ML20212H380
| ML20212H380 | |
| Person / Time | |
|---|---|
| Site: | Paducah Gaseous Diffusion Plant, Portsmouth Gaseous Diffusion Plant |
| Issue date: | 09/24/1999 |
| From: | Toelle S UNITED STATES ENRICHMENT CORP. (USEC) |
| To: | Kane W NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| References | |
| GDP-99-0173, TAC-L32121, TAC-L32122, NUDOCS 9910010093 | |
| Download: ML20212H380 (36) | |
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.6' USEC A Global Energy Company September 24,1999 GDP 99-0173 Mr. William F. Kane Director, Office of Nuclear Material Safety and Safeguards Attention: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Paducah Gaseous Diffusion Plant (PGDP)
Portsmouth Gaseous Diffusion Plant (PORTS)
Docket Nos. 70-7001 and 70-7002 Response to Request for AdditionalInformation on Paducah and Portsmouth Transmittals
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of 1999 Annual Update to Certification Applications (TAC Nos. L32121 and L32122)
Dear Mr. Kane:
This letter is in response to NRC's request, dated July 23,1999 (see the reference), for additional information specific to the 1999 Annual Update to the Certification Applications for PGDP and PORTS. The responses to the identified questions for PGDP and PORTS are provided in Enclosures 1 and 2, respectively. The RACs addressed in the referenced request are available at the sites for
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review.
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If you have any questions or comments on these responses, please contact John Lobre at (301) 897-3125. Commitments contained in this submittal are contained in Enclosure 3.
Sincerely, S. A.
^ l d Steven A. Toelle Nuclear Regulatory Assurance and Policy Manager D
9910010093-990924-PDR ADOCK 07007001 C
Reference:
Letter from Melanie A. Galloway (NRC) to Steven A. Toelle (USEC), Response to Steven A. Toelle Letter Dated June 14,1999, "10 CFR 76.68(a)(3) Decreased Effectiveness Reviews" and a Request for Additional Information for Paducah and Portsmouth Transmittals of 1999 Annual Update to Certification Applications (TAC Mos. L32121 and L32122, dated July 23,1999.
6903 Rockledge Drive, Bethesda, MD 20817-1818 Telephone 301->64 3200 Fax 301564-3201 http://www.usec.com Offices in Livermore, CA Paducah, KY Portsmouth, OH Washington, DC
Mr. William F. Kane September 2,4,1999 GDP 99-0173, Page 2
Enclosures:
- 1. United Ltates Enrichment Corporation, Paducch Gaseous DifTusion Plant, Docket No. 70-7001, Response to July 23, 1999 NRC Request for Additional-Information Conceming 1999 Annual Update to the Certification Application l
(TAC No. L32121)
- 2. United States Enrichment Corporation, Portsmouth Gaseous Diffusion Plant,
- Docket No. 70-7002, Response to July 23,1999 NRC Request for Additional i
Information Concerning 1999 Annual Update to the Certification Application
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(TAC No. L32122)
- 3. Commitments Contained in This Submittal ec:
Mr. Robert C. Pierson (NRC)
NRC Region III Office NRC Resident Inspector-PGDP NRC Resident Inspector-PORTS l
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3.
GDP 99-0173 Page 1 of 15 United States Enrichment Corporation Paducah Gaseous Diffusion Plant Docket No. 70-7001
' Response to July 23,1999 NRC Request for AdditionalInformation Concerning 1999 Annual Update to the Certification Application (TAC No. L32121) i l'
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GDP 99-0173 Page 2 of 15 Question 1.
RAC 96Cl20," Gas Treatment and Intermediate Gas Removal System (IGRS)"
The RAC relaxes monitoring requirements. Previously, the SAR required securing oxidants from off stream treatment ifonly one monitoring instrument was taken out of service. The RAC revises the requirement to only securing oxidants from off stream treatment when all instruments were taken out of service. That revision would allow operations with only one monitoring instrument in service. Why does that not increase the probability of an accident when reliance of detecting an explosive mixture can be with only one instrument? A failure of that one instrument then provides the opportunity for formation of an explosive mixture without detection.
Response
The safety analysis performed for this change was based on the inaccurate assumption that both monitoring instruments measured CIF and F concentrations. In fact, these monitoring devices 3
2 are not redundant, and the probability of an accident would have been increased. One of the instruments monitors CIF concentrations and the other monitors F concentrations.
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The procedure for the operation of the C-310 facility, CP4-CO-CN2010, requires that when one of these instruments is out of service, return of the gas that the particular instrument monitors is prohibited until it is returned to service. The SAR description prior to RAC 96Cl20 stated that return of both gasses was halted when either instrument was out of service. This is not necessary, and it has not been a plant operating practice, since an explosive mix can be prevented by controlling either concentration while the other gas is isolated from the cascade. The SAR change in RAC 96Cl20 was intended to describe more clearly the actual operating practice of monitoring the concentrations of these potentially explosive gasses as they are being returned from cell treatments. However, it was poorly worded and resulted in perraitting a mode of operation that would have been unsafe. A review of CP4-CO-CN2010 was conducted and it was determined that this mode ofoperation was never implemented (since the SAR change was never intended to change existing operations). Therefore, there has been no reduction in the level of j
safety of operation.
A new RAC will be completed to clarify the actual restrictions on returning cell treatment gasses to the cascade. This change will be completed by December 15,1999. ATRC-99-5097 was issued on August 26,1999 to document this problem and the associated corrective actions.
GDP 99-0173 Page 3 of 15 Question 2.
RAC 96C0151, " Autoclave Instrumentation" and RAC 97C113, "C-360 instrument Upgrade" The modifications in these RACs are installing new pressure transmitters and associated equiprnent in the autoclave and UF6 pressure systems for increased capacity, accuracy and resolution. While the new equipment is recognized to be more accurate, the RACs are not clear on the basis for stating that the limiting control settings and associated safety limits do not change as a result of these modifications. There is no discussion on the new setpoint calculations or methodology and no comparison to the old methodology. Provide more detailed discussion to support the USQD.
