ML20141F397
| ML20141F397 | |
| Person / Time | |
|---|---|
| Site: | Paducah Gaseous Diffusion Plant |
| Issue date: | 05/16/1997 |
| From: | Jennifer Davis, Ting P, Troskoski W NRC, NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| To: | |
| Shared Package | |
| ML20141F387 | List: |
| References | |
| 70-7001-97-201, NUDOCS 9705210297 | |
| Download: ML20141F397 (30) | |
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4 U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR MATERIAL SAFETY AND SAFEGUARDS Docket No:
70-7001 Certificate No:
GDP-1 Report No.
70-7001/97-201 Certificate Holder:
United States Enrichment Corporation Location:
Paducah Gaseous Diffusion Plant Paducah, Kentucky Dates:
March 17 - 21,1997 March 31 - April 4,1997 inspectors:
William Troskoski, Lead Inspector, NRC Headquarters Jack Davis, inspector, NRC Headquarters Sandra Larson, NRC Contractor Dennis Morey, inspector, NRC Headquarters Christopher Tripp, inspector, NRC Headquarters Joseph Wang, inspector, NRC Headquarters Approved By:
Philip Ting, Chief Operations Branch Division of Fuel Cycle Safety and Safeguards, NMSS l
9705210297 970516 PDR ADOCK 07007001 C
PDR 3
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s 2-EXEGMTJYJJiUMlWIARY UNITED STATES ENRICHMENT CORPORATION PADUCAH GASEOUS DIFFUSION PLANT NRC INSPECTION REPORT 70-7001/97-201 Areas insoected NRC performed an announced criticality safety inspectiori or tiie Paducah Gaseous Diffusion Plant (PGDP) in Paducah, Kentucky, on March 17 - 21, and March 31 -
April 4,1997. The inspection was conducted using staff from NRC Headquarters and one NRC contractor. NRC assumed regulatory oversight over the PGDP from the Department of Energy on March 3,1997. The focus of this inspection was to determine the overall level of nuclear criticality safety (NCS) through review of the NCS program implementation (as described in the certification application and j
the compliance plan).
j Major programmatic portions of the Nuclear Criticality Safety (NCS) program (IP 88015) which were reviewed at Paducah GDP included:
e Management and Administrative Practices for NCS e
Nuclear Criticality Safety Function j
e Configuration Control Program for NCS e
NCS Change Control Operating Procedures e
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Maintenance for NCS e
Nuclear Criticality Safety inspections, Audits, and Investigations
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Results e
No immediate safety issues were identified during this inspection.
e The Technical Safety Requirements (TSRs) specified in Section 3.11 were, in general, adequately addressed in the PGDP program, implementing procedures, safety evaluations, and plant conditions.
Sufficient management attention was not provided for determining overall NCS e
program readiness prior to the transition to NRC regulatory oversight. As a consequence, several program weaknesses were not identified by the facility management until after the transition period had been completed (Section 1.A).
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3-i e ~ Once a potentially significant NCS problem was identified, a broad, in-depth i
review of the problem was conducted. This is considered a strength j
i (Section 1.B).
e-Two Level IV violations were identified during this inspection involving the failure to (1) document a safety evaluation for the building C-335 used NAM and HEPA filter storage array'(Section 2.A), and (2) incorporate special fire fighting instructioas into the Pre-Fire Plan and update the emergency I
response training to incorporate special actions required for the fissile l
- control areas (Section 3.E).
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e Program weaknesses were identified in the areas of thoroughness of NCSE documentation (Section 2.C); flowdown of NCSA requirements to all affected q
i groups (Section 3.E); NCS plant change control (Section 4.A); and, configuration management assessments (Section 4.B). Some of these weaknesses were self-identified (Sections 1.B and C).
o Potential regulatory issues were identified conceming implementation of the i
10 CFR 76.68 review process.
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PGDP management's self-assessment program and planned follow up corrective 1
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actions for NCS is considered a strength.
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DETAILS 4
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introduction i
l A.
Regulatorv Background
. The NRC issued a Certificate of Compliance to USEC for the Paducah Gaseous Diffusion Plant (PGDP) on November 26,1996. The Certificate became effective on March 3,~ 1997. Certificate conditions require USEC to conduct its operations in j
accordance with the statements and reprecentations contained in the Certification
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Application (and revisions), the Compliance Plan (and revisions) and the approved J
TSRs. The Compliance Plan (CP) allows PGDP to come into compliance with NRC requirements and application commitments in accordance with the Plan of Action and
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Schedule provided in the CP. The greater majority.of NCS CP items were scheduled
. to be completed by March 3,1997.
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B.
Insocction Overview An NRC headquarters-based team conducted the first criticality safety inspection at PGDP from March 17 - 21, and March 31 - April 4,1997. Additional information provided by PGDP on April 14,1997 was also reviewed at headquarters. The purpose of the inspection was to ensure that PGDP had established and implemented appropriate measures to assure NCS and adherence to the requirements, programs and conditions of the certificate.
C.
Summarv of Insoection Observations The inspection team determined that PGDP had implemented a comprehensive NCS program. No immediate safety issues were identified, although two violations and several program weaknesses were observed by the team. In general, management had already been made aware of the program weaknesses through recent independent assessments and internal audits and surveillances of the NCS Program.
1.0 Management and Administrative Practice for NCS A. Internal Readiness Review Scope As discussed earlier, NRC assumed regulatory oversight over the GDPs from the Department of Energy (DOE) on March 3,1997. During the transition period from DOE to NRC regulatory oversight, an Internal Independent Audit and Regulatory Oversight Agreement (ROA) Appraisal Report, " Nuclear Criticality Satoty," KY/Q210, was issued on October 8,1996, to document: (1) the facility's compliance to the ROA, (2) overall NCS program function /offectiveness, and (3) the progress towards NRC readiness. The inspectors reviewed this document and discussed the findings and planned corrective actions with the PGDP management.
Observations and Findings Although the report noted that the NCS group manager had developed a detailed plan of action to enhance the NCS program in light of NRC requirements and expectations, discussions with this manager indicated that the plan of action was subsequently overcome by the large workload and limited staff required to develop the baseline NCS evaluations prior to the scheduled March 3,1997, transition date. Discussions with the Manager of Nuclear Safety indicated that while the 180-day transition period included about a 60-day margin to
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.r 5-l complete the scheduled program upgrades, the actual level of effort needed to complete all of the commitments turned out to be much greater than initially expected.-
Consequently, no formal assessment or audit of the depth discussed in Section 1.B below, was conducted to assure that all of the SAR 5.2 program commitments l
were adequately addressed. However, the NCS manager did note that once the problem with the implementation of administrative controls for the C-400 -
cylinder wash station was identified in February 1997, this shortcoming was recognized and several audits and assessments were initiated.
i Conclusion Sufficient management attention was not provided for determining overall NCS Program readiness prior to the transition to NRC regulatory oversight. As a F
consequence, several program weaknesses were not identified by the facility manegement until after the transition period had been completed.
