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Category:TEXT-SAFETY REPORT
MONTHYEARML20216G0111999-09-30030 September 1999
Year 2000 Readiness in U.S. Nuclear Power Plants
ML20206N2191999-04-30030 April 1999
Operator Licensing Examination Standards for Power Reactors
ML20205A5291999-03-31031 March 1999
Licensee Contractor and Vendor Inspection Status Report. Quarterly Report,October-December 1998.(White Book)
ML20211K2851999-03-31031 March 1999
Standard Review Plan on Power Reactor Licensee Financial Qualifications and Decommissioning Funding Assurance
ML20205A5991999-03-31031 March 1999
Licensee Contractor and Vendor Inspection Status Report. Quarterly Report,July-September 1998.(White Book)
ML17313A7791999-02-0505 February 1999
Safety Evaluation Accepting Licensee Rev to Emergency Plan That Would Result in Two Less Radiation Protection Positions Immediatelu Available During Emergencies
ML20203D0541999-01-31031 January 1999
Fire Barrier Penetration Seals in Nuclear Power Plants
ML20155A9281998-10-31031 October 1998
Licensee Contractor and Vendor Inspection Status Report. Quarterly Report,April-June 1998.(White Book)
ML20154C2081998-09-30030 September 1998
Licensee Contractor and Vendor Inspection Status Report. Quarterly Report,January-March 1998.(White Book)
ML15261A4681998-09-0404 September 1998
Safety Evaluation Supporting Amends 232,232 & 231 to Licenses DPR-38,DPR-47 & DPR-55,respectively
ML20203A1521998-07-31031 July 1998
Assessment of the Use of Potassium Iodide (Ki) as a Public Protective Action During Severe Reactor Accidents.Draft Report for Comment
ML20153D3371998-07-31031 July 1998
Assessment of the Use of Potassium Iodide (Ki) as a Public Protective Action During Severe Reactor Accidents.Draft Report for Comment
ML20236S9771998-06-30030 June 1998
Knowledge and Abilities Catalog for Nuclear Power Plant Operators.Pressurized Water Reactors
ML20236S9681998-06-30030 June 1998
Evaluation of AP600 Containment THERMAL-HYDRAULIC Performance
ML20236S9591998-06-30030 June 1998
Knowledge and Abilities Catalog for Nuclear Power Plant Operators.Boiling Water Reactors
ML20217Q7971998-05-0404 May 1998
Safety Evaluation Supporting Amends 227 & 201 to Licenses DPR-53 & DPR-69,respectively
ML20247E3951998-04-30030 April 1998
Licensee Contractor and Vendor Inspection Status Report. Quarterly Report,October-December 1997.(White Book)
ML20217F3801998-03-31031 March 1998
Risk Assessment of Severe ACCIDENT-INDUCED Steam Generator Tube Rupture
ML20202J3051997-11-30030 November 1997
Licensee Contractor and Vendor Inspection Status Report. Quarterly Report,July-September 1997.(White Book)
ML20197B0431997-11-30030 November 1997
Licensee Contractor and Vendor Inspection Status Report. Quarterly Report,April-June 1997.(White Book)
ML20211L2931997-09-30030 September 1997
Aging Management of Nuclear Power Plant Containments for License Renewal
ML20210K7801997-08-31031 August 1997
Topical Report Review Status
ML20149G9431997-07-31031 July 1997
Licensee Contractor and Vendor Inspection Status Report. Quarterly Report,January-March 1997.(White Book)
ML20210R2131997-05-31031 May 1997
Final Safety Evaluation Report Related to the Certification of the System 80+ Design.Docket No. 52-002.(Asea Brown Boveri-Combustion Engineering)
ML20140F0801997-05-31031 May 1997
Final Safety Evaluation Report Related to the Certification of the Advanced Boiling Water Reactor Design.Supplement No. 1.Docket No. 52-001.(General Electric Nuclear Energy)
ML20140J4301997-05-31031 May 1997