Response
The Autoclave Instrumentation and C-360 Instrument Upgrade modifications did not and were l
not meant to change the TSR limiting control settings (LCSs) or safety limits (SLs). The setpoint calculation methodology did not change. These modifications were a result of Compliance Plan Issue 3. The setpoint calculations were established using the setpoint change control procedure, CP3-EG-EG1070, at the PGDP. The setpoint change control procedure, used during the modifications in question, is based on (1) ANSI and ISA joint standard ANSI /ISA-S67.04-Part I, "Setpoints for Nuclear Safety-Related Instrumentation," and (2) Recommended Practice IS A-RP67.04-Part II," Methodologies for the Determination of Setpoints for Nuclear Safety-Related Instrumentation."
The methodologies utilized in the setpoint change control procedures, both current and previous revisions, have remained the same. The TSR LCSs and SLs are used as boundaries that are not to be crossed. To ensure that these boundaries are not crossed the setpoint change control procedure determines channel uncertainties (' measurable and immeasurable) associated with instrument loops to develop trip setpoints. The errors associated with instrument uncertainties are calculated by the square root of the sum of the squares method as described in the above standard and recommended practice. Also for added safety, and sometimes convenience, margins are added for insurance against breaching a TSR LCS or SL. With the combination of channel uncenainties and margins, setpoints are established. The setpoints ensure that, even with the channel uncertainties applied and after instrument drift with time, the trip settings will prevent a breach. Thus, the setpoints for the systems changed, but the TSR LCSs and SLs did nel change.
Thus, the setpoint calculations resulting from the modification used the same TSk limits and the same setpoint methodology to establish the setpoints for the new instrumentation. The replaced instrumentation for the modified systems included pressure measurement equipment, pressure indication equipment, and pressure switches. The setpoint calculations were then incorporated into the applicable TSR surveillance procedures that govern the testing and calibration of the
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GDP 99-0173 Page 4 of 15 systems. Post-modification testing was performed per the approved TSR surveillance procedures to ensure adherence to the applicable TSR LCSs and Sis.
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l GDP 99-0173 l
Page 5 of 15 Question 3.
RAC 97C0239," Problem Reporting Screening Committee" The change to SAR Section 6.8.2.4," Problem Reporting," deletes the screening committee. This would appear to decrease the effectiveness of the QAP. What committee (or position) is responsible to perform the screening specified, and where is that commitment documented?
Response
Through 10 CFR 76,76.93, USEC is required to " establish, maintain, and execute a Quality Assurance Program (QAP) satisfying... ASME NQA-1-1989,... or satisfying alternatives to the applicable requirements." NQA-1, Basic Requirement 16," Corrective Action,"in turn requires that conditions adverse to quality "be identified promptly and corrected as soon as practical," and that for significant conditions adverse to quality the " identification, cause, and corrective action shall be documented and reported to appropriate levels of management."
USEC committed to develop and implement a corrective action program meeting the requirements of NQA-1, Basic Requirement 16, through Section 2.16 of the QAP. Section 2.16 of the QAP defines responsibility for " development, maintenance, and implementation" of the corrective action program and that " Procedures for the corrective action process are established."
The QAP makes no commitment to establish or maintain a " Screening Committee."
Section 6.8 of the SAR provides a brief discussion of the system of audits and assessments at PGDP put in place to " ensure that the health, safety, and environmental programs... are adequate and effective ly implemented." Section 6.8.2, " Assessments," states that " Assessments are performed by management responsible for implementing sections of the QAP to assess the adequacy of the part of the QAP for which they are responsible.. " Following this statement are three subsections describing the types of assessments performed and a fourth, " Problem Reporting," which states that all plant employees "have the responsibility to write problem reports on safety, operating and noncompliance items." It then continues with " Problem reports are screened routinely to assign an owner.. as well as to determine ifidentified problems represent significant conditions adverse to quality (SCAQ) based on established criteria." These l
sentences did not and do not mention what entity (individual or group) " screens" the problems.
l The original version of Section 6.8.2.4 then continued with descriptive information regarding the entity screening the problems and when they met.
The original intent of SAR Section 6.8.2.4 was only to state that assessments are performed and
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assessment-identified problems are documented and evaluated for significance; as written, j
however, it did not specify where the programs resided to control that process. The change to l
6.8.2.4 eliminated the reference to the Screening Committee (still in use as a management tool) and substituted, as is customary throughout the Application, reference to the appropriate section of the QAP wherein the program commitments and requirements lie, i.e., QAP 2.16.
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GDP 99-0173 Page 6 of 15 Programmatically, the function of screening identified problems was not eliminated; it was and remains as described in 2.16 of the QAP and the Plant's implementing procedure.
Actual implementation of the corrective action process at PGDP is through procedure CP2-BM-
' CIl031," Corrective Action Program at PGDP," Section 6.0 Section 6 0 contains the necessary action steps to ensure conditions adverse to quality are corrected as soon as practical, and significant conditions have their " identification, cause, and corrective action" documented and reported to appropriate levels of management. The process of" screening" is twofold: an initial screen is performed by a " screening manager," which may be followed by a review by the Plant Shift Superintendent (PSS), and a final screen is performed by the " screening committee."
Specifically, Section 6.2 includes steps the screening manager (usually the initiator's immediate supervisor) takes to determine whether the problem needs immediate review by the PSS for operability concerns or potential reportability, etc. Once this initial screening is completed, further screening by the "screcninr, committee" is accomplished according to C11031, Section 6.5. The screening committee is defined (in CIl031) as a " committee comprised of functional organization representatives tasked with categorizing ATRs relative to their significance and assigning ownership..." Specific organizational responsibilities for the corrective action program are found in Section 5.0 of CIl031.
From the information provided above--specifically that Section 2.16 of the QAP does not 1
commit to a Screening Committee and that the screening process is fully defined in the implementing procedure-it can be stated that the change to SAR 6.8.2.4 did not constitute a
. reduction in effectiveness of the QAP.
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GDP 99-0173 j
Page 7 of l5 Question 4.
RAC 97C0275," Nuclear Criticality Approval (NCSA) Requirements for C-400 Cylinder Wash" The information provided in the RAC does not appear to support the conclusion that the following changes do not reduce the margin of safety (i.e., increase the likelihood of a nuclear criticality): the removal of borated wash solution, increasing the pan depth from 8 inches to 8.5 inches, and increasing the allowed enrichment beyond I wt% assay. Why don't the changes increase the probability or consequences of a nuclear criticality accident?