B. General Manager's independent Assessment SGDRA SAR Section 5.2.2.2, " Nuclear Criticality Safety Responsibilities," commits in part, "The General Manager has overall responsibility for NCS and approves the implementation of Nuclear Criticality Safety Approvals (NCSAs)..." SAR Section 6.8.1, " Audits," commits in part, "Where appropriate, audit teams are supplemented with on site and/or off site technical specialists." The i
inspectors reviewed a recent high-level management assessment of the NCS program, which included an outside consultant.
Observations and Findings On February 18,1997, a failure to meet the double contingency principle for the C-400 Chemical Cleaning Facility " Cylinder Wash" operation was identified by an NCS engineer. All cylinder wash operations were immediately terminated and management initiated an assessment to identify the root cause of the i-failure and to determine whether'there were generic implications for other plant operations. KY/A-578, General Manager's Independent Assessment Nuclear Criticality Safety, Revision 0, was issued on March 17,1997. The assessment was conducted by a high-level team, consisting of the General
. Manager, the Manager of Safety, Safeguards and Quality, the Manager of
' Nuclear Safety, and an outside consultant.
6-The assessment report was presented to the inspectors by the General Manager
'during the entrance meeting on March 17,1997. It identified a number of-deficiency findings that included, but were not limited to:
Use of administrative vs. engineered controls for both contingencies.
e Use and implementation of independent verification for complying with e
the double contingency principle (DCP).
Program weaknesses in NCSE/NCSA process to develop and implement e
controls for DCP.
Operations involvement with the process and implementation of controls.
o inadequate walkdown verification of DCP controls prior to management e
review.
Program weakness in timely surveillance to implement controls for DCP.
a The inspectors also noted that the assessment report identified a number of positive findings and concluded that the NCS program was fundamentally sound.
The report also identified a number of recommendations for improvement of the NCS program that encompassed the establishment of PGDP Management expectations, Operation's ownership of the NCS program, and NCS program improvements. At the conclusion of the inspection, facility management had not yet completed their evaluation of the report to identify the long term corrective actions and establish an appropriate schedule.
Conclusion Facility Management conducted a broad in-depth review of the cylinder wash NCS violation that resulted in a number of substantive findings and recommendations. This initial response is considered a strength; however, due to the importance of this itern, it will be tracked as inspector Follow Up item 97-201-01.
C. Other Facility Surveillance and Audit Findings Scope The inspectors reviewed other facility surveillance and audit reports to determine whether potentially significant conditions adverse to quality (SCAQ) related to NCS were being identified and addressed by management.
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,i Observations and Findings As an immediate response to the cylinder wash NCS violation, the NCS group manager initiated a plant-wide surveillance of all independent verifications for double contingency (SUR 97-31-22). The review focused on the
.l administrative controls which (1) used independent verification of information, and (2) relied on independent verification for both legs of
' double contingency. Surveillance walkdowns were performed for each of the above and problem reports were initiated for several discrepancies. The inspectors noted that the response appeared to be appropriate and timely.
In addition to the above assessment and surveillance, the Safety, Safeguards l
& Quality Division.(SS&Q) conducted a series of five routine surveillances of the NCS program between March 4-14,1997, to evaluate the overall program j
effectiveness. The inspectors noted that these surveillances also identified i
a number of deficiencies for which five additional problem reports were initiated.
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. On the morning of April 2,1997, the inspectors observed the Plant Operations
' Meeting in which the NCS manager presented a review of SCAQ issues on NCS.
This presentation discussed a corrective action plan with assigned l
responsibilities and completion dates in response to the NCS Management Assessment report KY/A.-578, the Nuclear Safety Assurance Assessment Report NAR-C97-01, and the Confirmatory Action Letter From NRC for C-400 Restart, p
among others. The inspectors observed that discussions among the PGDP management team demonstrated an awareness and sensitivity for NCS issues.
One example is a suggestion by the Engineering Manager to develop a program to encourage ownership for NCS by the plant operations staff.
l Conclusion The f acility demonstrated an ability to self-identify potential SCAQs related to NCS and surface those issues to senior plant management for root cause j
analysis and corrective action.-
2.0 Nuclear Criticalitv Safetv Function I
A. Identification of Operations Requiring NCS Evaluation Scope Safety Analysis Report (SAR) Section 5.2.2.2, NCS Responsibilities, requires L
that operations involving uranium enrichment to 1 wt% or higher and 15 grams i
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or more of assU be identified and evaluated for NCS prior to operations. SAR Section 5.2.2.2 also requires, in part, that [NCS Engineers] " verify sufficient information is documented to allow independent analysis, l
verifying credible process upsets related to criticality safety are promptly-identified and evaluated, verifying compliance with the double contingency principle, checking for accuracy, and verifying applicability of the calculational methods." The inspectors conducted discussions with plant
. management, technical and operations staff to determine how PGDP ensured that all such operations were covered.
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Observations and Findinas The inspectors were informed by PGDP personnel that the criteria specified in SAR Section 5.2.2.2 had been the historic criteria used throughout most of the i
plant's history, it was expected that this past practice resulted in individual organization managers having identified all covered operations and initiated appropriate NCS approval (NCSA) requests for evaluation. No additional actions had apparently been initiated to confirm that all covered operations were evaluated prior to the transition to NRC oversight.
During a walkdown of the building C-335 Fissile Control Area (FCA) on March 20,1997, the inspectors observed that used NAM (Negative Air Monitor) l filters and fixed HEPA filters were stored in a 3 X 4' array prior to characterization. It was subsequently determined that the FCA array was not covered by a documented NCS approval (NCSA) nor identified as exempt per the criteria of SAR Section 5.2.2.3. When informed of this issue, the NCS group analyzed the array and determined that it was safe. Proper evaluation of operations involving fissile material is an important element for assuring an effective NCS program. Therefore, the failure to evaluate the 3 X 4 FCA array in the C-335 Building for NCS is Violation 97-201-02.
o Conclusions 1
Independent plant tours and process documentation reviews determined that PGDP appeared to have identified and evaluated all of the major process operations involving fissile material for NCS. However, one violation was
-identified concerning the failure to evaluate a 3 x 4 storage array in the C-335 Building.
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B. NCS Evaluation Request Process SGDDD SAR Section 5.2.2.3 requires that each operation involving uranium enriched j
to 1 wt% or higher assU and 15 grams or more of 23sU is evaluated for NCS prior to initiation. The inspectors reviewed the NCS evaluation request process to J
determine whether appropriate organizational interfaces were properly established and functioning.
