ML20149F876
| ML20149F876 | |
| Person / Time | |
|---|---|
| Site: | Framatome ANP Richland |
| Issue date: | 09/12/1994 |
| From: | Edgar J SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER |
| To: | Pierson R NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| Shared Package | |
| ML19304C543 | List: |
| References | |
| JBE:94:0146, JBE:94:146, TAC-L21656, NUDOCS 9409280082 | |
| Download: ML20149F876 (183) | |
Text
)
GQ - O S]
a:
j 4
SIEMENS September 12,1994 JBE:94:0146 U.S. Nuc!sar Regulatory Commission Attention: Mr. Robert C. Pierson, Chief Ucensing Branch Division of Fuel Cycle Safety and Safeguards, NMSS Washington, D.C. 20555 License No. SNM-1227 Docket No. 70-1257
Dear Mr. Pierson:
Subject:
Additional information in Support of Siemens Power Corporation's License Renewal Application
Reference:
Letter, M. Adams to L J. Maas, " License Renewal - Request for Additional Information (TAC No. L21656)," dated June 14,1994 Enclosed herewith are Siemens Power Corporation's (SPC's) responses to the NRC staff's requests for additional information regarding SPC's August 1992 license renewal application.
The responses are in the form of specific answers to requests as well as revised license application pages, where applicable. In addition, enclosed for your information as requested, are copies of:
l 1.
Chapter 2," Radiation Protection Standards" from SPC's Safety Manual, EMF-30; 2.
The "1990 ALARA Committee Report", EMF-91-198; and 3.
"The ALARA Report" for 1992, EMF-93-091(P).
The 1992 Al. ARA report is a proprietary document and an affidavit attesting to that fact is enclosed.
Two copies of the following license application pages are enclosed as referenced in the responses to specific requests (changes are indicated by vertical lines in the right margins):
Chapter 1
- Pages 1-1 and 1-7 through 1-10 6
Chapter 2 - Pages 2-6 and 2-11 through 2-23
$pi Pages 3-2 and 3-6 I
k Chapter 3 Chapter 4 Pages 4-1 through 4-14 (complete revised chapter)
Y j
0 h$lfgg)g ap p, Siemens Power Corpolatbd g
j
/A//
/Y Nuclear Dm90n - Engineenng and Manufactonng facihty f/d(
/
2101 Horn D n+ Anar1 PO Rnx 130 Richland. WA 99352 0130 Tel (509) 375 8100 Fax (509) 375 8402 9409280082 940912 PDR ADOCK 07001257 PDR
)
s.
Mr. Robert C. Pierson September 12,1994 Page 2 l
)
Chapter 5 Pages 5-1 through 5-9 and 5-11 Pages 6-3 and 6-4 Chapter 6 Chapter
Pages 9-1,9-3 and 9-8 Chaptr-Pages 10-3 through 10-56 and 10-88 Chap.
Pages 12-1 through 12-13 plus Appendix B (complete revised chapter) l Chap *,13 - Pages 13-1 and 13-2 Chapter 14 - Pages 14-1 through 14-28 (complete revised chapter)
Chapter 15 - Page 15-30 If you have questions requiring this submittal of information, please call me at (509) 375-8663.
Very truly yours,
/3$
J mes B. Edgar Staff Engineer, Licensing JBE/cf Attachments c w/o incl:
C. A. Hooker NRC Region IV Walnut Creek Field Office
)
i o
i SIEMENS POWER CORPORATION'S RESPONSES TO NCR'S REQUEST FOR ADDITIONAL INFORMATION ON LICENSE RENEWAL (TAC NO. L21656)
AUTHORIZED ACTIVITIES, ORGANIZATION AND ADMINISTRATION, FACILITY DESCRIPTION, SPECIAL PROGRAMS, DECOMMISSIONING NRC Reauest AA-1 Chapter 1 should include a site map, similar to Figure Il-10.1, that identifies the specific locations listed in Table I-1.1.
1 SPC Response AA-1 A footnote has been added to Table 1-1.1 referring to Figure 11-10.1 for locations of authorized activities. Adding another site map to the application would be i
needlessly redundant.
NPC Reauest AA-2 During the December 1993 site visit, several waste container storage expansion areas were noted. The site map in Chapter i should show both current waste container storage areas and planned (over the next 10 years) expansion areas.
SPC Response AA-2 Figure !!-10.1 has been revised to show current and planned future waste storage areas.
NRC Reauest AA-3 The locations of all waste handling processes should be indicated in Table 1-1.1 and should be shown on the Chapter I site map.
SPC Response AA-3 Figure 11-10.1 has been revised to show areas where waste is packaged and inspected. In addition, Table 1-1.1 has been revised to indicate waste sorting, packaging and inspection as an authorized activity in the UO building.
2 NRC Reauest AA4 Section 2.1 should state that the Health Physics Component is responsible for review and audit of the environmental program, in accordance with License Amendment 21 issued November 1,1993.
SPC Response AA-4 Section 2.1.17 (item 5) of the April 24,1994 revision states that the Health Physics Component is responsible for " performing compliance inspections". Section 2.6.5 1
'e further elaborates on these inspections. SPC feels that the commitment is adequate as written.
NRC Recuest AA-5 This section (6.4.2) should be updated to reflect the current waste management engineering plan. Other waste management processes, including compaction, ash leaching, and recovery of uranium from HEPA filters should be described.
SPC Response AA-5 Section 6.4.2 has been updated to describe SPC's current and future waste management plans as described in its waste management engineering plan.
NRC Reauest AA-6 Section 7.1 should reference the April 1993 version of the NRC Decontamination Guidelines. The only change to the May 1987 version is that the final survey report should be sent to the Division of Fuel Cycle Safety and Safeguards.
SPC Response AA-6 The April 1993 version of the NRC Decontamination Guidelines was referenced in the Decommissioning Funding Plan submitted on August 15,1994.
NRC Reauest AA-7 Section 7.2 and Table I-7.1 should be updated to show the current decommissioning cost estimate. This cost estimate should include the cost to dispose of the onsite waste inventory, as instructed in NRC letter dated March 24, 1994.
SPC Response AA-7 Chapter 7 " Decommissioning" was revised and submitted on August 15,1994.
NRC Reauest AA-8 Section 9.1 should be updated to show the current ownership of SPC. The section indicates that 11.2 percent of the common stock of SPC is owned by i
Siemens AG; however, recent discussions associated with the waste container inventory have indicated that SPC is wholly-owned by Siemens KWU, Inc, which is a wholly-owned subsidiary of Siemens Corporation.
In addition, the Consolidated Financial Statements of Siemens Corporation, September 30,1993 j
and 1992, indicate (page 3, item 3) that "On October 1,1992, Siemens AG transferred the remaining 11 percent interest of a subsidiary of the Company."
Does this indicate the transfer of 11 percent of interest in SPC from Siemens AG to Siemens Corporation? How does this transfer affect the ownership of Siemens Power Corporation by Siemens KWU?
l 2
)
SEP 15 '94 03:56Pf1 P.2
)
i SPC Response M-4 Sections 1.1 and 9.1 have been revised tc update the ownership ctatus of SPC.
NRC Raouest M-9 The third paragraph of Section 9.1 should be updated to match Figure 1-2.1, which shows that the Manager, Safety, Securtty, and Ucensing, reports to the Richland Plant Manager and t.
9e Manager, Product Mechanical Engineering, reports to the vice President, Ei,.oring.
SPC Response l
M-9 Figure 1-2.1 and the third paragraph of Section 9.1 are not equivalent. Figure h2.1 shows the relationehlp of organizations with significant licensing or safety responsibility. The third paragraph cf Section 9.1 is simply descriptive of what i
organizations handle radioactive material, but is not intended to describe orgenizational relationships.
NRC Reauest M 10 Correct Section 9.4 to reflect the current metropolitan area population.
SPC Rannonna AA-t o Scotlon 9.4 has been updated to reflect the current population estirrate for the 1
metropolitan area (Benton and Frank!!n counties).
NRC Reauest M 11 Figures ll-9.2, li-i.3, th9.4, and 11-9.5 should be updated to reflect current conditions, and the source of the land use information on Figures 11-9.3,11-9.4, and 11-9.5 should be identified.
SPC Response M 11 Figure ll-0.2 has been updated. Figures 11-9 3.11-9.4 and 11-9.6 are still adequate.
SPC has not yet been able to conclus!vely identify the source of these figures.
We are, however, virtually cenaln they came from a Battet!e Pacific Northwest Laboratory land use survey document.
NRC Reauest AA 12 Correct the 1994 references in Section 10.1.2 and Figure ll-10.3.
SPC Emaponse M-12 Section 10.3.2 has been updated to 1994 and the figure reference has been corrected to 1110.3.
W^ b 3
W PAGE - O
NRC Reauest i
i AA-13 Figure 11-10.1 should be updated to show the current facility site plan, including the laboratory expansion in the UO Building. Figures ll-10.3 and 11-10.14 should 2
also be updated to show the new laboratories.
i SPC Response AA-13 Figures 11-10.1,11-10.3, and 11-10.14 have been updated for the current site, including the laboratory addition to the UO2 Building. These figures were submitted to the NRC May 18,1994.
NRC Reauest AA-14 The description of HVAC systems in Section 10.4 should include K-50, serving the SWUR room; K-52, serving Building 9; K-47, serving the ARF; and K-56, serving the gadolinia scrap recovery process in the ELO Building. These stacks are sampled and monitored, as indicated in Table I of the Environmental Report Supplement, and should be d3 scribed in Chapter 10.
SPC Response AA-14 Section 10.3 has been revised to describe the SWUR HVAC systems (see response to ENV-19). The K-56 system is not yet in place and will be described with the Gd scrap recovery amendment application (currently on hold). The K-52 system serves Building 9 which is licensed under Washington State Radioactive Materials Ucense WN-1062-1. Section 10.3 has been revised to describe the HVAC system for the Ammonia Recovery Facility.
NRC Reauest AA-15 Section 10.4.2 states that facilities are provided to load out excess ammonium hydroxide for sale offsite. If NH 0H is sold offsite, the licensee should propose 4
a limit on the amount of uranium present in the ammonium hydroxide that is sold.
SPC Response AA-15 SPC has not sold NH 0H offsite for a number of years. Prior to any future sale, 4
SPC will obtain NRC's agreement on a release limit for uranium in the NH 0H.
4 NRC Reauest AA-16 Section 10.4.3 should indicate the approximate flow rate of lagoon contents through the ion exchange process.
SPC Response AA-16 Section 10.4.3 has been revised to indicate a typical flow rate through the lagoon IX of 20 gpm.
4
1 NRC Reauest AA-17 Correct the dissolver vessel tank numbers of Table 11-10.2 to match the corresponding tanks on Figure 11-10.30. Show the location of the lagoon waste feed pump, P-658, on Figure 11-10.30.
SPC Response AA-17 Table 11-10.2 has been revised te make the dissolver tank numbers consistent with those on Figure 11-10.30. The location of the lagoon waste feed pump, P-658, has been added to Figure 11-10.30.
NRC Reauest 4
AA-18 Update Section 10.4.4 to include the relocation of the LUR centrifuge to the ELO Building, as described in the amendment application for the modifications to the GSUR process in the ELO Building.
SPC Response AA-18 Section 10.4.4 will be updated for relocation of the LUR Centrifuge when the Gd scrap amendment (currently on hold) is completed.
NRC Reauest AA-19 Section 10.4.4 should include the liquid waste treatment capacity of the LUR. This capacity should be expressed in batches per time period, assuming that each batch is 6,000 gallons, wnich is the volume of a precipitator tank.
1 SPC Response t
AA-19 Section 10.4.4 has been revised to state that the throughput of LUR is 10,000 gallons per day.
NRC Reauest AA-20 Section 10.4.5 should be updated to describe the disposal of uranium-contaminated sand at the U.S. Ecology disposal facility, as authorized in July 1990 by Amendment 9 to the U.S. Ecology special nuclear material license.
AA-21 The last sentence of Section 10.4.5 states that the greater than 20 mesh material is combined with the greater than 3/8-inch material, washed in a small cement mixer, and discarded as ground cover in the lagoon area. If this practice is being used, the license should include criteria for the discard of these solids on site and a sampling and analysis program to assure that the discarded solids meet the established criteria.