Response
Question 4 raised a concern by NRC that RAC 97C0275, Nuclear Criticality Approval (NCSA)
Requirements for C-400 Cylinder Wash" did not appear to support the conclusion that the following changes, removal of the borated wash solution, increasing the cylinder wash pan depth from 8 inches to 8.5 inches, and increasing the enrichment from 1.0 wt% 235U to 2.0 wt% 235U does not reduce the margin of safety.
Use of borated wash solution was an obsolete description in SAR Section 3.8.1.3. The borated wash solution was not being used in the cylinder wash operation under the previous NCSE/A 3974-04 Request 1875. Even when it was used as a wash solution prior to NCSA 3974-04 Request 1874, NCS did not credit the borated wash solution as an NCS control for double contingency. Therefore, since borated wash solution was not credited for safety as an NCS l
_ control, the nmin of safety could not be. reduced by this change in the.SAR.
Increasing the cylinder wash pan depth to 8.5 inches and increasing the enrichment of the 235 operation up to 2.0 wt%
U were changes made to the Criticality Accident Analysis Section 4.4.2.4 of the SAR. The USQD performed for the RAC, Safety Evaluation 98-029, Rev. O, determined that the margin of safety was not reduced since the new controls ensure double contingency and the margin of safety is defined by the implementation of the double contingency principle. The basis for this response is found in SAR Section 4.4, Criticality Accident Analysis.
Section 4.4 states that "In practice, a significant margin of safety is ensured by the NCS program which is described in Section 5.2." In Section 5.2.3.2, Process Evaluation and Approval, the NCS program requires that an operation involving uranium enriched to I wt% or higher and 15 j
g or more of 235U be evaluated for NCS prior to initiation. The evaluation is documented in an NCSE. The NCSE identifies and documents potential upset conditions presenting NCS concerns and analyzes each process upset condition to demonstrate double contingency. Since the NCSE.006.00 referenced in the RAC and USQD demonstrated that the change from 8 inches to 8.5 inches in the depth of the cylinder wash pan and the increase from 1.0 wt% to 2.0 wt% 235U j
operatica meet the double contingency principle, the margin of safety as defined by Section 4.4 of the SAR is not reduced.
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GDP 99-0173 Page 8 of 15
- Question 5.
RAC 97C0276,"Removeble Contamination Survey Frequency" What will be the criteria for performing contamination surveys in the lunchrooms on Friday and the following Monday? There is a possibility that contamination may occur on a weekend or holiday and that it would be undetected until after normal usage has resumed on a weekday. If the proper criteria is not used, that possibility could indicate a decrease in the effectiveness of the Radiation Protection Program.
Response
The criteria for performing contamination surveys in lunchrooms will not change. The applicable lunchrooms will be surveyed on Fridays and Mondays in the same manner and to the same criteria each time. The probability of contamination on plant holidays or weekends is less than on normal working days since there are fewer people on plant site and there is less work in i
progress. Also, the Radiation Protection Pr' gram requires anyone exiting an area controlled for loose contamination to survey themselves and their personal items for contamination. The controls in place have proven to be adequate to prevent a spread of contamination to lunchrooms by virtue of the fact that surveys performed on those areas since the implementation of the Program have not detected contamination.
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GDP 99-0173 Page 9 of15
- Question 6.
RAC 98C0052,"C-330 Purge Cascade"
- A. The test allowed by the RAC carilead to accumulation of explosive mixtures at a point of the cascade that would not be possible before the test. The safety evaluation states that the
- probability of an accident is uot increased because of undefined administrative controls. Explain why the probability of an accident is not increased and how the administrative controls are rigorously applied.
. Response
. The purpose of the test was to verify the feasibility of safely purgmg intermediate gasses from the top of the C-310 Tops Purge Cascade at a decreased product withdrawal rate by operation of only one high-speed cell. The test was also intended to validate theoretical work (a computer simulation) that indicated this approach would be successful.
The test of the ability of the C-310 Purge and Product Withdrawal Facility to remove R-114 from the cascade by purging (via high-speed cells), as a supplement to the normal manner by dissolution in the product withdrawal stream, was strictly controlled by procedure CP4-CO-
. CN6079,"C-310 Intermediate Gas Removal Test." This procedure was prepared explicitly for this test purpcsc and contained the following controls: (1) The test was performed under the supervision of a test team leader and Cascade Operations manager. Both personnel were
_ assigned for the test duration to overse-the conduct of the test; and (2) As a test prerequisite.
CIF and F usage was discontinued prior to the test. In addition, C-310 peak concentrations of 3
2 CIF and F were required to be stable at less than 0.2%, far below that required to support a 3
2 reaction with R-114.
.SAR Section 4.3.2.5.6 R-ll4/ Treatment Gas Reaction states "small quantities of R-114 are always present in an operating cascade due to inleakage from the many R-114 coolers used in various pieces of equipment. Larger quantities may be suddenly introduced by equipment failures." Later, the same section states " Strict administrative control is imposed on the amount of F and CIF added to the cascade at any one time." The following procedures contain portions 2
3 of the administrative controls applied: CP4-CO-CN2010, " Operation of the C-310 Product and Side Withdrawal System"; CP4-CO-CN2027, " Handling and Storage of Chlorine Trifluoride, Fluorine, and Mixed Gases in C-350"; and CP4-CO-CN2027a, " Transferring Mixed Gas or F2 from C-350 to the Cascade." Note that although Section 4.3.2.5.6 discusses some actions that
~ can be taken to mitigate the amount of R-114 on the cascade, it does not rely on control of R-114 concentrations (or location) to preymt or mitigate an R-114/ Treatment Gas Reaction. This is because random failures beyond the operators' control can result in R-114 concent ations that, given the presence of F and CIF, could reach explosive concentrations.
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GDP 99-0173 Page 10 of 15 i
Based on purge cascade computer modeling studies and the significant experience of the test development team with R-114 behavior in the C-310 cascade, the R-114 allowed to accumulate in C-310 did not " pocket" in a new location. Rather, the normal R-114 distribution was allowed to increase at all wet,tions where it is routinely present, with the R-114 peak concentration at approximaiety 14, nr.e location (stage) in the purge cascade. The amount and distribution of R-114 in the cascade.for this test was not significantly different from that which periodically occurs when a large R-14 bubble (from a cooler failure or inadvertent rapid return of surge drum
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gas laden with R-114) travels up the cascade to the C-310 stages. It is tb position of the authors and reviewers of RAC 98C0052 that the concentration of R-114 preser.t in C-310 is not a control relied upon for safety; rather, it is the concentration of the oxidants F and ClF relied upon for 2
3 safety. The test did not impact current controls on these oxidants, but provided for more strict control of their use. Therefore, the probability of an accident was not increased during the test.
