimum burnup requirements to verify certain assemblies could be placid into Region I1. 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 (NFSI. NFS sent a letter that listed assemblies by region and indicated which assemblies met the minimum burnup requirement for storage in Region )I. 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 failed to find the error for assemblv U38J on page 93-143. He notified the NMC of the error for U29J and the NMC wrote a variation to the rK; I L. Engineer 2 did not discover the second error and stated that the cause of the error was most likely dut 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 him to miss the burnup check for assembly U38J.

Engineer 2 and the NMC both signed the verification checklist on September 13, 1 993..

Discussions with Engineer 2 indicated that there have beer errors inlp.st NCTLs-but they had been ca*gjht 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 Hanullers placed assembly U38J into a Region It rack on September 26. 1993 in accordance with the NCTL.

On July 15. 1994. a non-licensed engineer (Engineer 3) discovered that fuel "ssernbly U38J was in a Region II spent Tuei rack. Ine fuel assembly had been in the Region It rack since September 26, 1993. The Fuel Handlers had placed the 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 Enrichmev for Region II Stordoe."

Engineer 3 discovered the error during preparations for moving fuel assemblies from Region I to Region II for the upcoming refueling outage on Unit 1 (B1 R06)_ 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. Ft Handlers moved fuel assembly U38J to Region I following an approved Nuclear Component Transfer List (NCTL)

The Fuel Handlers moved the assembly into Region I on Ju!y 16. 1994.

This event did not involve any inoperable systems aiid 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(al(21h)iBI. any operation or condition prohibited by the plant's Technical Specifications i9931R WWP! .080814-5?

LICENSEE EVENT REPORT (LER) TEXT CONTINUATION F ACU.ITY NAME DOCKET NUMBER LER NUMBSR PAGE TERSQ. NUIJMME ElZO BIRON NUCLEAR POkMR STATION BEI i ll TEXT EnMIi,.iutrvnI tneila $ro S onm frus) gd Wo xi ,o jo The itil ar.XI C. CAUSE OF EVEN':

The primary causes of this event were cognitive personnel errors. Both the NMC and the independent reviewer failed to use the approved method to verify assemblies meet the minimum burnup requirements for storage in Region II racks. it should be noted that use of the approved method would not guarantee this mistake would not recur because of a procedural weakness. The procedure wi!' be enhanced. There were also several contributing causal factors for this event that led to the cognitive personnel errors.

The current methodology for writing NCTLs is not well defined and relieb he:;.. or' =ne skills of the preparer. The preparer goes through many manual iterations on the NCTL until the mcst efficient sequence is found. This method is nut conducive to minimizing human error.

The methods to be used for verification are also not well defined. Many verification steps required by SAP 2000-3T2 can be done in several diffarcnt ways, as occurred during this event. And, some methods may not be as effective as others in catching errors or for performing verifications.

By not writing PIFs for failures found during independent verifications, the ibility to find and correct oroblems before they result iri higher level events such as an LER was minimized.

Although the Shuffle Works program is an effective program for its intended purpose, enhancement of the Shuffle Works program could help prevent errors of this type in the futu e "

A corrective action from a previous event was ineffective. Refer to the Recurring Events Search and Analysis sectiom for an explanation.

D. SAFETY ANALYSIS UFSAR Section 9.1.2.3. 'Safety Evaluation., says that "The largest reactivity increase occurs from accidentally placing a new fuel assembly into a Region It !Atorage cell with all other cells fully loaded. Under this condition, the presence of 300 ppm soluble boron assures that the infinite multiplication factor would not exceed the design basis reactivity for Region II With the recommended concentration soluble poison present (2000 ppm boron),

the maximum reactivity, K., is less than 0.95 even if Region It were to be fully loaded with fresh fuel of 4.2%

enrichment.

Byron Station normally maintair.s the boron concentration in the SFP at two thousand ppm and administrativeb*

controls the concentration to gre; ter than eight hundred ppm. At the time it was p~aced into tne SFP. fuel assembly U38J had a burnup of "9. 770 MegaWatt-Days per Metric. Ton-Uranium (MWDfMTIJ and an initial enrichment of 3.802% Thpreiore, the UFSAR analysis bounds the misplaced assembly and no safety significance existed while the asseinbiv was in the Region II iack.

~R i~ ! W;1 F

LICENSEE EVENT REPORT (LER) TEXT CONTINUATION FL.LT WiK VWCKIt iWM[R LIA W A.,

TlIAR _ a. JMluI4J R EYSP 3IM30 MM"tEAR POMR STATION

.. .ll l0 l00l l 4 ,I4 - I i olo 0i, TEXT Etsvp hIo" weawfat SyeM 1Os Mil we d.,iEdd a iuuto ai rfEI E. CORRECTIVE ACTIONS; Corrective Actions - Lonr Term

1. The NMC and the individual that performed .the independent verification were counseled.

2. The SED nuclear group will write a procedure that explains the methodology to be used to write NCTLs for a refueling operation. In addition, this instruction will give directions on when in the process verifications will be done and the preferred method for performing the verifications. NTS item 454-180-94-00600-01 tracks completion of this item.

3. The SED Nuclear group will determine the preferred method for performing each verification on the Nuclear Component Transfer List INCTLI Verification Checklist. BAP 2000-3-T I. SED wilg revise the checklist to:

al explicitly define the preferred method of each verification.

b) indicate whether alternate methods are allowed and explicitly define these alternate methods. These "methods will'be equivalent to the preterred method.,

c) organize the chocklist to distinguish important checks from less important checks.

d) provide cautions describing the pitfalls for each method.

NTS item 454-180-94-00600-02 tracks the completion of these items.

4. The SED nuclear group will pursue revisions to the Shuffle Works program that will allow it to perform more of the verifications the nuclear group presently does manually. NTS item 454-180-94-00600-03 tracks completion of this item.

5. Regulatory Assurance will issue PIF threshold guidelines that will require writing PlFs for errors caught during independent reviews. NTS item 454-180-94-00600-04 tracks completion of this item.

6. The SEQ nuclear group will revise the BAP 2000-3-TI fornim to include a column for recording both the assembly's burnup in addition to the minimum required burnup for storage in Region II.

NTS item 4F 1-180-94-00600-05 tracks completion of this item.

7. The SEO nuclear group will revise BAP 2000-3 to require a walkthrough of the entire refueling on -paper' tagboards. NTS item 454.180-94-00600-06 tracks completion of this item.

Interim corrective actions for the upcoming refueling outage on Unit I:

a The Station Reactor Engineer will discuss this event w;th al' members of the N'jclear Group and place this LER in the Nuclear Group Required Reading NTS item 0 454-180-94 0060007 tracks this item.

b. BAP 2000 3.T1 will be used prior to moving any fuEl into Region 1l This is presently a requirement of BAP 2000-3, so no NTS item is needed to track this action C A 'paper- tagboard will be used for a step by step waikthrough of the entire refuPling procedure NTS item 454 180-94-006-08 tracks this item 9qI' p' WPF - 4 "

LICENSEE EVENT REPORT (LER) TEXT C0i3tTINUATION

KAtI V OOCXFT IUMBER LEP W uERn

• TUIJC IRYALEAR " STATo rfirV EaW Mdfy UWAW&VOR

- S5a(drPd onm w d . t% ",Ip fill F. RECURRING EVENTS SEARCr AND ANALYSIS:

A search on ETS found one previous e-.,nt of a misplaced fuel assemnbly due to an error in an NCTL.

DVR 6-1 071. 'Fuel Transfer List Error.' documents this event. A review of the correctte, actions for this event Indicated that one of the corrective actions was not imclemented Corrective action to prevent recurrence.

fitm 28, states.

"BAP 2000.3 will be revised to require The use of a procpdijral checklist when developing the NCTL.

ThN. list will include."

"B The requirement to use a tag board Cprrently. SAP 2000 3 does not contain this requireiment. Discussions with the Station Re.actor Engineer MSAE).

at rho time of the event. Indicated that t*fis correcttve actiOn required a step by-step waikthrough of the entire refuehon evolution on the tagboards Howvevmr. Engineer 2 indicated that the intent of the corrective act*ni changed The intent changed to the use of a 'paw' tagboard as ormOsed to the use of-the physical tagbo1ards This would eliminate possb*.l errors from moving chips on tho r0 siil f.ai anagrds o It cannot be 4flerrmved why this rqiujitement was not mncorporatell into SAP 2000 3 A review of tri PJTS item Written to track complet"MO OF this corrective actxm in~dicated that the NrTS van not sapftir on esactiv wlat changes to SAP 2000 J weM needed. The NTS item simply stated to "devvelop apricuidirprcheckst th,-? %0ec!*U!s hoIw to prepare ian. CT..

At the"Ctime thoi checklist was develped. 'r f.11led to. incorporato this AP 2W00' 13ii~nsn~it J hmuorwfe MhIS c.orrectliv, action was ineffective G. COMPONENT FAILURE DATA.,

Thaerp vyj% no fasilpr r ompnflfl-t dtir,r, th, -,o

EXHIBIT B-7 Catawba Unit 1:

LER 413/90-0160-00 (April 19, 1990)

I Duke Power Company t(O.f) YQ-/4000 Catawb'a ,Yucleur Station P 0 Bar 256 Clocer. S.C 29710 m:,"! :-7 DUKE POWER "90 t.*-*' 11 Aý03:Cl2 Apr:il 18, 1990

5 Document Control Desk U. S. Nuclear Regulatory Commission Washington, D. C. 20555

Subject:

Catawba Nuclear Station Docket No 50-413 i6'i " ...

LER 413/91 I Gentlemen:

Attached is Licensee Event Report 413/90-16 concerning TECHNICAL SPECIFICATION VIOLATION AS A RESULT OF A MISSED REFUELING WATER STORAGE TANK SAMPLE DUE TO INAPPROPRIATE ACTION.

This event was considered to be of no significance with respect to the health and safety of the public.

"Very truly yours, (C

Tony B. Owen Station Manager kebNLER-NRC.TBO xc: Mr. S. D. Ebneter American Nuclear Insurers Regional Administrator, Region II c/o Dottie Sherman, ANI Library U. S. Nuclear Regulator Commission The Exchange, Suite 245 101 Marietta Street, NW, Suite 2900 270 Farmington Avenue Atlanta, GA 30323 Farmington, CT_ 06032 M & M Nuclear Consultants Mr. K. Jabbour 1221 Avenues of the Americas U. S. Nuclear Regulatory Co-mmission New York, NY 10020 Office of Nuclear Reactor Regulation Washington, D. C. 20555 INPO Records Center Suite 1500 Mr. W. T. Orders 1100 Circle 75 Parkway NRC Resident Inspector Atlanta, GA 30339 Catawba Nuclear Station

N ",;- 3@4

-C U.S NUCLEAR REGULATORY COMMISSION (9-3)APPOVED 0MS N40. 31-18"M LICENSEE EVENT REPORT (LER) EXPIRES: E/31I/

FACILITY NAMI 1) OCIKET NUMIBftER PA Catawba Nuclear Station, Unit 1 10 I5 1 0 0 10 14 113 I FoF 019 TITLE " Technical Specification Violation As A Result Of A Missed Refueling Water Storage Tank Sample Due To Inappropriate Action EVEMIT DATE (54 LER NUMBER IW REPORT DATE 171 OTHER FACILITIES INVOLVED 1SI MONTH DAY YEAR YEAR I RVI MONTH DAY YEAR FACILtTY NAMES OOCKET NUMBERISI N/A 0 510 10 101 1 S301o 5 90 910 - 01116 0 10 014 119 9 0 1510 0 101 rho falawmql (111 OPERATING THIS REPORT IS SIUMITTED PURSUANT TO THE REQUIREMENTS OF 10 CPR §: (Chc* anr or nort of Moog (le 15 20.4021b11 046. B.2alI,, 37 b POWER 20.441II61) SO.3E1.II1 1)73I2II~I 73.71 g)

LEVEL LI0 L0 20.4006(llI1141 50.3640121 50.73(1-iziliii OTHER fS".CW'y ,n Abs &7t

- j bdow..sd ,n Tet. NRC Fopr, 21*4*(*)

(1 (lX 5'0.73411(21411 50.13(ajl(lI~i*)lA1 36SA/

• 20.4061011111i41 O.S.(I1l,Hv w jjl 50.731*1121(IN?.

SO.?3(EII(21(I4IF- 71121,lIA E.IeZIlH 5A LICENSEE CONTACT FOR THIS LEM 112)

NAME TELEPHONE NUMBER AREA CODE R.M. Glover, Compliance Manager 810 p31[ 13 1i-I 3121316 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE ODECAlIEO IN THIS REPORT 113) 1 YTMMANUFAC.

CAS REPORTABL UESTM COPN T MANUFAC. RIEPORTABLE::. ..

CAUSE SYSTEM COMPONENT TUREA _To NPRO  : CAUSE SYSTEM COMPONENT TURER TO NPRDS

..........

......

___ _..

SUPPLEMENTAL REPORT EXPICTED 1141 TEAR MONTH O.ý EXPECTEO SUBMISSION DATE (15)

YES (tope.

j m ao, e EXPICTEO SUBMISIWON DATE) NO ABSTRACT LimlA to 1400 wee, i.e.. o.pv i-.Wr HAP siW'f* -w cr,

• typ.awonw iiuea 011)

During the period of February 5 through 26, 1990, samples for the Boric Acid Tank (BAT) and the 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. CHM contacted the Control Room Operator (CRO) 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 cause is assigned to deficient communications resulting from poor group interface between CHM and OPS. Corrective actions taken included CHM procedure revisions which will supply actions to take when T/S samples cannot be obtained 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.

U.S. NUCLEAR REGULATORY CO.MIW"ON MRC FW, 3" APPROVEo 0MB NO. 33150-104 LICENSEE EVENT REPORT (LER) TEXT CONTINUATION E1PIRES: 5131/1 NNUMBER(is PAGE 133 DOCKET NUMUER (21 VA-CILITY NAME (U1 o 15 1O 101o1 41t1j3 O9il 1-6-- 0 0 0l 2 OF O 0II Catawba Nuclear Station, Unit 1 rut-E(11 iw06w it .u NAC FM 364'l 4 BACKGROUND REFUELING WATER SYSTEM of borated (FW) System provides a large source The Refueling Water (EIIS:CBI to:

water and the necessary equipment System (ECCS) and the

1. Supply the Emergency Core Cooling System during the injection Containment Spray [EIIS:BE] (NS)

Accident (LOCA);

phase following a Loss of Coolant the Refueling Water Storage

2. Transfer the borated water between Tank (FWST) and Refueling Cavity; water by routing the water

3. Provide cleanup of the refueling

[EIIS:DA] (KF) System; through the Spent Fuel Pool Cooling and, water requirements and

4. Provide for various other borated miscellaneous flowpaths.

useable gallons is sufficient to provide a The FWST normal capacity of 395,000 assures:

This capacity volume exceeding 350,000 gallons.

water needed to increase the

a. The volume of borated refueling spilled water to a level that boron concentration of initially with the Reactor at assures no return to criticality rods [EIIS:RODI, except the most Cold Shutdown and all control (RCCA), inserted in the reactive Rod Cluster Control Assembly core.

to refill the Reactor vessel

b. The volume of water sufficient a LOCA.

after

[EIIS:VSL] above the nozzles (EIIS:NZL]

combined with ice melt and

c. A sufficient volume of water when spill in the containment Reactor Coolant [EIIS:ABI (NC) System to permit the initiation recirculation sump following a LOCA of the recirculation phase.

limit the radiation dose rate

d. A sufficient volume of water to Cavity to approximately 2.5 at the surface of the Refueling fuel assembly is transferred mrem/hr during the period when a over the Reactor vessel flange.

allow the station operator

e. A sufficient volume of water to valve [EIIS:V] alignment required to adequate time to complete the injection mode to the complete the switchover from the mode following a LOCA.

containment sump recirculation 0070

NRC FuMr 3J", U.JL NUCLEAR REGULATORY COUIMWION IS83 UCENSEE EVENT REPORT (LER) TEXT CONTINUATION APPROVED 08 NO. 315o-.o104 D(PlREm38/31/U FACILITY NAME 111 0GOCKET NUMBER (2) LEN NUMIER (61 PAGE I3 YEAR  ;:2;AL ION

!.*!! NUMOER rNUMSIERj Catawba Nuclear Station, Unit 1 o Isoo1olo0 111113 910 -- 01r6- 6 b0 O131OF 01 F 0 TMX M/fman pw ig iuqoWA%

_'_ A, M AM r--w M4J 117)

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

U.S. NUCLEAR REGULATORY COMMISGlON

RC Fonw 314A UCENSEE EVENT REPORT (LER) TEXT CONTINUATION APPROVEo OM, NO. 3150,14 EXPIRES

W31/8 FACILITY NAME Ill DOCKET NUMEER (2) LER NLMMEN (N PACE (M YEAR SEQUENTIAL SI4 1!3 NUNS NUME Catawba Nuclear Station, Unit I 0 151010 0 4 1 3 910 __01116 __ 0;0 0J OF 0

-TEXT fir tm e w is mqiumv um-I NAM Atw 3W'&J 4117 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 one sufficient boric acid to provide for refueling plus enough boric acid for the most reactive control rod Cold Shutdown immediately following refueling with 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:

i. 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

I NUCLEAR REGULATORY COMMISION UCENSEE EVENT REPORT (LER) TEXT CONTINUATION APPROVED COII NO. 3,150-41 EXPIRES: 8/31I/

DOCKET NUMER () LENI.*M N (Z PA G (Z p'ACILkTY NAME 11)

EUINTIAL 9OR .:::RVI!IiON YER NM BIER  ::NUE Is15 10p11 41113 91 011 0- 6 - 010 015 OF 0j Catawba Nuclear Station, Unit I 1KT (wmof ms A admd NAC Am* 3M4s IVA or the FWST, if the FWST was the seven day samples to be taken on either the BAT would need to be taken no later "declared borated water source" then the sample B that the than February 6. The OPS Shift Supervisor informed CHM Technician BAT was the borated water source.

Unit 1 was in Mode 6. All 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 />, by CHM personnel.

and analyzed required FWST and BAT samples were collected in No Mode, Core Defueled. CHM From February 11 through 26, 1990, Unit 1 was BAT and FWST samples.

personnel collected and analyzed all required CHM Technician C was informed by On February 24, 1990, Unit 1 was in No Mode.

ND Pump was on and that the the CR0 that the 1B Residual Heat Removal [EIIS:BP]

the FWST. At approximately 1439 Reactor cavity water was being pumped back to service, as a result of work hours, Diesel Generator (D/G) 1B was removed from list items related to the Outage.

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 entered Mode 6 on February 28, 1990. On OPS issued R&R 19-2838 on 1NV236B, Boric Acid to NV Unit 1 remained in Mode 6.

R&R 10-807 on A and B Boric Pumps Suction, for MOVATS testing and also issued Acid Transfer Pumps for the 1NV236B work. This action in combination with D/G by OPS that the BAT was 1B being out of service necessitated the determination, alignment. This change in inoperable, due to the unavailable BAT water source prescribed BAT sampling continued at the BAT status was unknown by CHM.

interval.

Mode 6. CHM Technician C 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 the FWST be placed in recirculation for contacted the Unit 1 CRO to request that was currently CHM Technician C was told that the FW pump the weekly sample.

and OPS was not able to state when the pump pumping down the Reactor cavity, the would be available. The CRO would check with the Shift Supervisor about at 0832 hours0.00963 days <br />0.231 hours <br />0.00138 weeks <br />3.16576e-4 months <br />, and there had situation. CHM Technician C called the CRO again the FWST At 1930 hours0.0223 days <br />0.536 hours <br />0.00319 weeks <br />7.34365e-4 months <br />, CHM Technician D discussed been no determination made. of the cavity that the draining status with the Unit Supervisor and was advised 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 collected as a result of the FW sample for boron analysis was due, but was not The FWST was last sampled at pump being in service for Reactor cavity draining.

0050 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> on February 26.