Observations and Findings The NCS evaluation request process defined in the SAR includes the initiation of a NCSA request (Part A) by the operating organization that includes a system description. The request is signed by the operating organization and forwarded to the NCS group for an evaluation (NCSE) per their internal procedures. The NCS group then prepares Part B of the NCSA based on the results of the NCSE in order to document the NCS approval (limits and controls) for the operation. The NCSA then goes through a formal review and approval process.
In two instances, the inspectors noted that Part A of the NCSA was signed by i
operations after Part B of the NCSA was signed by NCS. Thus the operational conditions described in Part A coulo have been changed without NCS update to the limits in Part B. The two instances were NCSA 3973-02 " Uranium Recovery System in C-409," and NCSA 3973-09, " Uranium Recovery System in C-400." This was brought to the attention of NCS etaff whom indicated that these were administrative changes that would not have affected the established limits.
Although not safety significant for thesa NCSAs, changes to the original assumptions must be reviewed by the appropriate staff so that potentially significant changes do not go unreviewed.
Conclusions The facility has established a NCS evaluation process that is consistent with
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the commitments in the SAR and appears to be functioning properly. However, isolated instances were observed involving a lack of attention to detail concerning whether all changes to the identified system description in Part A I
were reviewed by NCS to verify that the specified limits and controls remained valid.
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10-C. NCS Evaluations (NCSEs) and Documentation Scope SAR Section 5.2.2.2, Nuclear Criticality Safety Responsibilities, requires, in part, that [NCS Engineers] " verify sufficient information is documented to allow independent analysis, verifying credible process upsets related to criticality safety are promptly identified and evaluated, verifying compliance with the double contingency principle, checking for accuracy, and verifying applicability of the calculational methods."
The inspectors reviewed the following evaluations to verify that the above commitments were being adequately addressed:
(1) " Nuclear Criticality Safety Evaluation for Product and Side Withdrawal in the C-310 Building at the Paducah Gaseous Diffusion Plant";
(2) " Nuclear Criticality Safety Evaluation of the C-409 Uranium Recovery System at the Paducah Gaseous Diffusion Plant";
(3) " Nuclear Criticality Safety Evaluation for Pump Decontamination in the C-400 Alkali Tank";
(4) " Nuclear Criticality Safety Evaluation for Removal and Handling of Contaminated Equipment from the Cascade at the Paducah Gaseous Diffusien Plant;" and (5) " Nuclear Criticality Safety Evaluation of the C-400 Uranium Recovery System at the PGDP."
Observations and FindiD9s The technical adequacy of the evaluations was reviewed and several inadequately documented assumptions or conditions were identified by the inspectors. The product withdrawal evaluation required modified floor plates or temporary diking above the scale pits when withdrawing product enriched to 2 2 wt%, but the basis for not requiring them at lower enrichments was not stated. Also, results were used in the Equipment Removal evaluation to prove a scenario was subcritical, but the optimum conditions were not explicitly analyzed. NCS staff indicated that the reviewer knew the optimum case would be subcritical because the cases run had a k.nless than O.87. However, the inspectors reviewed the independent review documentation but found no comments relating to these calculations. As a final example, the C-409
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Uranium Recovery evaluation analyzed optimum reflection conditions for the-normal case, but reduced the reflection in the upset condition without 4
technical justification.
The inspectors also observed problems in the area of computer calculations.
The computer models in the C-409 Uranium Recovery and the Equipment Removai evaluations wars inconsistent with the model description given in the text in five cases.1 in the most significant case, the ' ext stated that concrete was placed below the model but a void region was sr.3wn in the input file. This case was re-run with the concrete by NCS personnei and the spacing limit s
determined by the calculation was unaffected. NCS procedures do not require all of the calculations ~to be reviewed which may indicate why these discrepancies were not found in the review process.
l The inspectors also reviewed the evaluations and determined that the
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contingencies and controlled parameters were not explicitly identified, i
although they could be determined by a skilled criticality safety individual, The evaluation of the C-409 Uranium Recovery System was an exception, with the two controls clearly defined. However, this evaluation did not show a trend, as evaluations completed at a later date were more ambiguous.
q Conclusions
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l NCSE assumptions were not always documented or justified and several discrepancies were noted between the computer models and the model
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descriptions. Neither of these concerns were explicitly identified by the independent review. These are considered to be further examples of what USEC identified as a program weakness as discussed in Section 1.0 of this report.
D. NCSE " Contingencies and Controls" Scope i
SAR Section 5.2.2.3, Process Evaluation and Approval, commits, in part, that "The NCS evaluation process involves:
(1) a review of the proposed operation and procedures, l
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' These cases are: (a) The presence of reflection in Appendix B of NCSA 3973-02; (b) Stated j
. dimensions in GEN-10;(c) The presence of reflection in GEN-10; (d) Concentration in Table j
B-10.of NCSA 3973-02; and (e) Water density in NCSA 3974-05.
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12-(2)' discussions with the aubject matter experts to determine the credible process upsets which need to be considered, 7
(3) development of the controls necessary to meet double contingency, and i
(4) identification of the assumptions and equipment (i.e., physical controls) needed to ensure criticality safety."
4 The inspectors examined selected NCSEs for accuracy and conservatism in
.describ ng plant condit ons. The nspectors con ucted a selected walkdown of i
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the plant to verify that the assumptions in the NCSEs/NCSAs reflect actual plant conditions. The selected NCSEs reviewed include:
(1) " Operation of Temporary Fissile Storage Area (TFSAs),"
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i (2) "On Site Generation, Handling, Accumulation, Staging, Transportation, l
and Storage of Potentially Fissile Waste Up to a Maximum 5.5 Weight Percent Enrichment,"
(3) "Five Gallon Drum Storage at 5.5 Weight Percent," and 1
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- " Storage of Specific Equipment items Received from K-25."
(4) c Observations and Findings 1
The inspectors observed that the TFSAs consisted of areas within the process buildings where waste drums are temporarily stored. Uncharacterized drums are stored in an array in a diked area which physically maintains the spacing requirements. Once waste drums are characterized and demonstrated to have L
i less than 120 grams of 285U they may be stored in a closely-packed array on the basis of NCSE-207, "Five Gallon Drum Storage at 5.5 Weight Percent," which demonstrates that an infinite array of closely-packed waste drums is subcritical with a safe mass limit of 120 grams. As an upset condition, mass i
control was assumed to be lost by placement of an uncharacterized waste drum i
in the middle of the closely-packed array of characterized drums. This is E
considered a credible upset condition because of the identical nature of the drums, the proximity of the characterized and uncharacterized arrays, and the possibility of misplacing a drum label. The array w::s modeled as a 5 x 5 x 3 square array of drums (stacked three high in the vertical direction) l
- containing a single waste drum _ in excess of the mass limit.