The results of the test validated the computer simulation results, showing that gas front locations and sizes could be accurately predicted. The test also identified that some modifications to the control systems will be required to reliably control the front locations.
B. The test leads to an increase in emissions of uranium resulting in exceeding the baseline effluent quantity and an increase in offsite doses from the uranium releases. Why is this not an undue risk to the environment or a decrease in the effectiveness of the environmental program or the Radiation Protection Program?
Response
With respect to the effectiveness of the environmental program, the proposed change did not change any aspect of the program desc~ ibed in Section 5.1 of the SAR. The number of high-r speed cells in operation in C-310 is not described in the SAR. None of the components in C-310 used to minimize, control, or monitor releases (alumina traps, sampling systems, etc.), were altered or deleted by the proposed change. Therefore, there was no change in the effectiveness of the environmental program.
With regard to the effect on the environment, there are no regulations on airborne releases of radionuclides with respect to environmental protection only. All regulatory limits are base upon effects upon human beings. While the, valuation of this plant change includes exposure to radionuclides through environmental patiyays: inhalation, direct radiation from ground deposition, consumption of vegetables, meat ano produce, etc., the effect on the environment itselfis notincluded. In the absence of regulatory standards for radionuclide release specific to environmental protection, this evaluation based conclusions of no significant risk to the environment on the evaluation of releases against existing EPA and NRC standards limiting the dose to members of the public.
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GDP 99-0173 Page11of15 As described in Section 5.1 of the SAR, the BEQ is an indicator of plant emissions and effluents versus historical levels. The BEQs do not represent regulatory limits. They are used as indicators that changes in emissions and effluents have occurred. When BEQs are exceeded, a detemiination as to the cause of the exceedance is required under certain conditions. In the case of the C-310 proposed test, BEQ exceedances were anticipated and, when the test was run, were documented in accordance with site procedures.
The original answer to Question 15.1 of PCR-C-98-0285, which addresses risk to public health and safety, was in part based on the evaluation of the effects of the releases from the proposed tests performed in accordance with SAR Section 5.1.3. This evaluation was based on an estimated increase in emissions from C-310 to 50 gU per day on an annual basis. The BEQ for uranium at the C-310 Purge Vent Stack provided in SAR Section 5.1 is I gU per day. Even though the test was scheduled for only 3 days, the estimated dose was evaluated on an annual basis to detem:ir.e if routine operation in this mode would result in a compliance issue. If so, the operational conccpt would require reevaluation prior to the test. This evaluation compared the estimated hicrease in dose to members of the public against the 10 CFR 20 constraint limit on the dose from airborne emissions and the 40 CFR 61, Subpart H, regulatory limit of 10 mrem / year. The total dose from the entire plant, based on 365 days / year elevated release rate, was projected to remain less than.05 mrem / year (.049 mrem / year versus.0052 mrem / year).
Therefore, this does not result in an undue risk to the environment.
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GDP 99-0173 Page 12 of15 Question 7.
RAC 98C0085, " Van-Stone Flange Spacers" The safety evaluation in the RAC indicates that the use of C-clamps is required when using spacers. In order to determine if the USQD is appropriate, please explain if the C-clamps used in the process would be controlled as AQ items and ifnot, why? In addition, does the use of the spacers affect the safety limit in TSR 2.4.2.37
Response
C-clamps were not introduced with the use of spacers. Prior to RAC 98C085, SAR 3.3.4.5.1 stated that the flanged joints on "B" line piping greater than 30 inches in diameter were reinforced with C-clamps unless the pipe flange was cut down to a 1 inch height and rewelded..
Therefore the C-clamps have always been considered part of the piping components listed in SAR 3.15.2.12. The C-clamps are controlled as AQ under Group ID 3261 in the Boundary
' Definition Manual Report for Buildings C-331, C-333, C-335, and C-337. As addressed by Jtem 7 in PCR-C-98-0632, Rev. O, and in the answer to Question 7 of the USQ determination in SE-98-044, Rev. O, TSR 2.4.2.3 is not impacted by this change.
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GDP 99-0173 Page 13 of 15 Question 8.
RAC 98C0099," Isolation of Meletrons" This RAC concerns, in part, the removal of meletrons from service on cells with assays of <1
' wt%. Although not credited for criticality safety below I wt% assay, does this relaxation of criticality controls increase the probability of an accident through an increase in the probability of coolant intrusion or pressure / containment loss on neighboring cells that might have assays
>l wt%?
Response
On systems >l wt%, the meletrons were credited for criticality safety to prevent potential moderation by RCW entering the UF6 system through the R-114 system. Since the time of RAC 98C099, the meletrons were removed from the AQ-NCS list through the appropriate NCSE revision and are not currently credited for criticality safety. Although not currently credited for criticality safety, removal of the meletrons from service would not increase the probability of coolar intrusion from a cell <1 wt% to a cell >l wt% due to the following reasons. In order for RCV oecome a moderator it must: (1) have a pathway to a system >l wt%; (2) not react with UF6;. ;d (3) have a driving force for transport into a potentially fissile environment. If any of i
these three elements are missing, moderation by RCW in this scenario is not feasible. In a shutdown cell, a driving force and pathway into cells >l wt% are missing. In a running cell, the RCW must travel through the barrier to reach cells >l wt% without reacting with UF6. The available RCW will react with UF6 before traveling through the cascade to a point >l wt%.
Travel of RCW through the UF6 system of on-stream cells is not feasible. Therefore, removal
.a of the meletrons from service will not increase the probability of this type accident as analyz;d
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per the SAR.
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GDP 99-0173 Page 14 of 15
' Question 9.
RAC 98C0100,"Use of Auxiliary Battery Chargers" l
' Why is there no increase in the probability of occurrence of a rnalfunction of equipment important to safety when a temporary charger will be parallel with the installed charger? The RAC does not address isolation and circuit protection along with fuse sizing and breaker coordination. Therefore, a system interaction leading to an increased probability of a malfunction cannot be discounted.