1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br />, CHM Technician B On March 9, 1990, Unit 1 was in Mode 6, and at FWST status. The Supervisor called the Unit Supervisor and asked about the and that OPS was planning to clear stated that the FW pump had been tagged out the tagout later in the day.

fiRC Fwm~ 30OA U.S NUCLEAR REGULATOPIV Cu-aO"M e UCENSEE EVENT REPORT (LER) TEXT CONTINUATION APPROVED 0MS NO, 31,0-oo*

• ACIUTY NAME (1) DOCKET NUMGEN (2 LIN NUMU R (0 PAGQ (M "I'"" 3 Q IN IALI mum NUM99M Catawba Nuclear Station, Unit I o is 10101 411 3 910 -011 1 6- 010 01 6 OF 0 r-Wc (if mww u ,

kmfzwvu mm f MACwyw 3MW lIM AM 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 CHM Technician D and said that the recirculation pump would not operate and asked if CHM could sample off of the FW pump. The CHM 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 lB D/G being out of service, 1NV236B 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 />.

OnMarch 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.

Z CFW3 '-- U.S. NUCLEAR REGULATORY COMMISWOI4 UCENSEE EVENT REPORT (LER) TEXT CONTINUATION A,.ROVEO ows NaO.3i5io4 EXP4DME: 8/31/n DOK1 NUMR t(2 LIM NUMIER (IP PAOG ({M

1. rACIUl'y HAW (1)

YEAA V I::SEGUGNTtAL RE:Vt5I0

,-.-- NUMIER NUMIER 1 3 01 7 OF 0 Catawba Nuclear Station, Unit 1 o015 10 10 1I 1i 4 j 910 011 1 6 0. 0 T1IXT (if nw mw ,

ip m~v u m ad~ftWWC F~ew m4 417I)r CONCLUSION violation is attributed to Inappropriate Action, as This Technical Specification involved not recognizing the need to ensure an a result of the individuals The Chemistry personnel, though having contacted operable borated water source.

the FWST in recirculation for OPS personnel on numerous occasions to place resolution to the problems when continuing sampling, did not pursue a timely interferences occurred. In addition, the information discussed by CHM personnel the T/S samples, was not carried out by OPS and OPS personnel, concerning timely manner to avoid missing a T/S sample. In the past, personnel in a the boron concentrations are provided Chemistry personnel have understood that requirements of T/S 4.1.2.5.a.1. The requirements of to OPS to fulfill the to the CRO by way of the Operator Aid Computer and 4.1.2.5.a.2 & 3 are supplied in PT/l/A/4600/02 E, F, & G, Periodic Surveillance procedures.

as required Therefore, Operations is responsible for the determination of OPERABILITY as CHM personnel concluded that if OPS did not place the stated in T/S 3.1.2.5.

during the period of March I through 15, OPS must have FWST in recirculation declared borated water source. In addition, CHM had maintained the BAT as the earlier in the outage that the BAT was the borated been told by OPS personnel water source. Communication between the groups is considered a contributing cause in that it did not achieve the necessary clarity and responsiveness to avoid the T/S violation.

of D/G lB and the tagout of 1NV236B necessitated the The inoperability BAT, due to loss of its boron injection flow path. This inoperability of the based on T/S 4.1.2.1b, which requires at least once INOPERABILITY was declared flow path is in its correct position. The per 31 days that each valve in the schedule for FWST and for the BAT is established in current Chemistry sampling If this schedule is followed as stated, regardless of concerns CHM procedures.

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 the requested action is not taken by an appropriate time, a request, that, if T/S violation will occur.

Chemistry Management Procedure 3.4.17 was also changed to include statements on which states that the inability to collect a FWST and BAT sampling enclosures same as being Out-of-Spec. A T/S Operability T/S sample is considered the Notification Sheet (Attachment 1 of Station Directive 3.1.15, Activities Operations) will be issued by Chemistry with a comment that Affecting Station the T/S sample is Out-of-Spec or unattainable.

of this event, emphasis will be placed on ensuring clear As a result on clear description of needed actions and clear communication, focusing understanding of the importance of such actions.

I .. -I - 7

NOW FOI ,uu6A . U UCLEAR REOULA1OR, C M -- rWUI

• "' UCENSEE EVENT REPORT (LER) TEXT CONTINUATION *ovuoVIO NO. 3150-104 E*piRES: 5/3JI/l FCILITY NAME III DOCKET NUMBER (2) '.ER NUMBER (51 PAGE (

f' , ,IE 3:I:*U N'ALL]' ISPONr NUM26ft Catawba Nuclear Station, Unit 1 o rI la Jo to f 4_ 1 j13 91 0 __ 01 j 6 "

107) o010 0 J8 OF 0 0

TEXTf (if ffW, Vef a AWqWVud. 06MW NC Awm 34',# tj1I 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.

N etc F-wm 31W ' U.S WUCLEAR REGULATORY COMkW*WIJ UCENSEE EVENT REPORT (LER) TEXT CONTINUATION "oveo GUS 14. 3W0,04 EXPIR5: S/31/U FACIUTY NAME (1 DOCKET NUMSER (2) LER NUMSER Mt PA[ (I3

,.-V.'jEGINTIAL'*WIV1T

~NV EIER P44U9M E Cacawba Nuclear Station, Unit 1o 0 Is 4 1f1 3 910 I0III6_4

-- 010 019 [O 0 TrCTr 1f *mm i*inmmt um . AM Am" J4 VI VtA SAFI*UY 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.

NMI FOR*M 301" GPO xq-'

NC Forn && ,U.$. GVO, 19SB-5ýo.184 00070

EXHIBIT B-8 Cooper Station:

LER 298/86-034-00 (December 18, 1986)

NRC FW,.. N U.S. NUCLEAR REGULATORY COMMISSION 1111131 APPROVED OMN NO 31604-lD LICENSEE EVENT REPORT (LER) IxP,041 5/3,1/

FACILITY NAME I11 DOCKET NUIIIIl[ PA Cooper Nuclear Station 70 1 I 01 0 0 2 1 8 oF 3

,TiT1 Storage of Fuel in the Spent Fuel Storage Pool with U-235 Loading in Excess of Technical Specification Limits due to Pellet Design Changes & Manufacturer Variances EVENT DATE 451 LEN NUMBER Il RIPORT DATE 171 OTHER FACILITIES INVOLVED III MONTH DAY YEAR YEAR NIL OMONTH DAY YEAR FACILITY NAM& DOCKET NU.ERIS 1 1881686 65 00341 21 886 °0 OPERATIPI THIS REPORT IS SUBMITTEO PURSUANT TO THE REOUIREMENTS OF 10 CFR A (CI ooPl ( 11I MOOS III N 20.4024b) 20.4"140 1110.731.0121101,1 73 ,lbI POWER LEV EL 1010 1oxi.

0o7,.2*

10.HTIIII 17771

[

i.l4f) EO.721.1121,l 73.711I OTE. ,, ,, A&,,,..,,

2'0.40l.IdII..ll~l 007)1.711t21 1 S8,.73t1l Il3lIlUl4AI .- d - , .. 1. NRC Xo.,

.T66A I 20.4051111111fr1~ 0073.i(ZI1E3I~ 0073.I(2Iil~lA)~l(03 20.400.11)II EO.)I*.I$ ltl~lldl IC.73(.1121(.I LICENSEE CONTACT FORl THIS LEN 112i NAME O TELEPHONE NW-REE AREA CODE D. L. Reeves, Jr. 141012 81215 1-1318 1 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (131 CAUSE SYSTEM COMPONENT MANUPAC TUNER REPORTARLE TONPROS CAUSE SYSTEM COMPONENT MANUFAC TURER VI TO EPORTABLE NRPUS SUPPLEMENTAL REPORT EXPECTED 1414f MO N OAI A YJ;AR I EXPECTED YSf! fit YPCTED RMSINCATE SDATE SUBMISSION ITEI CT /L' 1 0 rT,,*'n *F.WT*Af OP T'Vp*'fl* h.Wl IM1 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 I!,

contained a U-235 loading in excess of that allowed by Technical Specifications, paragraph 5.S.B. At the time of this discovery, a refueliug/major maintenance outage was in progress.

The 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 previous designs.

2) The fuel received for Cycles 7 and 10, while manufactured within approved desipn tolerances, included pellets of a density in excess of the nominal value.

(eneral FI putric his advised that while the 1-235 loading limit of 14.5 grams per axial centimet-r specitied h' Technical Specitlcations wa, exceeded, the average fuel enrichment was unchanged and, therefore, the reactivity of the fuel had not hEen Increased. Hence, the criticalirt-' cal]culations made in support of the high density luel rack npgrade remain ful lv applicable.

Corrective actio,: to bW taken will consist of a review oi Spent Fuel sW rage Pou; de.iglgn for fuel loading an! turther calculatinll, to update stouage OW prescribed In tit' CNS Tlech*il ,*l Spc f.IItions.

8612230 4 4 2 861218 PDR ADOCK 05000298 R. It S PDR

I U.S. NUCLEAR REGULATORY COMMISSION APPROVED O143 NO 31IW-OW4 1,431 LICENSEE EVENT REPORT (LER) TEXT CONTINUATION EXPIRES: 5/1/88 DOCKET HUISIR (21 FACILITY NAME III

~

WA'.J.117

- - £ i Cooper Nuclear Station o0 S l olO1 21918 6 I T KXT1~nAF AW l M"V 4 l , _

• Fý-

..... . ......

future While conducting an evaluation of fuel enrichment requirements to facilitate review of the existing CNS Technical Speci extended cycle (18 month) operation. a of General Electric Company (GE) to dptermine the extent fications wa made by the During the course of this Technical Specifica any revisions that might be required. Paragraph paragraph 5.5.B was identified.

tion review, an apparent violation of storage pool shall have a U-235 fuel in the 5.5.B states that, " . . . In addition, fuel grams of U-235 per axial centimeter of loading of less than or equal to 14.5 GE Type BP8DRB283, which had assembly". However, GE advised that the barrier fuel, grams, in a U-235 loading of approximately 14.6 been supplied for Cycle 11, contained limit. Hence, storage of the new fuel excess of the 14.5 grams per axial centimeter Specifications.

a violation of the Technical in the Spent Fuel Storage Pool constituted from GE on November 14, 1986, an evaluation was Upon receipt of this notification Pool.

been stored in the Spent Fuel Storage conducted of all fuel reloads that had fuel supplied for Cycle 7 was made that the On November 18, 1986, the determination 3, 1981 to April 27, 1981 and from February and stored in the Spent Fuel Storage Pool from Cycle 10, which was stored in the Spent Fuel Storage Pool the fuel 4upplied for than U-235 loading slightly greater July 23, 1984 to July 17, 1985, also contained the plant was in At the time of these discoveries, 14.5 grams per axial centimeter. on maintenance outage which had commenced a shutdown condition for a refueling/major October 4, 1986.

with the requirements specified in This event is being reported in accordance 14.5 fuel with a U-235 loading in excess of IOCFR5O.73(a)(2)(i) in that storage of 5.5.B of the CNS a violation of paragraph grams per axial centimeter constitutes It appears that this limitation is based upon the U-235 Technical Specifications. the fuel design parameters associated with loading which corresponds to the nominal conducted to support backfit of the Spent fuel type considered in the safety analysis fuel racks.

Fuel Storage Pool in 1978 with high density dated June 12. 1Q78, which provided Specifications, Amendment 52 to the CNS Technical was racks in the Spent Fuel Storage Pool, for installation of high density fuel The 14.5 grams per axial centimeter limit.

issued by the NRC with the aforertentioned basis for the to provide the technical criticality calculations which were performed These General Electric type 8DR283 fuel assemblies.

new design racks were based upon pellet density ol of 2.83 w/o and a nominal assemblies had an average enrichment a 6 inch The 150 inch fuel assembly design includes 95.01%theoretical density (TD). central 138 inches of. these top and bottom. The section of natural uranium at its 14.5 grams/centimeter value of 3.01 w/o. The fuel assemblies contain an enrichment 95.07. In establishing this nominal density of is based on this enrichment and the from nominal fuel assembly given to deviations value, however, no consideration was designed and ire within the tolerances considered in the fuel design parameters which may result from either licensed by C.F. These deviations from nominal parameters improvements.

manufacturing tolerances or design with an upgraded hv GE for Cycle II was manufactured In addition, the luel supplied As a result, the higher theoretical densit%'.

pellet design incorperating a :,lightlv With respect to the ftiel was excecded.

14.5 grams per axial centimeter limit limit wls exceeded due to manut-actiu ring provided lor- Cvcle. 7 and 10, the axial envelrpe.

toleranCes within tie approved design III3

Nl*FB, US. NULCLEAR REGULATORY COMMISSION

""4 LICENSEE EVENT REPORT (LER) TEXT CONTINUATION APPROVED oM N4o 35so-010, EXPIRES: I131/0 DOCKET NUMBES (21 LER NUMBER I1I PACE (3)

"iFACILITY NAME III SEQUENTIAL

$ft REVtI$N Y

Cooper Nuclear Station 0 151010101 21918 81 6- 01314 010 013 oF 013 TEXT1/ -me 900 ig AO- . uw*60"mm- NRC P.'" XKAsl '17 the man General Electric has advised that neither the pellet design change nor 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 lmit, coupled with a failure to recognize the impLct 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.

It's,

EXHIBIT B-9 Crystal River Unit 3:

LER 302/87-026-00 (December 1, 1987)

.. -. ,J %UC'LLAN 0GIOWL.ATOMI, Jj * ,O,

• O.0C-ia'.0 11m; LICENSEE EVENT REPORT ILERI SIT

• tjdI (R I ,U' 0D"O'* i5 ,

I st .. ga U or++ui 0 j 6l

• $lI.

, R. oD u I, s r Lo' a,- ~&r. . 13/

u: S t:] itl n,:rr "C -- L 1.VlIT I0.TI .I01

~

JOOC-ET OTMA ACLVII.S vOS IIJ t

- D.:'.

- 1 i ., , I

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• T....

IACJLPftll O/nll IJWVOL.VlO

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! I N/A 0 15O1 0 10 '

1 9 Io 7817 1-I l21,5 -

'A 10 Sdi0 1 A 0* 0 1Z I I

. a . *.....

.. .. i V ' -0 1 MQ..f.... *S ,c .. , . .... o o11,.. ..

4" i iif I 10 bll*73Wi2*

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, 0 W?11 .. .M1.114 73-U14.1 sJltS.511 COsIATC,? *05OI T.Im sli.ll l L. IuFFATT, NUCLEAR SAFETY SUPERVISuR MO. 7 COllAPLITe ON$I L-4I POO tr.. C ..I..OIkTP0 I. WII Oj1CI+Iq O .% '-Z 5lP1 13,12 TOa- " .-' I I.

.I. °O'+' 4I ..... . 0

' I. 11.1

• 'fC'tD I pVI . ril DA l, 0f On November 9, 1987, Crystal River Unit 3 was shut dcown in a refuelirng outage.

The reactor vessel was completely defueled to facilitate inspection of the oore flood valves. Fuel andz Control Rod assemblies were being moved in the spent fuel pools in preparation for the core reload. At 1715, while updating the a)ntrol room fuel location tag board, it was noted that a new fuel asseimby, w,.tih 3.851 pernent U-235 enrichmrent had been placed in the "A" Spent Fuel Pcol.

The fuel racks in the "A" Spent Fuel Pool are limited to storage of fuel assemblies with 3.5 ipercent or less U-235 enricfment. This event was caused by

j personnewl error. When move sheets were being preprxi to rove a fuel asnl y tfnn location M42 in the "B" Spent Fuel Pool to the "A" Spent Fuel Pool, location M43 was inadvertantly written instead of M42. The rmislocated fuel assembly was renmved froxn the "A" Spent Fuel Pool upon detection of its mislocation. Indepenient review of move sheets, prior to actual fuel movernt, hvs bexn impler*ented.

-- I. d 1 209004R 8 '120

'1 r* Aoc, (050(0302

• EN -11 LCS LICENSEE EVENT REPORT :LER)I .. '.,

121 il Jf*1 EI3 E! FIl ;tI l_

,

" '" .: "; fl,_,,Ic,"r. " ,* ul .- i ,ýs , l,,,:_

iev r of n g r , ' , :n0, r *- ov IvlW *lm l *Vl~~il I11'1 lltOIT i'l

• IOV-110 IVACILIT'll IlIVOL.i 0 19, Vll t~..

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i I* I lI I" III1 2 hi NIA@ti I0ll 1-mpol1 ri esI r o O ,lAtI TO4 ?w IVICVll "I"T OP* 10CF11 ' _.

1 5%lI 0 1 I*lkLIl *mliT I'T I*1, "/.,I * .. ,.=,

,Ct............ ens-.... i*. u L..:. IjuFFAIT: NUCLEAR SAFETY SUPERWISUR Co"O'k 2lTl ON# LIN& PCIR I&C.COP P&II.*il DIA, M-8ilI OPIt"T T--

T..6l IND&I' III, I i ll l

__________ ________ I i ' I ' = I S. . . . *, I  ! . ar. &: s ,

U ,V IN . ACT I b1.'I1,I On November 9,_ 1987, Crystal River Unit 3 was Shut down in a refueling outage.

Mhe reactor vessel was completely defueled to facilitate inspection of the core flood valves. Fuel and Control Rod assemblies were being mrv~ed in the spent fuel pools in preparation for the core reload. At 1715, while updating the cx)ntrol room fuel location tag boaird, it was noted that a new fuel assembiy, w;.th 3.851 percent U-235 enrichment had been placed in the "A" Spent Fuel Pool.

The fuel racks in the "A" Spent Fuel Pool are limited to storage of fuel assemblies with 3.5 percent or less U-235 enrichrent. This event wa+/-s caýused by a personrnel error. When move sheets were being prepred to move a fuel as-sbly tn3m location M42 in the "B" Spent Fuel Pool to the "A" Spent Fuel Pool, lc]ation M43 was inadvertantly written instead of M42. 7he mislocated fuel assembly was removed from the "A" Spent Fuel Pool upon detection of its rMislocatjon. Irdeperdent review of move sheets, prior to actual fuel movemernt, has been implemented.

1 2 ) 9 0048R A DOC 8 ' 1 2C, I

(,

0500030 r' I r.

I

"LICENSEE EVENT REPORT (LERJ TEXT CONTINUATION ..- ,to 0., -0 CRYSTAL RIVER UNIT 3 ____.. __"' '___"___"'_

oo3!02 3:7:--O0 1216 ji-010 013 o U13 Ris CsthSotyMler EVEa s it Thbis is the first occuarrencze of this type at Crystal River Unit 3.

EXHIBIT B- 10 Hope Creek Station:

LER 354/95-042-00 (March 25, 1996)

NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 (4-95) EKPIRES 04/30t98 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 (UIR) LICENSING PROCESS AND FED BACK TO INDUSTRY. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (T-6 F33), U.S. NUCLEAR (See reverse for required number of REGULATORY COD SSION, WASHINGTON, DC 21554-11, AND TO digits/characters for each block) THE PAPERWORK REDUCTION PROJECT FACtTy NAME ()U DOCKET NUMBER (2) PAGE (3)

HOPE CREEK GENERATING STATION 05000-354 1 OF 4 f*L (ru4)

Fuel Bundle Confirmed to be Misoriented during an Operating Cycle EVENT DATE (5) LER NUMBER (6) REPORT DATE (7) OTHER FACIUTIES INVOLVED (8)

I .=,= I *05000 12 2T 957 95 -- 042 -- 00 0g5 9 FACILI NAME DOCKETNUMBE 05000 OPERATING 5" THIS REPORT IS SUBMITTED PURSUANT TOTH REQUIREMENTS OF 10 CFR §: (Check one or mrno3 (9111)

MODE (9) 20.2201 (b) 20.2203(a)(2)(v) 50.73(a)(2)(i) 50.73(a)(2)( oyi P....VE(ER10 )20.2203(a)(1 ) 20.2203(a)(3)(i) 50.73(a)(2)(ii) 50.73(a)(2)(x)

L (10) 220a)(2)(i) 20.2203(a)(3)(i ) 50.73(a)(2)(i) 73.71

.... 20.2203(a)(2)(ii) 20.2203(a)(4) 50.73(a)(2)(iv) X OTHER 120.2203(a)(2)(iv) _________

________________~~~~i 50.36{c)(2) ________.NR 50.73(a)(2)(vii) Foi 366A Vofuntary Report IULCENSEE CONTACT FOR THIS LER (12)

NAME TEPKONt NUMBER Q-ncude Area Code)

Jeff Keenan, Licensing 609 - 339 - 5429

" COMPLETE ONE LINE FOR EACHOMPONENT FAILURE DESCRIBED IN THI REPORT (113)

To NE?=3 TO NflDS3

-UPLt IIrAL I REPORTE1 E EXECED MOTH D (if yes, complete EXPECTED SUBMISSION DATE). ____________________

. aK AL (LUmd to 14,0u spaces, I.e., approxmmatey1 sngie-spae ypewTten ,ine) k W1 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 correctly rotate a bundle when moving it within the reactor core. In 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 assessment of fuel movement practices will be completed prior to the next refueling outage.