Safety Evaluation Report Related to the Department of Energy'S Proposal for the Irradiation of Lead Test Assemblies Containing TRITIUM-PRODUCING Burnable Absorber Rods in Commercial LIGHT-WATER Reactors
ML20141J9391997-04-30030 April 1997
Safety Evaluation Report Related to the Renewal of the Operating License for the Research Reactor at North Carolina State University
ML20141C2411997-04-30030 April 1997
Circumferential Cracking of Steam Generator Tubes
ML20141A5791997-04-30030 April 1997
Proposed Regulatory Guidance Related to Implementation of 10 CFR 50.59 (Changes, Tests, or Experiments).Draft Report for Comment
ML20137A2191997-03-31031 March 1997
Licensee Contractor and Vendor Inspection Status Report. Quarterly Report,October-December 1996.(White Book)
ML20134L3601997-01-31031 January 1997
Standard Review Plan on Antitrust.Draft Report for Comment
ML20134L3631997-01-31031 January 1997
Standard Review Plan on Power Reactor Licensee Financial Qualifications and Decommissioning Funding Assurance.Draft Report for Comment
ML20138J2461997-01-31031 January 1997
Licensee Contractor and Vendor Inspection Status Report. Quarterly Report,July-September 1996.(White Book)
ML20135D5711997-01-31031 January 1997
Operator Licensing Examination Standards for Power Reactors
ML20133E9161996-12-31031 December 1996
License Renewal Demonstration Program: NRC Observations and Lessons Learned
ML20149L8261996-10-31031 October 1996
Reactor Pressure Vessel Status Report
ML20135A4981996-10-31031 October 1996
Historical Data Summary of the Systematic Assessment of Licensee Performance
ML20128P4381996-10-0909 October 1996
Safety Evaluation Accepting Review of Cracked Weld Operability Calculations & Staff Response to NRC Task Interference Agreement
ML20107F5611996-04-17017 April 1996
Safety Evaluation Providing Guidance on Submitting plant- Specific Info W/Respect to IST Program Alternatives Request
ML14183A6951995-09-18018 September 1995
Safety Evaluation Approving Relocation of Technical Support Ctr
ML20236L5971994-12-29029 December 1994
SER in Response to 940314 TIA 94-012 Requesting NRR Staff to Determine Specific Mod to Keowee Emergency Power Supply Logic Must Be Reviewed by Staff Prior to Implementation of Mod
ML20128Q0761994-11-0404 November 1994
Coordinating Group Evaluation,Conclusions & Recommendations
ML20149H0671994-11-0404 November 1994
Safety Evaluation Supporting Amend 27 to Amended License R-103
ML20149G4281994-09-28028 September 1994
NRC Perspectives on Accident Mgt, Presented at 940928 Severe Accident Mgt Implementation Workshop in Alexandria, VA
ML20149F7581994-08-25025 August 1994
Topical Rept Evaluation of WCAP-13864,Rev 1, Rod Control Sys Evaluation Program
ML20149F4151994-08-0404 August 1994
Safety Evaluation Concluding That Unit 1 Can Be Safely Operated During Next Operating Cycle (Cycle 14)
ML20149E8831994-08-0202 August 1994
Safety Evaluation Accepting Interim Relief Request IRR-03 Re Drywell Isolation Check Valves in Equipment Drain Lines & Reactor Equipment Closed Cooling Water Sys
ML20248C5731994-07-19019 July 1994
SER Step 1 Review of Individual Plant Exam of External Fire Events for Millstone Unit 3
ML20059J4591994-01-25025 January 1994
Safety Evaluation Supporting Request for Relief from ASME Code Re Inservice Testing Requirements to Measure Vibration Amplitude Displacement
ML20059H4991994-01-24024 January 1994
Safety Evaluation Accepting Revised Responses to IEB-80-04 Re MSLB Reanalysis
1999-09-30
[Table view]Some use of "" in your query was not closed by a matching "".