AA-22 Section 10.4.5 should also clarify that the screening operations take place in the leach pit.
l 5
i l
SPC Response AA-20 SPC's previous method of using the sand trench / leach pit for treatment of lagoon AA-21 solids is no longer in use. Section 10.4.5 has been revised to discuss the l
AA-22 disposal of uranium contaminated sand at the U.S. Ecology low level radioactive disposal site and to describe SPC's future program for processing lagoon solids.
?
i i
i l
I i
5 i
6
)
\\
l SIEMENS POWER CORPORATION'S RESPONSES TO NCR'S REQUEST FOR ADDITIONAL INFORMATION ON LICENSE RENEWAL (TAC NO. L21656)
ENVIRONMENTAL PROTECTION PROGRAM NRC Reauest ENV-1 Section 5.1.1 should identify the major radionuclides, and their chemical forms, that can be emitted in gaseous effluents.
SPC Response ENV-1 The major radionuclides that can be emitted with gaseous effluents are those in low enriched uranium; i.e., U-235, U-238, and a very small amount of U-234. The chemical forms of these radionuclides include UF, UO F, UO (NO )2, UO and e
22 2
3 2
U 0. Section 5.1.1 has been revised to add this Information.
3s NRC Reauest ENV-2 Chapter 5 should include a map that identifies the locations of all exhaust stacks that are monitored for gaseous emissions. This map should include all monitored ventilating systems described in Chapter 10.
i SPC Response ENV-2 Figure 1.1 in EMF-32, " Emergency Plan and Procedures", which is part of the license application by incorporation, shows the location of all exhaust stacks, except that for the analytical lab addition. This figure will be updated in the near future. For ease of drawing control, SPC prefers not to duplicate this figure in the license application. To facilitate locating the Emergency Plan figure, Section 5.1.1 in the license application has been amended to refer to it.
NRC Reauest ENV-3 Footnote 1 in Table I-5.1 indicates that action levels are calculated boundary concentrations based on individual stack concentration. This footnote or the text of Section 5.1.1 should indicate the method used to calculate the boundary concentrations.
SPC Response ENV-3 SPC uses version 1.0 (December 1990) of "TSCREEN - A Model for Screening Toxic Air Pollutant Concentrations" to calculate boundary concentrations. SPC prefers not to add this fact to Table l-5.1 in order to be able to change methods without having to apply for a license amendment.
7 1
l NRC Reauest ENV-4 To demonstrate compliance with 40 CFR 190.10, the action levels specified in Table I-5.2 indicate that action will be taken if the total emissions exceed 25 Ci alpha per calendar quarter.
Provide an up-to-date pathway analysis to demonstrate that this action level will maintain the annual dose equivalent to any member of the public within the limits specified in 40 CFR 190.10(a). This should be provided in the ALARA Report.
SPC Response ENV-4 Based on the boundary concentrations calculated by TSCREEN, SPC uses the COMPLY Code to estimate the dose to a person living and producing his own vegetables, milk and meat at the plant boundary. The effective dese equivalent for such a person is 0.4 mrem /yr or 2% of the 40 CFR 190.10 limit. This data will be presented in the 1994 ALARA Report.
NRC Reauest ENV-5 Section 5.1.2.2 indicates that process cooling wastewater may be disposed of by discharge to a seepage pond. If seepage ponds are used for this purpose, their locations should be shown on Figure 11-10.1. If not, the reference to seepage ponds should be removed from this section.
i SPC Response ENV-5 The reference to seepage ponds has been deleted from Section 5.1.2.2 as SPC does not use them.
NRC Reauest ENV-6 Section 5.2.1 specifies the frequency of sampling and analysis of air, soil, vegetation, and ground-water samples and the parameters for which these media are monitored. Sampling and analytical methods for each medium and each analytical parameter should also be specified.
SPC Response ENV-6 The sampling and analytical methods for the applicable parameters in soil, air, forage and ground-water are:
Ambient Air
- Sampling Method SPC Procedure EMF-1507,6.6 )
1
- Analytical Method-F F-ion Specific Electrode Forage
- Sampling Method SPC Procedure EMF-1507,6.5')
Analytical Method-F AOAC-F-Potentiometric 8
i 4
SPC Procedure EMF-1507,6.4 )
1 Soil Sampling Method Analytical Method-U Kinetic Phosphorescence Analysis SPC Procedure EMF-1507,6.11 )
1 Ground-water - Sampling Method
- Analytical Methods Cl EPA 300.0 NO(N)
EPA 300.0 3
NH (N)
EPA 350.3 3
pH Field instrument Gross alpha -
EPA 900.0 Gross beta EPA 900.0 L
1)
These procedures are located in SPC document EMF-1507, "Siemens Power Corporation-Nuclear Division Health Physics and Radiological Safety Procedures Manual."
i SPC prefers not to add this information to Section 5.2.1 in order to preserve its flexibility to change methods without having to apply for a license amendment.
SPC will, however, continue to use methods that are recognized and appropriately validated.
NRC Reauest ENV-7 Section 5.2.2 should be revised to include the additional analytical parameters in Safety Condition S-5 b of the current license,in accordance with Amendment 18 issued September 3,1993. These additional indicator parameters are chloride, l
nitrate-nitrogen, ammonia-nitrogen, and pH, in addition to gross alpha / beta.
SPC Response ENV-7 Other constituents to be measured and other wells to be sannpled from Amendment 18 have been added to Section 5.2.2.
NRC Reauest ENV-8 Figure I-5.2 should be revised to show test wells TW-6, TW-7, and M-21, in accordance with Amendment 18 to the current license.
SPC Response l
ENV-8 Figure I-5.2 has been revised to show test wells TW-6, TW-7 and TW-21.
NRC Reauest ENV-9 Section 5.2.2 should include LLDs for chloride, ammonia, and nitrate in the ground-water samples.
9
~
l SPC Response
)
ENV-9 The lower limit of detection (LLD) for chloride, ammonia and nitrate are 0.10 mg/l, 0.05 mg/l, and 0.10 mg/l, respectively. SPC does not believe these values are or should be license conditions and therefore has not listed them in Chapter 5.
NRC Reauest ENV-10 in Section 5.2.1, analytical LLDs should also be specified for uranium in the soil l
samples at stations 1 and 9, for fluoride in the air samples at stations 3 and 4, and for fluoride in vegetationhorage samples at stations 5 and 6.
SPC Response i
ENV-10 The LLDs for uranium in soil samples, fluoride in air samples, and fluoride in vegetation / forage samples are 0.5 g/g, 5 pg, and 10 pg/g, respectively. SPC does not believe that these values are or should be license conditions and therefore has not listed them in Chapter 5.
t NRC Reauest ENV 11 Section 1.4 of Appendix A to the Supplement to Applicant's Environmental Report states that soil samples are taken from between 1 cm and 5 cm beneath the l
surface of the topsoil, indicating that the topmost 1 cm of soil is removed before the samples are taken. Since uranium would be deposited on the soil surface, the soil sampling procedure should include soils from the top 1 cm. The sampling method should specify that the soil samples will include the topmost layers of soil.
SPC Response r
ENV-11 Section 1.4 in Revision 4 of EMF-14, " Supplement to Applicant's Environmental Report", sent to the NRC on July 25,1994, now states that soil samples are collected "...from the surface to between 1 cm and 5 cm beneath the surface of the topsoil."
NRC Reauest ENV-12 Radioactivity action levels are presented in Table I-5.1 for authorized liquid and gaseous emissions. Action levels should be proposed for the detection of liquid between the lagoon liners, as described in Section 5.1.3, and for increases in analytical parameters in groundwater above background levels.
SPC Response ENV-12 Action levels for the lagoon liner leak detection system are specified in Plant Operations Standard Operating Procedure P66,350. They are:
Lagoon 1 9,120 gallons / month Lagoon 2 6,566 gallons / month 10
Lagoon 3 8,500 gallons / month Lagoon 4 10,944 gallons / month Lagoon 5A 11,674 gallons / month Lagoon 5B 8,755 gallons / month Formally established groundwater action levels are cunently being developed as part of SPC's Part B Application to the Washington State Department of Ecology for its Dangerous Waste Final Facility Permit. They should be available for submission to the NRC by October 31,1994.
NRC Reauest ENV-13 Environmental air sampling station 4, shown on Figure I-5.1, has been moved from ESE of the plant to SSE and almost twice as far away. The Supplemental Environmental Report should state when and why this sampling station was moved, and discuss what effect this relocation has on the quality and comparability of fluoride data between the previous and current locations.
SPC Response ENV-13 in fact, the location of air sampling station 4 has not been changed. Figures 1-5.1 in the existing license and the renewal application were simply drawn to different scales.
NRC Reauest ENV-14 Ambient air sampling station 3 is located directly east of the plant, and station 4 is south-southeast. The air sampling program should include a provision that ambient air sampling will be performed on a day when the wind is from the west or the northwest.
SPC Response ENV-14 Ambient air at sampling stations 3 (prevailing wind) and 4 (toward Richland) is continuously sampled so there is no need to schedule sample taking for any particular day.
NRC Reauest ENV-15 Section 13.2 should discuss stack monitoring for fission products, as indicated in the ALARA report and the Environmental Report Supplement.
1 SPC Response ENV-15 The Fuel Services Building, which contains equipment used in irradiated fuel examinations and that is contaminated with fission and activation products is l
regulated under Washington State Radiation Materials Ucense WN-1062-1. For i
informational purposes a paragraph discussing the filtration and gaseous effluent sampling has been added to Section 13.2.
11
l i
i l
NRC Reauest ENV-16 Section 13.1 states that the background radiation level in this part of Washington State is approximately 60 mrem / year. The source of this formation should be provided or the background location where it is measured should be shown on Figure I-5.1 in Chapter 5.
SPC Response ENV-16 For the five year period 1987-1991 the background radiation level in this part of Washington averaged 77 mrem /yr according to the June 1992 "Hanford Site i
Environmental Report", PNL-8148/UC-602. Section 13.1 has been revised to "approximately 80 mrem /yr".
NRC Reauest ENV-17 Section 13.5 states that off-site sampling for uranium in soil has shown that 4
analyses are consistently at or below background. This background level should be provided, and the location of the background sampling station should be shown on Figure I-5.1, or the source of the background value should be provided.
SPC Response ENV-17 On July 7,1993 Section 13.5 was revised to state that "off-site sampling for uranium in the soil has shown no increasing trend in the concentration during this latest five year license period." It currently makes no statement regarding background concentrations.
i NRC Reauest ENV-18 The 1991 A. ARA Report includes, in Table 1, a column headed FP, fission products. Monitoring for fission products should be included in the environmental l
program, and the monitoring data should be reported in the semiannual effluent reports and included in the Environmental Report Supplement.
SPC Response ENV-18 The environmental monitoring program discussed in Chapter 4, Environmental Standards" of SPC's safety manual EMF-30, includes fission product monitoring.
Fission product monitoring is included in both the semiannual effluent reports and the environmental report supplement generated by SPC.
Fission product monitoring has been discussed in Section 13.2 in the answer to ENV-15.
i NRC Reauest ENV-19 Section 15.2.1.4.2 or 15.2.3 should include the SWUR Room exhaust stack K-50, if it is still in use (listed in Environmental Report Supplement Table 1).
12
f
$PC Response ENV 19 A discussion of both the K-50 (incinerator) and K-55 (Incinerator cooling shroud) j systems has been added to Section 15.2.1.4.2 and 10.3.2.
i i
i
(
I l
i i
I i
l t
i I
?
f i
13
SIEMENS POWER CORPORATION'S RESPONSES TO NCR'S REQUEST FOR ADDITIONAL INFORMATION ON LICENSE RENEWAL (TAC NO. L21656)
RADIATION PROTECTION PROGRAM NRC Reauest RP-1 Section 1.6.4 - Supply a justification for this exemption.