NRC FORM 366 (4-96)

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 II 2 OF 4 I955 -- 042 042 -- 00 OF 4 TEXTf moll space is ,8reI use adc~ronatcopies ot NH; f-orm 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 SEaU L 3 OF 4 95 -- 042 -- 00 TEXT (it more space is reqWef, use ad~tional copies of NRC Form 366A) (17)

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

",lel 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)

NNRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (4-95)

LICENSEE EVENT REPORT (LER)

TEXT. CONTINUATION FACIUTY NAME (1) DOCKET LER NUMBER (6) PAGE (3)

HOPE CREEK GENERATING STATION 05000-354 - imam OF 4 I4 95 -- 042 -- 00 copies of IRCForm366)(

TEXT(more space is requred,use adcona 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.

7ORRECTIVE 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- 1I McGuire Unit 1:

LER 369/94-005-00 (August 10, 1994)

L V

5 6 RECEA VE9:)

"94 AtUG29 P4:t5 August 10, 1994 U.S. Nuclear Regulatory Commission Document Control Desk Washington, D.C. 20555

Subject:

McGuire ea on Unit 1 Docke o. 50-369 Volu *ray Licensee 369/94-05 Probl :1.1-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.C. McMeekin' 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 940B180023 940810 PDR ADOCK 05000369 s PDR

bxc: B.L. Walsh (EC1IC)

P.R. Herran (MG01VP)

R.C. Norcutt (MG01WC)

K.L. Crane (MG01RC)

B.F. Caldwell (MG01VP)

R.W. 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 (EC13A)

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)

366 U.S. NUCLEAR REGULATORY COMMISSION 91? RM APPOVDaM.N501313ý55o0104 LICENSEE EVENT REPORT (LER)

FAC-ILITY NAME(1)

M*Guire Nuclear Station, Unit 1 TI-fLE(4) Boron Dilution of the Unit 1 Spent Fuel Pool During Drain Down and Decontamination of the Transfer Canal.

EVENT DATE[5) LER NUMBER(6) REPORT DATE (7) OTHER FACILITIES INVOLVED 6) pOwTH DAY YEAR YEAR SEQUENTIAL REVISION MONTH DAY YEAR FACILITY NAMES DOCKET NUMBER(S)

INUMER SNU 05000 07 11 94 94 05 0 08 10 94 05000 OPERATING THIS REPORT IS SUBMITTED PURSUANT TO REQUIREMENTS OF 10CFR (Check one or more of the followin (1111 I

McvDEc9) 11 20.4025(b) 20.405(c) 50.73(a) (2)(iv) 73.71(b)

PCX 100% 20.405(a) (1) (i) 50.36(c)(1) 50.73(a) (2)(v) 73.71fcl 20.405(a)(1)(ii) 50.36(c)(2) 50.73(a) (2)(vii) X OTHER

• geif n1 20.405(a)(l)(iii) 50.73(a)(2)(i) 50.73(a)(2)(viii)(A) and in Text NRC Form 36.A) 20.405(a)(1)(iv) 50.73(a)(2)(ii) 50.73(a)(2)(viii)(B) 20.405(a)(1)(v) 50.731a)(2}(iii) 50.73(a)(2)(x)

-EE CONTACT FOR THIS )

I I

ý q1E.p~ur P!JMARRR 11 Ii R-icky J. Deese, Manager, McGuire Safety Review Group AREA CODE 704 I 875-4065 11 CAUS-E SYSTEM COMPONENT MANUFACTURER REPORTABLE CAUSE SYSTEM CCOPONENT MANUFACTURER REPORTABLE TO NPRDS TO NPRDS SUPPLEMNTAL REPORT EXPECTED 14 EXPECTED MONTH DAY YEAR SUBMISSION TES (If es.,complete EXPECTED SUBMISSION DATE) X NO DATE(15)

ARSIMACT (Limit to 1400 spaces, i.e. approximately fifteen single-space ty-Fewritten lines (16)

Th3s report is being submitted voluntarily to provide information and lessons learned reý--rding 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 Lne 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.

IRC Form 366

LIoRM 3-56A U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OKB NO. 3150-0104 EXP IRES 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 (MNB8 7714), U.S. NUCE1AR REGULATORY CCOMISSION, WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE

_____________________________________oF frANACEMF7T AND BUDOFT WA~q$TNC:TrWNJ r9m FAiCILITY NAMEI) DOCKET NUMIBER(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.

EVALUATION:

Background

",'t-ve [EIIS:ISV] lKF-122, Fuel Transfer Tube Isolation, is located in the Spent Fuel Pool

') Transfer Canal and is used to isolate the SFP from the Refueling Cavity in the rz-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 Zost. 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 statement requires that all fuel movement be suspended if the Boron concentration is found to be below 2000ppm.

W's-ription of Event

ý.,is report is being submitted voluntarily to provide information and lesBons learned zregarding a Reactivity Management 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

,work 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.

-FOR 3b6A U.S. NUCLEAR REGULATORY COMItSSIO N 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 (MN1B]7714), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104),

OFFIC OF MANAGE.MENT AND-RIlF WASRTNG'MNq DC 2()Sn1_

FMCILITY NAME[1)

DOCKET NUMBER(2) LER ?ER(6) PAGE(3)

YEAR SEQUENTIAL REVISION NUMBER NUMBER Mccuire Nuclear Station, Unit 1 05000 369 94 05 0 3 OF 7 "The 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 seals 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 mister 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 po01 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 >/= 2 0 00ppm.

LEya* 36,A U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY CRB 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 COm9ENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (MNBB 7714), U.S. NUCLEAR REGULATORY CC*MISSION, WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE FACILITY NAME i) 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 conc lu5s ion This event is assigned a cause of improper Managerial Methods. The following is a list of examples/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 L....iuneralized 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 aimount 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.

£ 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 x-.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/O/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.

5 LEZRVOW 366A U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY cMB nO. 3150-0104 EXPIRES 5/31/95 LICENSEE EVENT REPORT ESTIMATED BURDEN PER RESPONSE 70 COMPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 RRS. FORWARD (LER) TEXT CONTINUATION COV9tENTS 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 NAMiE() DOCKET NUMBER(2) LER NUMBER(6) PAGE(3)

YEAR SEQUENTIAL REVISION NUMBER NUMBER Mccuire Nuclear Station, Unit 1 05000 369 94 05 0 5 OF 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 close direction to the employees involved to ensure adequate completion of the task aLssigned.

FRI 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

ý..--urn the SSS to operation as soon as possible. The Job Sponsor, Radiation Protection Tcchnician, Mechanical Maintenance Valve Supervisor, Work Window Manager, Maintenance Team

.m 1embers, and the Maintenance Team Support Technician, reviewed the situation; however, the

&aMountof 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 involving 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:

lmmediate: 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.

", 366A U.S. NUCLEAR REGULATORY COMMISSIOW APPROVED BY OMB NO. 3150-0104 L,]ýF, 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 COCMISSION, WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE FAC:1LITY NAM*(1) DOCKET NUMBER(2) LER NUMBER(6) PAGE(3)

YEAR SEQUENTIAL REVISION NUMBER NUMBER McGýuire Nuclear Station, Unit 1 05000 369 94 05 0 6 OF 7 Subsequent: 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.

r ined: 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

,Ep )R.66A U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY CM33 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 CO*flENTS 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 FAC ILITY 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 7 OF 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, the 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

_*-ib for the plant takes no credit for dissolved Boron in the pool under normal cdonditions. 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 e'vent.

EXHIBIT B- 12 McGuire Unit 1:

LER 369/91-0160-00 (November 25, 1991)

LICENSER EVENT REPORT (LER)

DOCKT NaKBE(2) PAH(31 FACILITY NA M(d) bMcGuire Nuclear Station, Unit 1 05000 369 1 OF 5 VITE(4) Qualified Fuel Assemblies Were Stored Improperly In The Unit 1 Spent Fuel Pool Due to A Defective Procedure.

EV~TDATE 5) LER NUMBER (6) REPORT DATE(7) oTH FACILITIES INVWOLVm) a YEAR SEQUENTIAL REVISION MONT2 DAY YEAR FACILITY NAMES DOCKET NUMBER S)

I4ONTH DAY YEAR NUMBER N/A 05000 NUMBER 10 24 91 91 16 0 II 25 91 05000 OPERATING NM THIS REPORT IS SUBMITrED PURSUANT TO REQUIREMENTS OF 10CFR (Check one or more of the follong) (1)1) 73.71(b)

"£400E(9) 20.402(b) 20.405(c) 50.73(a) (2) (iv) 50.36(c) (1) 50.73(a) (2) (v) 73.71(c)

EKAGM 0% 20.405(a)(1)(i)

X-EVEL(1O) 20.405(a) (1) (ii) 50.36(c) (2) 50.73(a)(2)(vii) (Specify in 20.405(a)(11(1ii) X 50.73(a)(2)(i) 50.73(a)(2)(viii)(A) Abstract below an~d in Twxt) 20.405(a)(I)(iv) 50.73(a) (2) (i) 50.73(a) (2) (viii) (B) 20.405 (a) (1) (v) 50.73(a) (2) (iii) 50.73 (a) (2) (x)

LICENSEE CONTACT FOR THIS LER(12) _________________________

_cry L. Pedersen, Supervisor, Safety Review Group AREA CODE 704 875-4487 CCOIPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT*.(13) ____ _..

REPORTABLE CAUSE SYSTEM COMONENT MANUFACTURER REPORTABLEJ CAUSE SYSTEM COMPONENT MANUFACTURER

__________________TO NPRDS TO NPRDS EXPECTED MONTR DAY YEAR SUPPLDENTAL REPORT EXPECTED (14)

SUBMISSION X NO DATE(15)

YES (If yes, complete EXPECTED SUBMISSION DATEY ALSTRACT (Limit to 1400 spaces, i.e. approximately fifteen single-space typewritten lines (16)

While reviewing Technical Specification Section 3.9.12, McGuire Reactor Unit personnel in a identified 11 fuel assemblies that had been stored in the Unit 1 Spent Fuel Pool This Limiting manner contrary to the requirements of Technical Specification 3.9.12.

Condition For Operation requires, in part, that fuel stored in Region 2 of the Spent for Fuel Pool shall undergo 16 days of decay, and if a checkerboard pattern is employed storage unqualified fuel, one row between normal storage locations and checkerboard locations will be vacant. The vacant row provision of the specification was not satisfied from March 23, 1990 through October 23, 1991. At the time of discovery at 1 (Power 0900 on October 24, 1991, Unit 1 was defueled, and Unit 2 was in Mode of Defective Operation) at 100 percent power. This event has been assigned a cause Procedure. The fuel assemblies in question were immediately moved to positions to establish the required vacant row.

LICENSEE EVENT REPORT (LE]) TEXT CONTINUATION DOCKET NUMBER(2) LER NUTMER(6) PAGE(3)

FACILITY NAME(i)

YEAR SEQUENTIAL REVISION NUMBER NUMBER L.gcGuire Nuclear Station, Unit 1 05000 369 91 16 0 2 OF 5 EVALUATION:

Background

The Unit 1 Spent Fuel Pool (SFP) is composed of two regions of high density storage racks

[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 aches 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 2 Storage.

range of TS Table 3.9-1, Minimum Burnup Versus Initial Enrichment for Region specified in TS Additionally, the TS requires that fuel not meeting the burnup criteria side of the Table 3.9-1 must be stored in a checkerboard fashion (empty locations on each spent fuel assembly) with an open row between the checkerboard and normal storage locations if stored in Region 2.

and fuel Free standing fuel assembly inserta, dummy assemblies, fuel storage racks assemblies are transferred within the same unit using procedure OP/O/A/6550/1I, Internal Transfer of Fuel Assemblies. Steps 3.1 through 3.6 of the procedure detail the process employed by the Reactor Unit (RU) Engineers in determining the fuel assembly storage locations. Enclosures 4.1, Internal Transfer Data Sheet and 4.4, Verification of semblies to be placed in Region 2, document the assembles initial and final locations, transfer dates, and required reviews and approvals.

Description of Event directed by step On March 13, 1990, RU Engineer A completed Enclosures 4.1 and 4.4 as 3.1.1 of procedure OP/O/A/6550/1I. 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 Fuel assembly locations were changed by RU Engineer A at the request of the Operations the next core Handling Supervisor to maximize available storage cells in preparation for off load scheduled during Unit 1 End of Cycle (EOC) 7. Procedure OP/O/A/6550/1I does not when final specifically address the necessity of generating a new Enclosure 4.1 or 4.4 fuel locations are revised. Consequently, the locations for 9 of the 11 qualified line assemblies originally recorded on Enclosure 4.1 on March 13, 1990 were deleted by through and the new locations were entered on the enclosure. Enclosure 4.1 was forwarded

LICENSEE EVENT REPORT (LER) TEXT CONTINUATION DOCKET NUMBER(2) LER NUMBER(6) PAGE(3)

FACILITY NAME(l)

YEAR SEQUENTIAL REVISION NUMBER NUMBER 05000 369 91 16 0 3 OF 5 M:cGuire Nuclear Station, Unit 1 to to the Maintenance Fuel Handling crew who transferred the assemblies in question locations specified by RU Engineer A. The records indicate that the assemblies remained 24, 1991.

in these locations until the event discovery date on October Conclusion a Technical Deficency This event has been assigned a cause of Defective Procedure due to is obscure. The because the procedural guidance provided by procedure OP/O/A/6550/ll phrasing of the procedure, procedure attempts to convey the intent of TS 3.9.12, but the the procedure in a direction

,pecially Enclosure 4.4, leads the individual completing For example, Enclosure 4.4 aat does not comply with the full requirements of TS 3.9.12.

2 of the Spent Fuel Pool are states: "Verify all fuel assemblies to be placed in Region 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".

Enclosure 4.5 requirements by checking the design and burnup documentation, the TS and will be satisfied. This is not the case. Also, although the "checkerboard pattern" is open row requirement is contained referred to in the procedure, the only reference to the (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 i.e. checkerboard array and requirements of procedure OP/O/A/6550/II and TS 3.9.12, row physical barriers. These requirements would prevent the violation of the open 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/II 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/II.

to either stop work on the individual performing the procedure and the procedure reviewer to verify that all TS procedure to retrieve the information from TS or to rely on memory 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.

There were no personnel injuries, radiation overexposures, or uncontrolled releases of radioactive material as a result of this event.

r II FACILITY NAME(1) r.Tr!RNSRK K"EN'T REPORT ILERI) TEXT CONTINUATION LIENE EVNTEOT(E)ETCNIUTOTE DOCKET NUMBER(2) ___________

LER ]*IBI* f 61

__________

IUB~6 ?G i

3 PAGE(3)

YEA SEQUENTIAL REVISION 11NUM13ER 11NUMBER H1 4 OF 05000 369 91 16 KcGuire Nuclear Station, Unit 1 0 5 CORRECTIVE ACTIONS:

Immediate: 1) RU personnel determined the cell locations necessary to re-establish the vacant row.

in

2) Maintenance Fuel Handling personnel moved the fuel assemblies question to new cell locations determined by the RU personnel.

will be Planned: 1) Procedure OP/O/A/6550/ll, Internal Transfer of Fuel Assemblies, revised by RU personnel to address all TS 3.9.12 requirements, require the specifically maintenance of the vacant row provision, and to as necessary completion of additional copies of Enclosures 4.1 and 4.4 to document changes in fuel assembly locations.

procedures

2) RU personnel will review and revise as necessary other have adequately involving fuel mqvement to ensure that the procedures addressed all acceptance criteria.

with

3) RU Training personnel will initiate additional training associated Reactivity Management.

SAFETY ANALYSIS:

in a checkerboard configuration TS 3/4.9.12.b (3) requires unqualified fuel to be stored in the Spent Fuel Storage Pool. In the event checkerboard storage is used, one row storage locations is to be kept vacant.

between normal storage locations and checkerboard evaluated the impact on criticality General office Nuclear Engineering (NE) personnel have Using the Keno Va module in the safety caused by the noncompliant fuel pool geometry.

that the loss of the SCALE III system of computer [EIIS:CPU] codes, NE has determined does not increase the vacant row between the checkerboarded and normal storage regions basis. Therefore, Spent Fuel Pool K eff beyond the value reported in the licensing the assemblies in the vacant Reactivity Management has not been jeopardized by placing considered in the row. Additionally, the Boron concentration in the pool, which was not

LICENSEE EVENT REPORT (LER) TEXT CONTINUATION DOCKET NUMBER(2) LER NUMBER (6) PAGE(3 FACILITY NKME(1)

YEAR SEQUENTALJ REVISION NUMBER NUMBER 05000 369 91 16 0 5 OF 5 4McGuire Nuclear Station, Unit 1 margin of above analysis, has been maintained at >/= 2000ppm and contributes an extra 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 356 EXPIRES: 4/30/g2 Estimated burden per response to comply With this information collection reauest: 50.0 hrs ForwarC comments regarding burden estimate 10 1he Recoros LICENSEE EVENT REPORT (LER) and Reports Management Branch (p-530). U.S. Nuclear Regulatory Commission, Washington. DC 20555. ano to the Paperwork Reduction Projecl (3150-01041. Office of Management and Budget, Washington. DC 20503.

DOCKET NUMBER (2) 1 PAGE f31 FACILITY WAME 11)

Millstone Nuclear Power Station Unit 2 01 51 01 01 013 13 1 611 j0d 0_4 TITLE (4)

Spent Fuel Pool Criticality Analysis Error REPORT DATE (7) OTHER FACILITIES INVOLVED I8)

EVENT DATE (5) LER NUMBER (6)

MONTI DAY YEAR YEAR  ;  : MONT DAY YEAR FACILITY NAMES

... *0 1

• 9*2 o*

MEAK3 MOTIAYYA 2i50 o 5 oo of001o1 I1 0 2lo 14,  ?] -1 1 120 o91o2lo 10 10 OPERATING "r*e*Dr*OT iR R~iINIrt *tRMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check one o,'

more of the iOiicOWiiyi fii MODE f(l 1 20.402(c) 50.73(a)(2)(iv} 73.71(b) 20.402(b) 50.36(ci)(1 50.73(a)(2)(v) 73.71(c)

POWER 20.405(a) (1)(i)

LEVEL 0R 310 20. 405(a) (I) (ii) 50.36(c) (2) 50.73. (a) (2) (vii) OTHER (Specity in 24a)(i-Abstracl below and in 50.73(a) (2) (viii) IA) Text. NRC Form 366A) 20.405(a)(1)(iiil 50.73(a) (2) (i) 20.405(a)(1)(iv X 50. 73 (a) (2)(ii) 50.73 (a) (2) (vi ii) (B)

20. 405 (a) f 11 iv ) 50. 73 1a) (2) (1ii) 50. 73 (a) (2) (x I LICENSEE CONTACT FOR THIS LER (121 Robert A. Borchert, Unit 2 Reactor Engineer, Ext. 4418 I

COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REP COMPONENT MANUFAC-EREPORTCAE SYSTE COMPONENT CAUSE SYSTE CAUSE TURER TO NPFS DI RI XBli:ii*',i,'iii~i: 19 1 I I i 1l4 SUPPLEMENTAL REPORT EXPECTED lie' "YES [if yes. comolete EXPECTED SUBMISSION DATE) 171 NO ABSTRACT (Limit to 1400 spaces. i.e., approximately fifteen single-space typewritten lines) ( 16 On February 14, 1992, ar 1415 hours0.0164 days <br />0.393 hours <br />0.00234 weeks <br />5.384075e-4 months <br />, with the plant in Mode 1 at 30% power, Northeast Nuclear Energy Company (NNECO) wdas notified by ABB-Combustion Engineering (ABB-CE) that a calculational error existed condition was in the criticalitv analysis for the Region 1 spent fuel storage racks. NNECO determined that this reportable as a condition outside of the desien basis of the plant. An immediate report was made to the NRC, of the spent fuel pool was verified to be in compliance with the plant and the existing reactivity condition Technical Specifications.

storage racks for The oriinal effective multiplication factor (Keff) calculated by ABB-CE for the Region 1 fuel and 4.5 weight percent enriched fuel assemblies was nominal dimensions, nominal spent fuel pool temperature The discovered error results in an underprediction of approximately 0.04 delta 0.9224 (without uncertainties).

the same conditions. An Keff. Revised calculations by ABB-CE indicate that Keff is actually 0.963 for investigation 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 to the NRC on April 16, 1992. These changes were approved by plant Technical Specifications were submitted the NRC on June 4. 1992.