Category:TOPICAL REPORT EVALUATION
MONTHYEARML20149F7581994-08-25025 August 1994
Topical Rept Evaluation of WCAP-13864,Rev 1, Rod Control Sys Evaluation Program
ML20059L1061994-01-12012 January 1994
Draft Topical Rept Evaluation of B&Wog Rept 47-1223141-00, Integrated Plant Assessment Sys/Structure Screening.... Applicant for License Renewal That Refs B&Wog Sys Screening Methodology Will Be Required to Develop Own Procedures
ML20059D1911993-12-30030 December 1993
Topical Rept Evaluation of RXE-91-005, Methodology for Reactor Core Response to Steamline Break Events
ML20058P2181993-12-10010 December 1993
SER Accepting Siemens Nuclear Power Corp Submittal of Topical Rept EMF-92-081, Statistical Setpoint/Transient Methodology for W Type Reactors
ML20058H9851993-11-26026 November 1993
Topical Rept Evaluation of WCAP-10216-P, Relaxation of Constant Axial Offset Control. Rept Acceptable
ML20059H8481993-11-0202 November 1993
SER Accepting Proposed Changes in Rev 3 to OPPD-NA-8302-P, OPPD Nuclear Analysis,Reload Core Analysis Methodology, Neutronics Design Methods & Verification
ML20134B4761993-10-30030 October 1993
Topical Rept Evaluation of Rev 3 to NP-2511-CCM Re VIPRE-01 Mod 2 for PWR & BWR Applications
ML20058M9851993-09-30030 September 1993
SE of Topical Rept, Transient Analysis Methodology for Wolf Creek Generating Station
ML20056G4171993-08-18018 August 1993
Topical Rept Evaluation of Rev 4 to OPPD-NA-8303, Transient & Accident Methods & Verification. Proposed Changes in Rev 4 Acceptable Except for Use of Cents Computer Code for Transient Analyses
ML20056E9661993-08-0606 August 1993
Sser Re Topical Rept HGN-112-NP, Generic Hydrogen Control Info for BWR/6 Mark III Containment Hydrogen Control
ML20056E3811993-08-0505 August 1993
Safety Evaluation of RXE-89-002, Vipre-01 Core Thermal- Hydraulic Analysis Methods for Comanche Peak Steam Electric Station Licensing Applications. Rept Is Acceptable for Ref in CPSES Core thermal-hydraulic Analyses
ML20056E3961993-08-0505 August 1993
Safety Evaluation of RXE-90-006-P, Power Distribution Control Analysis & Overtemperature N-16 & Overpower N-16 Trip Setpoint Methodology. Methodology Acceptable
ML20056E4681993-08-0505 August 1993
Supplemental Safety Evaluation for Topical Rept HGN-112-NP, Generic Hydrogen Control Info for BWR/6 Mark III Containments. Change Requests Consistent & Compatible W/ 10CF50.44 & Acceptable
ML20056E2571993-08-0505 August 1993
Corrected Safety Evaluation for Topical Rept RXE-91-001, Transient Analysis Methods for Commanche Peak Steam Electric Station Licensing Applications. Corrections Made to Second Sentense of Second Full Paragraph on Page Two
ML20056D9921993-07-29029 July 1993
Topical Rept Evaluation of OPPD-NA-8301,Rev 5, Reload Core Analysis Methodology Overview. Proposed Changes in Rev 5 Acceptable
ML20057A2661993-07-14014 July 1993
Topical Safety Evaluation of CEN-387-P, Pressurizer Surge Line Flow Stratification Evaluation. C-E Owners Group Analysis May Be Used as Basis for Licensees to Update plant- Specific Code Stress Rept for Compliance W/Bulletin 88-011
ML20056E1261993-06-29029 June 1993
Safety Evaluation of CENPD-382-P, Methodology for Core Designs Containing Erbium Burnable Absorbers. Rept Acceptable for Reload Licensing Applications of Both CE CE 14x14 & 16x16 PWR Lattice Type Core Designs
ML20057B5431993-06-26026 June 1993
Errata for Sser Re Topical Rept HGN-112-NP, Generic Hydrogen Control Info for BWR/6 Mark III Containments, for Use in Issuance of Final Approved Version of Topical Rept
ML20128B8101993-01-19019 January 1993