During a plant walkdown, the inspectors observed the storage area and noted l
that the drums are actually stacked four drums high, with no apparent j
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engineered or administrative limit on stacking' drums beyond three rows.
Thus, the analysis did not conservatively evaluate a credible process upset '
j condition or establish appropriate limits. The NCS group subsequently re-ran l
l their. calculations using a larger. array and demonstrated that a sufficient
- safety margin existed.
l The inspectors further observed that the inside height of the drum is 32 cm, I
.though the height of hydrocarbon moderated UF, solution in the drums used in the calculation was varied up to a maximum of only 25 cm, at which height the j
solution exceeds the established maximum safe k,,. Thus, the height was used j
to determined the maximum safe mass of 120 grams 23sU within the drum. The evaluation considered an infinite array of such drums, but did not consider the effect of an exceeding the safe mass within a single can in the array.
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Thus, the evaluation did not adequately consider ~a credible upset condition and effectively bound the system. Again, a subsequent review by the NCS group
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determined that a sufficient safety margin existed.
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j The inspectors observed the storage of specific equipment items in building C-331 and compared the observations with NCSE GEN-20, " Storage of Specific Equipment items Received from K-25." The K-25 equipment is stored mostly within the DOE Material Storage Areas (DMSAs). Although control of DOE
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equipment is not a USEC function, interaction of such equipment must be considered when establishing safety limits for USEC processes.
In this regard, the inspectors observed that NCSE GEN-20 requires a two-foot spacing between K-25 equipment and other process equipment without valid t
technical justification. Thus, the assumptions and controls are not
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adequately documented for USEC processes, and there is insufficient information to allow independent confirmation of the analysis. In most cases, there is a rope defining the boundary of the DMSA which is a minimum of j~
two feet from the nearest equipment. But in some cases, the equipment is within two feet of the rope, permitting violation of the spacing requirement without entering the DMSA. There is only an administrative control requiring i
that no equipment be' spaced within two feet of the K-25 equipment. Since the rope encompassing this buffer zone is not controlled by USEC, this configuration may lead operators to assume that equipment may always be placed just outside the rope. The actual position of the two-foot boundary is i
thus ambiguous and could lead to future violations.
Conclusions The NCSEs reviewed were generally technically adequate and were based on 3
consideration of credible upsets. Appropriate controls and limits were j
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14 established to prevent an inadvertent criticality. While several examples-were noted whereby an NCSE did not rigorously consider /model the "as-exists"
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system configuration, ~ subsequent analysis demonstrated that an adequate margin of safety existed.
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j E. NCSA Implementation l
Scope l
1-1 SAR Section 5.2.2.3, Process Evaluation and Approval, commits, in part, that "first-line management is responsible for implementing the conditions i
delineated in the NCSAs through the use of such tools as training, operating procedures, posting, and labels... land) for assuring the employees l
understand both the procedures and NCSA requirements before the work begins."
1 CP issues 5 and 6 titled, " Nuclear Criticality Safety Approval Documents,"
l and " Nuclear Safety Approval implementation" commits to update all old NCSAs and complete any new NCSAs by December 31,1996. Item 9b of issue 27 titled l
" Procedures Program" commits to incorporate all NCSAs into plant procedures 4
i (and complete the required training) by December 31,1996. In a letter dated March 18,1997 (Serial GDP 97-0035) to the NRC, USEC stated that the above item / issues had been completed. The inspectors conducted a walkdown of j
building C-335 to verify that NCSA requirements had been implemented in i
appropriate operating procedures and that these procedures were being followed by operations personnel.
I Qhservations and Findings During a walkdown of the building C 335 FCA for used NAMs (negative air machines) and HEPA filters, the inspectors observed 2" x 3" signs with fine print containing special fire fighting instructions required by NCSA GEN-09.
I' Because the print was so small the inspectors asked the operations personnel i
how the fire fighters would be aware of the special requirements. The inspectors were informed that the information should be captured in the Pre-Fire Plan for the building. Subsequent discussions with the Fire Services group indicated that the Pre-Fire Plan did not contain the NCSA 'special fire fighting instructions for the FCA and emergency response training did not
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incorporate the special actions specified in GEN-09. The failure to flowdown NCSA requirements into operating procedures and to train fire fighters on these NCSA requirements is Violation 97-201-03.
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15 Conclusions.
independent plant tours and reviews of CP closure evidence package (see 1
Section 6.0, "NCS Operating Procedures") determined that PGDP had identified the major NCSA requirements and incorporated them into plant procedures.
1 However, one violation was identified concerning the failure to include the special fire fighting instructions contained in GEN-09 into the Pre-Fire Plan and to. train the emergency response personnel.
I 4.0 Confiauration Control Proaram for Nuclear Criticalitv Safetv 1
A. Change Control j
t Scope J.
SAR Chapter 5.2 titled " Nuclear Criticality Safety," Section 5.2.2.8, Change Control, requires NCS approval of all changes to items relied on for l
criticality safety. Section 5.2.2.8 further states that, "the Change Control t
Board (CCB) verifies the required reviews have been performed before approval."
Chapter'6.3 titled, " Plant Changes and Configuration Management (CM),"
contains PGDP's commitmente concerning the Configuration Control Program for
' NCS. The inspectors reviewed the PGDP change control process to determine whether proposed changes would be adequately reviewed for their impact on NCS.
Observations and Findings J
The inspectors reviewed the Change Control Charter and related documentation which describes the change control process. - The inspectors noted that as of April 4,1997, NCS was not included in the approval membership of the CCB even though this committee has approval authority for changes impacting NCS [SAR l
6.3.5.2.4]. Discussions with the CM Program Manager and the NCS Manager indicated that they were aware of the deficiency and were in the process of j
developing a procedure (CP2-EG-CF1032) to address this issue.
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in later discussions with the NCS Manager it was determined that several instances were identified where modifications in progress did not have appropriate NCS review or representation. This is considered significant i
since Compliance Plan issue 27, Interim Regulatory Commitments, requires that I
procedures and controls be established to ensure appropriate reviews of j
changes to procedures, plant, or equipment design impacting safety systems to ensure adequacy of configuration controlicriticality safety, and nuclear safety. Therefore, implementation of a formal procedure that assures all J
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, proposed changes to items relied on for criticality safety are reviewed by NCS personnel will be tracked as inspector Follow Up item 97-201-04.-
3 Conclusion NCS involvement in the change control process as practiced to assure that i
4 SAR 5.2.2.8 commitments are effectively met was not fully described in current plant documentation. However, procedure CP2-EG-CF1032 is under development to define NCS representation in the process.