SPC Response RP-1 The request for tha exemption in Section 1.6.4 was granted to Siemens in our present license and many fuel fabricators have similar exemptions. There are literally thousands of packages containing uranium on site. Nearly all of these packages contain information on the contents (e.g. weight, enrichment, etc.) in the form of follower cards. Putting on and removing labels from each would entail a great deal of manpower, and subsequently result in higher doses. An entrance warning accomplishes the same purpose without unnecessary exposure.
NRC Reauest RP-2 Section 1.6.5 - For NRC to consider this request, SPC must provide all the requested information in 10 CFR 20.2002(a), (b), (c), and (d).
i SPC Response f
RP-2 The request has already been granted and SPC only wants its continuance. The requirements under which it was granted in the NRC letter of May 20,1982 (10 CFR 20.302) are not materially different from those in effect today (10 CFR 20.2002). 30 pCi/gm is a standard currently being used for decommissioning.
NRC Reauest RP-3 Section 1.6.8 -The cited document needs to be updated to the April 1993 version.
.SPC Response RP-3 Section 1.6.8 has been revised to reference the April 1993 version of the NRC decommissioning guidelines.
NRC Reauest RP-4 Section 1.6.10 - The exemption to 20.2203 should be explained and justified.
SPC Response RP-4 SPC withdraws its request for this exemption.
i 14 l
NRC Reauest RP-5 Section 2.1.7 - One of the responsibilities of the Manager, Plant Engineering, is to maintain and calibrate radiation protection instruments and equipment and criticality accident alarm system. Explain how this function is coordinated with the Manager, Safety, Security, and Ucensing.
SPC Response RP-5 Calibration of health physics instruments is placed on a PM schedule in the Maintenance Engineering (" Maintenance") computer system.
Each month Maintenance sends to the Health and Safety Technicians (HSTs), who work for the Manager, Safety, Security and Ucensing, a list of instruments that require calibration. Upon calibration, Maintenance places a sticker on the instrument designating the calibration and indicating when the next calibration is due. After calibration, the instruments are returned to the HSTs. Maintenance informs the HSTs of any instrument not turned in. Instruments not operating properly are turned into Maintenance for repair. Calibrations are required following any significant repair.
NRC Reauest RP-6 Section 2.1.17 (a)
What is the current staffing level and the number of authorized positions?
(b)
What is the turnover rate for the health physics component, and how does it compare to the plant-wide turnover rate?
(c)
Provide a copy of the Radiation Protection standard section of EMF-30.
(d)
The definition and structure of the Health Physics Component should be described.
SPC Response RP-6 The Health Physics Component described in 2.1.17 consists of one Health Physicist, who has been in this job since July of 1992. He is assisted by assignment of normally two of the senior Health and Safety Technicians.
There have been three Health Physicists at SPC in the last 10 years. The turnover rate for permanent plant employees is about 3 percent per year.
A copy of the Radiation Protection Standards (Chapter 2) of EMF-30 is enclosed for your information.
The organization chart shown in Figure 1-2.2 shows how the Health Physics Component fits in Safety, Security and Ucensing. The responsibilities of the Health Physicist are the same for those for the Health Physics Component.
Section 2.1.17 of the application has been revised to reflect this fact.
l 15
NRC Recuett RP-7 Section 2.2.5 - Some of the experience in radiation safety should be in fuel cycle facilities.
SPC Response RP-7 It is desirable, but not always possible to find a candidate with fuel cycle facility experience. Section 2.2.5 has been revised to include this preference.
NRC Reauest RP-8 Section 2.2.8 - Establish minimum qualifications for the Health Physicist Specialist.
SPC Response RP-8 The qualification statement in Section 2.2.8 has been revised to describe the qualifications required for the Health Physicist. As described in the response to RP-6, the Health Physicist is normally assisted by senior Health and Safety Technicians.
NRC Recuest RP-9 Section 2.2.9 - Describe the Health and Safety Technician Specialist environmental monitoring training and qualification program.
SPC Response RP-9 The Health and Safety Technician Specialist is the top level of the HST progression (excluding supervisor). The distinction between the Health and Safety Technician Specialist and Health and Safety Technician is that the Specialist either has lead responsibilities or performs specialized functions. The training program for Health and Safety Technicians, including Health and Safety Technician Specialists, consists of both classwork and hands-on training. The subjects covered are dependent on the technician's level. Additionally, there are certain experience requirements for all levels above trainee. The training program is described in SPC document EMF-1507, " Health Physicist and Radiation Safety Procedures Manual", Procedure 1.3, a copy of which is enclosed for your information.
Section 2.2.9 has been revised to better indicate the focus of the training program.
NRC Reauest RP-10 Section 2.2.10 - Describe how the Health and Safety Technician and the Health and Safety Technician Specialist positions differ.
SPC Response RP-10 Refer to the answer in RP-9.
16
NRC Reauest RP-11 Section 2.7 - Clarify the meaning of subitems 2.7(3) and 2.7(4).
SPC Resoonse RP-11 Originally those items meant that all items required to be reported to the NRC regarding radiation protection will be reported, with the exception of the items exempted in Section 1.6.10. SPC has withdrawn its exemption request in 1.6.10 and Section 2.7 has been revised to reflect this change.
NRC Reauest RP-12 Section 3.1.2 - Reference the internal procedure for preparing RWP's.
SPC Response RF-12 There is no specific procedure for writing Radiation Work Procedures (RWPs).
The approval sequence and contents are described both in Section 3.1.2 and in the Radiation Work Procedures Manual, EMF-897.
NRC Reauest RP-13 Section 3.2.1 - This area should be redefined in accordance with the new Part 20.
SPC Response RP-13 Sections 3.2.1 and 3.2.1.1 have been revised to define SPC's controlled and restricted area in accordance with 10 CFR 20.
NRC Reauest RP-14 Section 3.2.5 - Is the reference to 10 CFR 20.1201(b) correct or should it be 10 CFR 20.1206?
SPC Response RP-14 Section 3.2.5 has been revised to reference 10 CFR 20.1206.
NRC Reauest RP-15 in addition to the 1991 ALARA Report, copies of the 1990 and 1992 reports would I
be helpful.
SPC Response RP-15 Copies of the 1990 and 1992 ALARA Reports are provided for your information.
The 1992 Alara Report is proprietary. An affidavit attesting to that fact is enclosed.
l I
17
=
- l f
NRC Reauest i
i RP-16 Chapter 12 should be completely revised to demonstrate how the radiation protection program is designed to comply with 10 CFR Part 20 and ALARA.
j SPC Response r
RP-16 A revised Chapter 12 is included.
e f
l l
I t
-i h
I h
l a
i t
6 v
[
i i
i f
18 i
SIEMENS POWER CORPORATION'S RESPONSES TO NCR'S REQUEST FOR ADDITIONAL INFORMATION ON LICENSE RENEWAL (TAC NO. L21656)
EMERGENCY Pl.AN AND PROCEDURES (EMF-32)
NRC Reauest EP-1 Section 1 should include an enlarged map of the facility.
SPC Response EP-1 The facility map in Section 1 is 11"x17"in size, folded to 81/2"x11" as required by Reg. Guide 3.67. However, we can provide a larger sized map separately to NRC Headquarters and Region IV. Please let us know the size of the map you require.
NRC Reauest EP-2 Section 5.1 should describe the means to authenticate activation of the emergency response organization.
SPC Response EP-2 Section 5.1 will be revised to read:
"5.1 Activation of Emergency Response Organization Communication steps to notify and activate emergency personnel under each classification of emergency are indicated in this section. In addition, PERT personnel will be dispatched to any incident site, as soon as practicable, to assess the emergency and determine the validity of the initial report and to recommend the continuation of the emergency activation or the standing down of the Emergency Response Organization."
NRC Reauest EP-3 Section 5 should include criteria for requiring shut down of all or part of the facility and the approximate time required.
SPC Response EP-3 Section 5.3 will be revised to read:
"5.3 Mitigating Actions in all postulated accidents, the decision to shut down a single process or the entire facility will be made early in the response to the emergency by PERMT.
Existing Standard Operating Procedures for the safe orderly shutdown of equipment will be utilized by Operations personnelto perform this function. Most process equipment can be totally shut down within three hours, high temperature 19
calciners and furnaces within nine hours with several days needed for total cool down. However in all cases the facility was designed and built to be able to be evacuated with the equipment continuing to operate in a safe condition with no additional operator actions needed until safe reentry procedures can be utilized.
Additional mitigating actions for the postulated accidents are summarized below."
NRC Reauest EP-4 Section 5.4.1.1 should describe in greater detail the procedures for accounting for plant personnel and visitors.
SPC Response EP-4 The last pars. graph of Section 5.4.1.1 will be revised to read:
" Provisions for determinina and maintainina accountability: Accountability is
~
determined after building evacuation by the following procedure:
After all personnel have evacuated the building, PERT personnel will be dispatched to enter the building and ensure that all personnel have evacuated, then the exterior exit doors will be observed to prevent re-entry by unauthorized personnel.
Accountability is determined after site evacuation by the following procedure:
l l
When all personnel have assembled in the site evacuation assembly area, i
accountability of company employees is the responsibility of the department managers and is conducted using implementing procedures under the authority of the Staging Area Supervisor and Staging Area and Accountability Monitors with completion of a Personnel Accountability Check-Off List. All visitors and off-shift employees are required by security procedures to sign in and out on current..gn in rosters which will be recovered and used by the Staging Area Supervisor to account for all visitors and off-shift employees. All exterior fence gates will be shut to prevent re-entry by unauthorized personnel."
The revisione, described above, to SPC's emergency plan, EMF.32, will be made in the next general revision, expected by October 31,1994.
i 20 i
SIEMENS POWER CORPORATION'S RESPONSES TO NCR'S REQUEST FOR ADDITIONAL INFORMATION ON LICENSE RENEWAL (TAC NO. L21656)
NUCLEAR CRITICALITY SAFETY PROGRAM NRC Reauest NCS-1 in Section 4.1 numerous references are made to other sections of the application that either do not exist or that contain material other than that indicated. (For example, Section 4.1.1 references Sections 2.1.11, 2.1.13, and 2.2.5). Please correct the references, i
SPC Responso NCS-1 The references have been changed and a revised Chapter 4 has been provided with the application.
NRC Reauest l
NCS-2 include a complete description of the initiating events, accident pathways, margin to criticality, and the necessary NCS controls to satisfy the double contingency principle for each process in the facility. To illustrate the scope of the material that NRC is requesting to demonstrate NCS, the following example is provided:
The application describes a vaporization chest, in which a single cylinder is electrically heated to vaporize the UF solid, and a favorable geometric s
scrubber system. In addition, information is presented which indicates that there is a flow control valve between the hydrolysis system that is activated by a temperature-and pressure-control interlock system to prevent backflow of UF into the vaporization chest. Finally, the application states 6
that the ADU precipitation tanks are safe by geometry and that there is an accompanying POG system.
This description of the process and criticality safety analysis (CSA) is not sufficiently detailed to enable the staff to determine the adequacy of the safety of the system. For example, the application does not address the possible accident scenario in which the vaporization chest fills up with uranyl fluoride due to a break in the UF cylinder valve or through the 6
scrubber system.
This deficiency in the CSA was also noted in inspection report 92-08. In this inspection report Y was reported that uranyl fluoride was found in the vaporization chest N result of an accident sequence that eventually led to uranyl fluoride es.. ing via the process offgas system. This accident scenario, and others, should be clearly identified in the application for the staff to verify that the double contingency principle is actually being applied for all accident sequences in all processes.
21
4 Moreover, the application only vaguely describes the controls that are utilized to preclude the possible backflow of UF, from the hydrolysis system into the UF cylinder. Thus, it is unclear how the double s
contingency principle is satisfied. The application should, therefore, list the two unlikely failures that are required before a criticality is possible.
Finally, the information for the remainder of the ADU conversion process does not address any of the issues noted above. That is, there is no description of the equipment or discussion of accident pathways, and the description of the controls utilized to fulfill the double contingency is j
vague. Furthermore, additionalinformation is needed to demonstrate how j
a favorable geometric unit alone satisfies the double contingency principle.
r The entire application should be revised in an analogous manner in order for the staff to review the criticality portion. That is, for each system / process, the application should include a complete description of the initiating events, accident pathways, margin to criticality, and the necessary NCS controls to satisfy the double contingency principle for each respective process in the facility.
