ML20210D818
ML20210D818 | |
Person / Time | |
---|---|
Site: | Portsmouth Gaseous Diffusion Plant, Paducah Gaseous Diffusion Plant |
Issue date: | 07/23/1999 |
From: | Galloway M NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
To: | Toelle S UNITED STATES ENRICHMENT CORP. (USEC) |
References | |
TAC-L32121, TAC-L32122, NUDOCS 9907280032 | |
Download: ML20210D818 (9) | |
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j NUCLEAR REGULATORY COMMISSION I 2 WASHINGTON, D.C. 20555-0001
\\ *****j July 23, 1999 Mr. Steven A. Toelle Nuclear Regulatory Assurance and Policy Manager U. S. Enrichment Corporation 2 Democracy Center 6903 Rockledge Drive Bethesda, MD 20817
SUBJECT:
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 NOS.
L32121 AND L32122)
Dear Mr. Toelle:
This refers to your letter to Dr. Carl J. Paperiello dated June 14,1999, regarding Title 10 of the Code of the Federal Regulations (CFR) 76.68(a)(3) decreased effectiveness reviews and your transmittals of the 1999 annual updates to the certification documents dated April 14,1999, for the Padumb, Kentucky, Gaseous Diffusion Plant and Portsmouth, Ohio, Gaseous Diffusion Plant.
Your June 14,1999, letter explained your current practice to meet the requirements of the i
10 CFR '76.68(a)(3) decreased effectiveness reviews. Your staff performs unreviewed safety question determinations (USODs) to accomplish decreased effectiveness reviews of changes to safety programs in the Safety Analysis Report (SAR). You also requested Nuclear Regulatory l
Commission (NRC) interpretation as to whether or not your practice meets the requirements.
l Only the Office of the General Counsel (OGC) can interpret NRC regulations, and your request l
will be submitted to the OGC.
Meanwhile, my staff has had an opportunity to evaluate your current practice during our review of the annual updates to the certification applications dated April 14,1999. The annual updates submitted those changes you have evaluated as not requiring prior approval by the NRC as allowed by 10 CFR 76.68. Our review consisted of a review of each Request for Application Change (RAC) for the annual updates. We reviewed approximately 72 RACs for the Paducah annual update and have questions on 11 of those RACs. We reviewed approximately 81 RACs for the Portsmouth annual update and have questions on 13 of those RACs. Those questions are attached as a request for additional information (RAI) and demonstrate my staff's concerns about your current approach to the use of 10 CFR 76.68.
The majority of the questions raise a concern about the USODs performed by your staff. Your USOD process requires your staff to answer questions regarding increased probabilities or consequences of accidents or malfunction of equipment important to safety previously evaluated, or the possibility of creating an accident or malfunction of a different type, or reduction in the margin of safety as defined in the basis for any technical safety requirement from the 9907200032 990723 PDR ADOCK 07007001 C
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l S. A. Toelle, USEC
-2 modifications / changes to your certification applications. Based upon our review, it appears the thresholds your staff has set for stating there is no USO are too high. In addition, the technical i
bases for your determination of no USO is not always apparent.
A second concern does, in fact, question the appropriate use of the USOD to meet the requirement of the 10 CFR 76.68(a)(3) decreased effectiveness reviews. Chapter 5," Nuclear Safety Programs," of the SAR provides safety program descriptions. The USOD process does not review those programs to determine if there is a decrease in the effectiveness of any of those safety programs, nor does it otherwise assess whether there are instances where safety program effectiveness could be decreased despite a determination of no USO. Several of the RAI questions ask if programs such as the Radiation Protection Program and the Fire Protection Program are decreased in effectiveness as a result of the modification / change.
