ML20154D544

Insp Rept 70-0036/98-203 on 980804-07.Non-cited Violation Noted.Major Areas Inspected:Status & Adequate Implementation of Selected Areas of Criticality Safety Program Update & Followup on Previously Identified Insp Findings

ML20154D544
Person / Time
Site:	07000036
Issue date:	10/02/1998
From:	NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS)
To:
Shared Package
ML20154D515	List: ML20154D511 ML20154D544
References
70-0036-98-203, 70-36-98-203, NUDOCS 9810070259
Download: ML20154D544 (14)

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i OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS NUCLEAR CRITICALITY SAFETY INSPECTION REPORT i

i REPORT NO: 70-36/98-203 DOCKET NO: 70-36 y SNM-33 LICENSE NO:

4 LICENSEE: ABB Combustion Engineering P.O. Box 107 3300 State Road P Hematite, MO 63047 FACILITY NAME: ABB-CE Nuclear Fuels INSPECTION DATES: August 4 - 7,1998 l INSPECTORS: J. R. Davis, NCS Engineer ,

I Fuel Cycle Operations Branch APPROVED BY: Philip Ting, Chief Fuel Cycle Operations Branch '

Division of Fuel Cycle Safety and Safeguards. NMSS i

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l 9810070259 981002 l PDR ADOCK 07000036 C PDR ;_

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SNM-33 1 70-36/93 203 TAllLE OF CONTENTS I$ACKGROUND .. .. .. ...... . .. .... .. . . .............. .. 1 EXECUTIVE

SUMMARY

. ........ . . .... . .. . ... ... . ......... .1 Introduction . . . .. ........ . ... .. .. . ... .. .. .... .. .. .1 Major Results . . . . . . .. ... ..... . . . . . . .. . ... .. ... .1 REPORT DETAILS . .. ...... . . ...... ...... ... . ........ . . .. 2

01. PLANT ACTIVITIES . . . . . ...... ...... . ... .... . .... 2 01.1 'lecovery Area Uranyl Nitrate Feed Tank . . . ... .... 2 01.2 Incinerator RoofIIeat Exchanger . . . .. . .... .... .4

02. FOLLOW-UP OF PREVIOUS INSPECTION FINDINGS . . . . . . . . .6 02.1 IFI 98-201-01, RAAP-109 Issues . .... ...... .... .. 6 02.2 URI 98-201-02, UO2Screw Cooler . . . . . .. ...... .. ... 7 02.3 URI 98-201-03, Inaccurate Information SIU Issue . . . . . .... 8 02.4 IFI 98-201-04, Re-review of General SIU Classification . . . . 9 ITEMS OPENED, CLOSED, AND DISCUSSED ... . . . ........ .. .. . 11 MANAGEMENT MEETINGS . . . . . . . . . . . . . . . . .... ... . .. .. ... . .. . 12 ACRONYMS USED . . . ....... ..... .... .. ... ...... ... . . ... .. 12 NMSS FCis

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BACKGROUND Based upon the results of an annual criticality safety audit, and the inspection findings from the Nuclear Regulatory Commission (NRC) Inspection Report 70-36/96-202 (7/19/96). ABB-CE i recognized the need to enhance the administrative elements ofits Nuclear Criticality Safety (NCS) Program. In a letter dated September 20,1996, the CE site committed to an improvement plan which was appropriately titled. " Criticality Safety Program Update (CSPU)." This program i l

is intended to fonnalize and strengthen the criticality safety program at the plant site. taking into account recognized industry standards and methods of criticality control. Since that time, the iicensee has submitted quarterly (and now semi-annual) status reports of progress on this initiative. As part of this inspection activity. the NRC reviewed selected areas of program completion and reviewed preparation activities for the next scheduled effort, the Recycle / Recovery Process Area. l l

EXECUTIVE SUSD1ARY I l

Introduction The NRC perfomied an announced nuclear criticality safety inspection of the ABB Combustion i Engineering Nuclear Fuels facility located in IIematite, MO. from August 4 - 7.1998. The objective of the inspection was to determine the status and adequate implementation of selected I areas of the CSPU. and to followup on previously identified inspection findings and unresolved items. As a result of the inspection, one new unresolved item (URI) and one non-cited violation (NCV) were identified. The specifics of these findings and areas of review are fully developed in the Report Details; the major conclusions are summarized below.