SPC Response NCS-2 The request to identify all initiating events, all accident pathways, and all controls used for criticality safety, and the margin to criticality for each process in the facility is excessive. SPC firmly believes that the criticality safety analyses (CSAs),
and not the license, are the proper location of this process-specific criticality safety information, both from an operational and a regulatory standpoint. The CSAs are living documents that are created as new systems / processes are added to the plant and that are revised as existing systems are modified. SPC's CSAs are available for NRC inspection and review at its Richland facility. If NRC determines that SPC's CSAs are deficient, individually or categorically, it has ample avenues to effect change, whether by licensing actions, inspection activities, or technical interaction. The current SPC Criticality Safety Analysis i
Update Program is a case in point.
4 SPC's commitments relative to its nuclear criticality safety program are provided in Chapters 4 and 14 of the license.
Chapter 4 establishes the basic l
requirements, practices, and technical bases which underlie the program. Chapter 14 outlines the administrative practices, technical assumptions and programmatic requirements that are utilized to assure that the commitments in Chapter 4 are effectively and consistently met.
SPC establishes its criticality safety program at the operational level via the Nuclear Criticality Safety Standards within EMF-30, the SPC Safety Manual. EMF-30 requires that the criticality safety of nuclear material processing / handling systems be analyzed and controlled via a system of CSAs, criticality safety specifications (CSSs), and criticality safety limit cards (CSLCs). The CSA contains the system-specific criticality safety information such as initiating events, accident pathways, margin to criticality, and necessary nuclear criticality safety controls to satisfy the double contingency principle for that system.
22
.m l
.i Furthermore, in January,1993, SPC committed to an extensive program to update.
j the CSAs for this facility. This update is currently in progress. All of the existing
.l CSAs have been reviewed and grouped into process systems. Since that time an j
extensive effort has been expended to review the existing analyses, to update j
facility drawings and piping and instrument diagrams, and to complete accident analyses by a multi-disciplined team of experienced personnel familiar with the i
process systems. All of these activities are necessary inputs to a CSA. SPC has committed to completing updated CSAs for the 32 highest priority of the 65
~
process systems by early 1995; seven systems have been completed to date.
The remaining 33 process system CSAs will be reformatted to enhance process and accident descriptions and centrols following completion of the first 32.'
l i
For the first seven systems completed, these documents are 50 to 80 pages long.
l if the detail the NRC has requested is to be included in the application, then nearly half the material included in a CSA, up to 2,600 pages of information, would be entirely duplicated in SPC's license application.
4 Each CSA is subject to the revisions that are an ongoing necessity in the operation of a nuclear fuel fabrication facility. If overly detailed information is a part of the license, each of these modifications will of necessity be accompanied i
by an attendant change to the license. For example, inclusion of specific margins to criticality in the license will necessitate a license revision when a plant
[
modification changes the worst case K-eff of a system from 0.92 to 0.93, both of which are substantially below the margin allowed by the current license. Ukewise, any physical modification to a system that creates a new initiating event or j
modifies an accident pathway would require a license revision. The administrative.
burden required to keep this volume of information consistent.would be exhaustive and would do nothing to improve plant safety.
j r
However, some information obtained from this update process may.well be f
appropriately included in chapter 15 of the license application to make it more complete.- As CSAs for each process system are completed, SPC Licensing staff will review the information available and, as appropriate, update chapter 15 of the application. The first update to Chapter 15 should be available by the end of October.
t NRC Reauest I
NCS-3 The assignment of responsibility for initiation, review, and approval of NCS
(
documents and limits should be clarified and consistent. The responsibilities i
defined in Figure 1-2.3 are not consistent with those in the text. The Figure (December 21,1992, version) states that NCS standards are prepared by the Manager of Regulatory Compliance and concurred by the Manager of Safety, Security, and Ucensing. Section 4.1.2 states that criticality safety standards (presumably the same as the NCS standards) are prepared by the Manager of Safety, Security, and Licensing.
l l
f 23 j
SPC Response NCS-3 The responsible parties for preparation, review, and approval of NCS documentation are listed on the responsibility matrix in Chapter 2 of the application. Where appropriate, this matrix will be referenced throughout the remainder of the document.
NRC Reauest NCS-4 it appears from the application that the Safety Supervisor could be the immediate supervisor of the Criticality Safety Specialist because there is not an identified supervisor for the Criticality Safety Component. Are there any criticality safety training requirements, in addition to the basic criticality training which is described in Section 2.4, for the Safety Supervisor?
SPC Response NCS-4 The Safety Supervisor is,indeed, the supervisor of the Criticality Safety Specialists.
The Safety Supervisor has the educational background and experience required l
by the " Personnel Education and Experience Requirements"in Chapter 2. These education and experience requirements allow a person to adequately supervise a technical staff of experienced criticality safety specialists. An assigned Criticality Safety Team Lead reports directly to the Supervisor of Safety. The Team Lead provides technical oversight and direction for the Criticality Safety Specialists.
NRC Reauest NCS-5 The application should clearly define the terms that are used (e.g., nuclear criticality (i) criteria, (ii) safety analysis, (iii) safety standards, (iv) specifications, and (v) limits) and show how they relate to each other.
SPC Response NCS-5 Chapter 4 of the application has been revised to more clearly define key programmatic eierrents of the Criticesity Safety Program and to clarify how these elements relate to each other. Terrrs that could be confused have been defined.
24
SIEMENS POWER CORPORATION'S RESPONSES TO NCR'S REQUEST FOR ADDITIONAL INFORMATION ON LICENSE RENEWAL (TAC NO. L21656)
FIRE PROTECTION PROGRAM NRC Recuest FP-1 Section ll.2 of the NRC's " Technical Position on Fire Protection for Fuel Cycle Facilities" dated August 10, 1992, 57 Federal Reaister. pages 35607-35613, discusses the need to have a fire hazards analysis performed on fuel cycle facilities already in operation. The purpose of this fire hazards analysis is to
" reveal fire protection weaknesses or to confirm the adequacy of the protection i
measures." Section 111.9 of the guidance provides additional details on the expected contents of a fire hazards analysis. While a detailed emergency plan, including pre-fire plans does exist, no analysis appears to have been performed which addresses credible fire scenarios among other accident scenarios for determining off-site consequences. A thorough review of fire scenarios in and around the facility does not appear to have been performed per the NRC guidance. Provide a fire hazards analysis which provides information as detailed in Section 111.9 of the aforementioned NRC guidance.
SPC Response FP 1 SPC has performed an analysis which looks at credible fire scenarios among other accident scenarios for determining off-site consequences. This analysis is documented in the Emergency Assessment Resource Manual (EARM) which has previously been provided to NRC Headquarters when the Emergency Plan EMF-32 was distributed. This manual provides off-site consequences and source terms for all of the SPC postulated accident scenarios with potentially significant off-site impacts.
I The additional information required by Section 111.9 of the NRC guidance is i
detailed in the Pre-Emergency Plan which has been previously provided to NRC Region IV.
The EARM analysis revealed two fire scenarios with potentially significant off-site I
consequences; i.e., a fire at the UF cylinder storage area and a fire in the powder s
preparation area. Although both of these scenarios are highly unlikely, the SPC emergency plan has effectively planned for them.
NRC Recuest i
FP-2 The application states "Where moderation control is in place...high expansion form, dry chemical, or CO would be used to combat a fire." Since high 2
expansion foam is comprised of foam / water solution and air, provide information which indicates that high expansion foam will not prevent a moderation concem.
i 1
25 j
t SPC Resoonse FP-2 As stated in the application, Criticality Safety Analyses typically assume optimum moderator interspersed between discrete units of fi? site material and process equipment. Where prudent, high density foam is an allowed fire-fighting option, in certain locations foam is allowed where water is excluded because it is less likely than water to get inside some of the process equipment containing unmoderated uranium. The details of fire-fighting restrictions are included in the facility Pre-Emergency Plan.
[
t P
[
f r
l l
l h
26
SiemenS Power Corporaticn - Nuclear Division EM F.2
/
'i SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 U
PART I - UCENSE CONDITIONS REv.
CHAPTER 1 STANDARD CONDITIONS AND SPECIAL AUTHORIZATIONS 1.1 Corporate information The name of the applicant is Siemens Power Corporation (SPC).
App!icant is incorporated in the State of Delaware with its principal corporate offices located at 155 108th Avenue N.E., Bellevue, Washington 98009-0777. All of the common stock of Applicant is owned by Siemens Corporation, a Delaware corporation with headquarters in New York City. Siemens Corporation is, in turn, wholly-owned by Siemens AG headquartered in Munich, Federal Republic of Germany.
The Engineering and Manufacturing Facility of the Nuclear Division of SPC, for which this application is being made, is located at 2101 Horn Rapids Road, Richland, Washington 99352-0130.
1.2 Site Location
,m The Engineering and Manufacturing Facility is located in the state of Washington, county
(
of Benton, city of Richland. It is sited on a 320-acre tract, 0.9 miles west of the intersection of Stevens Drive and Horn Rapids Road within the north boundary of the city
's-of Richland. The site is on the south side of Horn Rapids Road which provides access.
All activities with special nuclear materials are conducted within a controlled access area.
The processing of uranium compounds is conducted primarily within the UO and g
Specialty Fuels (SF) Buildings with some development activity or pilot scale work in the Engineering Laboratory. Uquid waste processing is conducted primarily in the Ammonia Recovery (AR) and Lagoon Uranium Recovery (LUR) facilities which are sited within the waste storage lagoon area. Solid waste is packaged in the various facilities and stored in containers in designated storage areas while awaiting shipment to a low level radioactive waste disposal site or incineration of the combustible waste in the SF Building.
Storage of various uranium compounds is conducted in the various production facilities or in the Packaged Radioactive Materials Warehouse, Materials Warehouse and the UNH Drum Storage Warehouse. Storage of UF cylinders is conducted outside on pads s
adjacent to the UO Building, and packaged fuel elements or other uranium products are 2
stored outside or in a warehouse while awaiting shipment.
f I
i i
l l
(3
\\
t V
mcuoutuT amcatom oart:
NE NO Se tember 12,1994 1,1 I
SPC-ND.3330 947 (R 1107/92)
l Siemens Power Corporation - Nuclear Division EMF-2 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 PART I-LICENSE CONDITIONS g y, I
1.6.6.2 Material Control and Accountino
{
l SPC shall follow the special safeguards conditions given in the Safeguards Amendment SG-2 and the NRC approved Fundamental Nuclear Material Control Plan (FNMC) submitted in accordance with 10 CFR Part 74.31(b). The NRC approved FNMC Plan is:
i EMF-12(P), " Nuclear Material Safeguards Procedures Description for the Fuels j
Fabrication Plants." This document shall be maintained in a current and approved i
status and shall be properly implemented.
i 1.6.7 Auth0.ht;0r. at Reactor Sites j
SPC is authorized to possess fuel assemblies or fuel rods at reactor sites for the purpose of loading them into shipping containers and delivering them to a carrier for transport.
i 1.6.8 Authorized Release Guidelines i
SPC is authorized to release equipment, scrap or fa.cilities for unrestricted use, or for termination of license according to the " Guidelines for Decontamination of Facilities and j
Equipment Prior to Release for Unrestricted Use or Termination of Ucenses for Byproduct, Source, or Special Nuclear Material" as published by the U.S. Nuclear Regulatory Commission dated April 1993, 1.6.9 Authorized Criticality Alarm System Outane SPC is granted an exemption from 10 CFR 70.24(a) for the purpose of performing l
maintenance on the criticality alarm system. Sections of the criticality alarm system may i
be taken out-of-service provided that all movement or processing of fissile material in i
affected areas is halted for the duration of the outage. Health and Safety Technicians j
4 shall conduct periodic surveys of the areas during the criticality alarm system outage.
I i
1.6.10 &ldhqrized Workolace Air Samplina Adiustments l
I SPC is authorized to adjust Derived Air Concentration (DAC) limits and Annual Umit of
~
Intake (ALI) values in process areas to reflect actual physical characteristics of the airborne uranium.