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B. Configuration Management Assessments i
Scope
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SAR Section 6.3.5.6, Assessments, contains PGDP's commitments concerning the assessment program, as applied to the Configuration Management (CM) Program process. It requires an assessment of the CM Program's administrative requirements and related documentation, in accordance with six specified l
l elements. CP lscue 21, Action item 5, commits to developing procedures
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required to implement an assessment program to systematically evaluate the a
development and effective implementation of the CM Program elements and The inspectors reviewed the CM
.j related processes by December 31,1996.
assessment program process as applied to NCS to determine whether an adequate program had been developed in accordance with the SAR commitments.
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. Observations and Findinas The inspectors observed, on' April 4,1997, that no procedures had been F
developed which directly addressed assessments of CM Program development or related functional elements, as required. In response to this concern, the licensee submitted procedure UE2-QA-Cl1034, " Management Assessment" on April 14,1997, as evidence of such activity. Upon review of this document, the inspectors determined that UE2-QA-Cl1034 was not an implementing procedure, 1
but further supported the SAR commitment of establishing an assessment program. Specifically, Section 5.3.1 requires Division Managers to implement a management assessment program within their area of responsibility.
Discussions with Plant Management on April 4,1997, indicated that PGDP believed assessments as described by various other procedures adequately encompassed the CM Program and related elements. Since sufficient information had not been obtained to demonstrate how PGDP would fully implement SAR 6.3.5.6, this is identified as Unresolved item 97-201-05.
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. The inspectors also observed as of April 4,1997, that documentation of assessments completed on the development of the CM Program was' unavailable.
Discussions with plant personnel indicated that the auditors assigned to conduct the various assessments were expected to go back to the source.
l documents (in this case the.SAR) during development of the assessment / audit.
plans to assure that'all of the applicable requirements were met. One element in particular for this process is field verification of design requirements and documentation which includes verification directed toward the SAR system boundaries and their specific SSCs [SAR 6.3.5.6].
However, the inspectors determined that no baseline assessment of the CM Program had been completed to ensure these system boundaries and specific a
SSCs related to NCS are as documented in the Boundary Definition Manuals (BDM). In response to this concern, plant personnel indicated that these t
-documents had just been developed and were reasonably expected to be up-to-date. However, the inspectors noted that NCS signature review of the BDMs was not evident. Discussions with the NCS Manager indicated that although NCS had been involved with the development of limits associated with these documents, no baseline verification and review was completed for the final BDM records.
l Since' maintenance of these documents is an important element of the CM Program to ensure valid criticality safety configuration, this item is being tracked as inspector Follow Up item 97-201-06.
4 Conclusion It does not appear that management has defined its expectations in regard to how SAR commitment 6.3.5.6 is to be met. Although at least one procedure was identified which addressed assessments in general, it did not specifically satisfy the SAR requirement. Further, assessments had not been completed to i
baseline the adequacy of the CM Program and documentation was not available to demonstrate that the BDMs (which specify SSCs important to NCS) had been appropriately reviewed / approved. This is considered a program weakness.
i 5.0 Nuclear Criticality Safetv Chance Control Scope The inspectors reviewed the change control process to determine whether j
proposed changes were being reviewed and approved in accordance with 10 CFR 76.68. The review included the following evidence packages related to the NCS c
. change control process that were prepared by PGDP for closure of CP items and/or issues that were due for completion no later than December 31,1996:
]
f
i 4
18-(a) Issue 5, " Nuclear Criticality Safety Approval Documents," except for the 1
last item in the Plan of Action and Schedule (POA&S), which is to complete all aspects of Technical Safety Requirement 3.9 implementation and its '
associated tentacles for NCS which will be in place no later than March 3,
.1997.
i f
(b)_ lasue 6, " Nuclear Criticality Safety Approval Implementation."
j (c) Items 1(a) and 2(b) of issue 21, " Plant Changes and Configuration Management" (USEC Task ids _ CPI POA-C21.01a and cpl-POA-C2102b).
)
i (d)_ Items 3,4i 5, 7, and 8 of issue 22, " Maintenance Program" (USEC Task ids 4
CPI POA-C22.03a, CPI POA-C22.04a, CPI-POA-C22.05a, CPI-POA-C22.07a, and CPI POA-C22.08a).
(e) Item 9(b) of issue 27, " Procedures Program" (USEC Task ID CPI-POA-C27.09b).
(f) Item 2 of issue 29, " Quality Assurance Program implementation" (USEC Task ID CPI-POA-C29.02a).
Observations and Findinas The inspector reviewed several procedures pertinent to the 10 CFR 76.68 plant change process that included:
l (1) UE2-RA RR1036, "10 CFR 76.68 Plant Change Reviews," Change C, dated I
- December 23,1996, (2) UE2-OP-RR1034, " Control and Maintenance of the NRC Certification Documents,"
i l'
(3) UE2 OP RR1039, " Certificate Amendment Rsquests, Technical Safety Requirement Clarifications, and Technical Safety Requirement Basis Changes,"
(4) UE2 EG-NS1030, "Unreviewed Safety Question Determination," and (5) UE2-RA RR1035, " Regulatory Commitment Flowdown Tracking." The I
inspectors found that UE2-RA-RR1036 contains the Plant Change Review (PCR) form, which walks the plant staff through the PCR process to determine whether the proposed plant changes will requires regulator (i.e., NRC) or Plant Operations Review Committee (PORC) approval. The
--r-
.m-
=
19 inspector noted that item 7.2, Required Approvals, of the PCR form states:
If the answer to any of Questions 6,8,9.2,10,13,15.1 through 15.3, or 16 is Yes, the proposed change requires regulator approval prior to implementation.
10 CFR 76.68 authorizes the Corporation (USEC) to "make changes to the plant or the plant's operations as described in the safety analysis report without prior Commission approval provided all the provisions (emphasis added)" of 676.68 are met. The inspector reviewed the above questions against the specific requirements in 10 CFR 76.68 and identified the following concerns:
a.
Changes Which Decreases the Effectiveness of the Safety Program Without Prior NRC Review 10 CFR 76.68(3) states, "The changes may not decrease the effectiveness of the plant's safety, safeguards, and security programs." Question 13 of the PCR addresses the effectiveness of the Quality Assurance Program Description, Emergency Plan, Fundamental Nuclear Materials Control Plan (FNMCP), Physical Security Plan for the Transportation of Special Nuclear Material of Low Strategic Significance (" Transportation Security Plan"), and the Security i
Plan for the Protection of Classified Matter (" Classified Matter Plan").
These plans and programs are not part of the SAR in accordance to 10 CFR 76.35.
