ML20052H691
| ML20052H691 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 05/14/1982 |
| From: | Nichols T SOUTH CAROLINA ELECTRIC & GAS CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8205210344 | |
| Download: ML20052H691 (10) | |
Text
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.i SOUTH CAROLINA ELECTRIC & GAS COMPANY POST OFFICE BOX 764 COLUMBI A. S. C.
'218 May 14, 1982 Mr. Harold R. Dentcn, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Comnission Washington, D.C.
20555
Subject:
Virgil C. SumTer Nuclear Station Docket No. 50/395 FSAR Questions-Process Control Program
Dear Mr. Denton:
South Carolina Electric and Gas Ccmpany (SCE&G) hereby provides the responses to FSAR Questions 321.17, 321.18, 321.19, 321.20, 321.21, 321.22 and 321.23 concerning the Process Control Program as requested in a letter dated April 12, 1982 frcm Mr. B. J. Youngblood of the Staff.
If you have additional questions, please let us know.
Very truly yours,
/
d T. C. Nichols, Jr.
Senior Vice President Power Operations
'IE:'IOI:tdh cc:
V. C. Sumter (w/o attach.)
G. H. Fischer (w/o attach. )
H. N. Cyrus T. C. Nichols, Jr.
(w/o attach.)
M. B. Whitaker, Jr.
J. P. O'Reilly H. T. Babb D. A. Nauman C. L. Ligon (NSIC) 00I W. A. Williams, Jr.
1 R. B. Clary O. S. Bradham A. R. Koon l
l M. N. Browne G. J. Braddick J. L. Skolds J. B. Knotts, Jr.
B. A. Bursey NPCF File 8205210344 820514 PDR ADOCK 05000395
)
A PDR j
e.
Question 321.17 Describe how you propose to cmply with the interface requirements spelled out in the ChemNuclear Topical Report for their cement solidification system.
Response
Provisions have been nade to supply ChemNuclear with the interface l
requirements listed in their cement solidification systen Topical Report. These services include the following:
(1) General Electrical:
120A, 480/3 Phase /60Hz Service Water:
25 gpn at 80 psi minimum Service Air:
75 SCFM at 80 psi minimum Waste:
1-1/2 inch 150 pound ANSI flange connection Dewater:
1-1/2 inch 150 pound ANSI flange connection Vent:
A 4 inch stub connection on the building ventilatica header to interface with Chem-Nuclear's blower exhaust.
(2) The ChamNuclear control panel is located in the drunning station control room. The plant operator and the CNSI portable unit operator are in direct comunication with one another during a waste transfer operation.
(3) Prior arrangements for shipping of the solidified material shall be made.
(4) Preparations to accept cement shipnents shall be made prior to the arrival of the unit on site.
(5) A Radiation Work Permit (RWP) nust be issued to the CNSI operator before waste processirg begins, according to SCE&G's radiation protection procedures.
(6) Any clothing or equipnent for necessary radiation protection of the CNSI operator shall be provided.
(7) A controlled area already exists around the processing area includity the cement storage trailer.
(8) An area inside the Hot Machine Shop has been designated to be used for test solidifications.
(9) Crane services, torque wrenches, and other naterial necessary shall be provided for loading the disposable liners and preparing the solidified waste for shipuent.
(10) A forklift capable of 4000 pounds at 6 foot mcznent arm shall be provided as necessary to nove CNSI portable skids.
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-Question 321.18 Provide tables showing how the utility equipment, conponents, structures and services that interface with the skid-nounted cement solidification system comply with the applicable criteria of Regulatory Guide 1,143, Rev.1, October 1979, " Design Guidance for Radioactive Waste Mnagenent Systems, Structures and Coroponents Installed in Light-Water-Cooled Nuclear Power Plants" and Branch Technical Position EPSB 11-3, Rev. 2, July 1981,
" Design Guidance for Solid Radioactive Waste Management systems Installed in Light-Water-Cooled Nuclear Power Plants".
Responm:
Valves (GAI)
M Design and Fabrication Inspection and Testing Materials
- Diaphragm (Dia.)
ANSI B16.5 (1968)
ANSI B16.5 (1968)
AS1M (6/74)
Check (H O)
ANSI B16.5 (1968)
ANSI B16.5 (1968)
ASIM (6/74) 2 l
Dia. (auto.)
ANSI B16.5 (1968)
ANSI B16.5 (1968)
AS1M (6/74)
Plty (auto. )
ANSI B16.5 (1968)
ANSI B16.5 (1968)
ASIM (3/73) l Check ANSI B16.5 (1968)
ANSI B16.5 (1968)
AS7M (2/74) l Check (N )
ANSI B16.5 (1968)
ANSI B16.5 (1968)
ASTM (6/74) 2 Dia. (N )
ANSI B16.5 (1968)
ANSI B16.5 (1968)
ASTM (6/74) 2 D'a. (control ANSI B16.5 (1968)
ANSI B16.5 (1968)
ASIM (6/75)
N) 2 Valves (Westinghouse)
Dia..
ANSI B16.5 (1968)
ANSI B16.5 (1968)
ASIM (1/73) i ASME B&W Sec. III l
Dia. (auto)
ANSI B16.5 (1968)
ANSI B16.5 (1968)
ASTM (1/73)
ASME B&W Sec. III Check ANSI B16.5 (1968)
ANSI B16.5 (1968)
ASIM (11/71)
Globe ANSI B16.5 (1968)
ANSI B16.5 (1968)
AS1M (11/71)
- Dates listed under Materials refer to specification issue dates.
The AS7M Standard applicable to these dates apply..
Equipment Spec.