Safety Evaluation Accepting Methodology Described in Topical Rept RXE-91-002 Reactivity Anomaly Events Methodology for Reload Licensing Analyses for CPSES
ML20126E0381992-12-0909 December 1992
Safety Evaluation Accepting Topical Rept NEDC-31753P W/Ter Recommendations W/Listed Exceptions
ML20056D9351991-01-11011 January 1991
Topical Rept Evaluation Accepting Proposed Methodology for Fuel Channel Bowing Anaylses & for Referencing in Reload Licensing Applications W/Listed Conditions
ML20235Q7121989-02-22022 February 1989
Safety Evaluation Re Review of WCAP-10271,Suppl 2 & WCAP-10271,Suppl 2,Rev 1 on Evaluation of Surveillance Frequencies & out-of-svc Times for ESFAS
ML20206L9611988-11-23023 November 1988
Topical Rept Evaluation of PECO-FMS-0004, Methods for Performing BWR Sys Transient Analysis. Rept Approved,But Limited to Util Competence to Use Retran Computer Code for Facility
ML20205M0091988-10-25025 October 1988
Safety Evaluation of Topical Rept YAEC-1300P, RELAP5YA: Computer Program for LWR Sys Thermal-Hydraulic Analysis. Program Acceptable as Licensing Method for Small Break LOCA Analysis Under Conditions Stipulated
ML20204G8371988-10-18018 October 1988
Safety Evaluation Accepting Topical Rept 151, Haddem Neck Plant Non-LOCA Transient Analysis, Except for Issue of Feedwater Event
ML20155G7991988-10-12012 October 1988
Topical Rept Evaluation of TR-045, BWR-2 Transient Analysis Using Retran Code. Methods Described in Rept Acceptable for Reload Analysis When Listed Conditions Satisfied
ML20155G3201988-09-26026 September 1988
Safety Evaluation of TS NEDC-30936P, BWR Owners Group TSs Improvement Methodology. GE Analyses Demonstrated Acceptability of General Methodology for Considering TS Changes to ECCS Instrumentation Used in BWR Facilities
ML20155B0501988-09-22022 September 1988
Topical Rept Evaluation of Suppl 1 to NEDC-30851P, Tech Spec Improvement Analysis for BWR Control Rod Block Instrumentation. Analyses Acceptable to Support Proposed Extensions to 3 Months
ML20151K9921988-07-26026 July 1988
Topical Rept Evaluation of Nusco 140-1 Northeast Utils Thermal Hydraulic Model Qualification,Vol 1 (Retran). Rept May Be Generally Ref in Future Licensing Submittals.Further Justification by Util Required
ML20150D9651988-03-21021 March 1988
Topical Rept Evaluation of Rev 0 to TR-040, Steady State & Quasi-Steady State Methods for Analyzing Accidents & Transients. Util Methods Acceptable for Performing Reload Assembly Mislocation Analysis W/Listed Exceptions
ML20150D7671988-03-21021 March 1988
Topical Rept Evaluation of Rev 0 to TR-033, Methods for Generation of Core Genetics Data for RETRAN-02. Uncertainties in Input Parameters & Impact on Retran Results Should Be Determined for Qualification of Model
ML20236D2621987-10-21021 October 1987
Topical Rept Evaluation of CEN-348(B)-P, Extended Statistical Combination of Uncertainties. Rept Acceptable
ML20235D7041987-09-22022 September 1987
Safety Evaluation of Rev 0 to Topical Rept TR-021, Methods for Analysis of BWRs Steady State Physics. Rept,Methodology & Util Use of Methodology Acceptable
ML20239A5461987-09-0909 September 1987
Safety Evaluation Supporting A-85-11, Retran Computer Code Reactor Sys Transient Analysis Model Qualification for Use in Performing plant-specific best-estimate Transient Analyses at Plant
ML20215M3591987-05-0606 May 1987
Safety Evaluation Supporting Util Use of Suppl 1 to MSS-NA1-P, Qualification of Reactor Physics Methods for Application to PWRs of Middle South Utils Sys
ML20212M7781987-02-17017 February 1987
Topical Rept Evaluation of WCAP-10325, Westinghouse LOCA Mass & Energy Release Model for Containment Design - Mar 1979 Version. Rept Acceptable for Ref in Licensing Actions
