ML20214E691
| ML20214E691 | |
| Person / Time | |
|---|---|
| Issue date: | 03/20/1986 |
| From: | Parr O Office of Nuclear Reactor Regulation |
| To: | Shingleton R PACIFIC NUCLEAR SYSTEMS, INC. |
| References | |
| TAC-57810, NUDOCS 8603260379 | |
| Download: ML20214E691 (10) | |
Text
WflR 2 01986 Mr. Roger L. Shingleton, Vice President Pacific Nuclear Systems, Inc.
1010 South 336th Street Federal Way, Washington 98003
Dear Mr. Shingleton:
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION ON TP-04, REVISION 0 We are currently reviewing the Pacific Nuclear Systems, Inc. Topical Report TP-04, Revision 0 dated April 11 1985 covering the Pacific Nuclear System's Portable Solidification System. Our initial review reveals the need for the additional information indicated in the Enclosure.
In order to complete our review within the currently scheduled time, responses to these questions should be received by the NRC by April 18, 1986. Please advise Charles Nichols at 301-492-7694 if you cannot meet this, date.
The reporting and/or recordkeeping requirements contained in this letter affect fewer than ten respondents; therefore, CMB clearance is not_ required under P.L.96-511.
Sincerely, ei;j d bz Ojau Parr. M Olan D. Parr, Chief Plant, Electrical, Instrumentation and Control Systems Branch Division of PWR Licensing-B Office of Nuclear Reactor Regulation
Enclosure:
As-Stated Distribution Central File NRC PDR PEICSB Rdg. File F. Miragl.ia T. Speis G. Lainas C. E. Rossi H. Berkow D. Crutchfield R. Emch J. Wermiel J. Minns c/f C. Nichols
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LICENSING REVIEW QUESTIONS ON PACIFIC NUCLEAR SYSTEMS TOPICAL REPORT, TP-04, REVISION 0, PORTABLE SOLIDIFICATION SYSTEM Provide a revised version (Revision 1)lof the subject topical report which includes all the desired changes identified with vertical lines
- in the margin.
Revision 1 should incorporate all of the responses to the following review questions on the Revision 0 report:
1.
The following standards potentially apply to the licensing acceptability of this portable solidification system.
Identify, where appropriate, how the portable solidification system meets these standards, or explain why 2
the standards do not apply:
a.
NUREG-0800, Standard Review Plan, Section 11.4, Solid Waste Management Systems,-including Branch Technical Position ETSB 11-3, Design Guidance j
for Solid Radioactive Waste Management Systems.
b.
NUREG-0800, Standard Review Plan, Section 11.5, Process and Effluent Radiological Monitoring Instrumentation and Sampling Systems.
c.
10 CFR 20, Standards for Protection Against Radiation.
d.
10 CFR 50, Appendix A, General Design Criteria for Nuclear Power Plants.
e.
10 CFR 61, Section 61.55, Waste Classification.
f.
10 CFR 61, Section 61.56, Waste Characteristics.
g.
10 CFR 71, Packaging of Radioactive Material for Transport and Trans-portation of Radioactive Materials Under Certain Conditions.
.~
.---,n..,,.,
. ~.,, -. ~...
.---._..-,.-m_,-,
-2 h.
49 CFR 173, D0T regulations for packaging.
' i. Regulatory Guide 1.140, Design Testing and Maintenance Criteria for Nonnal Ventilation Exhaust System Air Filtration and Adsorption Units of Lignt-Water-Cooled Nuclear Power Plants.
j.
Regulatory Guide 1.143, Design Guidance for Radioactive Waste Management Systems, Structures and Components in Light-Water-Cooled Nuclear Power Plants.
k.
Regulatory Guide 8.8, Information Relevant to Ensuring That Occupa-tional Radiation Exposures at Nuclear Power Plants Will Be As Low As'Is Reasonably Achievable.
2.
Provide the design volumes, activities, physical and chemical characteris-tics of wastes to be processed by type of waste (e.g. sludges, resins, evaporator bottoms, dry active wastes).
Provide the range and limitations-on the types, compositions and~ physical properties of waste streams that can be processed and the bases therefor; the design volume and radio-nuclide contents of solidified wastes based on the types and quantities of wastes processed; the capacity or throughout of the system in relation to the expected waste solidification requirements of a 1200-MWe PWR and BWR; and the radioactive source terms used in the system design for snielding analyses and. calculations of normal effluents and effects of postulated accidents.
3.
Describe the type and size of solid waste containers which can be used and potential means for monitoring for removable contamination and for l
decontamination.
l I
l 4.
Provide the quality group classification of piping and equipment and the bases governing the classification; provisions incorporated in the system l
l l
c
i l
design to minimize leakage and facilitate operations and maintenance; provisions in the design or which should be made by users for containing and cleanup of overflows-and spills at radioactive. materials; and the design and initial and periodic testing of interfaces between the system and the plant (e.g., liquid radwaste line, compressed air line, and waste return line) in accordance with Regulatory Guide 1.143.