Maior Results

1) A URI was identified conceming the apparent failure to notify the NRC that the Uranyl Nitrate tank in the Recovery area is of single parameter control.

[Section 1.1]

2) A NCV was identified for the failure to maintain mass comrol in the Incinerator lleat Exchanger system. [Section 1.2]

3) A previously identified Inspector Followup Item (IFI) 70-36/98-201-01 concerning an inconsistency between the License and Procedure RAAP-109 has been closed.

[Section 2.1]

4) A previously identified URI 70-36/98-201-02 concerning the failure to analyze the UO, Screw Cooler for criticality concerns has been closed. [Section 2.2]

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5) A previously identified URI 70-36/98-201-03 concerning incomplete and inaccurate information provided to the NRC associated with the completion of subelements of the CSPU has been closed. [Section 2.3]

6) A previously identified IFI 70-36/98-201-04 concerning review of all identified safe individual units (SIUs) to ensure sufricient analysis and justification of classification has been closed. [Section 2.4]

REPORT DETAILS

01. PLANT ACTIVITIES 01.1 Recovery Area Uranyl Nitrate Feed Tank

a. Srops l

The inspector toured the Recycle / Recovery Area processes and reviewed tile l current progress with regard to the CSPU initiative to ensure that this NCS j higher risk area was being adequately addressed and proper safety was bemg l applied. l

b. Observations and Findines

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As part of the general Recovery / Recycle area review progress, the mspector  ;

questioned the licensee concerning the nuclear safety controls associated I with the Uranyl Nitrate Feed Tank. The inspector was informed by the I licensee that the tank utilized single parameter control on neutron poison (raschig rings). The inspector discussed the reliability of this control considering that the tank is oflarge unsafe geometry, concentration is not controlled or measured. and constituent feed to the tank would indicate exposure to fluorides. The licensee indicated that the requirements of ANSI /ANS 8.5-1986 are followed for use of raschig rings in this tank and that historical records dating back to the first use of these rings demonstrates no negative trend in any of the measured parameters to date. Further, there l would be no reason for maintenance or operations personnel to remove the rings, and an NCS posting is located on the tank specifying the rings as important to nuclear criticality safety control.

The inspector reviewed the raschig ring inspection records dating back to 1978 and compared them with the requirements of ANSI /ANS 8.5-1986.

The inspector determined that no statistically significant negative trend was evident in reduction of ring volume or ring isotopic content (specifically, Boron-10 content). The inspector also later determined that the fluoride Nuss Icis

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SNM-33 3 70-3C4203 exposure (corrosive) was typically found in insignificant quantities within

. this tank and although some etching of the rings is evident, such degradation does not exceed the proceduralized limit.

However, due to the inherent decreased safety margin associated with single parameter control, the inspector further reviewed the license application to determine if the use of such control was authorized for this equipment. The license application indicates that the process design philosophy used by Combustion Engineering to ensure nuclear criticality safety control utilizes the double contingency principle. License Application Section 4.1.l(a),

states that," Process design which does not meet these double contingency criteria shall be explicitly approved in Section 1.6 of this application."

However, the inspector determined that this tank was not explicitly approved in Section 1.6 of the Application. Further discussions with the licensee indicated that since License Application Section 4.2.1.3(e) allowed the use of raschig rings in nuclear criticality safety control, in effect, License -

Application Section 4.1.l(a) and Section 1.6 did not apply. However, the inspector pointed out that although Section 4.2.1.3(e) allows the use of raschig rings for criticality control, it does not specifically authorize such use as single parameter control. Discussions with more senior plant management indicated that they believed this issue was previously discussed with the NRC, but could not immediately locate documentation or remember the content or resolution of such discussions.