A final concern is captured in several of the Nuclear Criticality Safety Program questions in the RAl. That concem is your USOD responses appear to rely on continuing to meet double contingency as a satisfactory response to no increase in probability of an accident. If the modification / change has either decreased the reliability of one of the contingencies or substituted a less reliable contingency for a previous contingency, there is an increase in probability of an accident, and therefore, the use of 10 CFR 76.68 may not be appropriate.
l To further assist us in determining if you have used 10 CFR 76.68 appropriately, we have identified required additional information. An inappropriate use of 10 CFR 76.68 could be the basis for enforcement action. The additionalinformation, specified in the enclosed request, and the associated RACs should be provided within 60 days of this letter. Please reference the above TAC Nos. in future correspondence related to this request.
If you have questions, please contact Yen-Ju Chen at (301) 415-5615.
Sincerely, i
A 6%
Melanie A. Galloway, Chief Enrichment Section Special Projects Branch Division of Fuel Cycle Safety and Safeguards, NMSS Dockets: 70-7001/70-7002 Certificates: GDP-1/GDP-2
Enclosure:
As stated cc:
R. DeVault, DOE H. Pulley, PGDP J. Brown, PORTS i
i
c Request for Additionalinformation Application Dated April 14,1999 United States Enrichment Corporation Paducah Gaseous Diffusion Plant Docket 70-7001 Please provide the following information:
- 1. Reauest for ADolication Chanae (RAC) 96-C-0120. " Gas Treatment and Intermediate Gas Removal System (IGRS)"
The RAC relaxes monitoring requirements. Previously, the Safety Analysis Report (SAR) required securing oxidants from off stream treatment if only 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 Mcrease 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.
- 2. RAC 96-C-0151. " Autoclave Instrumentation" and RAC 97-C-113. "C-360 Instrument Uoarade" The modifications in these RACs are installing new pressure transmitters and associated equipment in the autoclave and UF, 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 Unreviewed Safety Question Determination (USOD).
- 3. RAC 97-C-0239. " Problem Reportina Screenino Committee"
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The change to SAR Section 6.8.2.4, " Problem Reporting," deletes the screening committee.
This would appear to decrease the effectiveness of the Qua'ity Assurance Program. What committee (or position) is responsible to perform the screening specified, and where is that commitment documented?
- 4. RAC 97-C-0275. " Nuclear Criticality Safety Acoroval (NCSA) Reauirements 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 1wt% assay. Why don't the changes increase the probability or consequences of a nuclear criticality accident?
l ENCLOSURE 1 j
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- 5. RAC 97-C-0276. " Removable Contamination Survev Freauency" r
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.
- 6. RAC 98-C-0052. "C-310 Purae Cascade" A. The test allowed by the RAC can lead to accumulation of explosive m..ures at a po;nt of the cascade that would not be pessible before the test. The safety evaluation states that the probability of an accident is not increased because of undefined administrative controls. Explain w hy the probability of an accident is not increased and how the administrative controls are rigorously applied.
B. The test leads to an increase in emissions of uranium resulting in exceeding the Baseline Effluent Quality 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?
- 7. RAC 98-C-0085. " Van-Stone Flanae 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 USOD is appropriate, please explain if the C-clamps used in the process would be controlled as AO items and if not, why? In addition, does the use of the spacers affect the safety limit in TSR 2.4.2.3?
- 8. RAC 98-C-0099. " 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 1 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 cdis that might have assays >
1 wt%?
- 9. RAC 98-C-0100. "Use of Auxiliary Batterv Charaers" Why is there no increase in the probability of occurrence of a malfunction 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 can not be discounted.
- 10. RAC 98-C-0136. " Remove < 1% Restriction for Meletron Outaaes" The statement that the Nuclear Criticality Safety Evaluation (NCSE) was revised to remove reliance on the meletrons for cells with > 1 wt% assay does not automatically mean that their removal does not increase the likelihood of nuclear criticality, or lead to a USO. The meletrons were previously credited in the SAR for 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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l Request for Additional Information Application Dated April 14,1999 United States Enrichment Corporation Portsmouth Gaseous Diffusion Plant j
Docket 70-7002 Please provide the following information:
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- 1. BAC 96-X-0274. " Delete Storace Container Lots" A
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 I
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 USOD?
- 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 be;ng 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 lon Exchange Process. Please provide bases for why increasing the pH levels (from 7.0 and 7.2 to a level that could be as high 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.