5.
Describe the estimated maximum radiation exposures to operating and maintenance personnel (and the basis for these estimates) and the measures taken in the design and in the operating, testing, and maintenance pro-cedures to keep radiation exposures ALARA,-including the uncrupling of the mixing equipment from the solidifying waste and the capping operation.
I 6.
A description should be provided of the potential radiation hazards and accidents and the procedures that may be used in abnormal situations, e.g. spills 'of radioactive materials and mixer motor or blade failure during solidification.
]
7.
Provide a detailed drawing of the overall system and a detailed piping and 4
instrumentation diagram.
8.
Clarify throughout the report whether the use of either ordinary cement or Envirostone is intended as the solidification medium or that only the use of Envirostone is intended.
Describe the system for storage, metering, and transfer of the cement and other additives.
9.
Describe the design, location, testing and maintenance of liner level' monitors, radiation monitors (in accordance with NUREG-0800, SRP 11.5),
and offgas system filters (in accordance with Regulatory Guide 1.140).
10.
Describe the information to be used and the criteria to be employed by the operator to determine that a homogeneous mix has been established, that accelerator should be added, that there is no free-standing liquid, and 4
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to select steps (describe steps) to be taken when it cannot be deter-mined that a homogenous mix has been established or that there is no free-standing liquid.
- 11.. Describe the means for determining the isotopic composition ard total activity of the waste product.
p
- 12. Describe Pacific Nuclear's conclusions regarding the long-term stability attributes (in addition to solidification and absence of free-standing liquid) of the solidified product, including the ettects of ridiation and decay.
-13.
Describe the means for assuring that the plant's waste storage tank has been adequately mixed and that the waste sample is an adequate repre-sentation of the waste; and that the plant's waste storage tank is completed isolated after the waste sample has been taken?
14.
In Appendix B, clarify whether the term " batch" refers to the. contents of the waste batch tank or the waste product container; the size of the
~
sample to be taken; that the test solidifications should be repeated until a satisfactory end product is obtained; and,the source of the Sample Proportion values to be used in the Sample Vsrification Worksheet.
t
'15.
Describe the quality assurance program for the design, construction and testing of the system in accordance with Regulatory Guide 1.143.
The following comments and suggestions are purely editorial:
- In the second paragraph on page 3-2, the usual symbol for volt or
[
volts is V, not Vt or Vts.
- In the second paragraph of Section 4.0, the sixth and seventh lines should be deleted.
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Mr. Roger 1. Shingleton, Vice President Pacific -Nuclear Systems, Inc.
1010 South 336th Street Federal Way, Washington 98003
Dear Mr. Shingleton:
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION ON TP-04, REVISION 0 We are currently reviewing the Pacific Nuclear Systems, Inc. Topical Report TP-04, Revision 0 dated April 11, 1985 covering the Pacific Nuclear System's Portable Solidification System. Our initial review reveals the need for the additional information indicated in the Enclosure.
In order to complete our review within the currently scheduled time, responses to these questions should be received by the NRC by April 18, 1986.
Please advise Charles Nichols at 301-492-7694 if you cannot meet this date.
1 The reporting-and/or recordkeeping requirenents contained in this-letter affect fewer than ten respondents; therefore, CMB clearance is not required under P.L.96-511.
Sincerely, Original MM%.
M Olan 07'kIIr.- Chief Plant, Elect-ical, Instrumentation and Control Systems Branch Division of P.c Licensing-B Office of Nucit tr Reactor Regulation
Enclosure:
As Stated Distribution Central F Ai hRC PDR 7 PEICSB Rdg. File F. Miraglia T. Speis G. Lainas C. E. Rossi H. Berkow D. Crutchfield R. Ench J. We'rmiel J. Minns.
C. Nichols head' 01)p PEICSB PE.I C'$ B PEICSB CNichcis:as JLeraiel 0DParr 3/21//86 3/26/86 3/ g /86 W
LICENSING REVIEW QUESTIONS ON PACIFIC NUCLEAR SYSTEMS TOPICAL REPORT, TP-04, REVISION 0, PORTABLE SOLIDIFICATION SYSTEM Provide a revised version (Revision 1) of the subject topical report which includes all the desired changes identified with vertical lines in the margin.
Revision 1 should incorporate all of the responses to the following review questions on the Revision 0 report:
1.
The following standards potentially apply to the licensing acceptability of this portable solidification system.
Identify, where appropriate, how the portable solidification system meets these standards, or explain why the standards do not apply:
i a.
NUREG-0800, Standard Review Plan, Section 11.4, Solid Waste Management l
Systems, including Branch Technical Position ETSB 11-3, Design Guidance for Solid Radioactive Waste Managemant Systems.
b.