AMENDMENT APPLCATION DATE:
PAGE NO.:
September 12,1994 1-7 SPC.ND3330 947 (R-1/07/92)
[
i Siemens Power Corporation - Nuclear Division sue.2 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 l
t i
PART I-LICENSE CONDITIONS
- ggy, I
TABLE l-1.1 Specific Locations of Authorized Activities )
l l
1 l
Location SNM Authorized Activity SF Building Pu and PuO -UO Storage and repackaging.
2 2
contaminated waste i
UO (up to 19.99 wt%
Storage, blending, pressing, 2
U-235) sintering, fuel rod loading and I
downloading, fuel rod welding, fuel element assembly; process tests; associated quality control activities.
Uranium Compounds (up to Waste storage, sorting, 5 wt% U-235) incineration, packaging, and associated quality control activities.
l UO Building Uranium Compounds (up to All operational steps of fuel l
2 l
(including Powder 5 wt% U-235) manufacturing from UF -UO2 6
Storage) conversion to packaging l
finished fuel elements, scrap recycling and reprocessing; i
process tests; waste sorting, 1
packaging, and inspection; associated quality control I
activities.
i UO (5 to 19.99 wt% U-235)
All operational steps of fuel 2
manufacturing involving UO ;
2 including associated quality i
control activities.
j l
ELO Building Uranium Compound (up to All operational steps of fuel 19.99 wt% U-235) manufacturing involving
- uranium compounds; _
inc!uding process tests and scrap reprocessing.
PDTF Building UO (up to 5 wt% U-235)
Hydraulic flow tests involving 2
single fuel elements.
.i 1)
The locations described in this table are shown on the site plan, Figure ll-10.1
]
AMENDMENT APPLCATON DATE:
PAGE NO.:
September 12,1994 1-8 SPC-ND3330 947 (R-UO7/92) l
L i
Siemens Power Corporation - Nuclear Division sup.2 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257
.[
PART I-LICENSE CONDITIONS REv.
I TABLE l-1.1 Specific Locations of Authorized Activities (Cont.d)')
i Location SNM Authorized Activity l
Packaged Radio-Uranium Compounds (up to Storage of closed and j
active Materials 5 wt% U-235) externally free-of-significant-l Storage Building contamination containers of j
product, scrap, and waste materials; and the loading of such containers into shipping containers.
Temporary Storage Uranium Oxide (up to 5 wt%
Storage of a planar array of i
Facilities U-235) closed containers of oxide pellets which are externally free of significant contamination.
l Materials Warehouse Uranium Compounds (up to Storage of closed and O
5 wt% U-235) externally free-of-significant-contamination containers of l
product, scrap and waste materials; and the unloading of such containers from shipping containers.
UNH Drum Storage Storage of Uranyl Nitrate Storage of Uranyl Nitrate i
Warehouse solutions (up to 5 wt%
solutions in a single tier of J
U-235 and less than 140 closed 55 gallon drums free of gU/t) significant contamination.
Laundry Facility Uranium Compounds (up to Cleaning of contaminated 5 wt% U-235) protective clothing and equipment.
UF Cylinder Storage UF, (up to 5 wt% U-235)
Outside storage of UFs s
Areas cylinders (full and empty).
UO (up to 19.99 wt% U-Outside storage of fuel Packaged Fuel 2
Storage Areas 235) packed for shipment; the transport containers are closed, sealed and properly.
labeled for shipment.
1)
The locations described in this table are shown on the site plan, Figure 11-10.1 AMENDWENT APPLCATON DATE:
PAGE NO.:
September 12,1994 1-9 SPC-ND:3330.947 (R-irD7/92)
'l j
1 SiemenS Power Corporation - Nuclear Division-sur-2 i
SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 PART I-LICENSE CONDITIONS REV.
U TABLE l-1,1 Specific Locations of Authorized Activities (Cont.d) j j
Location SNM Authorized Activity i
Packaged Waste Uranium Compounds (up to Outside storage of contaminated Storage Areas 19.99 wt% U-235) materials (including low level i
waste and incinerator ash) which are packaged, sealed, labeled and externally free of contamination.
Process Chemical Uranium Compounds (up to Transfer, mixing, sampling, i
Waste Storage 5 wt% U-235) storage and solar evaporation of Lagoon System contaminated liquid wastes.
Retention Tanks Uranium Compounds (up to Interim storage of potentially 5 wt% U-235) contaminated liquid wastes.
l t
High Uranium Solids Uranium Compounds (up to Transfer of uranium bearing I
C Pond 5 wt% U-235) solids, leaching for uranium l
recovery.
l Solids Trench Uranium Compounds (up to Transfer and storage of 5 wt% U-235) contaminated solids awaiting leaching or burial.
Lagoon Uranium Uranium Compounds (up to Recovery of uranium from waste Recovery 5 wt% U-235) solutions, i
i Ammonia Recovery Uranium Compounds (up to Removal and recovery of Facility 5 wt% U-235) ammonia from uranium l
contaminated liquid wastes.
Lagoon SA IX Uranium Liquid Wastes (up to Filtration and ion exchange of
)
Process-ARF Bldg.
5 wt% U-235 and less than uranium liquid wastes.
140 gU/t concentrations in filters and resins)
~ Any Permanent or Uranium solid waste (up to 5 Sorting and compaction.
Portable Building wt% U-235) j having HEPA filtration and isokinetic sampling.
9 The locations described in this table are shown on the site plan, Figure 11-10.1 AMENDMENT APPLCATON DATE:
PAGE NO.:
September 12,1994 1-10 SPC-ND3330 947 (R-is07/92) l
Siemens Power Corporation - Nuclear Division sur-2
(
') SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 LJ PART I - LICENSE CONDTTIONS REv.
programs include training, fire fighting restrictions, criticality alarm coverage and associated records systems.
4.
Performing Criticality Safety Analyses for designs and procedures, including second-party reviews.
5.
Providing professional advice concerning matters within the component's cognizance.
6.
Membership in the Plant Emergency Response Management Team (PERMT).
7.
Performing compliance inspections.
8.
Preparing Criticality Safety Specifications and Limit Cards.
All Criticality Safety Analyses shall be reviewed by a second party who shall be knowledgeable of the technical data and qualified in the techniques of criticality physics.
Second party reviews shall be arranged by the Criticality Safety Component, and may be (m
(")
either from within the component or by an outside reviewer. All nuclear Criticality Safety Analyses and reviews shall be documented, and documents shall be held until six months following the termination of the processes, equipment, or facilities to which they apply.
2.1.17 Health Physics Component / Health Physicist The Health Physics Component (which includes the Health Physicist) resides within the Safety organization. The responsibilities of the Health Physics Component include the following:
1.
Providing technical bases, criteria, and methods related to health physics.
2.
Providing for outside sources for aid a sd special services related to health physics and emergencies.
3.
Preparing and updating the Radiation Protection Standards section of the Company Safety Manual (EMF-30).
j 4.
Establishing radiological protection programs in accordance with criteria and standards provided by tne Manager, Regulatory Compliance. Such programs include air sampling, contamination and radiation surveys, bloassay in-vivo examinations, and associated records systems.
/'N 5.
Performing compliance inspections.
i autwoutwT appucaton oart:
Pa E NO :
%M@wR1m g
I sPC-ND.3330 947 (R-107S2)
Siemens Power Corporation - Nuclear Division EMF-2 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 PART I - UCENSE CONDITIONS REv.
2.2.1 Manaaer. Safety. Security. and Licensino The minimum qualifications of the Manager, Safety, Security, and Ucensing shall be a BS i
degree in a technical field with 10 years experience in the nuclear energy field, of which l
four shall have been in positions with nuclear safety responsibility.
{
2.2.2 Manaaer. Reaulatory Compliance The minimum qualification for the Manager, Regulatory Compliance shall be a Bachelor's degree in science or engineering, plus eight years experience in the nuclear or environmental safety fields.
2.2.3 Staff Enaineer. Ucensina i
The minimum qualifications shall be a Bachelor's degree in science or engineering, plus at least five years experience in the nuclear field or which three years experience shall have been in safety-related or safeguards fields requiring significant interaction with regulatory agencies.
2.2.4 Supervisor. Safety The minimum qualifications shall be a Bachelor's degree in a technical field, with five years experience in safety-related fields (industrial, radiological, health physics, or nuclear).
2.2.5 Supervisor. Radioloalcal Safety The minimum qualifications shall be a Bachelor's degree in a technical field, with five years experience in radiation safety, preferrably in fuel cycle facilities, or, in the absence of a degree, then 10 years experience shall be required.
i 2.2.6 Industrial Safety and Health Specialist findustrial Hvoienist)
The minimum qualifications of at least one member of the Industrial Safety and Health j
Component shall be a Bachelor's degree in science or engineering with two years experience in industrial safety or health.
2.2.7 Criticality Safety Specialist L
The minimum qualifications of at least one member of the Criticality Safety Component, as well as for each second-party reviewer, shall be a Bachelor's degree in science or engineering with two years experience in nuclear criticality safety analysis.
O i
i AME' OWE N M O' M ggptgmhgpj2,jgg4 g,g j j
NE E sPC-ND3330W7(A 907/92)
Siemens Power Corporation - Nuclear Division eur.2 i
] SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257
)
PART I-UCENSE CONDITIONS
- ngy, l
2.2.8 Health Physicist j
~
The minimum qualifications of the Health Physicist shall be a Bachelor's degree in science or engineering with five years general experience in radiation protection, or at least two years of radiation protection experience allied with nuclear fuel fabrication.
2.2.9 Health and Safety Technician Specialists The minimum qualifications shall be a high school diploma with ten years experience in radiation and chemical monitoring. They shall have passed the SPC Health and Safety Training and Qualification Program or shall have had the equivalent prior training. They must demonstrate the ability to perform specialized functions of SPC's radiological and environmental safety programs such as safety training, sample collection and control, and data correlation.
2.2.10 Health and Safety Technicians The minimum qualifications of certified Health and Safety Technicians shall be a high I
school diploma with two years of radiation and/or chemical monitoring experience, or four
(
years of similar experience in lieu of a high school diploma. Health and Safety Technicians shall complete a formal SPC training program, or shall have had equivalent prior training. They shall be proficient in SPC's radiological and chemical safety programs, criteria. specifications, procedures, and routines.
2.2.11 Environmental Enaineer The minimum qualifications of at least one member of the Environmental Engineering Component shall be a Bachelor's degree in science or engineering, and at least one year's experience in the environmental field.
2.3 Safety Review Committees 2.3.1 Health and Safety Council r
SPC has established the Health and Safety Council which convenes monthly at SPC's fuel manufacturing plant in Richland, Washington, to review various aspects of the safety program, includirig:
O SPC-ND3330 947 (R-1/07/92)
I SiemenS Power Corporation - Nuclear Division EMF-2
(
SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 w
PART I - LICENSE CONDITIONS ggy.
1.
Industrial safety practices and trends.
2.
Radiological safety practices and trends.
3.
Criticality safety practices and trends.
4.
Adequacy of emergency planning and procedures, including results of tests and drills.
5.
Overall safety awareness and attitude of employees and programs for promoting improvements.
6.
Unusual occurrences and accident investigations, including recommendations to prevent recurrences.
7.
Status of Council related action items.
I Membership of this Councilincludes:
b)
(
1.
Richland Plant Manager (Chair).
2.
Supervisor, Safety (Secretary).
3.
Manager, Safety, Security, and Ucensing (Co-Chair).
4.
Vice President, Engineering, Nuclear Division.
5.
Appropriate managers within these and other organizations.
6.
Key safety engineers and specialists.
Designated members of the Council make monthly inspections of buildings and grounds for housekeeping and safety practices, and report the findings to the Council at the monthly meetings. Findings are assigned to individuals for resolution and are held open until resolved.
2.3.2 ALARA Committee An ALARA (As Low As Reasonably Achievable) Committee maintains awareness of trends in employee radiation exposure and radioactivity content of effluent releases. The Al. ARA l
Committee shall convene at least semi-annually and shall issue a formal report at least annually to the Health and Safety Council reviewing employee exposures and effluent release data to determine:
AMENDMENT APPLCATON DATE:
PAGE NOc September 12,1994 2-13 SPC-ND.3330 947 (M/07/92)
SiemenS Power Corporation - Nuclear E.; vision eur.2 (7
SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257
/
)
PART I - UCENSE CONDITIONS
- ngy, 1.