1 Response' The auxiliary battery charger does not share overcurrent protection, input or output, with the existing main battery charger. The auxiliary charger connects to a separate supply circuit and its output connects directly to the battery terminals, so additional loading caused by the auxiliary
. charger is not subjected to the main battery charger overcurrent devices. Therefore the statement provided in the USQD, Question 4, (May the condition increase the consequences of a malfunction of equipment important to safety?) "The main battery charger and all other DC systems remain connected and the addition of auxiliary battery chargers is supplementary and presents no increase in malfunction of the DC system" remains valid.
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i GDP 99-0173 Page 15 of 15 i
Question 10. RAC 98C0126,"Reinove < 1wt% Restriction for Meletron Outages" The statement that the NCSE was revised to remove reliance on the meletrons for cells with >
lwt% assay does not automatically mean that their removal does not increase the likelihood of nuclear criticality, or lead to a USQ. The meletrons were previously credited in the SAR for I
criticality safety. What effect does the removal of this equipment have on the probability of nuclear criticality and what is the relative reliability of the original compared to the revised criticality control systems?
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Response
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The basis for the response in the Safety Evaluation, SE98-095, Rev. O, that the probability is not increased by removal of the meletrons as NCS controls is found in SAR Section 4.4, Criticality Accident Analysis. Section 4.4 states that " occurrence of a criticality accident is theoretically possible in a number of areas at PGDP, but the probability is very low. In practice, a significant margin of safety is ensured by the nuclear criticality safety (NCS) program which is described in Section 5.2."
As described in response to Question 4, regarding RAC 97C0275, the NCS Program, Section l
5.2.2.3, requires that NCSEs demonstrate double contingency for each process upset identified i
for an operation. Section 5.2.2.3 also states that " application of this principle ensures that no single credible event can result in an accidental criticality or that the occurrence of events
. necessary to result in a criticality is not credible." This establishes nyo equivalent methods for ensuring double contingency: no single event leading to a criticality,.r alternately, showing the occurrence of events are not credible. Before the change to remove the meletrons from NCS controls, the NCSE for the cascade, KY/S-243, Rev. 2, relied on the meletrons and several other systems to argue an upset condition is not credible. The change to the NCSE to remove the meletrons changed the argument for the upset scenario from arguing incredibility by relying on 4
multiple systems to make the probability of the event less than 10 to arguing double contingency by relying on two independent systems demonstrating that no single event could lead to a criticality. As described in the SAR, Section 5.2.2.3, and Section 4.4, these are two equivalent methods to ensure double contingency. Since the margin of safety as dermed in Section 4.4 is not reduced, and Section 5.2.2.3 demonstrates that the new argument without the meletrons for double contingency is equivalent to the previous argument used in the NCSE with the meletrons, the probability of occurrence of a criticality as dermed in SAR Sections 4.4 and 5.2.2.3 is not increased.
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GDP 99-0173 Page 1 of 18 1
United States Enrichment Corporation Portsmouth Gaseous Diffusion Plant Docket No. 70-7002 Response to July 23,1999, NRC Request for AdditionalInformation Concerning 1999 Annual Update to the Certification Application (TAC No. L32122)
,e GDP 99-0173 Page 2 of 18 Question 1.
RAC %X-0274, " Delete Storage Container Lots" A.
Have the facility pre-fire plans been revised to indicate the potential presence of combustible contaminated waste in various locations in the waste storage lots to maintain the effectiveness of the fire safety program?
B.
Why does the accident evaluation having the potential release from a fire in a storage lot bounded by a single container adequately address the increase in consequences question in the USQD?
Response
A.
No, because the amount of combustible material and the potential for a fire are very low. The combustible materials are in metal containers and any fire would typically be isolated to a single container. The Fire Protection Program and Waste Management Program provide adequate controls to prevent and control fires in such areas. The Safety Evaluation was reviewed and approved by the Authority Having Jurisdiction (AHJ), see SAR Section 5.4 for a description of the AHJ qualifications and ' responsibilities, which provides assurance that any potential impact on the fire protection program was considered.
B.
The accident evaluation does not postulate a fire involving multiple containers
- because there is no significant combustible loading in the yards and the fire potential is very low. As noted above, the combustible materials are stored in metal containers. This is the basis for the SAR discussion and is discussed in the last l
paragraph on page 2 of the Safety Evaluation. In addition, there is very little radioactive material in these containers. The material is waste with only residual low-level contamination.
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GDP 99-0173
- Page 3 of l8 Question 2.
RAC 97-X-0119, " Heavy Metal Filtrate" A.
Please provide bases for why. modifying the maximum Bio-D tank uranium concentration from 0.25 ppm to 5.0 ppm and maximum inventory in solution from 13.65 grams to 230 grams uranium for the 12,000-gallon tank and 2.0 grams to 20 grams U-235 for the 1,000 gallon tank does not reduce' the effectiveness of the plant's safety program by increasing the risk of a criticality event and not being consistent with the As Low' As Reasonably Achievable (ALARA) principle as
-applied to environmental effluents.
B.
'Please provide bases for why the increased potential for buildup of uranium-bearing precipitates does not increase the risk of a criticality event. Please identify the NCSA section/s that address precipitation of uranium out of solution and stratification in tanks and piping in the Bio-D Facilities and the Technetium Ion Exchange Process.
Please provide bases for why increasing the pH levels (from 7.0 and 7.2 to a' level
- that could be as hi,gh as 9.0) would not increase the likelihood of precipitating out uranium thus increasing the probability of a critical excursion thereby reducing the effectiveness of the plant's safety program.
Response
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-A.
-In both cases, a' criticality event is not possible due to low uranium mass and 1
concentrations in the tanks. Since a criticality event is not possible for either condition, there can be no increased risk. In addition, the revised SAR limits were based on revised NCSA/E analyses, which continued to show double contingency with significant margin.~ The discharges to the plant sewage system are controlled
~ to assure that the effluent from the sewage treatment plant will be less than the j
applicable concentration guides in 10CFR20, Appendix B. The sewage plant effluents are maintained less than the BEQs identified in SAR Section 5.1 and provide no significant dose (<0.01 mrem /yr) to the public. ALARA considerations are fully met.
- B.