U.S. NUCLEAR REGULATORY COMMISSION APPROVED OMB NO. 3150-0104 NRC Form 366A EXPIRES: 4,30/92 (6-89)I Estimated burden per response to comply with this request:

collectionburden information regarding 50.0 nrs. Forward Recordls LICENSEE EVENT REPORT (LER) I comments estimate to the TEXT CONTINUATION and Reports Management Branch (l-530). U.S. Nuclear Regulatory Commission. Washington. DC 20555. and to the Paperwork Reduction Project 13150-0104). Office of SManagement and Budget. Washington. DC 20503 f2) LER NUMBER (61 PAGE (31 DOCKET NUMBER FACILITY NAME Ill YEAR . % ____"

Millstone Nuclear Power Station Unit 2 051100O 1336 9 2 0o113 011 021 OF 014 TEXT (11 more space is required. use additional NRC Form 366A sl (17)

I. Description of Event an On February 10, 1992, at approximately 1130 hours0.0131 days <br />0.314 hours <br />0.00187 weeks <br />4.29965e-4 months <br />, Northeast Utilities (NU) was notified by independent contractor that a higher than expected effective multiplication factor (Keff) was calculated for the Region 1 fuel storage racks. On February 11, 1992, NU notified ABB-Combustion Engineering 1992. at (ABB-CE) of the potential error in the spent fuel pool criticality analysis. On February 14, (NNECO) was 1415 hours0.0164 days <br />0.393 hours <br />0.00234 weeks <br />5.384075e-4 months <br />, with the plant in Mode 1 at 30% power, Northeast Nuclear Energy Company 1 spent fuel notified by ABB-CE that a calculational error existed in the criticality analysis for the Region 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 The weight percent U-235. Region 1 was designed to allow fuel assembly storage in every location.

Region 1 storage racks contain a neutron poison material (Boroflex), and have a nominal center-to-center pitch of 9.8 inches.

their (b) Region 2 is designed to store up to 728 fuel assemblies which have sustained at least 85% of design burnup. Fuel assemblies are stored in a three-out-of-four array, with blocking devices Region 2 installed to prevent inadvertent placement of a fuel assembly in the fourth location. The storage racks have a nominal center-to-center pitch of 9 inches.

storage The original effective multiplication factor (Keff) calculated by ABB-CE for the Region 1 fuel spent fuel pool temperature and 4.5 w/o enriched fuel assemblies racks for nominal dimensions, nominal is 0.9224 (without uncertainties). T]pe discovered error results in an underprediction of approximately 0.04 delta Keff. Revised calculations bv ABB-CE indicate that Keff is actually 0.963 for the same not affected conditions. Evaluations by ABB-CE have confirmed that the Region 2 fuel storage racks are by the error.

of the NNECO determined that this condition was reportable as a condition outside of the design basis made to the NRC. and the existing reactivity condition of the spent fuel plant. An immediate report was movement in the pool was verified to be in compliance with the plant Technical Specifications.' All fuel poison spent fuel pool had previously been restricted due to the observed degradation of the neutron Region I fuel storage racks. No automatic or manual safety systems were required to material in the respond to this event.

I1. 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 1 and thus a lower energy group. This resulted in an overestimation of the neutron absorption in Region calculated Keff.

and unpoisoned Second, ABB-CE used a geometric buckling term corresponding to a sparsely populated array as an approximation of buckling in the poisoned configuration. This approximation also contributed to a lower calculated Keff in Region 1.

Ill. Analvkir of Event This event is being reported in accordance with 10CFR50.73(a)(2)(ii)(B), which requires the reporting of design any event or condition that results in the nuclear power plant being in a condition outside the basis of the plant.

U.S. NUCLEAR REGULATORY COMMISSION APPROVED OMB NO. 3150-0104 16-SQ)Form 366A NRC EXPIRES: 4/30/92 Estimated Oureden per response to comply with this leon request: 50.0 hrs. Forward LICENSEE EVENT REPORT (LER) coinlorma Durden colction c omments regardinlg estimate to the Records TEXT CONTINUATION I and Reports Management Brancn (p-530). U.S. Nuclear Regulatory Commission, Washington. DC 20555 and to the Paperwork Reduction Project (3150-0104). Oflice o0 Management and Budget. Washington. DC 20503.

(2) LER NUMBER (6) PAGE (3)

FACILITY NAME (1) DOCKET NUMBER YEAR 3f03A E U .

Millstone Nuclear Power Station Unit 2 01 510 0 01313 1 912- 010131- 1Oi 013 OF 014 (17)

TEXT (If more space is required, use additional NRC Form 366A s)

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 concentraton 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 anhalyses 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. Region 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 will be stored in a 3 out of 4 array in Region B, wýith blocking devices installed to prevent inadvertent placement or storage of a fuel assemblh in the fourth location. Reeion C is the new 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 figure 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 Storace Racks Manufacturer: Combustion Engineering Model: Hi-Cap Spent Fuel Storage Module EIIS Code: DB-RK-C490

U.S. NUCLEAR REGULATORY COMMISSION I APPROVED OMB NG 3150-013Z4 EXPIRES 4/30/g2 Estimated bur0en der resDonse IC comply wilt' this information collection request: 50.0 nris. Forward LICENSEE EVENT REPORT (LER) comments regarding burden estimate to the Records TEXT CONTINUATION and Reports Management Branch (D-530). U.S. Nuclear Regulatory Commission. Washingion, DC 20555. and to the PaperworK Reduction Project 13150-0104). Office oD Management and Budget. Washington, DC 20503.

PAGE (3' more space is reau;red. use additional NRC Form 366A'S) ztt C)

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0 0

0~

tI-It z 0

EXHIBIT B-14 Oconee Unit 1:

LER 269/96-001-00 (February 7, 1996)

U.S. NUCLEAR REGULATORY COMM3SSMIO APPROVED BY OMS NO. 3150-0104 NiC FORM 366 ERIM0TED MDinm F11 To Munon?mw T* K0ATo PAGE I PM0mI*0 PfaSANF1 ClUMlNo~(u COLLECT" OSMST "a N" SoD0u nTONrl400 KENNW.M W Lima CoToa 1ROMI Lima CDLIMOMEM u'Ox LICFNSEE EVENT R"-PORT (LER) mm .mO T m e4M NON N.NX (49 n t11 TWM

%.L =WUI0 PQUTM 00032 WAS"Tft K X5660SV. AM PaK an uiouc PMACT 01564i@ , Owt OFNiUMUM Us K=1.

9 r.ve for rOquaed nfuber of W*iro OC205a d*,psfchtwacteis for each block)

• ,I WOOELIFN 05000 269 1 OF 17i oconee Nuclear Station, Unit-One

" Mispositioned Fuel Assembly Due To Inadequate Self Checking and Management Direction 0^1Ig~S ~ LEMI NUAiEM 161 I AI III UOR U~THM F~AUjIR3 uiVOiVU IN T4M TE1 I WX.L IV*IS I*i. U UAY T n"lm r in__nm

• " conee. Unit Two 05000 270 01 08 96 96 01 00 02 07 96 Oconee, Unit Three 05000 287 MODI* 0 :20.22](0151, 20.4zU*ilAW140 X 50.73,h- 12MU 1B) 50./34*I;0iV0#

20.22034aH3(iI 4 50.734a)(2)iu (A) 50.7316142) .

- O-11011 O 20.22031aI11 0203,HJ)2.20lIH, 50.ita"14110 3.1ll LEE

' 20.220300121ta) 20.220.a)43 SO. 731aifiv) OTHER 22 ,n P&C Foe m 30A i U-510.362312

!0~a~lw 1 50. /J3aiI two-u L. V. Wilkie, Safety Review Manager (803) 885-35'18 TOUU TOWNS L FlA I EXP* IýT 1141 B*n[ TtvI.., UP, Mi.i UAI EA i0. EPA'P'I SUBAWSS-OI 4conne EXPECTED SU.SSION DATE).

- X I On December 14, 1995, with all three Oconee units at 1loo Full power, %

fuel handling team performing a fuel assembly (FA) inspection in the Unit 1&2 spent fuel pool (SFP) inadvertently left the FA unattended and suspended inside the SFP mast. It was discovered oa January 8, 1996, by fuel handling personnel during check outs for planned fuel movements. The FA was reinserted into the SFP rack. The primary safety significance of the event was the potential uncovering of the FA during a postulated event requiring actuation of the Reac'or Coolant Make-up function of the Standby Shutdown Facility (SSF). which uses the SFP as a water source. An engineering analysis concluded that the f-tel cLadding would not be breached during an SSF event with this FA in the mast. Therefore, IOCFRlOO limits would not have been exceeded and the Final Safety Analysis Report (?SAR) analysis consequences would have bounded the event. However, having an unattended FA in the mast is outside the intent of Technical Specification

)-a 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.

U.S. NUCLEAR 616GULATORY COSAUMOMM LIC N E EVENT REPORT (LER)

TEXT CONTINUATION OOCW I3 EUM In PAM 9 IV MOAK Is)

I 05000 "'i sum 1 2 OF 1.7 uclear Statione gai t One

'"~Nee

.a- ~A~Fo'm3~A*11Z gma~g 269 196 01. 00 IýI Pro "'r ~ -

in addition to a Spent 7t1 Pool (SFP) REItS:NDI where spent fuel is stored in racks submerged under borated water. Oconee Nuclear Station has an Interin Spent Fuel Storage Facility on site- There spent fuel is stored in dry containers, thus the term -dry cask storage" is used.

ruel h&ndling activities at Oconee are performed by members of a dedicated fuel handling maintenance crew. The fuel handling supervisor is a previously licensed Senior Reactor Operator. The crew's work activities are primarily fuel handling activities and plant crane [EXXS:RNJ maintenance. A significant portion of the fuel handling crews scheduled work involves shufflinig pent fuel assemblies in the SFP and support of dry The minimum crew number for operating the cask s*orage activities.

xn the SFP is one bridge operator and one refueling bridge (EnZSF-MI spotter. Puel Handlers are qualified to Fuel, Handling actAvities per Employee Training Oualiftcation Standards.

for OPiO/A115.O6/.01 * (JFuel r. C0qponent Handling)- is the "B0U, TO0 procedure using the fuel handling birdge. It is an ".Information Use" procedure which, policy, is has no sign-offw, is performed from memory. and..by management not required to be at the job location.

Niormally, OP/O/A/1S03/09 'Documentation of Fuel Assemblies i/or Component make Shuffle Within a ST Pool) is the 'WHORM TO procedure used to miscelLaneous fuel movements. 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 cof each move as Lt --s made.

Technical Specification 3 6 provides required prerequisites for fuel handling in the SFP. Or e requirement is that the SF1 filtered ventilation system [KIIS:VF) must be :perable, or fuel handling must be suspended. The SFP filtered ventilation system is considered inoperable whenever the fuel recexvinng bay door it open.

The Standby Shutdown Facility (SSF) [EIIS:NBI is designed to maintain the of an plant in a safe shutdown -ondition for a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> period in the event Appendix R fire, a turbin building flood. a security event, a statLon blackout when the turbine driven emergency feedwater [EIIS:BA) pump fEIaSrpi is inoperable, cc a tornado which renders the 4uxiliary service water and emergency teednter systems inoperable. The SSF Reactor Coclant (RC) makeup pump [ErIS:Cl1 takes water from the SFP inventory in order to make-'; to the Reactor Coc.anz System RCS) (EIIS AS) through the reactor coolart pump seals. En m:dLtifon. SFP -oolinq may a;so be lost during an

p= FORM3U.S. NtlEA"R FEG_*ATOAY CO:_-"*"*

LICENSEE EVENT REPORT (LER)

"I41O NAME 1ear SMato FAC*TY

'

TEXT CONTINUATION 06000 LEN

=WKI I "I lip AG M 3 OF 17

-Oconee Niuclear Station, Unit One 269 I96 01 00 wi4.-a ofA9

• WM~ffi 1171 TeXTWai SSF event such that boil-off of SFP water will also contribute to the loss depletion of SFP inventory. The design basis of the SSF system will allow to within one foot from the top of the SFP racks of the SVP inventory assuming no action to refill the SFP. T?.chnical Specification 3.1.8.4 each unit when the RCS is requires the SSF RC Makeup System be operable for at or above 250*F.

which started on During Unit 1 EOCL6 (End of Cycle 16) refueling outage, fuel assembly (FAI was observed uov. 2. 199S and concluded Dec 10, 1995. a grids damaged. As part of the root cause to have four intermediate spacer A desired to perform a visual inspection of FA evaluation, Reactor Engineer adjacent to -the damaged NJO ST CFA-8), the fuel assembly which,.had been assembly in the reactor core for the fuel. cycle A contacted the, On December 14, 1995, at about. 0900 hours0.0104 days <br />0.25 hours <br />0.00149 weeks <br />3.4245e-4 months <br />, Reactor Engineer

-A-S.- The, request was Fuel Handling Supervisor for support in inspecting Subsequently. one -of the planned. tasks initially denied, due to workload.

and the Fuel Handling Supervisor contacted was deferred several hours after lunch.

Reactor Engineer A to schedule the inspection for A entered Unit Around 1300 hours0.015 days <br />0.361 hours <br />0.00215 weeks <br />4.9465e-4 months <br />, two Fuel Handlers and Reactor Engineer FA-8. A pre-job briefing was 162 Spent Fuel Pool (SVP) to inspect Reactor Engineer A and Fuel Handler A but it covered only perfor-rrT. Lwtween Reactor Engineer A had no procedure the basics of what needed to be done.

evolution, and, since the inspection did not or !-ovement enclosure for this Reactor Engineer A felt that involve leaving an FA in a new SFP location.

he did not need one.

since he had Fuel Handler A thought Reactor Engineer A had a procedure listed in the normal fuel called the control room to verify prerequisites Reactor Engineer A stated that he called the control handling procedures.

However. Reactor Engineer A stated that he did not room out of habit.

that fuel handling activities were about inform the control room operator to take place.

which :s the Fuel Handler A operated the Unit 1&2 SFP bridge by memory, normal practice. Fuel Handler A stated that he felt comfortable doing 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.

-asr directed Fuei Handler A to SFP rack location K40. and directed in progress. During this operation iup/down) while video taping was

0C FO1M 6434 U.S. NUCLAR TO CORinO LICENSEZ EVEN REPORT (LER)

TEXT CONTINUATION

-FACKV NAME III DOKET LAMNMIMII 131 05000 ,M ,mM"f 4j OFi1 OcneNcerStte<Ud n 269 96 01 O00 IE~1l'1ff 'we wmcu 'S i m NW Fom 364k 1171 evolution, the mast and FA-B were moved several feet east to improve the available lighting. Also,-Reactor Engineer A requested Fuel Handler A to rotate the fuel mast 90 degrees and back while FA-B extended below the mast. After some scratches were noted on FA-B's lower end fitting,. A-S was returned to its proper l.ocation and lowered into the storage rack.

For comparison, Reactor Engineer A decided to look at another PA selected at random from the same cycle. Reactor Engineer A directed Fuel Handler A to SFP rack location L44 to pickup FA 100697 (FA-71 and directed mast operation (up/down) while the FA was video taped. After observing similar scrat.-hes on FA-7, Reactor Enginee& A stated that he had seen enough.

At this point neither Reactor Engineer A nor Fuel Handler A specifically stated a need to lower FA-7 prior to proceeding.

OPIO/A/IS06/01, Limit and Precaution 2.27.directs personnel to..notl* eaye portable underwater lights and camera&-.in 'close proximity to irradiated fuel assemblies when not being Used. jTerefore, Reactor, Engineer..began to raise the video camera. Due.to the need to wipe down the pole and ,cable attached to the camera as it is raised,, this, task requires .two people.

However. OP/O/A/1506/Ol, Limit and Precaution 2.22 directs personnel to turn off the Bridge hydraulic piimp to prevent overheating when a Bridge is idle for 15 minutes or greater and the hoist is not engaged. In this e-se the hoist was engaged, but during the investigation it yeas learned that -he FueL Handling-Supervisor has issued standing directions to turn off the pump even if the hoist is engaged.. When the hydraulic pump is off. most of the control panel indications are either de-energized or go to a default state.

n accordance with these instructions, Fuel Handler A stopped the hydraulic pump. left the control console, and &eaisted Reactor Engineer A with pulling up and wiping down the video equipment. Once the camera was secured, Fuel Handler A returned to the control console and de-energized the bridge. During interviews. Fuel Handler A stated that he believed that he had lowered the FA back Into the fuel rack and did not look at the control console indications to confirm this.

At 1342 hours0.0155 days <br />0.373 hours <br />0.00222 weeks <br />5.10631e-4 months <br />, Fuel Handlers A and B exited the Unit 1L2 SFP with Reactor Engineer A. This left FA-7 suspended and unattended i.t, the mast.

No fuel handling tasks in the Unit 1&.2 SFP occurred over the next several weeks.

U.S. NUCLEAR REGULATORY COMMSSIO "MRCFORM 30A LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION DOCKIEF LMf NUOMS Is) PAGE 13)

FACILIT NAME III

- 06000 T .&UF17LGi 269 96 01 DO Oconee Nuclear Statione Unit Oonee 1171 0 two is r@QWUWae*6anw -0 of NRC F-i 3W)J rT Fuel Handlers. A and C On January 8. 1996, at approximately 1030 hours0.0119 days <br />0.286 hours <br />0.0017 weeks <br />3.91915e-4 months <br />, for loading a dry cask later

.4' entered the Unit l&2 SFP to 9tart preparation When Fuel Handler A energized the bridge and started the in the week.

indications and realized hydraulic pump, he observed the control console assumed that chat a FA was in the mast. Fuel Handlers A and C initially by other members of the crew.

the FA had been left in the mast recently FA in the open rack at decision to lower the Fual Handlers A and C made the to determine location L44 to allow an identificatl-of-LbtEhe Min order ce ano to trace the last Kinown movement to determine who whore it should was responsible.

rack, Fuel Handler C while Fuel Handler A lowered FA-7 into the storage of the discovery and him called the Fuel Handling Supervisor and informed at L44. Fuel in the empty rack that Fuel Handler A had lowered the FA 1&2 SFP rack location FA as NJO6E7 at Unit Handlers A and C identified the L44.

called ýhe Rotating Equipment.