The inspector notes that the NRC requirement and safety significance of explicitly identifying and authorizing single parameter control processes recognizes the substantially-decreased safety margin afforded by such control methods and that such requirements allow the Agency to ensure appropriate heightened awareness and necessary management attention to these higher risk items. Since the licensee hasjust recently initiated the CSPU in the Recovery area and is just now compiling records and documentation, the failure to explicitly authorize the single parameter Recovery Uranyl Nitrate Tank in Section 1.6 of the Application is identified as URI 70-36/98-203-01.

c. Conclusions It appears that the licensee has not explicitly identified the Uranyl Nitrate Tank as a single parameter control in Section 1.6 of the License Application, as required. There may be reason to believe that previous NRC-licensee discussions concerning this equipment have taken place, but no documentation was immediately available.

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. .01.2 Incinerator IIeat Exchanger Mass Accumulation Incident

a. Scope l t l While conductmg mspectwns of the NCS program during the week of August 7 - 10,1998, it was brought to the attention of the NRC inspector that an unexpected accumulation of uranium-laden ash was discovered in the roof-mounted heat exchangers supporting the trash incinerator which was not in accordance with applicable nuclear criticality safety approval l requirements. The inspector reviewed the licensee's actions to ensure that l recovery from this incident was completed in a safe manner and that adequate actions were taken to prevent recurrence.

b. Observations and Findings Apparently, while performing maintenance and troubleshooting on the plant trash incinerator associated with an inadequate offgas flow, plant personnel  !

discovered that a greater than expected amount of uranium-laden ash had accumulated in the roof-mounted heat exchanger. Differential pressure i measurements across the heat exchanger revealed a 12-inch pressure drop. 1 Subsequent gamma measurements and visual inspections revealed a considerable amount of hold-up across the air-to-air heat exchangers.

Although the exact distribution and density of the material in inaccessible j portions of the heat exchanger was unknown, leading to some uncertainty in j i

the mass, the licensee estimated that no more than 300 grams of U-235 existed.

The inspector independently calculated an estimated mass from lab sample results of the material and compared the information with the licensee's calculations. The inspector also visually inspected the roof-mounted equipment and mass accumulation through inspection ports on each end of the heat exchanger ducting and concluded that the licensee's estimate was reasonable. The inspector reviewed the NCS analysis supporting this operation and discussed the abnormal condition with the NCS expert and the cognizant process engineer. The inspector agreed with the licensee's contention that the as-exists condition was safely suberitical and would remain so, since the incinerator was shut down (i.e., no additional mass accumulation) and the vessel integrity was intact (moderation control). Lab results of samples taken indicated that the material moisture level was 0.02 percent by weight supporting this conclusion. Further discussions revealed that the licensee had not conducted an annual inspection of the duct work as required by NCS analysis. It appears that the requirement to NMSS rCIS

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l SN N1-33 5 70-36M-203 l-l l annually inspect the roof-mounted duct work was unintentionally remosed l from the previously-revised procedure without NCS knowledge or approval.

License Application Section 4.1.4, Written Procedures, states. in part, that

" Operations irivolving the handling and storage of special nuclear material l shall be performed according to approved, written procedures.

Administrative controls and passive barriers which are relied upon for

! criticality safety shall be described in appropriate procedures." Contrary to

the above, as of August 5,1998, administrative controls relied upon for

criticality safety of the incinerator system ducting were not described in l appropriate procedures. Specifically, the administrative requirement to annually inspect the incinerator duct work for fissile material accumulation was not described in the appropriate procedure leading to an unexpected accumulation of uranium mass in the heat exchanger duct work.

l The inspector continued to follow the event recovery progress and attended the initial meeting of the event investigation and root cause team. The inspector noted that the licensee was proceeding cautiously and j conservatively and with the appropriate risk prioritization. Information obtained from the licensee on August 14.1998, indicated that the unit was removed intact from the roof and transferred inside the facility for clean-out to ensure moderation control and health protection. Clean out yielded approximately 33 Kgs total of uranium-laded ash containing approximately l 70 grams U-235 which is well below the system limit of 800 grams U-235.