- A't the low uranium concentrations involved in the Bio-D process, there is no potential for precipitation of uranium from a pH as high as 9. Most of the uranium is either discharged within permit limits to the sewage plant or removed when the biomass sloughs off the coal substrate. As such, there is no potential for significant buildup in the Bio-D system. The SAR Section 4.3.4.1.1 addresses the fact that the amount of uranium in the 12,000 gallon tank is less than 230 grams. Since the PORTS site is limited to production of 10% enrichment, this would typically be less than 20 grams U-235. However, the NCSA/E includes requirements for sampling I
GDP 99-0173 Page 4 of 18 Question 2 Response (continued) and/or NDA of certain sections of the process equipment to verify that no buildup is occurring. As noted in the RAC, this commitment was added to the SAR. As a result, there is no increase in the potential for buildup of uranium in the system. In fact, the addition of the periodic sampling /NDA monitoring provides additional
. assurance that no significant buildup would occur.
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GDP 99-0173 Page 5 of18 l
Question 3.
RAC 97-X-0218," Operation of the Lube Oil and Hydraulic Oil Systems" The change in the range of the lube oil differential pressure trip point to include a range up to a 3 psid would indicate that the response to a loss oflube oil would take longer to initiate.
Does this increase in time to trip a compressor motor cause any increase in consequences from an accident or an increase in probability of an accident?
Response
No, the change does not result in any increase in consequences from an accident because it does not initiate an accident, in and ofitself. As discussed in the USQD, page 4, the change does not affect the timeliness of the lube oil trip with respect to its mitigative function; the lube oil trip will still occur, with the associated cell trip, long before needed to allow for draining of the lube oil supply tank and reducing the cell pressures to below annospheric pressure. The change does not introduce any additional ignition sources or change the flash point of the oil. Therefore, there is no increase in the probability of a fire-caused release of UF. Additionally, as discussed in the USQD, the 2 psid was a nominal value with the SAR 6
range stated in Chapter 3 remaining unchanged. The trip point range takes into account the potential for instrument and set point tolerances. The change to the accident scenario
. discussion was to clarify some confusion regarding the range stated in Chapter 3 versus the number stated in Chapter 4. The credited action for containing a fire is the operation of the sprinkler system, which will prevent any consequences. In addition, the system in question is not identified as safety significant (i.e. it is not a Q or AQ system). The system's primary function is to protect the process equipment compressor bearings and compressors from damage in the event of a lube oil supply loss for any reason.
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C GDP 99-0173 Page 6 of18 Question'4.
RAC 97-X-0236, " Waste Stream for the X-705" A.
' The existing SAR does not discuss a 200 g U-235 sampling limit associated with the mix / feed tanks. In addition, the existing SAR does not include processing of drums or portable containers through the Oil and Grease Removal unit. Justify how the inclusion of the sampling limit and the processing ofdrums or portable containers do not raise a USQ through the possibility of an accident of a different type or an increased risk of a criticality accident.
B.
The USQD for SAR changes states that unknown quantities of non-listed organic contamination are being processed in the X-705 facility. The organics may lead to
- some moderation. Compare these v.bstances with the analysis basis from the NCSE.
Why does the presence of such materials not lead to an increase probability of an NCS accident (USQ)?
, Response.
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- A.
The existing SAR discussed the fact that uranium concentration and enrichment was sampled and measured for each bach of material processed, but it provided no limit.
The NCSA has had controls for the amount of U-235 processed through the mix / feed E
-tanks to Heavy Metals Precipitation. As such, there was no change to the way that solutions were sampled, batched and processed in the Heavy Metals Precipitation operation. The 200 gram U-235 batch limit has significant conservatism. The processing of drums and portable containers through the Oil and Grease Removal unit is subject to the same sampling and transfer restricti_ons~as all solution processed through the OGRU; therefore, there is no accident of a different type or an increased risk of a criticality accident. SAR Section 3.3.1.3.1.5 described the fact that field decontamination solutions could be fed into the OGRU system.
B.
The analysis basis in the NCSE assumes all materials are water solutions; the materials being processed are aqueous solutions with some organic constituents (e.g.
j urine, citric acid, etc.). The small quantities of organic materials processed would not contribute materially to any higher moderation; PORTS analyses have shown that only uranium / oil mixtures need special consideration from a moderation control perspective.
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GDP 99-0173 Page 7 of 18 y
Question 5.
RAC 97-X-0381, " Underground Storage Tanks" A.
.Have the facility pre-fire plans been revised to indicate the presence of permanent above' ground diesel storage tanks (AGST) at X-710 and X-1007 and a temporary
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AGST at X-1020 to maintain the effectiveness of the fire safety program?
B.
The USQD in the RAC did not include an accident evaluation taking into account the proximity of the tanks to Q and AQ systems. How does the RAo not increase the probability or possibility of the malfanction of equipment important to safety?
Response
A.
The storage tanks in this RAC install spill controls required by the OEPA.' The only change from a fire protection standpoint is that two tanks that were underground are replaced with two tanks above ground (X-710 and X-1007). The new tanks are doub!c-walled and have leak detection devices. In addition, the tanks were installed
. in accordance with current Fire Codes. As such, the conclusion of the USQD was that there was no greater likelihood of fire than was present with the previously existing tanks. The two tanks r~elocated above ground are'small tanks (120 and 265 gallons). The tank serving the X-710 is included in the X-710 pre-fire plan. The
~ other permanent tank services an administrative building (X-1007) which does not store or process uranium or other NRC-regulated materials. The temporary AGST i
installed for the X-1020 services a building that does not contain either uranium or other NRC-regulated materials. The temporary tank was installed with diking and other fire protection requirements during its use. Since this installation was temporary until the permanent tank modifications were completed, it was not included in the facility pre-fire plans. (It has since been taken out of service).
Therefore, the effectiveness of the fire protection program in preventing a nuclear accident is not diminished. Both facilities are equipped with sprinklers and covered by the Fire Protection program oversight of administrative buildings.
B.
.The tanks in question have no proximity to Q or AQ systems that une of any concern based on the small volume of the tanks; this is evaluated in the USQD. The USQD, y
page 3, paragraph 4, explains that no other equipment important to safety is located near the proposed sites of the temporary or permanent above ground storage tanks.
In addition, the tanks are double walled and have leak detection instrumentation so that a tank rupture and fire are extremely unlikely.
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.A GDP 99-0173 Page 8 of18
. Question 6.
RAC 97-X-0489," Load Alarm Setpoints" How.does increasing the load alarm setpoint not.cause an increase in probability or consequence of an accident considering the increased response time for an operator to respond to a compressor failure causing the load alarm?