ML20210N7331987-02-0404 February 1987
Safety Evaluation Supporting CEN-161(B)-P,Suppl 1-P, Improvements to Fuel Evaluation Model. Mods to Fission Gas Release & Fuel Thermal Expansion Models Acceptable
ML20215B2331986-12-0404 December 1986
Corrected Page 1 to 861031 Topical Rept Evaluation of Rev 2 to STD-R-05-011, Mobile In-Container Dewatering & Solidification Sys (Mdss). Word Effective Inserted Before Words Pore Sizes in First Line of 4th Paragraph
ML20214C5221986-11-14014 November 1986
Topical Rept Evaluation of Rev 0 to TR 020, Methods for Analysis of BWR Lattice Physics. Collision Probability Module Code Acceptable for BWR Fuel Lattice Calculations
ML20213F6531986-11-10010 November 1986
Safety Evaluation of Rev 2 to Vol 3 of XN-NF-80-19(P), Exxon Nuclear Methodology for Bwrs,Thermex:Thermal Limits Methodology Summary Description. Rept Acceptable for Ref in Licensing Applications
ML20207A8281986-11-0505 November 1986
Suppl 3 to Topical Rept Evaluation Re Submittal 2 to Rev 3 to CEN-152, C-E Emergency Procedure Guidelines. Rept Acceptable for Ref
ML20215N6901986-11-0404 November 1986
Topical Rept Evaluation of BAW-10155, FOAM2 - Computer Program to Calculate Core Swell Level & Mass Flow Rate During Small-Break Loca. Rept Acceptable W/Listed Restrictions Re Ranges of Core Flow Rate & Pressure
ML20215N3921986-10-31031 October 1986
Topical Rept Evaluation of STD-R-05-011, Mobile In-Container Dewatering & Solidification Sys (Mdss). Rept Acceptable for Ref in License Applications
ML20211D6921986-10-16016 October 1986
Safety Evaluation of Nusco 140-2, Nusco Thermal Hydraulic Model Qualification,Vol II (Vipre). Rept Acceptable for Establishing Input Values & Selection of Correlation Options & Solution Techniques for Calculations
ML20206S6511986-09-15015 September 1986
Topical Rept Evaluation of Addenda 3 to WCAP-8720, Improved Analytical Models Used in Westinghouse Fuel Rod Design Computations/Application for BWR Fuel Analysis. Rept Acceptable for Ref in Licensing Applications
ML20212N2001986-07-23023 July 1986
Topical Rept Evaluation of Rev 1 to XN-NF-85-67 (P), Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel. Rept Acceptable as Ref for Application to Jet Pump BWR Reload Cores,W/Listed Conditions
1994-08-25
[Table view]Some use of "" in your query was not closed by a matching "". |
Text
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4 ENCLOSURE l
SAFETY EVALUATION REPORT
~. .
j Report Number: TP-02-P, Rev. 1 Report
Title:
Nuclear Packaging Dewatering System Originating Organization: Nuclear Packaging, Inc., Federal Way, WA Reviewed by: Meteorology and Effluent Treatment Branch, DSI, NRR i
1.0 INTRODUCTION
TheNuclearPackagingDewateringSystem(theNUPACSystem)utilizesdewatering l _
equipment and disposable waste containers to dewater radioactive spent bead and powdered resins, and filter precoats. The dewatering process uses an air-driven positive displacement pump to obtain a continuous suction on a waste container to remove the bulk of free water. Then, the air blower recirculates air through the waste container and water separator to facilitate drying of the resin. These processes remove pumpable liquid from the waste container to a predetermined end point in accordance with the NUPAC process control program to meet the free standing liquid criteria set forth in Section 61,56(a)(3)of10CFRPart61. Vacuum gauges are provided at each waste outlet connection and manifold. The water removed from a waste container is
, returned to the user's liquid radwaste system.