At 1130 hours0.0131 days <br />0.314 hours <br />0.00187 weeks <br />4.29965e-4 months <br />, 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 l&2 ;SFP.

the event. The video tape At 1230 hours0.0142 days <br />0.342 hours <br />0.00203 weeks <br />4.68015e-4 months <br />, a meeting was held to discuss had shown the FA being put from 12/14/35 was reviewed to see if the tape concluded that FA-7 had been back down in the pool. The personnel present All three Oconee units were in the fuel mast from 12/14/9S until 1/8/96.

at 100 % full power throughout this period.

allow depletion of the SFP The design basis of the SSF system will inventory to within one foot from the top of the SFP racks assuming no the SFP. A concern was raised that FA-7 could have been action to refill for heating to clad failure uncovered by an SSF event, with the potential However, no analysis existed with resultant release of fission products-the severity of the releases to determine if clad failure would occur or if Thus there analysis o.- 10CFR100.

would exceed limits from e -- r the FSAR its intended have been unable to perform was a concern that the SSF : iht one considered past inoperable. Therefore.

function and would need to be evaluation an operability action item from the 1230 meeting was to start of expected clad temperatures and potential which would include calculation releases.

as the Station Manager)

The Maintenance Superintendent (who was acting staff meeting at 1330 discussed the event during the 5ration Manager's and assumed the hours. The Operations Superintendent was at the meeting control room knew of the event.

U.S. NUCLEAR REGULATORY COMNISSIWM FC OWR66 F4 FACXL NAME II*.I LICENSEE EVENT REPORT (LER) r "EThPITIMTI~l IATIflM

, r^, *.*,,. ...........05000 WA j LER NUMBER 9 PAGE 13)

PAGE 13*

6 OF 17 "00 1 1&

I96 "-1O A!Oconee Nuclear Station, Unit One 70;4 0 96 copms of NRC Fwm 36W 117)

TEXT Uf rmow sw*e'* eq*sk,udeuddiiwn

!

te Maintenance Superintendent, the At 1500 hours0.0174 days <br />0.417 hours <br />0.00248 weeks <br />5.7075e-4 months <br />, afte." the staff meeting, tkLng Supervisor went to inform the Rotating Equipment Manager and Fuel. Handli ONS NRC Resident Inspectors of the event.

Superintendent.

After briefing the senior resident, the Maintenance Rotating Equipment Manager, and Fuel Handling Supervisor discussed the procedures situation and decided not to continue with fuel handling until were revised to prevent this event from reoccurring.

and the Site VP At about 1800 hours0.0208 days <br />0.5 hours <br />0.00298 weeks <br />6.849e-4 months <br />, the Senior VP of Nuclear Generation Event discussed the event and decided to iniiate a Significant rnvestigation Team (SEIT).

the discovery Throughout this period, the control room was not informed of mast. On 1/9/96, at about 0630 hours, an NRC resident of the FA in the This was for the event.

asked control room operators about the log entry time Operations shift. had heard about the event.

the first direction .'

At-08oo hours, this event was discussed in the daily site Site management present discussed isuou:s related to. past

.meeting.

The information available at that time was operability and reportability.

insufficient to reach a conclusion.

event. Notes At 1414 hours0.0164 days <br />0.393 hours <br />0.00234 weeks <br />5.38027e-4 months <br />, a log entry was-made in Unit I Log about the Control Room Senior Reactor Operator were added on Reactor Operator (RO), in 1&2 (SRO), and Unit Shift Supervisor's turnover sheets not to move fuel completed.

and/or 3 SFP until'after the SEIT investigation was Issues Discussions of operability and reportability issues continued.

with Technical Specifications iTS) and FSAR discussed included compliance apply TS that potentially analyses of fuel damage and resultant releases. and the Spent Fuel Pool, in this case are 3.8. Fuel Movement and Storage in 3.18, Standby Shutdown Facility.

TS 3.6 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 progresp and, therefore, the TS was not exceeded.

operable for each unit at or TS 3.18.4 requires the SSF RC Makeup System be During an SSF event the SSF RC makeup pump takes above 250OF in the RCS.

suction from the SFP and can allow depletion of the SFP inventory such that Preliminary engineering calculations indicated FA-7 would be uncovered.

with resultant release of fission possible heating to clad failure products. This could result in dose consequences beyond tha licensing basis.

U.S. MUMEAR REGULATORY COMMS*"O LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACKMTY NAME II) DOCKET LER NUMMER IS) PAGE (31 050 -- A--M- 7 17 OCon Nuclear Station, Unit One 269 96 01 00 l

TJT ft?more 'W&C0 q~d ume a covs or NftC Foom 366.41 (17) analW At 1700 hours0.0197 days <br />0.472 hours <br />0.00281 weeks <br />6.4685e-4 months <br />, a decision was made to make a I hour NRC Emergency Notification System call, based on management's conclusion that these consequences represented an u.-analyzed condition that could significantly compromise plant safety. The notification was made at 17S5 hours.

On 1/10/96, the SEIT arrived on site and began an investigation. On presented its preliminary findings in a formal exit with 1/12/96, the SNIT site management and the Senior VP of Nuclear Generation.

a FA One concern raised by the SEIT was the interpretation that leaving of met the requirement to suspend fuel handling. A survey in the mast defined industry practices revealed that all of the other sites contacted an assembly was supported by the fuel fuel handling to include any time These other sites interpreted "suspension of handling bridge or crane.

until any FA fuel movement" to mean that fuel movement should be continued in a raised position could be moved to a safe location and lowered.

Applying- this more conservative interpretation, of "fuel handling- resulted FAM-7was in in the-conclusion that TS 3.8 should be applied the entire time the fuel mast. Since the fuel receiving. bay.door was opened at various times during the period, making. the filtered ventilation system inoperable.

would mean that the intent of TS 3. X.12 was not met.

the new interpretation the results The operability calculations aria analysis were completed and the "Safety Analysis" section of this are discussed in more detail in The analysis showed that FA-7 would not be damaged and would not report.

result in off- site releases exceeding 10CFRI0O 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 SSF event is not in compliance with the intent of TS 3. 18.

condition that Therefore. in addition to being reportable as an unanalyzed could significantly compromise plant safety. this event would also be reportable as a condition outside the intent of Technical Specifications.

In response to the SEIT preliminary concerns. "Short Term" actions were initiated to enhance programs, policies, and procedures to address the SEIT recommendations and observations. These aere primarily aimed at those cask items needed to resume limited fuel shuffles in preparation for dry storage and new fuel receipt prior to a refueling outage on Unit 2, currently scheduled fcr late March. !996.

On Feb 1, !9.6. the SEIT issued its final report. .he root causes as the ro-t causes listed below.

identified are the same

FORM 3A U.S. NUCLEAR **E*ULATORY COMMISSIO

• 45e LICENSE EVMET REPORT (LER)

TEXT CONTINUATION DOCKET LER NUIBER () PAGE 130 FACtITY NAME I I!

05000 J MI 8 OF 17 Unit One 269 96 01 00 Oconee Nuclear Station, usg ,Mvjwl cope of NRC Fvi 366A4 117) r-oXT ppf no.m X #qwW.r*

intended The root causes of this event are related to inadequate barriers type of error. Two root causes for the to minimize the potential for this event have been determined:

Handler A to The first root cause of this event is the failure of Fuel from a This was a skill based error resulting self-check his actions.

while performing routine actions using an momentary memory lapse Information Use procedure The second root cause to this event is the lack of management expectations The lack of for formality in all aspects of the fuel handling process.

was exhibited in the following actions, which were in accordance formality

-his type of work in the with management's expectations at the time for spent fuel pool, leading up to the leaving of the FA in the mast:

.onduct of

1. The failure to write a.id process a work request for the this activity.

to

2. The perception that no task specific procedure was required conduct this activity.

3. 0P/O/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 jub location. Performing proceduxes from memory will increase the risk of human error. Requirements of OP/0/A/1506/01 were not met in that:

a) The Control Room was not specifically notified that fuel handling was in progress in the Spent Fuel Pool iSFP1.

b) Fuel Handler A rotated 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 from the fuel grapple are included in the procedure but the omission of those steps resulted in FA-' being left suspended inside the fuel mast.

3U.S. NUCLEAR REGULATORY COMM=--iOW MAC FORM 368A W LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION DOCKET LEM NJMSMR InI PAGIE 131 FACLITY NAME 111 G5000 "'

MJM W 9 OF 17 269 96 _101 00 Oconee Nuclear Station, Umit. One I n~u~vd.gaeadhbw.I coesof NRC FAMw 36seA 117) sacGa TEX idle d) Limit and Precaution 2.22 directs that *When & Bridge is for 15 minutes or greater and the hoist is not engaged, turn This off the Bridge hydraulic pump to prevent overheating-"

condition was 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 coamon fuel handling practice.

4. Inadequacy of OP/O/A/1506/O1 (Fuel & Component Handling) in that it did not provide steps for the fuel handler to verify that the fuel bridge mast was empty prior to shutting down the bridge.

for the

5. The failure to provide an adequate pre-3ob briefing evolution of The pre-job brief ing did not address roles and responsibilities During most of- the activities, Fuel the individuals involved.

Engineer A. This Handlir A was acting under the direction of 'Reactor of Fuel Handler A: for.

poten~ially led to an expectation on the parethe PA. Reactor Engineer Reactor Engineer A to instruct him to lower it was not his responsibility to ensdre that FA-7 was lowered A felt back into the SFP racks.

Past industry and site experience was reviewed to determine if this event is recurring. It was concluded that industry operating experience has not events. SER .91 been used effectively at Oconee to prevent fuel handling events that occurred 1S. as an example, identified fuel misDositionina verif.cation and wicnin tne incustry due in part uo inauequace indlependent Oconee reviewed the SER. revisec. refueling self -verification techniques.

and evaluated procedures, enhanced methods of fuel har.dlers conmmunication.

However, these corrective actions were training in response to this SER.

events that occurred in ineffective in preventing four fuel mispositionincl 1992 through 1994.

An operating experience review was performed using the Oconee Problem handling Investigation Process (PIP) data base in the area of fuel activities to icok for similar events wit' root causes similar to this fuel handling events event. Attachment A to this report summarizes past and the related NRC vLolations.

The first root cause (self-verification as it relates to fuel handling work practices) has contributed to four events resultinq in three NRC v.,iolations at__Qcone'e during the period of 199, throuqn 1995.

N: FORFM U.S. NUCLEAP RIEGULATOY CNM5O LICENSES EVENT REPORT CLER)

TEXT CONTINUATION FACT WAME

,E III DOCKET NUNN" 161 PAGE (3)

-105000 5L A NE ...10 OF 17 Oconee Nuclear Station, Unit One 269 96 0l 00 TEXT W so m a nreqzer. mer *'hwl PC Fawm 356A45117) os*, 0 N@

The second root ciuse (lack of management expectations for formality in all aspects of the fuel handling process) has also contributed to 4 fuel handling events at Cconee (particularly PIP 1-094-0707 and the associated NRC violation of August 2. 1994).

Therefore, it is concluded that this event is recurring with respect to both root causes. The repetitive nature of these fuel handling events demonstrate the lack of ful- use of lessons learned from previous events and application of too narrow a scope for corrective actions.

.here were no radioactive releases. personnel in]uries or over exposures, or NPRDS reportable equipment problems associated with this event.

COURECIT ACIIONS tImmediate

1. Fuel Handlers ,loweredthe fuel assembly into a Spent Fuel Pool (SFP) storage rack location..

2. Mechanical Maintenance management suspended-fuel handling activities pending procedure changes.

Subsequent E.Engineering calculations were performed and this event was analyzed with respect to the potential for exceeding design basis releases.

?lanned "Step by step procedures wi"ll be required for all f..;eL* "novements.

-. A procedure checc>!.st " be provided to assure that the fuel mas is returned to a prcper end state at the conclusion of fuel handlinr Formalized pre-,=b briefings for all fuel related activities in the SFP will be iplemiented.

4. Appropriate pei-:nnre. correcti'e actrins wi*l be ýa-en in 3c--ordance with Duke Power :::-es A Self !nit.ated z:;ni.a Aud-t SITA) wtýL! be pertDrd t provide Sbroader review :,- : 2_'

andlinq and -Ier SFP Z>-.*Ps .A -ork processes.

NRC FONR 36. . U.S. NUCLEAR AEGULATORY COM0ISSK LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACIJl P NAME M ' DOCKET LER NUMER 161 PAGE (38 05000 Nu" I g*I* 11 OF 17 Oconee Nuclear Station, Unit One 269 96 01 00 MIT f#firim speco 'a reouwawd. use eobonaI copos of NRC Form 366.4)1171 Planned corrective actions 1 through 5 are concidered Commitments to the NRC. They are the only items included in this report intended to be NRC Commitments.

SAEMI AHALX515 The consequences of the failure of a fuel assembly (FA) in the spent fuel pool (SFP) are analyzed in the Final Safety Analysis Report (FSAR), Section 15.11.2.1, "Single Fuel Assembly Handling Accidents". The FSAR accident scenario is a radioactive release from all 209 fuel rods. This accident is assumed to occur under at least 9 feet of water for iodine retention. The dose calculation with the FSAR initial condition assumptions of release inventories and conditions yields a dose of .66 rem whole body and 174 rem thyroid at the size boundary.

During an event requiring the Standby Shutdown Facility ISSF) Reactor Coolant (RC) makeup pump. FA NJO6E7 (FA-7) would have been uncovered by the decreasing inventory of the SFP. A heat up calculation of air-cooling of the FA has been performed%using the actual decay time after shutdown 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 P.

metal water reaction 2200 deg F.

enhan..ed fission gas release from the 2450 deg F.

U02 pellet matrix zircaloy melting 3400 deg F.

This calculation shows that cladding integrity would be maintained and no effluent radiation release occurs. Therefore, the existing analysis in 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 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after su.bcriticality, was assumed in the mast and 'ooled by air and radiation This analysis determined a maximum cladding temperature of .Q00 legrees F 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.

jBC FORM 36" U.S. NUCLEAR REGULATORY COMPMSbM LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACAM TNAM 11) DOCKET LER NUMBER 16) PAGE 9 I W YEPAaJ.A KlY' l 12 OF 17 Oconee Nuclear Station, Unit One 269 96 01 00 rWe.XT 1ff more ssa*e 4

• d. uLoe atonaW cooas of AfRC Form 36641 1171 Depletion of the SFP inventory removes the majority of the shielding fro the spent fuel assemblies such that direct radiation shine from the spent fuel will become significant. Fowever. the SFP walls 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.

Since the SFP inventory must be eventually replenished remotely, having FA 7 in the fuel mast does not impose any additional restrictions to the operability of the SSF RC makeup system.

During the time period of interest, no spent fuel was moved in the SFP.

Since the fuel mast provides a positive mechanical lock for the spent FA and the SFP bridge is seismically designed, no additional. potential for a 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 is 3.3 E-04. 'For the 25 day period FA-7 was ir the fuel-rest, the probability becomes 2.3 9-Os. Furthermore, typical PRA ca]"culationa utilize a 24 hour*

minimum time for the system relied upon to mitigate the accident.- In this case a time in the -range of 36-40 hours would have been available before 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-? was in a static, stable position such that the probability of fuel damage :_y another mechanism (collision.

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 SSP event actually did occur, an extensive period of up to 36 to 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> woild have been available for compensatory actions to ae taken prior to uncovering FA-7. Additionally, calculations show that FA-7 wo,-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,

NRC'FORM 366A U.S. NUJCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FAC4JTY NAME III 10

.05000DOCKET ILEA NJMSE 13PA06 13I Oconee Nuclear Station, Unit One 269 96 01 00 TEXT 01avwe sow# rec mar.usa adONrM o copes of NRC Formn 366&40I 1)

ArrACHf"E? A if I

-t ocogeFul PjcmC~nL Exentm el . npncrlptiori 1-092-0723 Wronigfile_.*x *se.mnly (FA) was placed into Unit I Reactor Core aurLng 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 *Wronqq FA was placed into Unit J. Reactor Core during refueling activities as 'a result of inadequate se4kf-.-cheCk and, independent "verification .Changes to the refueling,.procedure .were implemented as. corrective actions to prevent recurrence  !

1-094-0707 Refuelina x*etwgnce was altered at:the requ~et.of reiator engineers to Observe nucAeat instrumentation 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 the refueling procedure to prohibit sequence deviations without the use of a procedure change or test procedure.

1-094-0714 A -rono FA was placed into I';ait I Reactor. Core during refueling acivities as a result of ina-.quate self-check and independent verification. Corrective actions to prevent recurrence involved changes to procedures and methods of independent verification.

PIP a PROBLEM INVESTIGATION PROCESS

_ U.S. NUCLEAR REGULATORY COMMISSION 140C FORiF 36M (LER)

& LICENSEE EVENT REPORT f FACIUTY NAME 11)

Oconee Nuclear Station, Unit One COP" W--

TEXT CONTINUATION

.

DOCKET 06000 269 96 LEN NUMIE (6) 01.

-tgu 00 PAGE in 14 OF 17 i-I .XTto 71"Waco is Wan0W Ua "M*Do-* "

,ift- EILIgz3. vIif/Conceprns 2LaMieff ~Df 2-092-0024 A FA and control rod was damaged while the FA was being positioned for repair. The procedure was no, reviewed prior to the move and the control rod and FL were damaged. Corrective and actions to prevent recurrence involved procedure changes pedestal modifications.

3-092-0470 Bent spider assemblies causes delay in removal of burnable poison rods from two fuel assemblies. Zt could not be ct previous determined whether the damage occurred as a result fuel handling activities by Duke or by the fuel -endor.

Corrective actions involved manufacturing a component sizing template to be used by Qua4lity Assurance during the component inepection performed upon unloading of the-ne fuel assemAlies.. from.its 2-093-0431 An intermediate grid strap became torn and separated caused FA during.. refueling operation.. This type -o daqmag.is when the grid straps of adjacent assemblies esnagreach other, during fuel movements made in the core. Corrective*actions to procedure.

prevent recurrence involved changes to the refueling VA grid strap damage.

to provide new guidance to prevent the 3-094-0204 A dummy control rod assembly located in the deep end ofcore fuel transfer canal was struck while transporting the support assembly. This was a result of inadequate self-check of clearances. Crane control and water clarity problems contributed to the problem. Transport had to be halted to the transfer perform inspections of the core support assembly, storage racks. Corrective canal liner plate, and the fuel

-_o actions to prevent recurrence involved procedure changes incorporate preventive measures.

1-095- 1429 During reactor defueling activities, Spent Fuel Pool (SFP several times bridge hoist and grapple operation was hampered due to unexpected interference with consolidated fuel canisters. This interference problems in disengaging .rom Luel assemblies. Corrective actions involved moving the that 1s consolidated fuel canisters to an area of th* SFP outside of the off-load area.

damage on

. -095- 1462 FA NJO776 was found to have significant struc:-ral

n --he southw-9t four consecutive intermediate spacer grids cor.ier. No fuel rod damage was found or suspected.

U.. NUCLEA I&GULATORY COMUISSM

[ FIMA LICENSEE EVEMT RZPORT (LER)

I rrvYT rnMTINIJATInNl TEX C-..uK i m I 9*0K II NFACOlTY NAME I1l MA-F-C I1 ,AG0r ocone Nuclear Station, unit one

-I 269 196 01 010 15 OF 17 zrpszran QP2VTq'V 21M 3l-C95-01184 During set up of the B&W Fuel Reconstitution, elevator parts The sheared/fell from the elevator Into the cask area.

elevator part apparently sheared-when it contacted the cask design deficiency and worker attention to area wall. -levatOr detail contributed to this event.

atoroaw 2-M9)-0055 Several new fuel assemblies were received and placed in Root cause*

cells tzna wore not in accordance with qrpcedure.

to CoILoW procesure and iLattention to detail.

were Callure The fuel com;onmnts were not adversely affected-in tho "r FA. Corrective .ctaois Q414 A control rodwas t~ierted invoivea procedure .changes and persocnei training to. prevent recurrence.

A contractor personnel faile tdo -ollow procedural requireaents which for handling fuel rods during reconstitution activities.

-to' resulted. in severely bent- fuel rod and subsequent. cbsl~eng IV integrity. This resulted in a, NmC level the fuel cladding violation (PIP 2-N93-09171 for the failure of contractor personnel to follow proced.;;l requirements.

pert orymed I-M96-0002 A sequence within the I&- Insert shuffle procedure was incorrectly resulting in the qt~aosition of a&_IhmLe plDg La The verification process LdentLZLed ana corroc-ed the SrP.

this discrepancy. No corrCtiLVe actions to prevent recurrence were Identified.