Response
As stated in the USQD, the load alarm setpoint does not increase the probability of an accident since it is not an accident initiator or directly related to an accident initiater. As noted, the load alarms simply monitor the condition of the plant motors. As stated in the USQD, for the bounding scenarios, the SAR does not credit operator response to load alarms with preventing or mitigating an accident. In the case of a compressor failure, the load alarm will actuate upon the compressor failure (e.g. deblade) and the change in setpoint for the stripping or purge sections does not effectively change the operator response time. However, this is primarily an operational issue.
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GDP 99-0173 Page 9 of I8 4
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~ Question 7.'_
RAC 97-X-0492, "DC Control Power Description"-
The RAC does not discuss isolation and separation between load groups of adjacent process units, fuse sizing, and breaker coordination. How are these issues addressed such that the
. temporary jumpering of one bcttery's loads to an adjacent battery does not increase the -
probabihty or increase the consequences of a new or previously analyzed accident?
Response
The load controls of the battery (bank) were evaluated in the EnF neering Evaluation i
referenced in the USQD. The conclusion of the evaluation was that the load protection was sized adequately to handle the potential load for two battery (banks). Thus, there was no change in the load protection features with the jumpering in place. Since there were no changes required to the existing protective controls, there was no discussion of this issue in the USQD.
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GDP 99-0173 Page 10 of 18 l.
Question 8.
RAC 97-X-0533,' Removing the Benedicts Equation from NCSA" i
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' Yourjustification in Question 1 of the USOD for this RAC states "With implementation of the revised NCSAs, postulated criticality events are no more likely than those previously evaluated in the SAR." This statement is not necessarhy true. Demonstrate explicitly that the removal of using the Benedict Equation to ensure that the UF remains in a gaseous state 6
does not increase the probability of nuclear criticality.
Response
As noted in the USQD, the SAR still retains use of the Benedict equa: ion in conjunction with control of cascade temperatures and pressures to maintain UF in a gaseous state. The 6
gaseous diffusion process cannot function without maintaining UF in a gaseous state. The 6
TSRs fcr Moderation Control acknowledge in the Basis Statements that deposits ofsolid UF.
"are an expected result of normal operations" (e.g. due to not remaining within the Benedict equation parameters). However, the Basis Statements go on to explain that unmoderated deposits of 7% enrichment or less cannot achieve criticality and that an unmoderated deposit l
with assays between 7% and 20% would require a mass to achieve criticality that would I
exceed the amount of UF available in the portions of the cascade operating at such assays.
6 The conclusion of the TSR Basis Statements is that "for a freeze-out condition in the event l
. of a total loss of power to the cascade, criticality wodd not result." The only SAR change
[
was to Section 3.1.2.6, which addressed the 250 EBRolers.Tnere is no criticality concern for these coolers at below 20% enrichment, which is greater than the maximum permitted production enrichment of 10%. As a result, the use of the Benedict equation as an NCS control will not reduce the probability of a criticality; the removal of the Benedict equation l-as an NCS control will not increase the probability of a criticality. We recognize that i
including the above information would have strengthened the USQD.
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l GDP 99-0173 Page 11 of 18 Question 9'.
. RAC 97-X-0555, " Heavy Metal Precipitation" A.
Drawing number X-705-1606M, rev. 3,'which is intended to replace SAR figure 3.3.1.4-7a, contains a criticality warning note. What is the basis for this warning?
The' discussion paragraph of the USQD stated that the accident analysis is based on the total capacity of the Heavy Metals Precipitation Raflinate Storage (HMPRS) columns. In making the USQD, has the possibility of a criticality been explicitly
~
considered in both the caustic storage tank (V-200) and the mix / feed tanks (V-2 and V-3) or has the possibility of backflow from the mix / feed tanks to the caustic storage tank been considered as a potential criticality initiator?
3 B.
. Page 4.3-9 of the SAR, rev. 31, states that the raffinate in a mix / feed tank contains no more than 0.185g U-235/L and no more than 323 grams U-235 total. Given a 200 gallon tank volume, describe how 323 grams translates into 0.185 g U-235/L. Justify why there is no USQ with the increase in tank volume and corresponding increase in probability of an NCS accident given that field generated solutions may be directed into the mix / feed tanks.
Response
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., AH -The basis forthe "waming note" on the drawing is that normally at PORTS dikes are installed around all caustic tanks to limit spills to a relatively small area. In this case, a dike that would contain the tank contents cannot be installed (and was not installed)
I due to the potential for inadvertent collection of fissile material solutions. With respect to a potential criticality in the caustic storage tanks, a criticality has been
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considered as a potential and backflow prevention design with administrative controls in the NCSA have been in place to deal with this contingency. These controls were not changed in this RAC. The NCSA and NCSE are not reproduced in their entirety in the SAR.
' B.
As noted in your "A" question, the size of each batch treated by Heavy Metal Precipitation is governed by the size of the HMPRS columns. The 323 grams U-235 is based on the 0.185 g U-235/L in the 1748 L (~462 gallons) that could be contained in the HMPRS columns. As a result, the actual amount of U-235 that could be present in a 200 gallons mix / feed tank would be less (considerably less) than 323 grams U-235. It should be noted that the 323 grams is based on processing 100%
enriched uranium which is no longer permitted under the NRC Certificate; therefore, the margin is actually greater than indicated by the SAR analysis. Any field generated solutions must be sampled prior to addition to this system to meet the l
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Page 12 of 18 Question 9 Response (continued) j.
uranium mass requirements (as noted in your question 4, there is a 200 g U-235 limit on the l-feed / mix tanks). As a result, there is no increase in probability of an NCS accident.
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GDP 99-0173 4
Page 13 of18 I
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Question 10. RAC 97-X-0568,"NCS Revisions for Coolant Systems" A.
The USQD for this RAC contains the statement: "The Recirculating Cooling Water
.. (RCW) control of supply valves are administratively controlled and no credit is taken for any automatic actt:ation, therefore these items are not considered to be equipment important to safety." Are these valves providing an NCS ftknction and if so, are they
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identified as important to nuclear criticality safety so that there is no increased probability.of failure (i.e., a USQ)?
B.
The statement that there is no change in probability of criticality because double contingency is~mainteined is not necessarily correct. Justify the conclusion that the probability of occurrence of criticality is not increased by the removal of the requirement to maintain R-114 coolant pressure above RCW and process gas pressure, given that the R-114 pressure is one of several items credited in the double contingency discussion of the three accident scenarios in this RAC.