-Tne inspector determined that the PCR form addressed changes to the TSR bases
)
and Chapters 3 and 4 of the SAR through the performance of an Unreviewed Safety Question Determination (USOD). However, no part of the procedure directly addressed a determination as to whether the proposed change involved a decrease in the effectiveness of the plant safety programs described in Chapter 5 of the SAR, or a decrease in overall nuclear criticality safety.
This is a significant omission because the actual safety basis for many of the plant operations are provided in the NCSEs, which were not individually reviewed by the NRC staff during the certification process. Rather, the staff relied on the development and implementation of NCSEs and NCSAs in accordance with the commitments specified in SAR Chapter 5.2, Nuclear Criticality Safety, to provide the required level of nuclear safety for those items not directly covered by a TSR (TSRs are required for those instances where double contingency is not met).
UE2 EG-NS1030, Appendix E, provides 6 questions concerning accidents and malfunctioning of equipment described in the SAR and one question concerning
y e
e.,,.
20 -
' the technical basis for-the TSR to determine whether the proposed plant change constitutes an Unreviewed Safety Question (USO). These questions are not applicable for the majority of NCS activities because (1) the accident analyses of NCS events are based on the NCSEs which are not in'the SAR, and (2) there are no TSRs required for process operations that comply with'the double contingency requirements (other than those specified in TSR 3.11).
Since the specific double contingency schemes identified in the NCSEs are not t
.in the SAR or TSRs, the response to the seven USQD questions need not be addressed. Even if they.were addressed, the response to the questions would most likely be universal, because the probability and consequence of all double contingency criticality events is'an inadvertent criticality.
The effectiveness of plant safety to prevent an inadvertent criticality event is currently provided through the NCSE/NCSA process described in the SAR 1
2 1
Section 5.2, Nuclear Criticality Safety, it currently includes commitments made to follow certain ANSI standards, and meet specificd program elements and technical aspects, all of which would not be adequately addressed by the t
USQD process.
. The inspectors met with the Manager, Nuclear Safety, and the staff from Nuclear Regulatory Affairs to discuss this issue. They provided the i
inspectors _ background information regarding the philosophical and technical basis behind the methods used in the PCR process procedure. Specifically, Section 6.3.2 of the SAR clearly states how USEC plans to implement the 10 CFR 76.68(a) requirements, and 10 CFR 76.35, " Contents of Initial Application,"
i which includes the nuclear safety programs as part of the SAR. Section 6.3.2 of USEC's certification application considers only those programs / plans in 10 CFR 76.35, that are not in the SAR, as needing to meet 10 CFR 76.68(a)(3). The initial plant's response therefore was, "We (USEC) believe the process as described, achieve the effectiveness review of safety program changes as required by 10 CFR 76.68."
The Inspectors noted that the SAR Section 6.3.2 description of the change review process did not appear to be consistent with the specific requirements of 10 CFR 76.68. This matter will be tracked as inspector Follow-up Item 97-201-07.
b.
Changes to Certain Conditions of the Compliance Plan Without Prior NRC Review j
10 CFR 76.68(5) states that "The changes may not involve a change to any condition to the approved compliance plan." The " approved compliance plan" 4
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3,,,
j ;
refers to 10 CFR'76.35(b).which requires the application for an initial
- certificate of compliance to include a plan prepared and approved by DOE for Lachieving compliance with respect to any areas of noncompliance with the NRC's regulations that are identified by the Corporation as of the date of the application that includes:
.(1) A description of the areas of noncompliance; 1
l (2) A plan of actions and schedules for achieving compliance; and (3) A justification for continued operation with adequate safety and safeguards.
' Since the CP is a one time " exemption," authorized by Congress for a limited period from NRC requirements, the description of the areas of noncompliance and the justification for continued operations (JCO) provide the safety basis-for granting this exemption. Changes to the description of noncompliance may require changes to the ' CO, which in turn can affect the safety basis for the J
i certification. Therefore, allowing all changes to the description of noncompliance section of the CP without prior NRC approval has both safety and regulatory significance.
1 The inspector noted that the PCR process would require only those proposed changas wnich involved a change to the JCOs or to the POA&S sections of the CP.
be submitted to NRC for prior review. If the proposed change involved a change to the Description of Noncompliance, Requirements, or Commitment sections of the CP, no pr!or NRC approval would be sought according to procedure UE2-RA-RR1036.
The inspectors discussed this issue with the Manager of Nuclear Safety, and staff from Nuclear Regulatory Affairs. They responded by presenting to the inspectors a USEC letter to NRC dated December 11,1996 (Serial No. GDP-0200),
which documented a telephone conversation between the NRC Project Manager for PGDP and a representative for USEC Headquarters. The inspectors requested a copy of NRC official response to this letter since USEC had already implemented the agreement discussed in the telephone conversation with an NRC i
staff member. No response was provided to the inspectors on the request.
.Since DOE prepared and approved the compliance plan, including the description of noncompliance, any changes to the compliance plan, in accordance to 10 CFR 76.35(b), need to be approved by DOE. The inspector could not identify, in procedure UE2-RA-RR1036, any reference to DOE in this -
regard.
I
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The Inspectors noted that Procedure UE2-RA-RR1036 authorizes the plant to make changes to the description of noncompliance, commitments, and requirements sections of the compliance plan without prior NRC approval.
This does not appear to be consistent with the wording of 10 CFR 76.68(a)(5)
. or any docketed exemption granted by the NRC. This matter will be tracked as inspector Follow-up item 97-201-08.
Conclusions a.
Changes Which Decreases the Effectiveness of the Safety Program Without Prior NRC Review.
i j
The change review process described in SAR Section 6.3.2 and implemented by procedure UE2-RA-RR1036, Change C, does not provide specific guidance to j
plant staff on the requirements of 10 CFR 76.68(a)(3) regarding the j
determination that the changes may not decrease the plant's safety.
b.
Changes to Certain Conditions of the Compliance Plan Without Prior NRC j
Review USEC has developed and approved a procedure (UE2-RA-RR1036, Change C) to make.
changes to the description of noncompliance, commitments, and requirements sections of the compliance plan without prior NRC approval. This does not appear _to be consistent with the requirement of 10 CFR 76.68(a)(5) or any docketed exemption approved by the NRC.
6.0 NCS Ooeratina Procedures l
1ICDDR Item 9 of the POA&S se:: tion of CP issue 27, " Procedures Program," commits to issue the required operational policy statements ar.d implement new or updated l
procedures (including required training) to fully implement the Quality l
. Assurance Program or other activities identified in the application for level 2, 3, and 4 AQ-hlCS procedures (unless covered by item 8) by December 31,1996.
4 The inspector reviewed the procedures containing NCSA requirements for completion by March 3,1997, rather than by December 31,1996. This is based on NRC's letter to USEC dated December 18,1996, which stated that no action on NRC's part was required for Certificate Amendment Requests for due dates that would be completed prior to March 3,1997.