- 3. RAC 97-X-0218. "ODeraten 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 Juld indicate that the response to a loss of lube oil would take longer to initiate. Does this increase in time to trip a compressor motor cause any increase in con: equences from an accident or an increase in probability of en accident?
- 4. RAC 9/-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 of drums or portable containers do not raise a USO through the possibility of an accident of a different type or an increased risk of a criticality l
accident.
1 B. The USOD for SAR changes states that unknown quantities of non-listed organic l
contamination are being processed in the X-705 facility. The organics may lead to some 1
r 2
I moderation. Compare these substances with the analysis basis from the NCSE. Why does the presence of such materials not lead to an increase probability of an NCS accident (USO)?
- 5. RAC 97-X-0381. "Underaround Storaoe Tanks" A. Have the facility pre-fire plans been revised to indicate the presence of permanent above ground diesel storage tanks (AC..T) at X-710 and X-1007 and a temporary AGST at X-1020 to maintain the effectiveness of the fire safety program?
B. The USOD in the RAC did not include an accident evaluation taking into account the
. proximity of the tanks to O and AO systems. How does the RAC not increase the probability or possibility of the malfunction of equipment important to safety?
- 6. RAC 97-X-0489. " Load Alarm Setooints" 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 oporator to respond io a compressor failure causing the load alarm?
- 7. RAC 97-X-0492. "DC Control Power Descriotion" 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 battery's loads to an adjacent battery does not increase the probability or increase the consequences of a new or previously analyzed accident?
- 8. RAC 97-X-0533. "Removino the Benedicts Eauation from NCSA" Your justification in Question 1 of the USOD for this PAC 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 necessarily true. Demonstrate explicitly that the removal of using the Benedict Equation to ensure that the UF, remains in a gaseous state does not increase the probability of nuclear criticality.
- 9. RAC 97-X-0555. "Heaw 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 USOD stated that the accident analysis is based on the total capacity of the Heavy Metals Precipitation Raffinate Storage (HMPRS) columns. In making the USOD, has the possibility of a criticality bec i 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?
B. Page 4.3-9 of the SAR, rev. 31, states that the raffinate in a mix / feed tank contains no more than 0.185 g 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 USO 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.
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- 10. RAC 97-X-0568. "NCS Revisions for Coolant System" A. The USOD 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 actuation, therefore these items are not considered to be equipment important to safety." Are these valves providing an NCS function and if so, are they identified as important to nuclear criticality safety so that there is no increased probability of failure ( i.e.,
a USO)?
B. The statement that there is no change in probability of criticality because double contingency is maintained 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 severalitems credited in the double continency 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 the malfunction of items relied on for safety, or lead to new accident sequences. State how the safety controls can be revised without introducing different types of malfunctions or the consequences of their occurrence.
- 11. RAC 98-X-0038. "Calciner Accident Seauence 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).
B. This RAC proposes to use " frequently" to describe the number of times the oxide receiving can is checked for filling. No technical bases 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 saf. ' evaluation states that the relatively low oxide production rate is not sufficiently high eragh 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.
I
- 12. RAC 98-X-0137. *Chanaes 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.
- 13. RAC 98-X-0180 The SAR currently describes a limit of less than or equal to 10 ppm uranium in the alumina trap.
Why does dropping the limit not increase the probability or consequences of a criticality accident?
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[USEC A Glot:al Energy Company June 14,1999
)
GDP 99-0099 i
Dr. Carl J. Paperiello Director, OfHee 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 & 70-7002 10 CFR 76.68(a)(3) Decreased Effectiveness Revicws
Dear Dr. Paperiello:
The purpose of this letter is to request Nuclear Regulatory Commission (NRC) interpretation in accordance with 10 CFR 76.6 of 10 CFR 76.68(a)(3) to con 6rm that the United States Enrichment Corporation's (USEC) current practice of performing an unreviewed safety question determination (USQD) review to accomplish decreased effectiveness review of changes to safety programs in the Safety Analysis Report (SAR) meets requirements. As denned in SAR Section 6.3.2, the SAR is i
comprised of Volumes 1 and 2 of the Application for United States Nuclear Regulatory Commission Certi0 cation (the Application) for both PGDP and PORTS.