Discussions with the Licensee on September 15,1998, indicated that the incinerator NCS analysis was re-evaluated to be consistent with License Application Section 4.2.1.3(a) and new lower limits were added. Procedures for inspection and clean-out have been revised and approved to reflect quantitative acceptance criteria. In addition, inspections are now required to l be perfomied at each system clean-out in addition to the annual requirement. j Following these changes. the incinerator was placed back in service to '

alleviate trash build-up. This non-repetitive, licensee-identified and  ;

corrected violation is being treated as a non-cited violation. consistent with l Section VII.B.1 of the NRC Enforcement Policy and is identified as NCV 70-36/ 98-203-02. Final root cause investigation results and recommendations are expected to be completed in another month.

c. Conclusions The inspector determined that the licensee failed to ensure the proper i implementation of nuclear criticality safety requirements intended to preclude the excessive accumulation of fissile materialin the incinerator off-gas heat exchanger. However following discovery, the licensee took appropriate and )

timely interim corrective actions to ensure the continued safety of plant l ssiss n i

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SNM-33 6 70-3C98-203 operations and instituted a formal root cause investigation to ensure pennanent corrective actions are implemented to prevent recurrence.

! 02. FOLLOWUP ON PREVIOUS INSPECTION FINDINGS 1

02.1 IFI 70-36/98-201-01, l.icense/ Procedure inconsistencies

a. Scope This inspector followup item concerns the apparent disagreement between license conditions and implementing procedural guidance. The inspector l reviewed the concern and licensee actions to determine if this item could be l closed satisfactorily.

b. Observations and Findings ,

IFl 70-36/98-201-01 raised concerns over the apparent disagreement between License Section 4.2.4(k) and NCS procedure RAAP-109," Criticality Safety Program." Section 5.1.15, dated December 30,1997. License Section 4.2.4(k) states, " Process systems shall be designed to minimize the likelihood for accumulation of fissile material within the system. In addition, process procedures shall have provisions for verifying that fissile material has not accumulated within the system . Section 5.1.15 of RAAP-109 states, "The possibility of accidental accumulation of fissile materials in inaccessible locations shall be minimized through equipment design or administrative controls or included in the nuclear safety evaluation of the i process." According to this IFI, the procedural guidance is less stringent than the overriding license requirement.

The inspector reviewed the license section and the associated implementing procedure. RAAP-109 and found no such disagreement. Section 5.1.15 of RAAP-109 relates more closely to License Application Section 4.2.1.3(k) which states," Process systems shall be designed to minimize the likelihood for accumulation of fissile material within the system." Section 4.11 of RAAP-109 specifically repeats License Application Section 4.2.1.3(a).

Therefore, both of these sections of RAAP-109 fully support License Condition 4.2.4(k) and eliminate any inconsistencies or disagreements.

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. c. Conclusions No disagreement between license conditions and implementing procedural guidance was evident. Therefore, this item is closed.

02.2 URI 70-36/98-201-02, Failure to Analyze Conversion Process Equipment

a. Scope This URI concerns the failure to analyze the uranium dioxide (UO 2) cooler as part of the Conversion Area process. The inspector reviewed the supporting evidence to determine appropriate disposition of this item.

b. Observations and Findings l

URI 70-36/98-201-02 raised concerns that the UO2 cooler was in operation l even though no technical basis was available to demonstrate that the unit l would remain safely suberitical under all credible normal and abnormal conditions. The licensee stated that no analyses were completed for this piece of equipment because the unit was classified as a SIU in Table 4.5 of l License Application Chapter 4. URI 70-36/98-201-02 was raised because under certain postulated process upsets, it was not clear that the unit would remain suberitical or would remain within the geometric classification constraints of an SIU. URI 70-36/98-201-02 considered a moderated, loss of geometric control condition to be credible since the UO2 cooler consists of a screw feeder surrounded by a liquid water cooling jacket which also flows i down the shaft of the cooler. A breach in the container integrity would essentially produce a condition of optimally moderated fissile material in an unfavorable geometric container. j The inspector reviewed the classification criteria for SIUs and discovered that the licensee had actually identified two different types of SIUs (classic and calculated). Classical SIUs are based upon experimental data whereas calculated SIUs are derived from analyses using validated computer methods.