Trends in personnel exposures and effluents.
2.
If personnel exposures or radioactive effluents might be lowered under the concept of ALARA.
3.
If equipment for effluent and exposure control is properly designed, used, maintained, and inspected.
Their reports shallinclude review of required audits and inspections performed during the past year, and review of employee external exposures, bioassay results, unusual occurrences, effluent releases, in-plant airborne radioactivity, and environmental monitoring.
The membership of the Committee includes:
1.
Manager, Regulatory Compliance (Chairman).
2.
Health Physics Specialist (Secretary).
3.
Supervisor, Safety.
4.
Manager, Plant Engineering.
5.
Manager, Safety, Security, and Ucensing.
6.
Manager, Plant Operations.
7.
Manager, Process Engineering.
I 8.
Supervisor, Radiological Safety.
2.4 Trainina In addition to normal on-the-job training, employees shall be instructed in radiation protection and criticality safety requirements and procedures, industrial safety, fire protection, and emergency procedures. The degree of training shall be commensurate with each employee's position in the Company (related to general and special responsibilities), and with the extent of the employee's contact with radioactive and fissile 3
materials. The minimum safety-related training requirements and training course content for various employee positions shall be established by Safety, Security, and Ucensing.
Employee instruction shall be provided by personnel knowledgeable in the various training topics. All formal training shall be documented, and records maintained by Safety, Security, and Ucensing.
s AMENDMENT APPUCATON DATE:
PAGE NO.:
SPO-ND3330.947 (R-U07/92)
SiemenS Power Corporation - Nuclear Division EMF-2
/O SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 O
PART I - LICENSE CONDITIONS REv.
2.4.1 initial Trainina Employees shall be provided initial instruction adequate to allow them to safely start on-the-job training; they are provided the full instruction within two weeks after starting work.
Prior to assignment to independent operation, employees are required to have been instructed in radiation protection, health physics, criticality safety, fire protection, hazardous chemical safety, emergency requirements, and operating procedures appropriate to their positions.
2.4.2 Follow-up Trainina When changes are made in radiation protection, criticality safety controls (procedures, specifications, etc.), or in emergency procedures, each employee affected shall be promptly inforrr.ed and properly instructed.
Safety topics are routinely discussed in monthly safety meetings held by operating components. Additionally, each employee routinely working with special nuclear material shall receive annual refresher instruction as part of SPC's continuing program in radiation Q
protection and criticality safety awareness. The effectiveness of this annual refresher (j
training is determined by a written examination and a review of the correct answers to the questions at the end of the test.
2.4.3 Health and Safety Technician Trainina Health and Safety Technicians shall be given special training related to their radiation protection and chemical safety assignments. Previous training is accepted if considered equivalent to the SPC training program. Despite previous acceptable training, the Health I
and Safety Technicians are required to become proficient in SPC's radiation protection, chemical safety, and criticality safety programs, criteria, specifications, procedures, and routines, as demonstrated by successfully passing an SPC certification examination within six months after employment as a Health and Safety Technician. In addition, refresher training shall be provided to all Health and Safety Technicians annually.
2.4.4 Trainina Evaluations Employee awareness of, and conformity to, safety requirements and procedures, as well as the effectiveness of safety training programs, shall be evaluated at least monthly by the Radiological Safety Component for radiation protection, by the Health Physics Component for bioassay and in-vivo programs, by the Criticality Safety Component for criticality safety, and by the Industrial Safety Component for fire protection and hazardous I
chemical safety. Specialists in these components have the authority to require retraining of employees. These evaluations shall be documented along with actions required as a result.
v aut at NT apptc4 Tom oATE:
Se tember 12,1994 2-15 sPC-ND.3330 947 (R-1/07/92)
)
Siemens Power Corporation - Nuclear Division EMF-2 Q SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 o
PART I - LICENSE CONDITIONS ggy.
2.5 Operatina Procedures. Standards and Guides SPC conducts its business in accordance with a system of Standard Operating Procedures, Company Standards, and Policy Guides. SPC is committed to controlling i
activities involving special nuclear materials in accordance with these approved written procedures, standards, and guides. These documents shall be prepared, reviewed, revised, approved, and implemented in accordance with the Approval and Responsibility Matrix (Figure 1-2.3). P! ant and facility managers are responsible for assuring compliance with all pertinent radiation protection, criticality safety, health physics, fire protection, environmental protection and hazardous chemical safety procedures, specifications, and practices within their respective facilities. Violations of radiation protection procedures or criticality safety specifications which are of repetitive or serious nature, are subject to disciplinary action.
2.6 Internal Audits and insoections Audits and inspections shall be conducted to determine that plant operations are conducted in compliance with regulatory requirements, license conditions, and formal j
s procedures. These audits and inspections apply to radiation protection, criticality safety, hazardous chemical safety, fire protection, and environmental protection.
2.6.1 Radiation Protection Health and Safety Technicians shall perform daily inspections as a part of their normal l procedures for collecting air samples and performing radiation and contamination surveys of all areas of the plant where uncontained radioactive materials are stored, processed, or otherwise handled. Detected minor infractions of radiation protection procedures, I
exposure controls, and sound radiation protection practices are corrected on-the-spot as part of a continuing education effort in their field of expertise. Serious inf4 actions and noncompliance with license conditions shall be documented and distributed to the respective facility management, and to the Manager, Safety, Security, and Licensing.
The Health Physics Component shall perform monthly audits of radiation protection practices and exposure controls. These audits shall be made in accordance with a formal plan (found in EMF-30, Chapter 2). Results of these audits shall be documented, including any recommended corrective act:ons, and distributed to the respective facility management, and to the Manager, Safety, Security and Licensing. The Health Physics Component shall follow up on each detected discrepancy in subsequent a'idits until there is satisfactory resolution.
AMENDME NT APPLCATON DATE:
PAGE NO.:
SPC-ND:3330 947 (R-1/07/92)
Siemens Power Corporation - Nuclear Division eur.2
/m SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 k
PART I - UCENSE CONDITIONS ngy.
2.6.2 Criticality Safety While performing their daily duties, Health and Safety Technicians are instructed to be alert for infractions of criticality safety specifications and limits. Detected infractions shall be communicated to the Criticality Safety Component.
The Criticality Safety Component shall conduct monthly audits of the various criticality safety control systems, (e.g., moderation control, enrichment control, neutron absorber inspections, process alarms, and trips related to criticality safety, etc.). The Criticality Safety Component shall also audit new installations and modifications to equipment and processes prior to their operation with special nuclear material. These audits shall be performed in accordance with a formal plan (found in EMF-30, Chapter 3), and reports of the findings shall be submitted to the respective facility management, and to the Manager, Safety, Security, and Licensing. The Criticality Safety Component shall follow up on each detected infraction in subsequent audits until there is satisfactory resolution.
2.6.3 Hazardous Chemical Safety n
As part of their normal duties, the Supervisor, Safety, and the Industrial Safety and Health
' ')
Component shall make periodic inspections of all areas of the plant where hazardous
!s chemicals are stored, processed, or otherwise handled. Detected discrepancies of exposure controls and sound industrial hygiene practices are corrected on-the-spot as part of their continuing educational effort.
The Industrial Safety and Health Component shall also monitor the levels of the various chemicals that are present on the plant site. Results of these inspections shall be documented, including any recommended corrective actions, and distributed to the respective facility management, and to the Manager, Safety, Security, and Licensing. The Industrial Safety and Health Component shall follow up on each detected discrepancy in subsequent inspections until there is satisfactory resolution.
2.6.4 Fire Protection The inspection committee of the Health and Safety Council shall perform monthly inspections of selected areas of the Richland facility. These inspections shall include housekeeping and industrial safety. Results of these inspections shall be documented, including any recommended corrective actions, and distributed to the respective facility management. The inspection committee shall follow up on each detected discrepancy and recommended corrective action in subsequent inspections until there is satisfactory j resolution.
The Health and Safety Council shall monitor progress toward such j j
resolutions.
l h(~N '
The Supervisor, Safety shall ensure that a monthly inspection is made of all plant fire extinguishers in accordance with a formal plan (found in EMF-30, Chapter 1). Results of AuCNOMENT APPLCATON DATE:
' PAGE NO; September 12,1994 2-17 SPC-ND.3330 947 (R-UO7/92)
i l
Siemens Power Corporation - Nuclear Division EMF-2 l
SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 PART I-LICENSE CONDITIONS
- ggy, I
these inspections shall be documented, and any detected discrepancies corrected during I
the inspection or followed up on subsequent inspections until there is satisfactory
{
resolution.
I
~
2.6.5 Environmental Protection Ir spections The Environmental Engineering Component shall monitor the levels of regulated material released to the environment. Health and Safety Technicians are responsible for the field sampling under the guidance of the Industrial Safety and Health Component. Action l
levels and descriptions of actions to be taken as a result are found in EMF-30, Chapter
- 4. Results of this monitoring shall be documented and distributed to the Manager, Safety, Security, and Licensing.
The Health Physics Component shall perform quarterly inspections of environmental protection practices and exposure controls. These inspections shall be made in i
accordance with a written plan. Results of these inspections shall ce documented, j
including any recommended corrective actions, and distributed to the respective facility management, and to the Manager, Safety, Security, and Licensing. The Health Physics l
Component shall follow up on each detected discrepancy and recommended corrective i
action in subsequent inspections until there is satisfactory resolution. The Health and i
Safety Council shall monitor progress toward such resolution.
i 2.7 investinations and Reportino of Reportable incidents in addition to, and/or in-line with, the reporting requirements specified in other sections of this License, or in Regulatory Guide 10.1, the following reporting schedule shall be i
adhered to:
1.
Employee and former employee radiation exposure information shall be i
reported to said individuals in accordance with 10 CFR 19.13; l
2.
Employee exposures and monitoring information shall be reported to the NRC in accordance with 10 CFR 20.2206; 3.
Overexposures and excessive levels and concentrations shall be reported l
to the NRC in accordance with 10 CFR 20.2203; 4.
Theft, or loss of licensed material, shall be reported in accordance with 10 CFR 20.2201; 5.
Effluent monitoring information shall be reported to the NRC in accordance with 10 CFR 70.59;
[
I AMENOMENT APPLCATON DATE:
PAGE NO.:
j i
SPC-ND.3330.947 (R-1/07/92) g-3
.nv,-
y
-y-ev--
i Siemens Power Corporation - Nuclear Division EMF-2 O SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 PART I-LICENSE CONDITIONS
- ngy, 6.
In the event that the radioactivity in plant gaseous effluents exceeds 50 microcuries per calendar quarter, a report shall be submitted within 30 days to the NRC (Document Control Desk, with a copy to NRC Region V Office), identifying the cause for exceeding this value, and the corrective actions to be taken to reduce release rates; 7.
If parameters important to dose assessment of the public relative to gaseous effluents from the plant change, a report shall be submitted to the NRC (Document Control Desk, with a copy to NRC Region V Office) within 30 days identifying the changes in parameters, and providing an estimate of the resultant change in dose commitment; 8.
Reports of excessive radioactive contamination on packages of radioactive material, and excessive radiation levels external to the packages on I
receipt, shall be reported to the NRC immediately in accordance with 10 i
CFR 20.1906; l
9.
Accident reports on transportation of licensed material shall be reported to the NRC and DOT in accordance with 10 CFR 71.5(b) and 49 CFR 171.15 and 49 CFR 171.16.
i The Manager, Safety, Security, and Licensing, has the responsibility for investigating, recording, reporting, and following up on actions taken for reportable incidents in accordance with NRC reporting requirements.
2.8 Records in addition to, or in accordance with, the documentation requirements specified in other sections of this Application and 10 CFR 20.2108,20.2109, and 20.2110, the following records shall be retained:
1.
Personnel radiation exposure histories and determinations of personnel accumulated dose. (information on personnel prior radiation exposure histories is obtained in accordance with 10 CFR 20.2104.)
2.
Employee radiation exposures, external and internal, including dose evaluations.
3.
Health and Safety Technicians' radiation and contamination surveys, including room air and exhaust air monitoring.