1-1
r Observations and Findings j
The evidence packages for completion of procedures containing NCSA l
requirements (USEC Task ID cpi-POA-C22.07a) contained NCSA flowdown documentation for both December 2 and 31,1996.
The inspectors cross checked the evidence provided against the NCSA Index l
containing all existing NCSAs as of March 3,1997. For those NCSAs that were completed by December 31,1996, the evidence provided for procedures containing NCSA requirements and training appears to be complete. The inspector was provided an updated cross reference database sheets (dated L
April 2,1997) which relate plant procedures and training / module to each j -
NCSA.
j
' The inspectors followed up by selecting specific procedures such as procedure 4
CP4-CO-CN2007, Rev. O, titled, " Operation of Surge Drums," and conducted a walkthrough of this procedure for building C 331. The inspectors observed that the NCSA requirements are properly identified in this procedure.
r l-However, the SAR commitment was not stamped in the procedure as committed to in SAR Section 6.11.4.2. The inspectors discussed documentation requirements in the surge drum procedure with the Building Manager for C-331. The inspectors observed that the C-331 Area l Equipment Status and Control Room l.
Log, during the period from March 3,1997, through March 18,1997, had all the
~
required documentation entries in accordance to the plant's surge drum l
procedure (e.g., pressure and room temperature).
l Conclusion T,o Level 2, 3 and 4 AQ-NCS procedures (not covered by item 8 of the CP issue 2 / olan of Action and Schedule) have generally been updated in accordance with i
the commitments. However, cases are continuing to be identified by the plant and NRC whereby all of the NCSA requirements have not yet been identified in the operating procedures.
3 7.0 Maintenance for Nuclear Criticality Safety
)
Scope
)
The inspectors reviewed the evidence package ifor closure of item 5, of CP t
issue 22 (USEC Task ID CPI-POA-C22.05a), " Maintenance Program." PGDP's application commitments are contained in SAR Section 6.4 titled,
" Maintenance." The inspectors reviewed selected documents provided in the
)
evidence package.
,_y.+
...7-,
o..'.
24 -
Observations and Findinas The inspectors observed that the " Corrective Maintenance Program" procedure (MAP-CM-001, Rev. 2) and the " Work Control Process" procedures (MAP-CM-001 and CP2-GP-GP1032, Rev.1) are provided as supporting evidence for close out of item 5, Issue 22 of the CP. The introduction section for the evidence package relatirg to CP maintenance was signed by the Maintenance Function Manager, and 'Jated December 27,1996.
During a subsequent review of the referenced procedures, the inspector noted that MAP-CM-001, Rev. 2 had been deleted under the Plant Change Review (PCR) process on February 27,1997 (PDF #CG7-00140). The basis for this deletion, as stated in the PCR, was that the implementation of procedure CP2-GP-GP1032, Work Control Process, Rev.1, provided the accessary commitment requirements for Work Control, Post Maintenance Testing, Procurement, Receipt inspection, Control, and issuance of Q items, Repair Parts, Materials and Services as addressed in issue 22 of the Certificate of Compliance.
The inspectors reviewed both procedures MAP-CM-001, Rev. 2 and CP2-GP-GP1032, Rev.1, and found that not all items in proceduro MAP-CM-001, Rev. 2 had been incorporated into procedure CP2-GP-GP1032, Rev.1. Specifically missing in CP2-GP-GP1032 is a section on reviewing available corrective and preventive maintenance histories to identify problems and suggest solutions, and to maintain trends of corrective maintenance backlogs. Trending of maintenance activities is discussed in SAR Section 6.4.12.3, titied " Performance Data,"
in the application, and is addressed bs/ item 1 in the POA&S of issue 22 in the CP, which commits to the developrnent of a maintenance history and trend analysis program by September 30,1997. However, before this action is completed, the Certificate Holder has committed to the following in the Justification for Continued Operation (JCO) section of issue 22:
... Procedures are in place for corrective maintenance, preventive maintenance, schedules calibration, and control of work.... The site will continue to operato under these procedures until they are
. replaced or superseded by procedures compliant with the Application i
commitments.
The inspectors discussed the above issues with the Mechanical Maintenance Manager and the General Production Service Manager. The plant staff stated they can understand the observation of the inspectors in that the PCR was in error in that the deletion of the " Maintenance Program Procedure" was an administrative violation of the JCO section of the CP. The next day, plant staff presented the inspector with a copy of " Work Control Performance
i l
- Indicators" for February 1997, which demonstrate that the plant is currently monitoring work maintenance activities for trends. A problem report was also initiated.
Conclusions The plant is generally implementing the various maintenance program commitments specified in the SAR and Compliance Plan. However, one example was identified, whereby the completed PCR form to delete a procedure failed to address the interface with the Compliance Plan JCO commitments.
8.0 Validation of NCS Calculations Scope SAR Section 5.2.3.2 provides information concerning the use and validation of computer calculations for nuclear criticality safety evaluations. The inspectors reviewed KY/S-221, " Validation of the Paducah Gaseous Diffusion Plant Nuclear Criticality Safety Code System for the ENDF/B-IV 27 Group Cross Sections," dated January 1996, to determine the area of applicability and thoroughness of the validation study. The software configuration control program as discussed in SAR Section 5.2.3.2 was not examined.
Observations and Findings The Validation Report was found to be a significant improvement over the previous one (dated September 1993) in that the number of cases run, materials considered, and software options exercised were greatly expanded, and in that the source references for critical experiments used were given. This allowed for independent duplication of results for the first time. The report covered 66 benchmark experiments using enrichments varying from that of natural uranium. i to 5 wt%. Fissile compounds modelled included uranium-metal, I
UO F, UO, U 0s, UNH, and UF. A variety of reflectors and moderators in 2 2 2
3 4
addition to water and concrete were modelled. The validation was done using the ENDF/B-IV 27-Group cross section library only and thus the area of applicability is confined to the 27-Group library. Resonance self-shielding was used for lattice-cell and multiregion calculations and biasing was included in the validation.
The validation report was observed to have two weaknesses. The albedo option was not validated in the validation study, even though it has been used in several NCSEs. The NCS group ran several cases with and without albedos and found no statistically significant difference at > 20. Corrective measures
1 o,
. 26 -
have been taken to prohibit the use of this unvalidated option except for initial scoping calculations.
' Secondly, the maximum enrichment validated by the computer code was 5 wt%,
even though an enrichment of 7 wt% was observed to have been used in two technical justifications. No instance was found of analyses done at enrichments beyond the current operating assumptions of the facility.