Discussions with the licensee indicated that they classified the UO2cooler as a classical SIU since they didn't consider it credible to have a breach in the vessel integrity such that the fissile material would exist in an unfavorable geometric shape. Their reasoning was that the differential pressures between the two concentric vessels would effectively keep the UO2 confined to the inner cylinder under all credible container integrity breaches. Although the licensee agrees that such logic really constitutes a calculated SIU rather than a classical SIU, they do not believe that poorjudgement in this particular case suss tcis p-

s u s.n s wans.m constitutes a violation of requirements. 'I he licensee has since re-evaluated the UO 2cooler for the postulated scenario and has shown the cooler to meet the criteria of a calculated SIU. The inspector notes that a lack of criteria for determining what constitutes a calculated SIU contributed to this misapplication. The licensee has since made a commitment, as part of the CSPU. to re-evaluate the classification of all SIL!s to ensure that they were appropriately identified.

c. Conclusions A lack of established criteria for determining the differences between a classical and calculated SIU was a contributing cause to the misapplication. The licensee has now adequately addressed this issue, and URI 70-26/98-201-02 is closed.

02.3 URI 70-36/98-201-03, Inaccurate and Incomplete Infonnation

a. Scope This URI concerns inaccurate information supplied to the NRC under the CSPU communication commitment as related to the UO: cooler. The inspector reviewed the supporting evidence to determine appropriate disposition of this item.

b. Observations and Findings URI 70-36/98-201-03 raised concerns that no analysis was available to demonstrate adequate safety of the UO2 cooler even though a September 30, 1997 CSPU Quarterly Update to the NRC indicated that all analyses for the Oxide Conversion area had been completed.

The inspector reviewed the CSPU Plan and Schedule, dated September 20, 1996, and noted that the licensee committed to an analysis and evaluation update,if appropriate. Section ll.c states,"After the existing documentation is assembled, each category, process, or piece of equipment as appropria/c

[ italics added] will be reviewed to identify any assumptions, bounding conditions, upset conditions, contingencies, margin to criticality, and NCS barriers. The results of the review will be compared against requirements."

Section II.d states, "The nuclear criticality safety analysis will be updated if required [ italics added)." The Ji:ne 27,1997, CSPU Quarterly Update to the NRC stated in Section VI.l.b that," Activity in [the Conversion Area]is essentially 100% complete. The majority of the analyses of the conversion process have been redone using the latest verified and validated SCALE methodology " The September 30,1997, CSPU Quarterly Update miss icis

. e SN433 9 70-36/98-203 indicated that the Conversion Area Upgrade was complete.

Since the licensee had considered the UO2 cooler to be a classical SIU (as discussed in Section 02.2 above). no analysis or evaluation was deemed necessary for this piece of equipment in accordance with the original commitments. as stated in the CSPU Plan and Schedule. Although such a classification was later shown to be an exercise of poorjudgment. the inspector determined that it did not constitute inaccurate or incomplete information. Therefore, this item is closed without being cited as a violation.

c. Conclusions Although the UO 2cooler nuclear criticality safety status was not analyzed and documented at the time the licensee submitted a report to the NRC indicating that all upgrades to the Conversion Area documentation were complete, no inaccurate or incomplete infom1ation was submitted because the unit was initially classified as a classical SIU in accordance with License Application requirements and CSPU commitments. Therefore, this item is closed.