Issue Equitznent Date* Design and Fabrication Inspection and Testing Materials Tank (Q )
6/72 ASME B&W Sec. VIII ASME B&W Sec. VIII ASME B&W Sec. VIII Tank (PSRS) 6/72 ASME B&W Sec. III ASME B&W Sec. III ASME B&W Sec. III Tank (NBSRS) 6/ 75 ASME Section VIII ASME Section VIII ASIM Tank (WBCT) 1/74 ASME Section VIII ASME Section VIII ASIM Tank (PSR) 11/71 ASME Sec. III, ASME Section III, Hydraulic Inst.
HIS ASME Section III Standard (HIS)
Punp (NBSR) 1/74 HIS HIS ASIM Punp (WBC) 1/74 HIS HIS ASIM Ptmp (CD) 11/71 HIS HIS ASTM Filters 7/72 ASME Section VIII ASME Section VIII ASIM
y.
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Piping (GAI)
Spec.
Issue Sy tan Date* Design and Fabrication Insg a tion and Testing Materials t
Waste,Hp
-10/73 ANSI B31.10 AN5I B31.10 ASTM N, Air 10/73 MSI B31.10 ANSI B31.10 ASTM 2
Piping (Westinghouse)
Waste, N, 8/71 ANSI B31.10.
ANSI B31.10 ASIM 2
HO 2
- Um these dates for refering to the revisions of the standards listed.
Notes:
Chemical Drain O
=
Primary Spent Resin Storage PSRS
=
Nuclear Blowdown Spent Resin Storage NBSRS
=
WEC Waste Evaporator Concentrates
=
Boiler and Pressure Vessel Code B&W
=
)
i I..
t Questica 321.19 Describe how the plant design, as it relates to the cement solidification system, reflects consideration of the following design features intended to maintain occupational radiation exposures AIARA:
a Minimizing the length of piping runs b.
Avoiding low points and dead legs in piping c.
Using larger diameter piping to minimize plugging
Response
a.
The spent resin storage tanks, the waste evaporator concentrates tank, and the associated punps are locatal in the Auxiliary Building at floor elevation 412'. The waste hold-tp tank, the chemical drain tank, and the associated punps are located at floor elevation 374'.
The portable cement solidification area is located directly above this equipnent at floor elevation 436'.
By locating the equipment as such, the piping runs are minimized.
b.
The resin transfer lines are sloped to avoid low points in the piping. Also, 5 diameter pipe bends are used from the resin tanks to the solidification area. These long sweeping bends are necessary to avoid plugging that could occur in the inner wall of a bend. Instrunent lines are kept to a mininum to avoid dead legs. All lines will be flushed with make-up water prior to any maintenance activities.
c.
The resin and waste transfer lines are 2" nominal dianeter.
This size is sufficient for the desired flow rates without causing excessively large pressure drops that could result in line plugging.
I:
k' Question 321.20 Clarify diether heat tracing has been incorporated for tanks that contain evaporator concentrates that are likely to solidify at ambient temperatures.
Response
Heat tracing has been incorporated in the Waste Evaporator Concentrates Tank which is the only waste tank that contains concentrates likely to solidify at ambient temperatures.
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Question 321.21 Describe the equipment, conponents or structures and services you provich for containing radioactive spills that may occur in the cement solidification system.
Response
In addition to the process equipment discussed in Chapter 11.5.8.1 of the FSAR, a remote camera will be located in the waste processing area to view the liner and associated connections durirg a packagirg operation. Both the ChenNuclear operator and the plant operator will have the capability to observe the equipment during processing of waste from the drunmirg station control roan. If leaks of any kind or spills are observed, the operation in progress can be inmediately terminated. Any spill which may occur will be contained by permanent and/or portable curbing in the solidification area and in the truck bay when this area is used.
4;
- i Questions 321.22 Describe the plant inspection program to assure that cement and/or conditioniry chemicals are maintained at the proper quality during the time they are stored.
Response
ChemNuclear's solidification system uses Portland 1 cement, calcium hydroxide, and sodium sulfate. These chenicals have an indefinite shelf life as long as noisture is excluded. The cement is stored in a bulk trailer ard the other chemicals are stored 'n the radwaste packaging area in the Auxiliary Building. Prior ' o a full scale solidification, a successful sanple solidification is performed using naterials obtained fran the full-scale solidification supplies in accordance with the PCP. Unti?. an acceptable test solidification is obtained, full scale solidification will not take place.
In this manner, the quality of the solidification chemicals is assured.
f t;* M Question 321.23 Describe how the curie content and identification of radionuclides in each container are determined prior to shipment.
Response
The primry activity deteunination nethod will be to sanple the este stream (resins and liquid este) during transfer to a process container and analyze the sanple using the appropriate countirg instrumentation. An isotopic determination is nade of the radionuclides present a x1 the activity of each. Sumnation of the individual activities is used to calculate the curie content of the processed container.
For cases dere the primary nethod cannot be used, an alternate technique as set forth by Messrs. Bowman and Swindlel will be inplemented. The alternate method entails using the dose rate of the packaged este in order to calculate the curie content. The calculation considers the waste characteristics, geometry of the mste package, characteristics of the container and solidification media (if applicable), and the average gamna energy. For spent cartridge filters, this alternate nethod will be used to determine the curie content. The appropriate countiry instrumentation is used to analyze sanples taken frcxn the process stream and the effluent to> identify radionuclides present and the average ganna energy.
1 W. B. Bowman and D. L. Swindle, " Determination of the Content of Packaged Radioactive Wasta Usity Measured Dose Rates", Health Physics, Perganon Press, Volume 31, 1976.
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