[3 4.
Environmental surveillance data and waste discharge reports.
N]
AMENDMENT APPLCATON DATE:
PAGE NO.:
SPC-ND.3330 947 (R-907/92)
)
Siemens Power Corporation - Nuclear Division EMF-2
{
SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 1
f PART I - LICENSE CONDITIONS REv.
?
I 5.
Incident investigation reports.
l 6.
Employee indoctrination, training, and retraining reports.
l 7.
Facility and process acceptance tests, i
8.
Radiation safety, health physics, and criticality safety inspection reports.
1 9.
Richland Health and Safety Council meeting reports.
{
k 10.
ALARA Committee reports.
I 11.
Criticality dosimeter inspection reports.
12.
Reports of test results for the criticality accident alarm system.
l l
13.
Reports of test results for the emergency electrical power supply system.
14.
Radiation detection instrument maintenance and calibration; also for the neutron detectors of the criticality accident alarm system.
i 15.
Results of sealed source leak checks.
~i 16.
HVAC system monitoring and test results.
I 17.
Engineering Change Notices (ECN).
l
[i 18.
Nuclear Criticality Safety Analyses and revisions thereto.
j h
Such records shall be maintained for a minimum period of five years, unless there are I
legal or license requirements for longer retention periods for specific records.
i I
i s
j 4
Y AMENDMENT APPLICATION DATE:
PAGE NO.:
l SPC-ND.3330 947 (R-UO7/92) l
. \\
Ob R
%E m*
Siemens Power Corporation - Nuclear Division oj 5
w a i
m F
r Z
C To 9
SENIOR VICE PRESIDENT h$
g a
mo 5
GENERAL MANAGER NUCLEAR DIVISION 3O Ek
>r
-( O m
I 1
3a m
a 5; 5' f
DIRECTOR VICE PRESIDElli PA g
OUALITY EllGillEERlHG
-4 MANAGER g7 3
2!
mc Q
g ZO C
o CD m~
a I
I I
I m
m mW g
h Zh PLANT
- SAFETY, PRODUCT OUALITY MANUFACTURING MATEntALS &
N CONTROL ENGINEERING ScilEDULING OPERATIONS SECURITY MECilANICAL L
m HICitLANI)
& LICENSlHG ENGINEERiflG g-g g
O Z --
h ANALYTICAL PROCESS THAFFIC &
LA80R4100MES
~
ENGINEEnlHG WAllEllOUSitfG
~
PRODUCTDEV
& TEsitt G a
PLANT O
N
~
ENGINEERING M AIEUI^t 8 I
"y OESE AllCll M
Z
~
WA!iUFACit# Wit)
TEClitlOt OGY O
WASTE MANAGEMEf4T h
ENGillEEHING H
9 2
5 O
u E
a
-P M
m b
Un m 5
M N E
,N 71 to
i Siemens Power Corporation - Nuclear Division EMF-2 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227. NRC DOCKET NO. 70-1257 I
PART I-UCENSE CONDITIONS ngv.
FIGURE l-2.2 SAFETY, SECURITY, AND UCENSING l
}
I SAFETY PLANT SEC'lRITY Ew T NC INCUSTRIAL l
SAFETY &
UC"iNSING HEALTH l
I CamCaury
_I sAF m sAFEcuARos l
l ENVIRCNMENTAL HEALTH PHY3!C3 ENGNEERING HEALTH RECCRCS RADICLOGICAL SAFETY
. HEALTH & FETY SPEC:AUSTS e
HEALTH &
SAFETY l
M i
5 AMENDMENT APPLCATON DATE:
PAGE NO.:
September 12,1994 2-22 SPC-ND.3330.947 (R-1/07/92)
D-Cg3O] OIQ g tkWe.Ooi28 WC'gk9Og D
mEPto O C1mO
>g D
. 7C gm>y $>d 2>r r O9(DA Z
- CzE ag 7"UO g 7m 7Q* o*aNWN D
,>2H -[pmZgN oOZC QZg M<
he _
_ _ _ _,_E
.e
?.
m n
. r _.
i a
- i _
m 1
r r IE}e l _.
1 t
t I
m s
s'
_ r_._A e
- n a:j.gh _
t
& r r
n
.,t f
o k
c rs D F
_D_'C_ _A D
l
(
D- -
_D l
t I
& r r Y n
I t
n ei
]InC C- - -
C_
C O
c ar
/ 0
- _E~
_C / Mi A
4 4 d 2 %'
si A
A A
/ MC A A c
1 E N E
}g3 eA 1 A
ns X
a l
e n
1 C-C C
C' B
D D C
n T
r A
u M
4M- - - - -
e Y
m A A T
a l.
t CA-
- -Aa a
i E
m C
C A
C n
ID l
O e
l e
S CC_ -
C-e 1
l C
C C
C i
n O
I P
S I]!
r t
E 7
H 5
D- - - -
4'
_D D Y r
2 D
1 t
f t
s0 l
1 e -
l A.
1 n7 c ~ C' -
C D-
_D e
r r Y f
t l
r ico.
l
.-D l N A
1 l
V DD_&~O.r_A iat
_D r
y r r r e
O 1
1 t
I t
t t
t r k tec 1
I I
a o P
A- __
M D P
1._
rC A
a1i 3
e H.
1 l
C C'
2 c
1 u7 N2 I
l 2 E
CA_
at 1
n' d-H 1
eM U
pN G
D_D_
D- -
S.
_D S
r I
F l
t o
l B_i- -
O-i
_B
. a O O D 0
n i])l s
d n
e se lo
=
d s
n 4
l d
t a
a e
c ia is e
s S
s e
r s
fk~P.
=
P n
a o
s d
d.;
s a
r ta d
e e
C A
s e
d e
e W a
a p
s e
d e
b a
r n
e o
h
!Iy d
g n
c w
a d
n r
y c
d d
h a
n n
e e
C ia a
o Q
ti S
e e
ee ep te na s
s t
a 5
d H
s l
e i
e a
S s
r s l
c d
s e
d d
f e
r er n
eo l
a a
e d
l o
b O
s s
S S
p a
l S
n l
d re sn a
p co e
t g
e a
wis i
s u
i e
S C
a i
e k
- m. n m
ee ey y
e Ia.
su s
d oi l
l a
r y
y y
d~l v
e e
p c
e e
s le e
w ic ic l
fe tS se s
e er r
le Pi r
o a
c v
w A
lk n
r,
h D
a a
e C,
S y
e e
re e
s o
s l
t v
a o
s p
w e
c f
e s
a a
a a
t P
S W
e n
t nr e
g o
c
,a e
e S
S l
c T
i e
v a
n o
o
- s g
C n
n n
C C
C n
h g
e d
la d
=
- m. m h
m l
e s
t n
y y
a c
ml A.
4 e
l A
u a
o e
o s
a d
k g
c i
e e
J c
le e
a a
r s
e e
s los A D C o e A-o la ia la SE d
d d
c ec le ic it i
d m
P H
O v
O c
d o
c t
v a
a p
e a
e u
u u
e n
e n
n n n n
C Ce E
I E
E l
n f
E2WE*;ag9 b
,,P,,,o -
meU53C$ *y ao>u*
Nm T
c g
ot8."9 fa$S i
I l1 lllI1 l'
Siemens Power Corporation - Nuclear Division EM F-2 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 PART I-LICENSE CONDITIONS REv.
5.
Personal survey and protective clothing requirements; 6.
Personal dosimetry requirements; and 7.
A statement of the respiratory protection equipment required for entry into an airborne radioactive materials area.
3.2 Technical Reauirements 3.2.1 Controlled Areas The perimeter fence and outside office building walls define the controlled areas of the SPC facilty. Access to controlled areas shall be controlled by SPC security personnelin accordance with a formal, NRC-approved physical protection plan.
3.2.1.1 Restricted Areas O
All radioactive materials at SPC's Engineering and Manufacturing Facility sha!! be stored and processed within restricted areas. Each access point to restricted areas (as defined in 10 CFR 20) shall be posted in accordance with 10 CFR 20.1902. Additionally, RWP's for the respective areas shall specify the existing or potential radiological conditions and radiation protection measures required.
3.2.1.2 Clean.IntermediateandContaminated RadioactiveMaterials/ Radiation Areas With the possible exception of temporary stepoff pads, clean areas shall be separated from contaminated controlled areas by intermediate areas. Intermediate areas shall be identified, and their boundaries visibly marked. Personnel shall follow posted special i
procedures or restrictions when leaving one area and entering another.
3.2.1.3 Chance Rooms and Step-Off Areas l
Change rooms servicing contamination controlled area workers shall be divided into contaminated, intermediate, and clean areas to minimize the spread of contamination.
Step-off pads shall be provided when exiting contaminated areas. Separate toilet facilities may be located in contaminated, intermediate, and clean areas. Use of the toilets in contaminated and intermediate areas without removal of protective clothing shall be l
permitted provided a personnel survey is performed first.
Additional step-off areas may be established for maintenance work, temporary situations or conditions, or to accommodate personnel entry and exit not requiring the use of O
AMENDMENT APPLCATION DATE:
PAGE No; i
SPc-ND;3330.947 (R-v0W92)
l f
Siemens Power Corporation - Nuclear Division EMF-2
~ LPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 PART I-LICENSE CONDITIONS REv.
3.
Each on-line radiation detection instrument shall be checked for proper l
operation either by Health and Safety Technicians or by electronic j
surveillance daily (Monday through Friday for a normal work week). When daily checks are performed in a manner which qualifies as calibration, j
separate semlannual calibrations shall not be required; l
4.
Portable survey instruments shall be source-checked each shift they are used; 5.
Each AC-operated personnel contamination survey instrument shall be provided with an individual check source to allow personnel to source-check the instruments; l
6.
Calibration sources shall be traceable to the National Institute of Standards and Technology (NIST); and 3.2.4.2 Criticality Accident Alarm System T
See Chapter 1, Section 1.6.1.
f l
3.2.4.3 Criticality Dosimeters i
Criticality dosimeters shall be strategically located throughout the process facilities.
These criticality dosimeters shall be capable of measuring 0.1 to 10 rems of neutron-j radiation over a neutron spectrum of thermal to 2.5 MeV. The criticality dosimeters shall
-l be inspected at least annually to confirm their presence and undamaged condition, j
3.2.5 Radiation Exposure SPC shall strive to maintain external radiation exposures as far below the limits specified in 10 CFR 20.1201(a) as reasonably achievable. Radiation exposure records shall be i
reviewed by the Al. ARA Committee.
i In the event that it is necessary to exceed the exposure limits specifes in 10 CFR l
20.1201(a), exceptions shall be authorized in accordance' with 10 CFR 20.1206.
l J
Respiratory protective equipment shall be used for entry into areas where airborne radioactive materials are known to exist in excess of the occupational DAC.
l
[
[
a AMENDMENT APPLCATON DATE:
PAGE NO.-
SPC-ND:3330.947 (R.vo742)
SiemenS Power Corporation - Nuclear Division
~
EM F-2
(')' SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 L/
PART I - LICENSE CONDITIONS ngy.
Chapter 4 Nuclear Criticality Safety Nuclear criticality safety shall be assured through both administrative conditions and technical practices. Administrative conditions include establishing the responsibilities for nuclear criticality safety; providing adequate and skilled personnel; and preparing written standards and procedures, process analyses, materials and operational controls, operational and incident reviews, and emergency procedures. Technical practices include exercising control over the mass and distribution of significant quantities of special nuclear material (SNM) and over the mass, distribution, and nuclear properties of all other materials with which the SNM is associated.
4.1 Administrative Conditions i
Administrative conditions define:
le The lines of responsibility for assuring all criticality safety aspects of the process l
are reviewed, documented, and approved by management; I
(')s The design approach employed in the definition of all processes involving the e
(
handling and storage of the SNM; The required safety and procedural documentation, the written procedures, and e
the postings employed to describe the approved processes for handling and storage of the SNM; Criticality safety practices for the marking and labeling of the SNM; and e
The location and capabilities of criticality alarm systems.
e 4.1.1 Responsibilities j
! The responsibilities and authorities for nuclear criticality safety, as well as the professional
{ requirements for criticality safety personnel are described in Chapter 2.