NUREG-0800, Standard Review Plan, Section 11.5, Process and Effluent Radiological Monitoring Instrumentation and Sampling Systems.
c.
10 CFR 20, Standards for Protection Against Radiation, d.
10 CFR 50, Appendix A, General Design Criteria for Nuclear Power L
Plants.
e.
10 CFR 61, Section 61.55, Waste Classification.
f.
10 CFR 61, Section 61.56, Waste Characteristics.
g.
10 CFR 71, Packaging of Radioactive Material for Transport and Trans-portation of Radioactive Materials Under Certain Conditions.
L
h.
49 CFR 173, D0T regulations for packaging.
- i. Regulatory Guide -1.140,~ Design Testing and Maintenance Criteria for Normal Ventilation Exhaust System Air Filtration and Adsorption Units of Lignt-Water-Cooled Nuclear Power Plants.
- j. Regulatory Guide 1.143, Design Guidance for Radioactive Waste Management Systems, Structures and Components in Light-Water-Cooled Nuclear Power Plants.
k.
Regulatory Guide 8.8, Information Relevant to Ensuring That Occupa-tional Radiation Exposures at Nuclear Power Plants Will Be As Low
~
As Is Reasonably Achievable.
9 2.
Provide the design volumes, activities, physical and chemical characteris-tics of wastes to be processed by type of waste (e.g. sludges, resins, escporator bottoms, dry active wastes).
Provide the. range and limitations on the types, compositions and physical properties of waste streams that can be processed and the bases therefor; the design volume and radio-nuclide contents of solidified wastes based on the types and quantities of wastes processed; the capacity or throughout of the system in relation to the expected waste solidification requirements of a 1200-MWe PWR and BWR; and the radioactive source terms used in the system design for snielding analyses and calculations of normal effluents and effects of postulated accidents.
3.
Describe the type and size of solid waste containers which can be used and potential means for monitoring for removable co~ntamination and for decontamination.
4.
Provide the quality group classification of piping and equipment and the bases governing the classification; provisions incorporated in the system
. design to minimize leakage and facilitate operations and maintenance; provisions in the design or which should be made by users for containing and cleanup of overflows and spills of radioactive materials; and the design and initial and periodic testing 'of interfaces between the system and the plant (e.g., liquid radwaste Ifne, compressed air line, and waste return line) in accordance with Regulatory Guide 1.143.
l S.
Describe the estimated maximum radiation exposures to operating and maintenance personnel (and the basis for these estimates) and the measures taken in the design and in the operating, testing, and maintenance pro-cedures to keep radiation exposures ALARA, including the uncoupling of the mixing equipnent from the solidifying waste and the capping operation.
6.
A description should be provided of the potential radiation hazards and
-accidents and the procedures that may be used in abnormal situations, e.g. sp111s of radioactive materials and mixer motor or blade failure during solidification.
l 7.
Provide a detailed drawing of the overall system and a detailed piping and instrumentation diagram.
8.
Clarify throughout the report whether the use of either ordinary cement l
or Envirostone is intended as the solidification medium or that only the use of Envirostone is intended. Describe the system for storage, j
metering, and transfer of the cement and other additives.
9.
Describe the design, location, testing and maintenance of liner level monitors,radiationmonitors(inaccordancewithNUREG-0800,SRP11.5),
and oftgas system filters (in accordance with Regulatory Guide 1.140).
- 10. Describe the information to be used and the crite la to be employed by the operator to determine that a homogeneous mix has been established, that accelerator should be added, that there is no free-standing liquid, and
m-z,_
m _._.
.-_ _ m.u
. to select steps (describe steps) to be taken when it cannot be deter-mined that a homogenous mix has been established or that there is no free-standing liquid.
- 11. Describe the means for determining the isotopic composition and total activity of the waste product.
- 12. Describe Pacific Nuclear's conclusions regarding the long-term stability attributes (in addition to solidification and absence of free-standing liquid) of the solidified product, including the ettects of radiation and decay.
- 13. Describe the means for assuring that the plant's waste storage tank has been adequately mixed and that the waste sample is an adequate repre-sentation of the waste; and that the plant's waste storage tank is completed 1solated after the waste sample has been taken?
14.
In Appendix B, clarify whether the term " batch" refer.s to the contents of the waste batch tank or the waste product container; the size of the sample to be taken; that the test solidifications should be repeated until a satisfactory end product is obtained; and the source of the Sample Proportion values to be used in the Sample Verification Worksheet.
- 15. Describe the quality assurance program for the design, construction and testing of the system in accordance with Regulatory Guide 1.143.
The following comments and suggestions are purely editorial:
- In the second paragraph on page 3-2, the usual symbol for volt or volts is V, not Vt or Vts.
- In the second paragraph of Section 4.0, the sixth and seventh lines should be deleted.
_