.-.- ents Sim 91-15 This report descrLbes six industrv fuel ML9VOa.tOtLoLnl during retu" .nq and defuielingl activities as a result of Ln procedures. independent VerCIfca~ton. and inadequacies training.

Oconee's reVLiW of thLi event reDUltd in chanqie t.) rtf*uoling procedure changes and methods of cr~wunicatiofon evvnts hat SER %4- This report lew:rLbes six speci~fc industry human performance deflciencLeS wMLLe haninq r*ector -ore reSUIted Lin JLCUS t A or 4ther :or* -_mP*onekn.

.-oI*nentf tht damaqe. -iamaqe ro rovfudLnq "tqpment. %ndior incrrasa.1

.imi rn pors-ttISl fi~r lamAge I- luol )r '-hor -orot

U.&. NUCLEAR REGULATORY C3MMSO qc FCrM 3eUA LICENSES EVENT REPORT (LER)

TEXT CONTINUAT1O NAME Ml t OW 'n 16 OFM SFACUTY

-i 05000 -" I 1mTaR o mn mimomes 16 OF171 96 OI1 00 01ooree Nuclear Stat on. Unit One 26 MYT ow ffmwspace. U"~a~~twfcc -a of NRC Fvnw 366V I11I7) -

Oconee into the incorporated operations fuel industry lessoa these handling Lentsplan. and lessons learned ZI 94-3 This report describes potential probIems resulting fro.

inadequate oversight of refueling operations and inadequate performance on the part of refueling personnel based op four industry events.

oconee's -- vzew of this report remslted in no recommended actions based on actions taken with SIR 94-4.

IN 944-13, Sup. I* This report describes an industry event involving unauthorized movement of a defective spent fuel rod.

ocone~ o s rev ie w of thi s report: re sul ted ..in no r ecommended actions.

t= Level1 IV vieslakriesl 1110YAnbMn 2..L, 19901 One example of a failure to adequately implement a refueling procedure that resulted in a FA being placed in thwe wvm location in Root causes were operator errar and poor vjialnlaty In tn.e the core.

srP. corrective actions to prevent recurrence involved 7ounselLng the bridge operator.

=E LevelI 1M vLiQ+/-alaimLL2D +/- &1X 12, 1IL jne e*xampLe of failure to adequately Ilement & refueling procedure rhat resulted in a FA being placed in the.vror..snant rum, lewat.._

Rort causes were insufficient a-.tention to detaiL. LnouflICLent procedure detail and communlcCaton errors. Corrective ac inns to prevent recurrence involved procedural -hangeA and fu@L ,andlinq training S"i L&YAz "M '"IALLM IAbx " 19911

MOM SONA 364A.S. UCLL'.R REGULATORY COMANSSIM LICENSEE KVWa REPORT (LBR)

TEXT CONTINUATION FACILIT NAME (I I I OC"IT LEK WRIGME M ~ PG

.05000 [lI 17 OF 17 tOconee Nuclear Station, Unit One 269 I- 96 1 01 7 00 fIWr i Rmw sw@io 4e0uqa *sg M copm oN*fC F0om Il11n 1O Two examples of failure to implement refueling procedures that resulted in two fuel assemblies being placed in the wr.um ladeation in the core. Root cause was inadequate self-checking. "Corrective actzons to prevent recurrence involved procedural changes. (Covered by PIPs 1-092-0723 and 1-092-0724)

P-fuelina sequence w&a altered to observe nuclear instrumentation responsee vltfOut proper docueatnation and procedural control- This was perfor at the reqWest of Reactor Engineeri*n personnel.

,Covered by PIP 1-094-0707)

MC T-xml 1Y( vin+/-ar*n X=L civil 2nA&LXJAUU 19141 A VA retrieved from the vronr ap-em fuel A and pla-ed Ln the reactor core. Root causes were Inadequate, saLf-tcheck,an -ndependent verification. This was the fourth occurrence of failuro to identify and adequately verify FA Locations. Corrective actions-t' prevent recurrenco involved procedural changes an4 persocal training.

(Cover*d vy, PIP 1-094-0714 And PIP 1-094-07.07.

EXHIBIT B- 15 Oyster Creek Unit 1:

LER 219/87-006-00 (February 24, 1987)

gCL w.LO* _ 01R_

t 6MtATO)Y CO(--OM I

"UNS EVENTo "

APPROVPO 1LR OU 01So U6 "WIUCENSEE EVENT REPORT (LERI 'ex" '

DGET I2 (

PAGSafIN

1. w ' ism"j~nl10Creek, Unit 1 Oyster 1 l nI nlol )l1 11 91,l r 1 I T" - TECHNICAL SPECIFICATION VIOLATION CAUSED BY IMPROPER STORAGE OF HIGHER ENRICHM*ENT D

TO Lift PmUON?

f tlE 3OUVITIAL I1 InV" R -w "O"T" DAY AI YEAR OTh FACLTY WAO Pac -

-- s----*SN-

011tI~

OOCKlY MUOl1ll3 I

!

tll I05

• 11 21187870 610 r.VihD PUMA5I 10 TWIN9M.""NOM6 DIP c CPR 00 -W or MW, st HI m

-..:: ::::::.

' N:::::::::::

? - i .----- :b m U 2A iS md SOTHR111 m .n* a 't7,.Ib Molotov 2t0 R WEII.)l NIMI nei"4w. OW #0 I .

LI)CSNIE CONTACT F06 THIS L611 M) 04AMI [ 1 I !b!1I II [j1u, . I IT .lWEI 1 1 ICOJMS0 APIA'4f4

!

6 1 019 91711 -14 t6 1319

-!

Hari S. Sharma, Core Engineer

. _____1__Ii COINIPETIN ONE __I L109 Pool KAZ4 WANUP

[]

"7COVNEW0* T PAILUME o1ac*SiO MID IftT REPORT 1130 1..

INI~~APIUPAC- REPCOITAII.....

. ..

SYSTEM CO#APONEP1' TUNtER TO NMO .....

COMouo"s[ MIUFC EOTA CAUSE CAUSE SYST11s L.LL..L. ~~ ~

____- ~ .~... L.~ ...LL.L.

~ ~ ~ . ...........

EXPECTED %MONTW DAY YPA SJPPLEMIINTAL REPORT EXPECTED t141 SSAMISSIOP4 DATE l15t YES UXPPTIit v XIJU#SIOM c~N A4S fuel stored in Oyster Creek Technical Specification 5.3.1(C) specifies that the average planar the fuel pool storage racks shall not exceed a maximum enrichment of 3.01 wt% U-235. Contrary to the above, reload fuel bundles planar enrichment supplied by General Electric Company (GE) having an average pool during the lR outage of 3.19% U-235 were temporarily stored in the fuel error in not in 1986. The cause of the event is attributed to personnel and in not of the new fuel performing a thorough safety analysis for storage fuel storage prior to recognizing a conflict with the Technical Specifications in the spent fuel pool.

procedures, revising Corrective actions will consist of revising the refueling I

on stored the Technical Specifications to raise the enrichment limitations fuel, and reviewing the occurrence with engineering personnel.

r r)H f\ [ C) K 7 *1

qU I WA *tREGUILATMV CO-Mf PWW~~~ 304 o -01 0'o UCENSEE EVENT REPORT (LMR) TEXT CONTINUATION A-*oOE rrFR[v I/II DOCK97 ki~p-04 En LEN WJWWM M PAM t

,O=U" IGMMAMi I)

'SAN U6NIAL Ov" Oyster Creek, Unit 1 o ji I121 9 817 -- 1OOI 6-10 1 0 1 2oF 013 9f MWTfET aa *ol M06 OWWA twmwI DATE OF DISCOVERY review of The violation was discovered on January 21, 1987 during a subsequent to the the Oyster Creek Technical Specifications for potential changes related new fuel design.

IDENTIFICATION OF OCCURRENCE in the Fuel with an average planar enrichment of 3.19 wt% U-235 was stored Technical Specification 5.3.1(C) spent fuel pool beginning February 27, 1986.

storage facility shall not states that the fuel to be stored in the spent fuel is wtZ U-235. This event exceed maximum average planar enrichment of 3.01 reportable under IOCFR50.73 (a)(2)(0)B.

CONDITIONS PRIOR TO DISCOVERY mode in At the time oa' occurrence, the plant was operating in a coastdown time of discovery, the plant was at preparation for the 11R outage. At the for Cycle 11 operation. All the fuel approximately 20% power starting up fpr bundles which exceeded the Technical Specification enrichment limitations removed from the spent fuel pool and loaded storage in the fuel pool had been in the core.

DESCRIPTION OF OCCURRENCE of A total of 204 GE P8DRB299 fuel bundles, with an average planar enrichment 3.19 wt% U-235 and a bundle average enrichment of 2.99 wt% U-235. were received for in 1986. At the time of fuel receipt, the dry storage vault had a capacity 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 P8DR8299 fuel in the spent fuel pool had been transferre! '-o the core.

APPARENT CAUSE OF OCCURRENCE The cause of this occurrence is attributed to personnel error. The safety operation of the plant analysis which was prepared was orienten toward the safe It did not take into using the higher enrichment fuel during the next cycle.

Fuel pool account that the new fuel could conceivably be stored in the spent (only dry storage was considered). Had this possibility been envisioned, the need for a Technical Specification change would have been recogniZed.

U.a PCLIAX 181A.ATONY COWWM LICENSEE EVENT REPORT (LER) TEXT CONTINUATION POVIOown Nor-410 MAM ! P44 0WEAN WM"Mt"LI to-OhUTY NAAM fit r 22Gf rI-~IIi L n . In- 'In Oyster Creek, Unit 1 to0 1510 1 1 I--I"u

-- Ig--Iv I"1I:011" "

controls were A contributing factor Is this event is that procedural a precaution regarding inadequate. The refueling procedure (205.0) contains enrichment, however, the bundle the Technical Specification restriction on fuel to ensure compliance with the refueling procedures do not require verifications in the spent fuel pool. Had enrichment restriction associated with fuel stored not have been stored in the such a verification been performed, the fuel would spent fuel pool.

ANALYSIS OF OCCURRENCE AND SAFETY ASSESSMENT Poison Racks (HDPR)

The initial criticali*y analysis of the High Density U-235. The analysis also did assumed a uniformly enriched lattice of 3.01 wt, in Regulatory Guide 1.13. The not take credit for burnable poisons as allowed to be 1.33 which resulted in lattice K-infinity for the analysis was determined bundles in the spent a HOPR cell K-effective of less than 0.95. The P8DRB299of 1.22 as determined by fuel pool have a maximur cold, uncontrolled K-infinity cell K-effective did not the fuel vend6r. Therefore, it is expected that thethe HDPR criticality reach or exceed 0.95. However, a re-evaluation of for burnable poisons. The analysis is currently being performed taking credit report.

results of this analysis will be submitted in a supplement Corrective Actions planar enrichment of Currently, there are no fuel bundles with an average Corrective actions will greater than 3.01 wt% U-235 in the spent fuel pool.

consist of the following:

added that

1. Fuel movement procedures will have appropriate controls ensure Technical Specification compliance in this area.

a Technical

2. Based upon the results of the HDPR re-evaluation, 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 licensingsafety basis docunents and associated restrictions when performing reviews.

SIMILAR OCCURRENCES None (0288A)

I I

mUMuclearSu Post Othi<e Box 388 SouthNew Jeirsey 08731-0388 Route 9River.

Forked 609 971-4000 Writer's Direct Dial Number:

February 24, 1987 U.S. Nuclear Regulatory Commlssion Document Control Desk Washington, DC 20555

Subject:

Oyster Creek Nuclear Generating Statton 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 7Tele Vice President and Director Oyster Creek PBF: KB: 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 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

'f'

GPu Nudear Post Office BoxCorporation 388 Nuclear Route 9 South Forked River. New JeFsery 08731-0388 609 971-4000 Writers Direct Dial Number:

February 24, 1987 U.S. Nuclear Regulatory Co*m* ssion Document Control Desk Washington, DC 20555

Subject:

Oyster Creek Nuclear Generating Station Docket Ne. 50-219 Licenso' Event Report copy of Licensee Event Report (LER)

This letter forwards one (1)

No.87-006.

Very truly yours, Pete r---ie]r Vice President and Director Oyster Creek PBF:KB: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 Commission 7920 Norfolk Avenue, Phillips Bldg.

Bethesda, MD 20014 Mail Stop No. 314 NRC Resident Inspector Oyster Creek 4uclear Generating Station Forked River, NJ 08731 'I'

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 INFOWTION NOTICE 94-13: UNANTICIPATED AND UNINTENDED MOVEMENT OF FUEL ASSEMBLIES AND OTHER COMPONENTS DUE TO IMPROPER OPERATION OF REFUELING EQUIPMENT power All holders of operating licenses or construction permits for nuclear reactors.

The U.S. Nuclear Regulatory Commission (NRC) is issuing thisfrom information notice to alert addressees to potential problems resulting 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 DescriDtion 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 original location in the reactor core. The licensee suspended fuel handling had not and investigated the event. The licensee determined that the grapple 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 assemblj that was being moved to a fuel sipping can was inadvertently lowered, instead of raised, striking another core fuel component. The potentially damaqed fuel assembly was then moved to the 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.

'93-B81, issued The AlT documented its f'.ndings in NRC lnpec'iýn Report 59-27 ne1 October 21, 1993. The All con-lud-,d that m*stakes mad-. b, '  ;,÷r 940?- 32tt '-

tvays;unee axio

-am

~ Atfbu~ba n04 n was.~ ef fecdl Iy t opei,4ors were nott awareli eiffl 1.ly. the: personnel-.,

' t de_.t. lingep act jties

'S w ott .

tomost ins e of .the reqbiremet. to

• lsure lien engaging and 1 tihg fuel assemblies.

isual~ly verify gra*p ectations and provide

-the AlT-:fou" th*at .4nai t di.. not communocate 4 proper- dversight of fuell handling actiities.

peach BottoM Events With Unit 3 shut down for refuelf.tg on September 23,, 1993, a fuel assembly could not be fully inserted into its spent fuel rack cell. It was thought that t*e fuel assembly h4d swellnd due to trradiatl in-the core, and the fuel astembly was successfully placed in a different- cell'. It was further postulated that there might be some debris in the. cll, and that the cell should be checked at soe future date. On September; 4, 1993, another fuel assembly became stuck in its. spent fuel rack cell. TheV icensee evaluated the material condition of the fuel assembly, calculated an allowable lifting force, and conferred with the fuel vendor. The licensee innc-eased the load limit of the refueling hoist and the fuel assembly was freed from the rack

- with no damaoe- to the fuel assembly- Subsequent examinations 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 corv;ucted 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.

• _.souehanna 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 fuPe assembly from a peripheral location in "he core. The personnel involved realized they had removed the.wrong assernly and they inappropriately dccided to return the assembly to its prior positwon in the core. The appropriate action, per licensee procedures, would have '-*en to place the bundle in the spent fuel pool and secure fuel ha ,'n; a'--:,ip; until the cause of the error was detco r:-- a-d c rrect.

j .. . .. .ary F...a...3 of,*225 . 19

"z oire a. fis 'isembly into the ýpre during ob 6;, 9"3, wh s s t of 'the ober.Inq~-4n e99, wh..droto4te 2o toý 2Ct b S Stl hgenough thehigh hit the to-*

blado guidetoC00te .Th 1 cesee sUsPecflT

• ...Th sideO of the,-'ctorvesset th~evess$*. b&cs*e rfueing9 it was not te6 the mast The ties, revised theassociated proedt,-a the iasdwi fuel b"dl1fl .equipment creload was resumed after survell aIce on atte~pttQpg to grapple 1 ,successfully conducted. On October ZR, 1993, while the fuel fuel assembly in the fuel pool, the personnel perfovminf to the pool for Wadino activities heard two loud bangs and observed bitbbls S 0seconds. Subsequent inspection ire~ialed that ont section Of the bymast the fro Unit 2 was bent. The licensee believes during that the mast.was weakened the October 27 event.

i*oct with the reactor vessel that occurred On.: October 29, 1993. the NRC dispatched an AIT to the site to review the

, eve . The AIT documented its findings in Inspection Report 50-W87/93-80, that facility management did issued on December 21, 1993. The AIT concludedactivities and that inadequate not maintain proper oversight of refuel floor corrective actions were implemented in the past for problems with the fuel handling equipment. The AIT also concluded that the licensee fuel handling procedures were adequate for the proper completion of the fuel handling could be made to increase the activities, although certain improvements awareness of the operators concerning potential problems.

Nine Mile Point Event on Nine Mile Point Unit 2 was shut down with refueling in progress into the spent the core November 1, 1993, when a blade guide was moved from the grapple and fuel pool. The contractor refueling orerator disengaged There was no procedura, observed the correct light indication 'n the bridge.

the Senior Peactor requirement to visually verify disengagement or for to verify Operator Limited to Fuel Handling (LSRO) or the spotter The refueling operator noticed increased drag after the disengagement. cm (9 in) toward the approximately 23 refueling bridge crane had been moved that the blade next location. At that time, licensee personnel determined to its The bridge was returned guide was still engaged on the grapple. (positive previous position, the blade guide was lowered and disenqagei pyoceided t, .oie the verification was obtained this time), and the operator afs'b 1 low.?ri*r t*-i: 'A' next component, which was a fuel assermbly. While

t1a z. 4aperf therel r each:

seveal ttp* l.- ft.

Is"ues MOMd Is sua li5

~1ICtI0Sof bnrappIing wWe the mAt.,h tobe!rified by ýralsifl and 0oat* the blade ot verify dtSengageet aferrleaSing the LSRD of the 2to0 -the .rfueling operator did not nOtify Ipr isponse et (remaining co"Wed. to the blade g914e while ridg). Also contributing to the e$

it the fact that the

. i a refueling bridge trolley bearin*g about which be was opt Cthat vt.t'm than the hadling ment of the expectations blade guide. License. review regardin the supervision of refueling acivit~ie *ad not been clearly expressed to the ISROs Refue0tig activities are safety-sign.ificant operationS that are not conducted on aIrotne basis. In addition, fuel handling activities are often performed of licensee personnel. As a by contractor personnel under the supervision with the fuel handling result, fuel-handling personnel may not be familiar fuel handling operations equipSrit or may feel that their experience in for procedural use and adherence.

pemits then to ignore some requirements managietemt attention and Either of these situations could require increased performance of fuel overs.ght by the licensee to ensure proper and safe handling activities.

Reouuletions Appendix B to Part 50 of Title 10 of the Code of Federal to control procedures (10 CFR 50) requires licensees to have appropriate to be taken during operktton aCtivities affecting quality (such as the actions followd. In are used and of refutling equipment). and that the procedures a trai*lng progrIm tf;r a4dtio'n, I0 CFR 50.120 requires licensees to implement to ensure that those various categories of nuclear power plant personnel and abilities to perform their personnel have the necessary knowledge, skills, personnel (including assigned Jobs competently. This rAle applies to the of the refueting contractors) who operate nr supervise the operation

,n wh'ch the equipment. The cases dtscusstd in this notice include sitvilio-us! F ) .

licensees failed to conduct appropriate traininq in the r.j,'

equipment. particularly witN respect to deplqn rodif'ca"'.v, " °.

TIt-,o ,,p '"

, P. 4!; .

4 'i '.g -' .1 controls for the fuel 'rasl

IN 94-13 February 22, 1994 Page 5 of 5 wtre variously not aware that handling personnel involved in certain instancis from expected results, cease deviatiors Management expected them to identify abnormal condition is encountered, and notify operations when an unexpected ur unexpected or abnormal conditions.

operations and/or plant management of action or written response. If This Informatiton notice requires no specific in this notice, please contact you have any questions about the information or the appropriate Office of one of the technical contacts listed below, manager.