The NRC requirements contained in 10 CFR 76.68 require USEC te review all changes to the plant or plant operations to determine whether the changes require prior NRC approval. These requirements, in part, state:
(a) The Corporation may make changes to the plant or to the plant's operations as described in the safety analysis report without prior Commission approval provided all the provisions of this section are met... (3) The changes may not decrease effectiveness of the plant's safety, safeguards, and security programs.
SAR Section 6.3 describes how NRC requirements are implemented to control changes to all sections of the Application. USEC developed proccdure UE2-RA-RR1036. "10 CFR 76.68 Plant Change Reviews" to comply with the requirements of 10 CFR 76.68 and to flow down the commitments of SAR Section 6.3. SAR Section 6.3.2 allows changes to be made to the SAR provided all provisions of 10 CFR 76.68 are met. Changes to the SAR are evaluated with a written safety analysis and a USQD review. The USQD determines, in part, whether a proposed change to the SAR involves an increase in the probability or an increase in the consequences of an accident
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I Dr. Carl J.Paperiello GDP 99-0099 June 14,1999, Page 2 previously evaluated in the SAR. USEC believes performance of the USQD for changes to the SAR satisfies the requirements of 10 CFR 76.68(a)(3) by dete mining whether a proposed change to the safety programs described in the SAR involves unreviewed safety questions. Those proposed changes that involve unreviewed safety questions would be sent to the NRC for prior approval.
Those proposed changes that do not involve unreviewed safety questions would not be sent for prior NRC approval.
USEC's approach in using the USQD to determine whether a change would decrease the l
etrectiveness of SAR safety programs appears to be consistent with previous NRC staff positions on handling changes at fuel cycle facilities. Attachment I to SECY-96-079, dated April 16,1996, equates no reduction in effectiveness of safety programs to no increase in risk and no unreviewed safety issues (see Comment 9 in Attachment I to SECY 96-079). The NRC further states "Ifit could be demonstrated (through an ISA) that there were no previously unreviewed safety issues and no increased risk, then the changes would be allowed with subsequent notification to the NRC." USEC determines by performing a USQD review for changes to the SAR whether the change involves an increase in the probability or an increase in the consequences of an accident previously evaluated in the SAR, and, hence, whether risk has been increased or an unreviewed safety question has been created.
Volume 3 of the Application contains some safety, safeguards, and security programs. These programs do not fall under the requirements of 10 CFR 76.68. Changes to these programs are reviewed for decreased ef fectiveness in accordance with SAR Section 6.3.3.1.
USEC believes that the USQD satisties the requirement for a decreased effectiveness review of plant safety programs contained in the SAR. This approach appears to be consistent with past NRC staff positions on this issue. USEC requests that NRC confirm that USEC's current practice meets the requirements of 10 CFR 76.68(a)(3).
Ifyou have any questions regarding this matter, please contact Mark Lombard at (301) 564-3248.
There are no new commitments contained in this submittal.
Sincer,ely, n
5.A.
I Steven A.Toelle Nuclear Regulatory Assurance and Policy Manager cc:
Mr. Robert C. Pierson, NRC liq P.L Hiland - NRC Region III Office NRC Resident Inspector - PGDP NRC Resident inspector - PORTS
" S.' A. Toelb, USEC
- July 2p,1999 niodifications/chingas to your cartification cpplications. Bas:d upon our rsvisw, it appears the thresholds your staff has set for stating there is no USO are too high. In addition, the technical bases for your determination of no USO is not always apparent.
A second concern does, in fact, question the appropriate use of the USOD to meet the -
requirement of the 10 CFR 76.68(a)(3) decreased effectiveness reviews. Chapter 5,' Nuclear Safety Programs, of the SAR provides safety program descriptions. The USOD process does
~
not review those programs to determine if there is a decrease in the effectiveness of any of those safety programs, nor does it otherwise assess whether there are instances where safety
' program effectiveness could be decreased despite a determination of no USQ. Several of the
- RAI questions ask if programs such as the Radiation Protection Program and the Fire Protection
- Program are decreased in effectiveness as a result of the modification / change.