3 The review of the NUPAC System, which was conducted in accordance with Section 11.4 of the Standard Review Plan (SRP), included the waste container internal design drawings, descriptive information on the dewaterfng operation, 8509180019 850906 DR TOPRPEPfVffy p
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equipment description, process control program, and qualtiy assurance program. The dewatering process treats " wet" radioactive waste to meet requirements in NRC Branch Technical Position, ETSB 11-3, Revision 2 in SRP Section 11.4. The process is not intended to meet the waste stability form or classification requirements in 10 CFR Part 61.
Nuclear Packaging submitted separate topical reports on High Integrity Containers for NRC review and approval. In these reviews by the HMSS staff, the structural integrity of the NUPAC containers is being evaluated to ensure long-tern isolation of low-level radioactive waste from the soil environment.
2.0 EVALUATION The design and operation of the NUPAC System are described in detail in the NUPAC Topical Report, TP-02-P, Rev. O and Rev. 1 dated August 6, 1984 and June 28, 1985 respectively. In the staff's evaluation of the NUPAC System, the staff considered:
(1) The process control program to assure complete dewatering of " wet" solid I
radwaste. , ,
- (2) Design provisions incorporated in the equipment and system design to reduce leakage and control and monitor releases of ' radioactive effluents to the environment.
I (3) The quality assurance program for the design, fabrication and testing of the system.
(4) Typical interfaces with the reactor plant.
(5) l!aste container internal design.
(6) Provisions to control potential exothermic reaction in dewatering ion exchange resin.
(7) Radiation protection design features.
The NUPAC System consists of a dewatering waste container, a dewatering pump, an off-gas vent unit, a container level indicator, a waste fill head, a water separator with water chiller unit, an air blower, a relative humidity instrument, a control panel, and interconnecting piping and valves.
After " wet" radwaste from the user's plant is charged into a NUPAC waste container, dewatering is achieved with continuous suction on a waste container provided by the dewatering pump. The residual free water in the waste container is removed by recirculation of drying air provided by the air blower.
Various types and numbers of filters are used within the waste container in different configurations to retain spent resin and filter precoat materials.
Water removed from the waste container is retur,ned to the user's liquid radwaste system.
The dewatering pump is operated for given time intervals in accordance with the NUPAC process control program. The pumping time may range from eight to
~, , , , , , .
~
~
l sixteen hours depending upon type of wastes and waste containers. After most of the free water in the waste container has been removed, drying air is' continuously recirculated in a loop from the air blower to the waste container through the water separator to remove any residual free water in the waste
. container. The NUPAC System is provided with temperature instrumentation which is interlocked to automatically shut down the dewatering process on high air temperature due to potential exothermic reaction in dewatering ion exchange resin.
I The waste container is considered dewatered when the volume of collectable liquid and relative humidity in the recirculating drying air meet the acceptance criteria specified in the NUPAC process control program. A relative humidity instrument and monitor are provided to remotely and continuously monitor the waste container outlet air. This instrument is used to establish positive end point to the dewatering process.
The topical report describes NUPAC generic Process Control Programs (PCPs) for dewatering spent bead resin and filter precoats for dewatering to ensure that the dewatered waste containers meet the free standing liquid criteria set forth in Section 61.56(a)(3) of 10 CFR Part 61., The PCPs are developed based on the actual test results on drainable liquid obtained from dewatering waste
~
containers and from the subsequent road tests on dewatered waste containers.
The staff reviewed the NUPAC dewatering test procedures and its results, detailed dewatering operating and maintenance procedures, and acceptance criteria described in the topical report. The staff finds the NUPAC acceptance criteria and dewatering test results meet the free standing liquid criteria in 10 CFR Part 61 and NRC Branch Technical Position ETSB 11-3, Rev. 2 and therefore, the staff finds the NUPAC PCPs to be acceptable.