Nuclear Reactor Regulation (NRR) project Brian K. Grimes, Director Division of Operating Reactor Support Office of Nuclear Reactor Regulation P. L. Eng. NRR E. M. Kelly, RI Technical contacts: (215) 337-5183 (301) 504-1837 J. R. White, RI L. E. Nicholson. R1

(?IS) 337-5114 (2;5) 337-5128 At t chmenrt y Issued NRC of Recer, ly rf a ," Notic -S L6.ist

Attachment IN 94-13 February 22. 1994 Page I of I LIST OF RECENTLY ISSUED NRC INFORMATION NOTICES InforsutIon Date od Subject Issuance Issued to Notice No.

02/09/94 All holders of OLs or CPs 94-12 Insights Gained from for nuclear power reactors.

Resolving Generic Issue 57: Effects of Fire Protection System Actuation on Safety Related Equipment All holders of OLs or CPs Turbine Overspeed and 02/08/94 for nuclear power reactors.

94-11 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 02/03i94 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 01/0 1/94 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 01/31/94 All holders of OLs or CPs 93-26, Grease Solidification for nuclear power reactors.

Supp. I 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 auth'or-zeD' so01 lY for the Revised 10 CFR Dart 20 se 3 P- so,;rý -s, OL . OperatIqq L.Cese CP = Construction Permit

UNITED STATES NUCLEAR REGULATORY CL*"IISSION OFFICE OF NUCLEAR REACTOR REGULATION WASHINGTON, D.C. 20555 June 28, 1994 1: UNANTICIPATED AND UNINTENDED NRC INFORMATION NOTICE 94-13, SUPPLEMENT MOVEMIENT OF FUEL ASSEMBLIES AND OTHElR COMPONENTS DUE TO IMPROPER OPERATION OF REFUELING EQUIPMENT permits for nuclear power All holders of operating licenses or construction reactors.

is issuing this information The U.S. Nuclear Regulatory Zommission (NRC) event involving unauthorized notice supplement to alert addressees to an It is expected that recipients will movement of a defective spent fuel rod. to their facilities and consider review the information for applicability problems. However, suggestions actions, as appropriate, to avoid similar not NRC requirements; therefore, no contained in this information notice arerequired.

specific action or written response is Bac kiround and Unintended The NRC issued Information Notice (IN) 94-13, "Uranticipated Due to Improper Operation of Movement of Fuel Assemblies and Other Components that could result from Refueling Equipment," to alert addressees to problems and inadequate performance on the inadequate oversight of refueling operations part of refueling personnel. IN 94-13 described various refueling events that and Nine Mile Point.

occurred at Vermont Yankee, Peach Bottom, Susquehanna, controls over, and operation These events demonstrate the importance of proper event at the Waterford Steam of, refueling equipment during use. A recent potential for fuel damage or Electric Station (Waterford) demonstrates the equipment that is not personnel hazards which could result from fuel-handling use.

properly stored and not secured from unauthorized Qescription of Circumstances operating at 100-percent power On February 18, 1994, the Waterford plant was o,*ect hanging from the when a senior reactor operator found an unknown Health physics fuel-handling machine in the fuel-handling buildinr. fuel pool area and found technicians measured radiation levels in the spent with vise them to be normal. Licensee personnel ;-omotely secured the object levels were .2 to .7 Sv/hr grips and determined that underwater radiation from the object. A Combustion (20 to 70 R/hr] at 15 centimeters [6 inches]as a fuel rod encapsulation tube.

Engineering employee identified the object The licensee posted a security No visual damage was apparent on the tube.

the event to the NRC.

guard in the spent fuel pool area and reported 9406220075 P~k ' C*

rI /

IN 94-13, Supplement 1 June 28, 1994 Page 2 of 3 and determined that the tube The licensee reviewed fuel storage records removed from an irradiated fuel contained a defective fuel rod that had been time, the tube had been placed in a assembly several years earlier. At that in the spent fuel racks. The licensee center guide tube in a grid cage stored fuel-handl~ng area and interviewed reviewed computer access records-for the relevant personnel about the event. Personnel who may have had access to the regarding the event. The fuel-handling machine completed questionnaires had used the fuel-handling 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 not used tube. However, the refueling director had tube was hanging from the that he would have noticed if the encapsulation records indicated that hoist at the time he used the machine. Surveillance attached to the fuel-handling the fuel rod encapsulation tube must haie become tool sometime between February 11 and 18, 1994.

tube showed that the Design drawings of the cap of the fuel rod encapsulationinner diameter of the end of outer diameter of the cap was about equal to the become bound in the the fuel-handling tool. Apparently, the cap had the top of the spent fuel fuel-handling tool when the hoist was lowered to completely removed from the rack and, when the hoist was raised, the tube was grid cage.

operations for previous Although contractors had performed the fuel-handling, to perform the fuel refueling outages, Waterford personnel were scheduled licensee speculated that handling for the March 1994 refueling outage. The the March outage one of the people assigned to fuel-handling activities.for while practicing the use may have inadvertently lifted the encapsulation tube physics staff before of the hoist. Personnel were required to notify health records showed that accessing the refueling machine; however, health physicsNo keys or special time.

no one had made such a notification during this the fuel-handling machine.

knowledge was needed to access the controls of electrical breakers and Electrical power could be obtained by closing two The licensee questioned pushing one switch that were located on the machine. use of the several employees, but ,)o one admitted to unauthorized fuel-handling machine.

the computer that As an interim corrective action, the licensee deenergizedin a locked power a breaker controls the fuel-handling machine by opening a means to prevent the develop control center. The licensee planned to (1) lifted by the fuel rod encapsulation tube from being inadvertently procedure warning fuel-handling tool, (2) add a precaution to the operating the storage location, and operators not to lower the fuel-handling tool over proficiency training.

for (3) add hoist manipulations to the lesson plans.

Discussion fuel and core Procedures governing the use of equipment for handling operation of that components may not prevent unauthorized or unintended the equipment. Precautions such as locking out breakers that energize in highly visible areas fuel-handling equipment and the placement of placards equipment is forbidden declaring that unauthorized opPratiOn of fuel-handling

IN 94-13, Supplement 1 June 28, 1994 Page 3 of 3 used without proper authorization.

may help ensure that the equipment is not in an area where accidental Additionally, storing the fuel-handling machine will not impact stored fuel or other movement of the hoist or grapple to the prevention of inadvertent fuel movement or

-;'-'-components may contribute of fuel and core damage. Management attention and oversight of the operation that fuel and core component handling equipment is important to ensure movement and that plant components are protected from damage or unauthorized to radiation.

personnel are protected from unnecessary exposure or written response. If This information notice requires no specific action this notice, please contact you have any questions about the information in Office of Nuclear te technical contact listed below or the appropriate P:actor Regulation (NRR) project manager.

atbL-Cd# ',

Brian K. Grin.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 I of 1 LIST OF RECENTLY ISSUED NRC INFORMATION NOTICES

1nfoii~al~flDate of Issuance Issued to 1Notice No. Subject Accuracy of Information 06121/94 All U.S. Nuclear Regulatory

.94-47 Commission Material Provided to NRC during the Licensing Process Licensees.

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 14-45 Potential Common-Mode for nuclear power reactors.

Failure Mechanism for Large Vertical Pumps 06/16/94 All holders of OLs or CPs 94-44 Main Steam Isolation for nuclear power reactors.

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 06/07/94 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 t icense CP = Construction Pc mit

EXHIBIT B-17 Three Mile Island Unit 1:

LER 289/98-002-01 (April 3, 1998)

GPU Nuclear, Inc.

Route 441 South Post office Box 480 NUCLEAR -Middletown, PA 17057-0480 Tel 717-944-7521 April 03, 1998 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555

Subject:

Three Mile Island Nuclear Station, Unit I (TMI-l)

Operating License No. DPR-50 Docket No. 50-289 Lice F en) No.98-002, Revision I not sampled in accordance with the On February 4. 1998, GPU Nuclear determined that the Spent Fuel Pool was Surveillance Requirement (SR) specified in Table 4.1-3, item 4, which requirements of the Technical Specifications of work activities determined that no sample was taken requires sampling monthly and after each makeup. A review was found to be reportable in accordance with 10 CFR following a water addition on January 23, 1998. This condition A.subsequent analysis determined that the filling 50.73(a)(2)(i)(B) as a.condition prohibited by Technical Specifications.

significantly.

activity could not have diluted the boron concentration 3, 1998. Attached is Revision 1 of LER 98-002, which This condition was reported to the NRC by letter dated March the reason for this event, the extent of the problem provides additional information that addresses the following items:

safety consequences and implications of the event, and the associated with the missing operator aid, the assessment of the corrective action section.

The event did not affect the health and safety of the public.

if you have any questions regarding this matter.

Please contact Adam Miller, TMI Licensing at (717) 948-8128 Sincerely, Ja~i anebas' Vice President and Director, TMI AWM cc: TMI Senior Resident Inspector Administrator, Region I TMI-I Senior Project Manager File 98048 9604130278 980403 PDR ADOCK 05000289 S PDR

NuoaZhR IJW3?TORX CMGC~ss:

N royN 366A,' -u.s.

(4-951 LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION DOCKZ. LKf K3SIZ (6) PAGS (3)

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NUC Y EAR SEQUENTIAL REVISION*

-- NUMBER NUMBER Three Mile Island, Unit 1 05000289 98 -- 002 -- 1 5 OF 6 IL (17)

(If more space is required, use additional copies of NRC Form 366A) over a approprIate because no major replenishment of pool water is expected to take place The short period of time." This bases appears to be consistent with the TMI-1 TS bases.

TMI-1 staff is evaluating if a request to revise the current surveillance requirement is appropriate.

VII. Corrective Actions:

A. Corrective Actions Taken:

(SFP)

1. A new Operator Aid has been posted at the valve that is used tp fill the Spent Fuel Pool tracks the need from the Reclaimed Water System. In order to ensure the Shift Supervisor 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.

to

2. The Primary Auxiliary Operator Turnover Checklist has been revised to include a requirement to the Spent Fuel notify the Chemistry Department of sample requirements if a water addition are Pool has either been initiated or completed durihig the shift. The contents of this checklist 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 by Chemistry Departments. The review will be documented and the documentation maintained 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 of the check of Operator Aids will be performed. Each Shift Supervisor will select five (5)

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 physically 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 MORM366A 14-95)

V.S. 1HUC.AD. RZR7LL!ORX COMWSSXO znxq Font 366k

-

(4 -95)

LICENSEE EVENT REPORT (LER)

TE.XT CONTINUATION DX= I- LEa Hulna= (6) P 1'= (3) r7f = HAMN () NUMB= (2)

A n NUMBER NUMBER tYt*trssZ*f U.S.* win.]UgGUZ,.L*On"v SEQUENTA

-TAYEAR R EVISIOuN 98 -- 002 -- 1 6 OF 6 Three Mile Island, Unit I 05000289 7 rf more space is required, use additional copies of NC Form juO A i%-'I insure

3. All of the Operating Procedures which contain Operator Aids will be reviewed to 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.

of Technical

4. All licensed personnel will be given training on an overview of the contents 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 (E-IS), System Identification (SI) and Component applicable, as Function Identification (CFI) Codes are included in brackets, [SUCH] where required by 10 CFR 50.73 (b)(2)(ii)(F).

VRC FORM 365A (4-95)

I*

• C 366 U.S. M R p =RX C0ISSIO1% ARP2aVED BY COB NIo. 31.o-0104 KC 04/30/98 4-95) 3EXPI.R ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS MANDATORY INFORMATION COLLECTION REQUEST: 50.0 HPRS.

REPORTED LESSONS LEARNED ARE INCORPORATED INTO THE LICENSEE EVET REPORT (LER) LICENSING PROCESS AND FED BACK TO INDUSTRY. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (T-6 F33), U.S. NUCLEAR iSee reverse for required number of rev e r3 r Eo e lREGULATORY COMMISSION, WASHINGTON, DC 20555-0001, AND TO See THE PAPERWORK REDUCTION PROJECT (3150-01041, OFFICE OF dIgit5/characters for each block)

MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.

.IT 'Du.T*hUB I-MuT (2) 1AE (3)

Three Mile Island, Unit 1 05000289 1 OF 6 MI.,SSED SPENT FUEL POOL SAMPLE FOLLOWING A WATER ADDITION DUE TO UNFAMILIARnY OTHM N'ITH SAMPLING REQUIREMENTS AND A MISSING OPERATOR AID U

-r(iza

-un (6)

DOCKET NUMBER MONTH 2

DAY 04 YEAR 98 YEAR 98 I SEQUENTIAL NUMBER 002 2

[

REVISION NUMBER)OETNBR 1

MONTH 04 DAY 0

YEAR 98 FACILITY NAME FACILITY NAME DOCKET NUMBER M1IQRZRWTS O* 10 C*i a+: (Check one or more) (III THIS B1POBT 1S SUBMITTEI PIURSU*" TO TXH wine tQ% N 20.2201(b) 20.2203(a) (2) (v) X 50.73(a) (2) (I) 0.13(a2) (viii) 50.73)a) (2) (ii) 50.73(a) (2) (x)

T T fl - 20.2203(a) (1) 20.2203(a) (3) (i) 100 20.2203)a) (2) 1il 20.2203(a) (3) (ii) 50.73(a)(2)(iii) 50.73(a)12)(iv) OTHER 20.22031a) (2) (7i) 20.2203(a)(4) 7

50. 3(a) (2) (v) Specify in Abstract 20.2203(a) (2) (iii) 50.36(c) (1) or in NRC Form 50.below 50.36(c) (2) 20.2203(a) (2) (iv)

.L'111_ *8- " IR TNIS LEr. [12)

NAME ITELEPHONE NUMBER (Include Area Code)

Adam Miller, TMI Licensing Engineer (717) 948-8128

_ 7 W I T W r U FO R uEAC C M E W_ A L *E D S T z N MS R P ( M3 I Yes, (If yes, "rA?

compilete (Limit to 1400 EXPECTED SUBMISSION spaces, i.e.,

DATE).

approximateLy i/ sirngi February 4, 1998, GPU Nuclear determined that the Spent Fuel Pool was not sampled in accordance with the 4, which requirements of the Technical Specifications Surveillance Requirement (SR) specified in Table 4.1-3, item no sample was requires sampling monthly and after each makeup. A review of work activities determined that in accordance taken following a water addition on January 23, 1998. This condition was found to be reportable determined with 10 CFR 50.73(a)(2)(i)(B) as a condition prohibited by Technical Specifications (TS). An analysis that the filling activity could not have di'uted 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 to replaced and the Primary Auxiliary Operator (AO) Turnover Checklist has been revised to include a requirement either been notify the Chemistry Department of sample requirements if a water addition to the Spent Fuel Pool has initiated or completed during the shift. Additionally, licensed personnel will be given training on Technical I Specification section 4 requirements.

and safety of There were no adverse safety consequences from this event, and the event did not affect the health the public....

9804130281 980403

m366A6 "

U.. JCLZAA P1joLAim;X camuisiioa I (4-9 - LICENSEE EVENT REPORT (LER)

"TEXT CONTINUATION iL= MtAG- (6) PjA= (3)

OX KAMczrf(1)

SEQUENTE REASL Y EAR ER 05000289 NUMBER KNrUvM1 98 002 1 2 OF 6 Three Mile Island, Unit 1 Form 366A) (17)

Tcr (It more space is required, use additionaJ copies of VRC I. Plant Operating Conditions before Event:

to and during the event described in this TMI- 1 was operating at 100% steady state power prior LER.

Inoperable at the Start of the Event and II. Status of Structures, Components, or Systems that were that Contributed to the Event:

None.

III. Event

Description:

that the Spent Fuel Pool Water [DA]* be The TMI-1 Technical specifications Table 4.1-3.4 requires 1104-6 "Spent Fuel Cooling System" sampled monthly and after each makeup. Operating Procedure completion of a Spent Fuel Pool water addition requires notification of the Chemistry Department at the the addition was completed.

that a sample must be taken between 24 to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after to the Spent Fuel Pool on 01/23/98 (From 0918 Contrary to thcse requirements, a water addition was made addition was made to the Spent Fuel Pool to 1705) without the required follow-up water sample. Another then sampled on 01/28/98 at 0430 and again on 01/27/98 (From 1410 to 1817). The Spent Fuel Pool was sample requirement for the addition that was on 01/29/98 at 0830. These samples exceeded the 48-hour performed on 01/23/98.

a Staff Chemist noticed that During a routine review of work activities by the Chemistry Department, and 01/29/98. He recognized that these samples of the Spent Fuel Pool had been obtained on 01/28/98 the Spent Fuel Pool after each addition.

samples were taken to comply with the requirement to sample for samples after filling the pool to the The Staff Chemist identified a possible lack of formal tracking the scope of the potential problem by Manager, Radwaste and Chemistry. The manager investigated data and the computerized Control Room reviewing two months of spent fuel pool boron concentration following an addition to the pool on Logs. lie found that there was no spent fuel pool boron data Process) Form (T1998-0066) to document 01/23/98. The manager submitted a CAP (Corrective Action the missed sample.

NRC FORM 366A (4-95)

..S. 1MUr.ZAR BZGDX.&=P C02.SS31in 106366i.

LCENSEE EVENTr REPORT (LER) yFRISEQU N LAL REVISION NUMBER NUMBER 05000289 98 -- 002 1 3 OF 6 Three Mfile Island, Unit 1 --

of V'RC Form 366A) (17)

TZ=! (If more space is requ-ed, use additional Copjies IV. Identification of Root Cause the Spent Fuel Pool. This The Primary AO was notified at the shift turnover meeting of the intent to fill system on the secondary task is coordinated with processing (i.e. purifying) water with the ECOLOCHEM Water Storage Tank plant. As water is processed on the secondary plant it is transferred to the Reclaimed Fuel Pool. Towards the end on the primary plant. This tank is then used as the source tank to fill the Spent system which would in turn of the shift the AG was notified of the intent to shutdown the ECOLOCHEM Fuel Pool. When the require securing the filling of the Reclaimed Water Storage Tank and the Spent in his logbook and also notified Primary AO terminated the filling of the Spent Fuel Pool he made an entry to the Control Room. Contrary the Control Room. He did not convey any sample requirement inform-ation did not notify the Chemistry to the requirements of the Operating Procedure the Operations Department did not track the need for Department .of the need to sample the Spent Fuel Pool and the Shift Supervisor was aware that the Spent the water sample on the Shift Supervisor's Turnover. The Shift Supervisor Specification section 4 Fuel Pool fill had been performed, but he was unaware of the Technical sampling requirements.

does not require the operator The task of filling the Spent Fuel Pool is considered a routine evolution that For this reason an Operator to actually use a copy of the Operating Procedure to perform the evolution.

Pool in order to remind the Aid is affixed to the wall directly behind the valve used to fill the Spent Fuel Aid was missing from the Operator of the notification and sampling requirements. However, the Operator no Operator Aid available to wall on 0 1/23/98 when the Spent Fuel Pool was filled. Therefore there was at the completion of the serve as a reminder to the Operator to notify Chemistry and the Shift Supervisor fill process.

was made to the Spent There has been one previous occurrence, June 13, 1996, where a water addition the first that resulted in an LER Fuel Pool without the required follow-up sample being performed. (This is Tech Spec Surveillance). As a due to the recent change at TMIL-1 conc~erning the reportability of a muissed Procedure 1104-6 "Spent result of the previous occurrence, the procedure guidance contained in Operating Department of the required Fuel Cooling System"e was enhanced to require notification of the Chemistry until the sample is taken and sample and to track the need for a sample on the Shift Supervisor's Turnover was added to indicate that analyzed. As part of the procedure enhancement, an enclosure to the procedure an Operator Aid is posted at the Reclaimed Water supply valve.

sample of the Spent Fuel Pool Factors which contributed to the failure to obtain the Tech Spec required following a water addition are:

"* Pertinent information not transmidtted

"* Required procedure/document not followed

• Installed Operator Aid not provided (i.e. missing)

NRC FORM 366A (4-95)

U. S. U.ZAa RZ3LATOR cSSZcV U.S. mx=ZAR WCOUIATORX C064C331ca

rood 3S66 LICENSEE EVENT REPORT (LER)

'r*YT ON'TINUATION

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NUIMXR (2)

M YTEAR SEQUENTttALu092- 1 OFB6 4EVISION 05000289 98 -- 002 - 1 4 OF 6 Three Wie Island, Unit 1 T (If more space 25 required, use acditional copies of NRC Form 366A) (17) relies on As stated these are all contributing factors. We incorporate a work ethic that 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:

no The failure to obtain a sample of the spent fuel pool following its fill on January 23, 1998 had adverse safety consequences.