02.4 IFI 70-36/98-201-04, Suflicient Documentation for all Plant SIUs

a. Scope This inspector followup itern was raised to ensure that the licensee has sufficient analysis orjustification for the status of all SIUs and necessary controls to maintain them as SIUs. The inspector reviewed the supporting evidence to determine appropriate disposition of this item.

b. Observations and Conclusions IFI 70-36/98-201-04 originated from a concern over the appropriateness of identifying the UO cooler 2 as a classical SIU since credible postulated upset conditions could cause the UO2 cooler to exceed the limits and controls associated with a classical UO2 cooler. It was determined that the licensee exercised poorjudgment in identifying the UO 2cooler as a classical SIU even though later analysis of the postulated upset condition demonstrated that the UO2 cooler met the requirements of a calculated SIU.

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The inspector discussed the concern with licensee management to determine if they had any plans or schedule for a re-review of all identified SIUs. The inspector was presented with a March 31.1998. CSPU Semi-Annual Update letter which committed the licensee in Section VI.3 to add the reevaluation of SIUs to the scope of the CSPU. The schedule Ibr completing these evaluations will be performed as associated areas are analyzed or as additional time is available. Since the next area scheduled fbr review is the Recycle / Recovery Area. the inspector determined that the reevaluation of SIUs was consistent with a risk-based approach and that the licensee's commitments were adequate and sufficiently documented to close IFI 70-36/98-201-04.

c. Conclusions The licensee has formally committed to add reevaluation of all identified SIUs to the scope of the CSPU and has prioritized the reevaluation within a risk-focused approach. Theretbre this item is closed.

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. ITEMS OPENED. CLOSED. AND DISCUSSED Onened

, . URI 70-36/98-203-01 Apparent failure to identify to the NRC as required by license l condition, that the Uranyl Nitrate tank in the Recovery area is of single contingency control L NCV 70-36/98-203-02 Failure to maintain double contingency control in the Incinerator Heat Exchanger system Closed i

l IFI 70-36/98-201-01 inconsistency between the approved License and Procedure j RAAP-109 i I

URI 70-36/98-201-02 Failure to analyze the UO2 Screw Cooler for criticality _ j Concerns  :

URI 70-36/98-201-03 Incomplete and inaccurate information provided to the NRC associated with the completion of subelements of the CSPU IFI .70-36/98-201-04 Review of all identified SIUs to ensure sufficient analysis for -i justification of their status Discussed IFI 70-36/98-201-07 This itens concerns the adequacy of the current Change l Control Management Tracking system. The inspector began a review of this IFI, but did not have sufficient time to make a determination of adequacy due to other priorities. Therefore, this item remains open.

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sNM.33 12 70-36/98-203 MANAGEMENT MEETINGS Exit Meeting Summarv l The NRC Inspector met with ABB Combustion Engineering management throughout the inspection. An exit meeting was held on August 7.1998. No classified or proprietary information was identified. The following is a partial list of exit meeting attendees:

Combustion Engineering Nuclear Fuels Dr. Bruce Kaiser, Vice President, Fuel Operations Robert Sharkey, Director, Regula'ory Affairs Gil Page, Director, Ceramic Manufacturing Earl Saito, Ilealth Physics Specialist Michael Eastburn, Nuclear Criticality Safety Specialist Robert Freeman, Nuclear Criticality Safety Decialist Nuclear Regulatory Commission c

Jack Davis, Nuclear Criticality Safety, NRC Headquarters ACRONYMS USED i i

AN S . . . . . . . . . . . . . . . . . . . ... . . . . . . . . . . . . . . . . . . . . . . . . . . American Nuclear Society i

ANSI . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . American National Standards Institute CE................................ . .......... . . . . . . Combustion Engineering CSPU . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Criticality Safety Program Update i l i Q . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . H ead q uarte rs (N RC)

IFI.................... ........ .. . .......... ... . . . . . Inspector Followup item NCS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Nuclear Criticality Safety NCV........... .. .. . . ............ . .. . .... . . . . . . . Non-Cited Violation NRC . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Nuclear Regulatory Commission SIU ......... . ............. . . .. ... .. ... . . . . . . . . . Safe Individual Unit UO...................

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........... ..... .. ... .. ..... . . . Uranium Dioxide U RI . . . . . . ..... ................ ........... ........ . . . . . . . . . . Unresolved item

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