Although a higher enrichment was analyzed for the K-25 Equipment in the C-331 Building, the NCSE governing this area (GEN-20) did not use computer
-calculations because the enrichment would have been outside the validated range. The inspectors examined the plots of neutron multiplication vs.
j enrichment in the Validation Report. Since there were no error bars associated with the data, it was not possible to determine whether there was a statistically significant trend in the bias as a function of enrichment.
t s
Conclusions The Validation Report has been revised significantly and is adequate in meeting the computer calculation requirements of SAR Section 5.2.3.2 for i
enrichments up to 5 wt% 23sU.
j 9.0 Basis for Criticality incaoabilitv of 6 wt% 23sU l
Scope
]
SAR Section 5.2.4.14 (July 1996) states "The technical justification for the j
statement that uranium enriched to 6 wt% or less will not support a nuclear chain reaction without the presence of moderator has not been properly documented or is not retrisysbie." The technical justification for that statement was subsequently documented in KY/G-614, " Technical Justification for Subcriticality of Unmoderated Uranium Enriched to 6 wt% '5U," which was reviewed by the inspectors for technical adequacy.
Observations Spheres of UO F, under full water reflection and at theoretical density were 2
modelled to represent the most reactive credible configuration of material and calculations were performed using a number of different H/U ratios. The inspectors independently confirmed the atomic number densities and the mass density of the homogeneous mixtures used in the calculations. The insoectnre U
also~ performed bounding calculations which showed that an effectively infinite mass of dry UO F, cannot go critical.
2
1 os. *
=
, The technical justification stated that since oxygen is a more effective moderator than fluorine, UO F represents a more reactive configuration that 2 2 solid UF, or UF. Although oxygen has a much higher moderating ratio that 4
fluorine, the theoretical density of UF,is greater than that of UO F -
2 2 Although the assertion that UO F is the more reactive compound was therefore 22 not adequately justified, the inspectors found this to be the case based on independent calculations. Furthermore, the calculations were performed at 7
)
wt% 2asU, even though the Paducah SCALE code has only be validated to 5 wt%.
The inspectors found the argument for allowing extrapolation to 7 wt%,
I without compensating adjustments to the margin of suberiticality to be unconvincing.
The certificate holder is committed to ANSI /ANS-8.1, which states in part
)
4.3.2 "The area (s) of applicability of a calculational method may be extended
... by making use of trends in the bias." However, no attempt was made in the i
technical justification to extrapolate the bias into the region of extended enrichment. Since the system is very subcritical for reasonable masses of material, this is not considered to be a substantial safety concern.
Conclusion KY/G-614 provides the technical justification for asserting that uranium j
enriched to 6 wt% or less is incapable of supporting a chain reaction without the presence of moderator. However, the SCALE code used to conduct the supporting calculations was only validated to 5%.
- 10. Basis for Criticality incanability of Gaseous UF.
SCDDD SAR Section 5.2.4.15 (July 1996) states "Section 5.2.3.1 states that UFe in the gaseous phase, at pressures and temperatures existing in the enrichment cascade equipment, is incapable of supporting a nuclear chain reaction even when intermixed with hydrogenous material (i.e., HF)...This justification has not been properly documented or is not retrievable." The inspectors reviewed KY/G 615, " Technical Justification for Suberiticality of Gaseous UFe" which was written in November 1996, for technical adequacy.
Observations To allow for a conservative treatment of the system, the maximum credible 1
configuration of a spherical mass of UF. with full water reflection was utilized. Maximum operating cascade pressure is presumed to be 45 psia (at 1
2 9
4
d,e s
+
28-product withdrawal just prior to liquification), at which minimum gas temperature is 195"F. These conditions have been verified to produce the highest credible density of UF,in the cascade. Calculations were performed at different temperatures and pressures and at different H/U levels.
The atomic number densities and resultant k,,, were confirmed by the inspectors through independent analysis. The results show that for P < 50 psia, the system is subcritical at all moderation levels and k,,, has a maximum at H/U = 0. For systems above this pressure, the neutron multiplication increases with increasing moderation and can reach a maximum k,,,> 1. The inspectors determined that this is due to a phase change of HF to a liquid at
~50 psia, resulting in a hydrogen number density approximately three orders of magnitude greater. However, the maximum cascade pressure is limited by the high discharge pressure safety system, which trips the Normetex pumps at 42 psia. Thus both the safety system must fail to trip and there must be a significant buildup of moisture in the cascade in order to reach a critical gaseous configuration.
Conclusion The technical justification has been determined to be adequate.
- 11. E.xit Meeting The inspectors met with plant management on a daily basis throughout the inspection. period. An exit meeting was held on April 4,1997. No classified or proprietary information was identified.
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PARTIAL LIST OF PERSONS CONTACTED Lockheed Martin Utility Services 1.
S. A. Polston, General Manager
^
H. Pulley, Enrichment Manager W. E. Sykes, Nuclear Regulatory Affairs Manager S. R. Penrod, Operations Manager
+
V. J. Shanks, Chemical Operations C. Dean, Nuclear Criticality Safety Manager United States Enrichment Corooration f
I J. M. Brown, Engineering Manager J. A. Labarraque, Safety, Safeguards and Quality Manager l
United States Deoartment of Energy G. A. Bazzell, Site Safety Representative Nuclear Reaulatorv Commission K. G. O'Brien, Senior Resident inspector i
J. M. Jacobson, Resident inspector e
1
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'6,4e?
l
-. ACRONYMS USED CAAS Criticality Accident Alarm System CCB Change Control Board t
CM-Configuration Management CP Compliance Plan DCP Double Contingency Principle i
DMSA DOE Material Storage Area DOE Department of Energy FCA Fissile Material Control Area i
FNMCP Fundamental Nuclear Material Control Plan HEPA High Efficiency Particulate Airfilter ids inventory Differences
-IP Inspection Procedure JCO Justification for Continued Operation NAM Negative Air Monitor l
NCS Nuclear Criticality Safety NCSA Nuclear Criticality Safety Approval NCSE Nuclear Criticality Safety Evaluation NCS-AQ Nuclear Criticality Safety Augmented Quality NMSS Office of Nuclear Material Safety and Safeguards i
QAP Quality Assurance Program PCR Plant Change Review PDF Procedure Development Form PDR-Public Document Room i
PORC Plant Operations Review Committee
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POA&S Plan of Action and Schedule l
PGDP Paducah Gaseous Diffusion Plant ROA Regulatory Oversight Agreement SAR Safety Analysis Report j
SCAQ Significant Conditions Adverse to Quality i
SS&Q Safety, Safeguards & Quality SSCs Systems, Structures, and Components TFSA Temporary Fissile Storage Area
)
TSR" Technical Safety Requirement USEC-
_ United States Enrichment Corporation USQ' Unreviewed Safety Question USOD Unreviewed Safety Question Determination
,