Section 5.4.1 of TMI-1 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-l 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 (4-95)

NuoaZhR IJW3?TORX CMGC~ss:

N royN 366A,' -u.s.

(4-951 LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION DOCKZ. LKf K3SIZ (6) PAGS (3)

-Ac1.Mr RAM (1) (2)

NUC Y EAR SEQUENTIAL REVISION*

-- NUMBER NUMBER Three Mile Island, Unit 1 05000289 98 -- 002 -- 1 5 OF 6 IL (17)

(If more space is required, use additional copies of NRC Form 366A) over a approprIate because no major replenishment of pool water is expected to take place The short period of time." This bases appears to be consistent with the TMI-1 TS bases.

TMI-1 staff is evaluating if a request to revise the current surveillance requirement is appropriate.

VII. Corrective Actions:

A. Corrective Actions Taken:

(SFP)

1. A new Operator Aid has been posted at the valve that is used tp fill the Spent Fuel Pool tracks the need from the Reclaimed Water System. In order to ensure the Shift Supervisor 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.

to

2. The Primary Auxiliary Operator Turnover Checklist has been revised to include a requirement to the Spent Fuel notify the Chemistry Department of sample requirements if a water addition are Pool has either been initiated or completed durihig the shift. The contents of this checklist 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 by Chemistry Departments. The review will be documented and the documentation maintained 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 of the check of Operator Aids will be performed. Each Shift Supervisor will select five (5)

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 physically 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 MORM366A 14-95)

V.S. 1HUC.AD. RZR7LL!ORX COMWSSXO znxq Font 366k

-

(4 -95)

LICENSEE EVENT REPORT (LER)

TE.XT CONTINUATION DX= I- LEa Hulna= (6) P 1'= (3) r7 r.-TT

= AMN () NUMB= (2)

A n NUMBER NUMBER tYt*trssZ*f U.S.* win.]UgGUZ,.L*On"v SEQUENTA

-TAYEAR R EVISIOuN 98 -- 002 -- 1 6 OF 6 Three Mile Island, Unit I 05000289 7 rf more space is required, use additional copies of NC Form juO A i%-'I insure

3. All of the Operating Procedures which contain Operator Aids will be reviewed to 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.

of Technical

4. All licensed personnel will be given training on an overview of the contents 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 (E-IS), System Identification (SI) and Component applicable, as Function Identification (CFI) Codes are included in brackets, [SUCH] where required by 10 CFR 50.73 (b)(2)(ii)(F).

VRC FORM 365A (4-95)

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 Probability and Consequences of CriticalityEvents 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 a

the arrangement of the racks and the practice of moving fuel assemblies one at time. Thus, this type of criticality scenario will also have a relatively low probability.

Appendix C Assessing the Probabilityand Consequences of 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 "f'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 the set of criticality To illustrate the concept of an envelope of criticality, consider a rack of fuel scenarios that involve fuel mispositioning (the placement in a reduction in the assemblies with inappropriate burnup/enrichment or age),

of these concentration of soluble boron in the pool water, or a combination occurrences. In order to determine the envelope of criticality for these scenarios, and the one would begin by specifying a particular pool and rack configuration, be a fresh most reactive fuel assembly that could be placed in the pool (this may fuel assembly). Next, one would identify the possible range of fuel of fuel mispositioning events. Then, one would determine the combinations a

mispositioning events and soluble boron concentrations that will yield of Keffective of exactly 1 (or, if a factor of safety is used, some lesser value the envelope of Keffective such as 0.95). The set of these combinations would be criticality in the pool, for these scenarios.

undertaken Discovery in this case suggests that no entity in the United States has a fuel pool.

the calculations necessary to determine the envelope of criticality in witness During depositions of NRC Staff witness Dr Laurence Kopp and CP&L how they Dr Stanley Turner, Orange County's attorney asked these witnesses above. Both would determine the envelope of criticality in a fuel pool, as defined CP&L's witnesses' responses indicated that neither the NRC Staff, CP&L nor contractor Holtec has given significant attention to developing a thorough here.

understanding of the potential for criticality scenarios of the type discussed 2.5 The Potential for Mispositioning of Fuel power Appendix B reviews the record of fuel mispositioning at US nuclear County.

plants, drawing from documents that are currently available to Orange of These documents almost certainly do not reveal the full historical record relevant events, for reasons that are explained in Appendix B. Nevertheless, in a fuel pool of Appendix B shows that fuel mispositioning, involving placement or age, is a one or more fuel assemblies with inappropriate burnup/ enrichment likely occurrence.

the Most of the relevant events described in Appendix B directly involved The other relevant mispositioning of one or more fuel assemblies in a fuel pool.

or fuel handling events involved fuel handling errors that affected a reactor core, mispositioning of errors that occurred in a fuel pool but did not directly lead to a fuel. These other events are relevant because they show that ongoing are likely to administrative controls related to fuel handling and management

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 m;ispositioning 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 1

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.

' 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 of A study by the Westinghouse Corporation sought to estimate the probability 2 This study examined a generic, soluble boron dilution at PWR plants.

composite" plant. It sought to estimate the probability of diluting the soluble

"*

37 percent boron concentration in the fuel pool from 2,200 ppm to 1,380 ppm (a 7 per reactor-year. The reduction), yielding a probability estimate of 3.8x10-July study did not summarize the historical record of relevant events, such as the 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 an previously been separated by gates. In addition, this study was performed by 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, 4 These CP&L provided the results of some calculations related to criticality.

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 Criticality Events in Fuel Pools Page C-8 a souble boron In its response of June 14, 1999 to the RAI, CP&L asserted that less than concentration of 400 ppm would be sufficient to maintain Keffective calculations were 0.95 if this mispositioning event occurred. No supporting provided.

were provided by The results of some additional calculations relevant to the RAI CP&L in a letter of October 15, 1999, to which was attached a letter of October 11, Holtec 1999 from Holtec. These results were supported by a proprietary The proprietary document which provided some details about the calculations.

document is not cited here.

CP&L's For the mispositioning event postulated in the April 29, 1999 RAI, (with a 95%/95 %

additional calculations showed that Kinfinite would be 0.9916 5 These calculations assumed confidence level) in the absence of soluble boron.

U-235) the placement of one fresh PWR fuel assembly (enriched 5 wt% in by Figure 5.6.1 of surrounded by PWR fuel of the maximum reactivity permitted Kinfinite the proposed new Tech Specs. CP&L also calculated that the maximum would be 0.9352 if the soluble boron concentration were 400 ppm. Further boron calculations showed a maximum Kinfinite of 0.8671 (0.7783) for a soluble concentration of 1,000 (2,000) ppm.

CP&L In a variant of its calculation that assumed an absence of soluble boron, adjacent to the assumed that the one fresh PWR assembly is placed in a PWR cell PWR and BWR storage racks. Assuming that this assembly is surrounded by that Kinfinite BWR fuel'of-the maximum permitted reactivity, CP&L calculated would be 0.9932 (with a 95 %/95% confidence level).

were reported Some related calculations were performed by the NRC Staff, and Ulses to in an internal NRC Staff memorandum of November 5, 1999 from Tony Ralph Caruso. 6 This document is hereafter described as the "Ulses event in which Memorandum". The calculations assumed a fuel mispositioning the difference 5 A fuel pool can contain a relatively large array of fuel. Thus, pool between Keffective and Kinfinite will be relatively small for many often discussed situations. As a result, the approach to criticality in a fuel pool is largely follows in terms of the value of Kinfinite. The discussion in this appendix that practice.

6 A copy of this document is provided herewith as Exhibit C-2.

Appendix C Assessing the Probabilityand Consequences of Criticality Events 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-1I 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 4

neutrons), then one fission in the first generation would lead to 2.1x10 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 7

full power, 3 billion Joules of energy per second.

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

)

DEPOSITION OF: LAURENCE I. KOPP, Ph.D.

November 4, 1999 DATE: at 2:15 p.m.

Commencing PLACE: Goya Conference Room Four Points Sheraton Hotel 37611 U.S. Highway 19 North Palm Harbor, Florida 34684 Dale E. DeFranco, RPR REPORTED BY:

Notary Public State of Florida at large ORIGINAL

2 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 that's why we decided this week to actually do a 1

2 calculation and see if would be true for Shearon Harris.

for the entire rack.

3 And we found we are subcritical Under what circumstances, if any, and 4 Q. Okay.

if any, does the NRC 5 under what regulatory requirements, of errors in controlling boron G require the reporting in the water of fuel storage pools?

7 concentration 8 A. I'm not sure if there would be any requirements If the boron concentration were a 9 for reporting that.

were in tec specs and if that 10 minimum boron concentration during the surveillance interval, there 11 were violated 12 would be a certain amount of time where one could level.

13 reborate and get back up to the required minimum I guess reportable unless 14 And that would not be really in time. There's a certain interval is one did not borate 16 where you come back within regulations.

A. I see. And if you correct it with appropriate 17 it's not a reportable event; is that what 18 intervals 19 you're saying?

20 A. Right.

To the extent that boron dilution events 21 Q. Oi1ay.

to the NRC, does the NRC keep any 22 are reported events that you 23 centralized record of boron dilution 24 know?

would be the same as the LER's for fuel 25 A. It

36 There would be the LER's as far as I 1 misplacements.

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 with controlling boron concentrations in fuel 5 experience 6 storage pools?

7 A. Not that I know of.

MS. CURRAN: I'd like to ask-the court reporter 8

to mark as Exhibit 10 an October 25th, 1996 letter 9

from Timothy E. Collins, Acting Chief, Reactor 10 System Branch, Division of System Safety and i1 12 Analysis, NRC, to Mr. Tom Green, Chairman Westinghouse Owner's Group.

Subject:

Acceptance 13 14 for Referencing of Licensing Topical Report 15 WCAP-14416-P, Westinghouse Special Fuel Rack 16 Criticality Analysis Methodology.

Attached to this cover letter is a Safety 17 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.)

Q. Dr. Kopp, are you familiar with this document?

22 23 A. Yes, I am.

If you would turn to page 10 -- actually page 24 Q.

a continuation of a discussion that starts on page 25 10 is

37 1 8, Section 3.7 entitled Soluble Boron Credit Methodology; 2 isn't that correct?

3 A. Yes.

Q. If you look at the second full paragraph on 4

I'd like to ask you about a sentence 5 page 10 of the SER, "However, a boron dilution analysis will be 6 that reads:

performed for each plant requesting soluble boron credit 7

time is available to detect and 8 to ensure that sufficient before the 0.95 k effective design 9 mitigate the dilution

0 basis is exceeded and submitted to the NRC for review."

11 In parentheses, "Ref, dot, 29."

Can you explain to me what is meant by this 12 13 sentence and the reference to Ref 29?

This is the new methodology that I spoke 14 A. Yes.

This is one of the reasons for updating the 15 of earlier.

This is a recent approval we gave for 16 Grimes letter.

soluble boron in spent fuel pools. And 17 crediting partial are allowing, not for Shearon Harris, but for 18 since we 19 some reactors, credit for soluble boron under normal to meet .95, this would now require a new 20 conditions accident to be evaluated which would be the boron 21 22 dilution event.

such as Shearon Harris, which 23 For other plants, take credit for soluble boron during normal 24 do not the fact that they calculate the five percent 25 conditions,

38 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 Eval'uation 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 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 sp-ecific spent fuel pool. That's why that boron dilution 17 event was never approved or accepted. It was a generic 18 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 of my I, LAURENCE I. KOPP, do hereby certify that I have read the foregoing transcript any, hereto, find it to deposition testimony and, with the exception of additions and corrections, if be a true and accurate transcription thereof.

1//

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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.

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errors in the Please read the transcript of your deposition and make note of any Please sign and transcription on this page. Do NOT mark on the transcript itself.

date the transcript on PAGE Please return both Errata Sheet and transcript to:

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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 Is/

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 RPF GHolahan JWermiel RCaruso Aulses JStaundenmeier -RLandry UShoop FEitawila DEbert DCarlson S (:DSSA RC *O 11/ /99 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 bumup 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 U0 2 fuel enriched to 5 w/o U 235 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 rnisloading 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 staff's 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:

Typic'al 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 burnup 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.vthe 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 1 and Problem 2 CASMO MCNP SCALE' Problemi 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 1 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

1. "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," ORNLUTM- 12667, Oak Ridge National Laboratory, March 1995.

5. W. C. Jordan, et. al., "Validation of KENOV.a, Comparison with Critical Experiments" ORNL/CSD/TM-238, Oak Ridge National Laboratory, 1986.

• .#

Spectrum of W 15x15 Fuel in Poisoned Rack KENO- VI Results 4.0 Error bars represent +/- 1 sigma 3.0 uncertainty t2~)

-a S2.0 1.0 10 10' Energy (eV)

Figure 1 KENO Predicted Spectrum for W 15x15 Fuel Assembly

Spectrum in Outer Boral Sheeting KENO VI Results 5.0 4.0 Error bars represent +I- 1 sigma uncertainty S3.0 0

2.0 1.0 10 100t 102 n 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/ 15x15 w Ofluxes for global unit region 1 region 2 region 3 region 4 Ogroup flux percent region 5 region 6 flux percent flux pdrcent flux percent flux deviation percent flux percent deviation deviation deviation deviation 1 0.000E+00 0.00 1.376E--04 5.36 8.973E-06 1.932E-05 deviation 18.11 19.59 1 .823E-05 18.58 2.150E-05 12.41 2 0.000E+00 0.00 4.190E-04 3.46 5. 856E-05 8.68 5.653E-05 3 8.31 4.404E-05 9.44 4.438E-05 8.41 0.000E+00 0.00 1.267E-03 1.92 1.656E-04 5.23 1.544E-04 5.92 4 1. 281E-04 5.35 1.475E-04 4.97 0.000E+00 0.00 4.204E-03 1.14 5.437E-04 3.46 5.072E-04 3.55 4.983E-04 3 .51 5 0.000E+00 0.00 2.834E-03 4.957E-04 2.91 1.37 3.175E-04 3.97 3.393E-04 3.74 3.345E-04 6 0.000E+00 0.00 8.974E-04 4.01 3.336E-04 3.86 2.41 9.913E-05 5.97 1.171E-04 7.88 1.041E-04 9 0.000E+00 0.00 3.574E-03 6.83 1.040E-04 6.43 8 1.33 4.377E-04 3,59 4.251E-04 3.72 3. 972E-04 0.000E+00 0.00 4.386E-03 4.34 4.402E-04 3.67 9 1.18 5.304E-04 3.18 5.120E-04 3.19 4 . 895E-04 0.000E+00 3.44 5.272E-04 3.33 0.00 6.307E-03 1.08 7.926E-04 3.15 7.232E-04 3.15 10 6.767E-04 3.19 7.520E-04 2.96 0.000E+00 0.00 1.103E-02 0.78 1.355E-03 2.44 1.291E-03 2.43 1 .246E-03 2.54 1.271E-03 2.40 11 0 000E+00 0.00 1.178E-02 0.74 1.464E-03 2.26 i.336E-03 2.36 12 0.000E+00 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 0.000E+00 0.00 3.03 7.950E-04 2.93 1.595E-03 1.71 2. 171E-04 5.22 1.834E-04 5.38 1.810E-04 14 0 .000E+00 5.70 1.772E-04 5.20 0.00 7.130E-03 0.92 8. 294E-04 2.75 7.465E-04 2.97 7.295E-04 15 0 000E+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 0.000E+00 0.00 6.581E-04 2.97 5.505E-03 0.92 5. 951E-04 2.66 5.580E-04 2.90 5.222E-04 17 0.000E+00 2.90 5.567E-04 2.73 0.00 3.273E-03 1.15 3 484E-04 3.02 3.119E-04 3.32 2.897E-04 18 0 . 000E+00 3.31 3.040E-04 3.06 0.00 2,444E-03 1.36 2. 262E-04 3.42 2.102E-04 3.45 2.065E-04 19 0.000E+00 3 .42 2.197E-04 3.25 0.00 1.374E-04 2.80 3 954E-05 7.46 3.452E-05 6.89 3.002E-05 20 0.000E+00 7.71 3.751E-05 8.08 0.00 5.568E-04 2.75 4. 471E-05 6.49 4.308E-05 6.56 3.613E-05 21 6.99 4.229E-05 6.87 0.000E+00 0.00 4.168E-04 2.97 3.427E-05 6.93 2.959E-05 7.91 3.034E-05 22 7.48 3.174E-05 7.49 0.000E+00 0.00 7.767E-04 2.22 5. 679E-05 5.29 6.221E-05 5.63 5.160E-05 23 0.000E+00 5.67 5.866E-05 5.56 0.00 8.810E-04 2.08 6. 311E-05 4.74 6.017E-05 4.94 5. 510E-05 4 .97 24 0.000E+00 6.113E-05 4,85 0.00 9.433E-04 1.98 5. 403E-05 5.21 5.279E-05 5.27 5. 165E-05 5.02 25 0. 000E+00 5.716E-05 4.83 0.00 7.081E-04 2.24 4. 488E-05 5.09 3.932E-05 5.31 3.755E-05 26 0.000E+00 5.45 3.815E-05 5.11 0.00 6.778E-04 2.21 3.444E-05 5.51 3,357E-05 5.09 2 .928E-05 27 0.000E+00 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 0.000E+00 15.39 4.106E-06 14.27 0.00 9.516E-05 5.12 6.096E-06 12.92 4.348E-06 14.18 3 879E-06 29 0. 000E+00 14.43 4.893E-06 13.21 0.00 1.080E-04 4.50 5.983E-06 13.01 5.313E-06 15.04 5.081E-06 30 0 .000E+00 12.86 6.356E-06 11.50 0.00 2.454E-04 3.50 1.201E-05 8.66 1.019E-05 11.42 1.107E-05 31 0. 000E+00 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 12.02 32 5.699E-06 12.93 0 000E+00 0 .00 1.513E-04 3.91 6.783E-06 10.57 6.677E-06 10.90 6.587E-06 33 11.27 7.080E-06 9.90

0. 000E+00 0.00 1.739E-04 3.52 6.496E-06 9.67 6.806E-06 9.95 7. 721E-06 10.56 6.509E-06 10.88 34 0.000E+00 0.00 4.281E-04 2.47 1.835E-05 6.12 1.563E-05 6.45 1. 648E-05 35 6.63 1.735E-05 6.08

0. 000E+00 0.00 6.916E-04 1.90 2.472E-05 5.58 2.395E-05 5.19 2 .208E-05 36 0 000E+00 5.36 2.412E-05 5.32 0.00 6.888E-04 1.78 2.515E-05 4.84 2.490E-05 5.24 2.035E-05 37 6.30 2.428E-05 4.73 0 000E+00 0.00 5.795E-04 1.93 1.896E-05 5.63 1.813E-05 5.20 1 .732E-05 5.52 1.871E-05 5.04 38 0. 000E+00 0.00 3.240E-04 2.18 1.036E-05 7.29 8.607E-06 7.85 1.001E-05 7.77 9.824E-06 8.23 39 0 .000E+00 0.00 3.261E-04 2.37 9. 701E-06 8.07 8.411E-06 7.75 7 . 653E-06 40 7.96 9.549E-06 7.42 0.000E+00 0.00 1.468E-04 3.28 3.917E-06 10.32 4.053E-06 10,54 41 3.024E-06 12.74 3.141E-06 11.64 0.000E+00 0.00 3.566E-04 2.29 1.058E-05 6.57 9.020E-06 7.10 8.858E-06 7.00 9.430E-06 6.96 42 0.000E+00 0.00 3.604E-05 5.88 1.009E-06 27.70 8.304E-07 19.11 8.564E-07 25.72 8.251E-07 21.12 43 0.000E+00 0.00 3.968E-05 5.42 9.087E-07 18.02 1.018E-06 16.41 8. 949E-07 30.82 6.629E-07 20.54 44 0 .000E+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