J A final concern is captured in several of the Nuclear Criticality Safety Program questions in the RAl. That concern is your USOD responses appear to rely on continuing to meet double
)
- contingency as a satisfactory response to no increase in probability of an accident. If the
)
modification / change has either decreased the reliability of one of the contingencies or substituted a less reliable contingency for a previous contingency, there is an increase in probability of an accident, and therefore, the use of 10 CFR 76.68 may not be appropriate.
To further assist us in determining if you have used 10 CFR 76.68 appropriately, we have identified required additional information. An inappropriate use of 10 CFR 76.68 could be the basis for enforcement action. The additional information, specified in the enclosed request, and the associated RACs should be provided within 60 days of this letter. Please reference the above TAC Nos. in future correspondence related to this request.
If you have questions, please contact Yen-Ju Chen at (301) 415-5615.
Sincerely, Melanie A. Galloway, Chief Enrichment Section Special Projects Branch Division of Fuel Cycle Safety and Safeguards, NMSS Dockets:. 70-7001/70-7002 Certificates: GDP-1/GDP-2 h
Enclosure:
As stated cc:
R. DeVault, DOE j
- H. Pulley, PGDP
/
J. Brown, PORTS DISTRIBUTION: (Control Nos. 590S.510S)(TAC Nos. L32121,L32,122)
- Dockots 70-7001/70-7002
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- 10. RAC 97-[ 0568. *NCS Revisions for Coolant System" l
A. The USOD 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 actuation, therefore these items are not considered to be equipment important to safety." Are these valves providing an NCS function and if so, are they identified as important to nuclear criticality safety so that there is no increased probability of failure (l.e., a USO)?
B. The statement that them is no change in probability of criticality because double contingency is r intained is not necessarily correct. Justify the conclusion that tha probability of occurrence of criti ity is not increased by the removal of the requirement to maintain R-114 coolant pressure above W and process gas pressure, given that the R 114 pressure is one of severalitems credited in e double continency discussion of the three accident scenarios in this R AC.
C. Justify the asse that removal of the reliance on maintaining R 114 coolant pressure above RCW and process s pressure does not increase the consequence of the malfunction of items relied on for safety, or d to new accident sequences. State how the safety controls can be revised without introduci different types of malfunctions or tne consequences of their occurrence.
- 11. RAC 98-X-0038. "Calciner Accid t Seouence Casr. C-25d*
A. The safety evaluation supporting t RAC states that the maximum oxide production rate is 24 kg per 8-hour shift. This is based on a d rate between 7 and 7.6 liters per hour (as documented by the supporting analysis). Justify why t assumption does not increase the probability of a criticality accident given that the feed rat ould be as high as 15 liters / hour (as stated in Part A of the NCSA).
B. This RAC proposes to use " frequently" to desc e the number of times the oxide receiving can is checkad for filling. No technical bases for checki a frequency has been provided. How will
" frequently" be impk'mented in the field to ensure t re is ne increase in probability of an accident?
C. The safety evaluation states that the relatively low on production rate is not sufficiently high enough to affect the controlling oxide level setpoirits Be use the rate of oxide production is dependent on the feed rate, it is not clear whether there is increase in probability of an accident at the highest credible feed rates.
1
- 12. RAC 98-X-0137. "Chsaaes to the Classified Material Protection Pla CMP)"
Why was Addendum 3 entitled, " Master Security Plan For Microcomputer S ems Processing Classified,"
removed without at least referencing the fact that it is available on plant site?
is would appear to decrease the effectiveness of the CMP.
- 13. RAC 98-X-0180 The SAR currently describes a limit of less than or equal to 10 ppm uranium in the alumin trap. Why does dropping the liniit not increase the probability or consequences of a criticality accident l
l DISTRIBUTION: (Control Nos. 590s,510S)(TAC Nos. L32121,L32122)
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