C.
! Justify the assertion that removal of the reliance on maintaining R-114 coolant pressure above RCW and process gas pressure does not increase the consequence of j
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- - -the malfunction ofitems relied on for safety, or lead to new accident sequences.
l State how the safety controls can be revised without introducing different types of malfunctions or the consequences of their occurrence.
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Response
A.
The safety action is to shut the RCW supply valve when the cell motors are ofTor the UF inventory is evacuated. When the valves are closed, there can be no pressure in 6
the condenser greater than the RCW pressure; any valve leakage will not be capable afpressuring the condenser whh RCW pressure greater than the vapor pressure of the coolant (R-114). Therefore, the' valves are not credited with providing a leak-tight
- functim and with the retum valves open cannot result in an RCW pressure exceeding the R-114 pressure. The NCSA requires the presence of R-114 in the coolant system or the condenser must be drained of RCW and remain drained. As such, there is no function of the RCW supply valve credited for NCS other than its normal process
~ function.-
B.
It appears from this question that a misreading of this RAC change has occurred.
)
This RAC change is to clearly make the SAR agree with the controls required in the
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l GDP 99-0173 Page 14 of 18 Question 10 Response (continued) l TSRs on moderation control and provide for the control of pressure differential between the RCW and coolant by more readily demonstrable means. The moderation control TSR identifies that the R-114 pressure must be maintained above the RCW pressure (for cells with a deposit greater than a safe mass) or the RCW condenser must be drained. The RAC simply identifies in the SAR and TSR Basis Statement the action already identified in the TSR: the option of draining the RCW from the condenser if the coolant pressure cannot be maintained above the RCW pressure. The requirement has not been removed; the NCSA provides more detailed controls that explain how this condition is maintained.
C.
The basis control of maintaining the coolant pressure above the RCW pressure has not been removed. The RAC simply adds tSe TSR LCO of draining the RCW from the condenser as an action that provides equivalent protection when the R-114 must be drained from the coolant system. In addition, the NCSA simply describes in more detail the controls assuring that the coolant pressure is higher than the RCW pmssure.
The system design and operation have not changed; therefore, there are no new accident sequences, no different types of malfunctions of safety equipment and no new accident consequences.
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c GDP 99-0173 Page 15 of18 Question 11. RAC 98-X-0038,"Calciner Accident Sequence Case C-25d" A'
The safety evaluation supporting this RAC states that the maximum oxide production rate is 24 kg per 8-hour shift. This is based on a feed rate between 7 and 7.6 liters per hour (as documented by the supporting analysis). Justify why this assumption does not increase the probability of a criticality accident given that the feed rate could be as high as 15 liters / hour (as stated in Part A of the NCSA).
l B.
This RAC proposes to use " frequently" to describe the number of times the oxide receiving can is checked for filling. No technical base:; for checking frequency has been provided. How will " frequently" be implemented in the field to ensure there is' no increase in probability of an accident?
C.
The safety evaluation states that the relatively low oxide production rate is not sufficiently high enough to affect the controlling oxide level setpoints. Because the rate of oxide production is dependent on the feed rate, it is not clear whether there is no increase in probability of an accident at the highest credible feed rates.
Response
A.
The maximum production rate allowed by the SAR, as revised, is 24 kg/8hr shift.
= The actual production rate established by procedural controls is below that amount.
The NCSA/E evaluated an upset condition where the feed pumps ran at maximum capacity (significantly grecter than the SAR limit) and still met double contingency.
I It should be noted that the SAR and NCSA/E assume 100% enriched uranium is being processed; the NRC Certification only allows production of 10% enriched uranium in NRC-regulated space. As such, there is very substantial margin in the SAR analysis and in the NCSA/E.
B.
As noted in the USQD, the basic NCS control is double verification that the receiving can is in place; there is also a TSR Surveillance Requirement for the same verification to be performed. Protection against overfilling of a can is provided by Q SSC controls. The checking frequency provides defense-in-depth and the frequency remains the same at once every 30 minutes. The change from eight times per can to frequently simply allows for different can filling times. No basis is required on number of check times; only the periodicity matters.
C.
The oxide can level probe is set to actuate an alarm at 80% of the can height.
Assuming a minimum time to fill a can of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> based on a 24 kg/8 hr rate. the alarm would provide approximately 24 minutes for an operator to respond if the L,1
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GDP 99-0173 Page 16 of18 Question 11 Response (continued) automatic system did not function to shut off the calciner feed. Also, as noted above, the checks every 30 minutes would also provide defense-in-depth for this issue.
Finally, the L'AR and NCSA/E evaluations assume an enrichment of 100% while the X-705 is currently limited under the NRC Certificate to 10% enrichment of production. As such, there is no increase in the probability of an accident at the highest credible feed rates.
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r GDP 99-0173 Page 17 of 18 Question 12. RAC 98-X-0137," Changes to the Classified Material Protection Plan (CMP)"
Why' was Addendum 3 entitled, " Master Security Plan for Microcomputer Systems Processing Classified," removed without at least referencing the fact that it is available on
. plant site? This would appear to decrease the effectiveness of the CMP.
Response
The " Master ADP Security Plan for Microcomputer Resources Processing Classified Information" is referenced in Section 19.4 of the Classified Matter Plan.
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- GDP 99-0173 Page l8 ofl8
_ Question 13. RAC 98-X-0180 l
The SAR currently describes a limit ofless than or equal to 10 ppm uranium in the alumina i
trap. Why does dropping the limit not increase the probability or consequences of a criticality accident?
Response
l The deletion in the first paragraph of Section 3.1.4.3.5, where the statement, "(i.e. material
['
- having <10 ppm uranium)", was deleted, is not part of RAC 98-X0180. This RAC does not change any concentration limits or any limits associated with the alumina traps. It simply removes the reference to NCSA-0333_016 as a basis for alumina trap loading and assay limits. The SAR has no 10 ppm uranium limit for alumina traps. The previous change that was questioned was not related to material in the alumina traps but rather to the uranium concentration in the gas being trapped.
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GDP 99-0173 j
Page1of1 Commitments Contained.in This Submittal
- 1. A new RAC will be completed to clarify the actual restrictions on returning cell treatment gasses to the cascade for PGDP. This change will be completed by December 15,1999.
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