No airborne or liquid radwastes are released to the environment from the NUPAC i
dewatering operation. The dewatered liquid radwastes are routed to the user's liquid radwaste system and resin drying air is vented to the user's off-gas system. The NUPAC System is designed to prevent uncontrolled releases of radio' active materials by monitoring liquid levels in the waste container by a level indicator. During the waste filling operation, the operator is required to be stationed near the control panel and visually monitor the waste transfer process observing the video monitor and the radiation monitor provided. Curbs or other means to contain inadvertent spills and overflows will be provided by the user with floor drains routed to the user's liquid radwaste collection systems.
The consequences of a waste dewatering containe,r failure releasing radioactive materials to a potable water supply is site dependent and will be evaluated for individual license applications. The staff finds th'e NUPAC dewatering process and waste container design meet the requirement of Section 20.106 of 10 CFR Part 20, Section 50.34a of 10 CFR Part 50, and General Design Criteria 60 and 64 of Appendix A to 10 CFR Part 50.
The design, procurement, fabrication, testing and operation of the NUPAC System is accomplished under prescribed quality assurance requirements which conform, to the extent practicable, with the guidelines provided in Regulatory Guide 1.143 " Design Guidance for Radioactive Waste Management Systems, Structures and Components Installed in Light-Water-Cooled Nuclear Power Plants." The quality assurance program defines and controls those elements of NUPAC and their suppliers' performance which affect the quality of the NUPAC System.
The design and arrangement of the NUPAC System components are based on ,
maintaining the operator radiation exposure as low as is reasonably
_ achievable. The topical report provides a list of specific design and operating features which were incorporated to minimize personnel radiation exposure. All active components are located so they can be easily accessed for maintenance. ' All pumps, valves and piping can be flushed prior to '
inspection and maintenance.
i The staff also finds that NUPAC has adequately identified interface information and requirements which users should provide.
Upon completion of the staff revicw by NHSS of the NU?AC topical reports on High Integrity Containers, a separate Safety Evaluation" Report will be provided to supplement this evaluation.
l*
3.0 CONCLUSION
Based on the foregoing evaluation, the staff finds the NUPAC Topical Report, TP-02-P, Rev. 1, to be acceptable.
The bases for our acceptance is our conclusion that the NUPAC Dewatering System is designed and can be operated in accordance with current guidance of applicable regulatory guides, standard review plans, branch technical positions, and Federal regulations.
The capability of the plant radioactive waste treatment systems to meet the requirements of Appendix I to 10 CFR Part 50 with the NUPAC System in operation is site dependent and will be evaluated for individual license applications. In addition, the packaging and shipping of all processed wastes including waste classification in accordance with the applicable sections of 10 CFR Parts 61 and 71, and 49 CFR Parts 170-178, will be determined for individual license applications. The consequences of a potential waste container failure releasing radioactive materials to a potable water supply is also site dependent and will oe evaluated for individual license applications. , ,
The staff concludes that the HUPAC Topical Report is acciptable for reference in future license applications for light water reactors. Any application incorporating this report by reference should include the following information:
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l
- 8-(1) An) exceptions or deviations from the NUPAC Topical Report, Rev. 1, dated June 19S5.
(2) Interfaces between the plant and the NUPAC System.
(3) Location ano arrangement drawings of the NUPAC System in the plant incitding curbs or other means to contain inadvertent spills and overflows.
(4) The waste classification program to demonstrate that the solid waste product is classified in accordance with 10 CFR Part 61, Section 61.55 and NRC Branch Technical Position on Waste Classification.
(5) Description of the solid waste product container to be used.
(6) The capability of the plant radioactive waste treatment system to meet the requirements of Appendix I to CFR Part 50 with the NUPAC System in operation. .
(7) The plant site information on potable water supply.
i 9 e e
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