4.1.2 Eautoment and Process Desian Philosophy Before any operation utilizing SNM is begun or changed. it shall be determined that the entire process will be subcritical under both normal and credible abnormal conditions l
l
- within the technical criteria specified in Section 4.2. The process design, inasmuch as the j l
handling and storage of the SNM are concerned, shall incorporate sufficient factors of safety such that at least two unlikely, independent, and concurrent changes in process conditions are required before a criticality accident may occur.
\\
)
x._/
WE NDMENT FPUCMON DME:
September 12,1994 PAGE NO.:
41 t
L i
SPC-ND 3330 947 (R-90792)
1 SiemenS Power Corporation - Nuclear Division EMF-2
/*
SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 k_
PART I - LICENSE CONDITIONS REv.
Where practical, SPC uses passive engineered features, such as favorable geometry equipment, as the preferred method of criticality control in order to reduce the dependence on administrative procedures. When necessary, active engineered features and administrative controls are used for criticality control. in these cases, controlled parameters are clearly specified and approved by management.
4.1.3 Procram Documentation The documents listed in this section are key elements of the administrative portion of the Criticality Safety Program. These documents are prepared and approved in accordance with Figure 1-2.3.
4.1.3.1 Criticality Safety Standards SPC shall establish and maintain a system of written Criticality Safety. Standards. The purpose of these standards is to present the policies, administrative practices, and criteria, etc. for criticality safety for SPC. These standards are used to help ensure that a criticality accident will not occur and to reflect both the regulatory requirements and
{J
's accepted industry practices.
These standards shall be kept current by annual review and updated as appropriate.
4.1.3.2 Criticality Safety Analyses Criticality Safety Analyses (CSAs) shall be performed on all applicable operations.
Documentation of each analysis shall be sufficiently detailed such that an independent reviewer can reconstruct the analysis and bases for the conditions presented. All CSAs l
shall be reviewed by a second party who is knowledgeable of the technical data and qualified in the techniques of criticality physics. Second-party review shall be arranged by the Criticality Safety Component and may be conducted by SPC or contractor personnel who meet tha professional requirements specified in Section 2.2.7.
All CSAs and reviews shall be documented and the documents shall be held a minimum of six months following the termination of the processes, equipment, or facility to which they apply.
4.1.3.3 Criticality Safety Specifications Criticality Safety Specifications (CSSs) shall be prepared based on limits established in the CSAs and shall be in a standardized format containing the following information:
i work location (s), equipment description, SNM description (element, isotope, enrichment, 7N autwoueur appoc4 Tom o4TE:
September 12,1994 PAGE NO 42 i
SPc ND:3330 947 (R-197S2)
SiemenS Power Corporation - Nuclear Division eur.2
[] SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 U
PART I-LICENSE CONDITIONS
- ggy, form, density), operation (s) involved, limits, moderator and reflector restrictions, spacing restrictions, criticality safety controls, date, and approvals.
CSSs shall be prepared when the Criticality Safety Component determines an analysis has plant-wide applications, when requirements from several analyses need to be combined into a single location for administrative convenience, or when administrative controls not specified on a limit card are required.
CSSs shall be maintained in an up-to-date manner to reflect currently approved processes and/or equipment. Additionally, they shall be reviewed annually and updated as appropriate. Copies of current CSSs shall be maintained in designated mini-libraries located near working areas of the processing and laboratory facilities.
4.1.3.4 Criticality Safety Limit Cards Each work area where SNM is handled or stored shall be posted with a Criticality Safety Limit Card which contains the criticality safety limits directly controllable or observable by operators. The limit card normally includes the type and form of material permitted, allowable quantity (containers, pieces, weight, or volume), restrictions on moderators, and y) required spacing from other SNM. The limit card also has a title and number such that the card is directly traceable to the CSA that supports it.
4.1.3.5 Written Procedures Operations in which nuclear criticality safety is pertinent shall be governed by written procedures. All persons participating in these operations shall be required to be familiar with the procedures. Process Specifications together with the applicable CSS and/or Criticality Safety Limit Cards shall be part of the basis for written operating procedures.
4.1.4 Criticality Safety Practices 4.1.4.1 Confirmation of Analysis Assumptions After a CSA is complete and prior to the introduction of the SNM into a new or modified operation or process, the Criticality Safety Component shall inspect the facility and equipment and confirm that the specified controls are in place and functional. The results of these inspections shall be appropriately documented.
A modified process is defined as one invoMng a change in equipment design, SNM amount and/or configuration, or any process controls which requires revision to the applicable CSA.
ha s i-i- ;2,is94 w
m,
+a
.mm-sPc-ND 3330 947 (41/07/92)
SiemenS Power Corporation - Nuclear Division eup.2
[') SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 V
PART I-LICENSE CONDITIONS agv.
4.1.4.2 Preoperational Startuo end Testina Equipment may be operated under engineering control during preoperational startup and testing before being turned over to Operations for production work. Operation under engineering control shall require adherance to the same criticality safety practices as normal production operation.
4.1.4.3 Fissile Content Determination The fissile isotope content (enrichment) of all incoming SNM shall be verified by nondestructive analysis or laboratory analysis of representative samples prior to the i conduct of any activity other than storage.
4.1.4.4 Special Nuclear Material Control l
l The movement of the SNM shall be controlled. The CSSs describe materials control I practicen Workstations, administrative!y controlled only on the basis of mass of material, shall be halted as follows:
C 1.
No more than one safe batch shall reside at a work station at one time.
A safe batch !s defined as no more than 0.45 of the minimum critical mass of the materialin process; 2.
No more than one safe batch shall be moved at one time when introducing or removing material from a workstation; 3.
Individual safe batches shall be spaced a specified minimum distance apart; and 4.
A record shall be maintained of the SNM inventory at each mass-limited workstation.
See Section 4.2 for a discussion of the parameters (e.g., geometry, moderation, or fixed neutron absorbers) used for control of systems containing SNM.
l 4.1.5 Markina and Labelino of SNM 4.1.5.1 Labelino of SNM Insofar as practical, all SNM in the plant shall be identified with distinctive labels. When possible, the label shall be applied to the outer container of the SNM. When such nU)
(
aucuoutuT amcuos cars:
September 12,1994 PAGE No.:
44 sPC-ND.3330 947 (R-1/07/92)
1 i
Siemens Power Corporation - Nuclear Division sup.2 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 701257 PART I-LICENSE CONDITIONS
- ggy, i
labeling is not practical, such as for fuel pellets outside of containers, each item shall be identified by coding posted in the immediate vicinity.
The label or sign, as appropriate, shall show the type of material, form, enrichment, gross, tare and net weights, element or fissile isotope weight, fuel identification number, date, and initials of the person preparing the label.
l Except for waste, materials of different enrichment, physical and/or chemical form, and isotopic content shall be kept segregated until combination of different materials is required in the process.
4.1.5.2 Postino of Special Nuclear Material LWlons Each location where the SNM is handled, processed, transported, or stored shall be identified by a distinctive symbol which is observable from all approaches at a distance at least equal to the spacing requirement from other SNM. For work locations separated from other areas by partitions, walls, etc., posting on the opposite side of the separator shall indicate the spacing limit requirements for the SNM in that area.
In addition to the symbol, each work area shall be posted with a Criticality Safety Limit Card containing the criticality safety limits directly controllable or observable by operators.
The limit card normally includes the type and form of material permitted, allowable.
quantity (containers, pieces, weight, or volume), restrictions on moderators, and required i
spacing from other special nuclear material. The limit card also has a title and number i
such that the card is directly traceable to the CSA(s) that supports it.
4.1.6 Criticality Alarm System With the exception of items exempted from coverage by the criticality accident alarm system in Section 1.6.1 of this application, all SNM shall be located such that two -
detectors in the alarm system are capable of detecting a criticality originating in the material. The alarm system is described in Section 10.6. The system is capable of i
detecting, with two detectors, a criticality that provides an absorbed dor,c in soft tissue of 20 rads of combined neutron and gamma radiation at an unshielded distance of two maters from the reacting material within one minute.
This capability meets the requirements of 10 CFR 70.24(a)(1).
w m,ogm ameum ous:
septemoer 12,1994 paos wo.:
4-5 SPc ND3330 947 (R-147/92)
i Siemens Power Corporation - Nuclear Division EM F.2 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 i
PART I - LICENSE CONDITIONS REv.
4.2 Technical Practices 4.2.1 Double Continaency Policy Process and equipment designs and operating procedures incorporate sufficient factors of safety to require at least two unlikely, independent, and concurrent errors, accidents,'
equipment malfunctions, or changes in process conditions before a criticality accident is possible.
i 4.2.2 Limits on Maximum Multiolication Factors The maximum evaluated neutron multiplication factor at normal and credible abnormal conditions shall not exceed k, as defined below.
The maximum allowable multiplication factor, k,, shall be calculated from the expression:
k, = k - a k - A k e
u m
O Where:
k The value of k that results from the calculation of benchmark
=
e g
experiments using a particular calculational method. The value represents a combination of theoretical techniques and numerical data.
Ak The uncertainty in the benchmark experiments, including random
=
u and systematic errors (blas) within the range of parameters encountered in the equipment design.
Ak A safety margin to assure suberiticality. A value of 0.05 shall be
=
m used for normal conditions. A value as low as 0.03 may be used for abnormal conditions if justified by a sensitivity analysis.
4.2.3 Limits Based on Geometry Wherever practical, reliance shall be placed on equipment designs which physically limit the dimensions of units containing SNM. Safe dimensions may be established by utilizing the following safety factors:
i 1.
The k of the unit may be established by using the guidelines given in Section g
4.2.2; or O
AMENDWENT APPLCATON DATE:
deplernDer 12,1994 PAGE NO.:
4-6 SPC-ND:3330.947 (R-907/92)
)
Siemens Power Corporation - Nuclear Division EMF-2 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 PART l-LICENSE CONDmONS REv.
2.
Critical dimensions multip ied by the applicable safety factors given in Tables 1-4.1 and I-4.2 may be used..
Where applicable, dimensional limitations shall include an allowance for fabrication i
tolerance and/or potential dimensional changes from corrosion or mechanical distortion.
4.2.4 Limits Based on Neutron Absorbers j
Criticality safety may be assured through the use of neutron absorbers such as cadmium or boron. Neutron absorbers may be fixed or blended / mixed into the SNM such as is the
]
case with soluble absorbers or absorbers blended into urania powders.
1 4.2.4.1 Fixed Neutron Absorbers l
Fixed reeutron absorbers may be used provided the following conditions have been met:
1.
Neutron absorbers are designed and fabricated as an integral part of the equipment 2.
Inspections to verify the continued integrity of the equipment and neutron absorber structure are performed on established time frequencies sufficient to insure their effectiveness; 3.
Results of these inspections and the basis for the inspection frequencies are recorded and audited; and 4.
Viable alternatives to the use of fixed neutron absorbers to assure criticality safety do not exist.
4.2.4.2 Soluble or Blended Neutron Absorbers Soluble / blended neutron absorbers may be used provided the following conditions have been met:
1.
Neutron absorbers are shown to be uniformly distributed in the SNM; 2.
The required minimum content is confirmed by analysis; and
-l 3.
The process does not have any credible mechanisms to selectively remove the neutron absorber from the SNM.
l nue,ousu? menouout:
deptemoer 1z,1w4 PAGE NO.:
4*7 i
SPC NO3330.947 (R-vo7/92)
']
Siemens Power Corporation - Nuclear Division eur.2
[G] SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET N PART I-LICENSE CONDITIONS REV.
i 4.2.5 Limits Based on Concentration Control Reliance for primary criticality control may be placed on concentration controls in areas where geometry control is not practical, and where the nature of the process and operations maka violation of the concentration limit unlikely even after failure of any single control.
Concentration control may be applied to both overmoderated and undermoderated accumulations of material as described below.
i 4.2.5.1 Concentration Control-Solutions The concentration of the SNM dispersed or dissolved in another medium may be limited to prevent criticality, provided that the following cor.ditions have been met:
1.
The concentration limit shall not exceed 50 percent of the minimum critical concentration in the system being evaluated; 2.
The concentration limit shall assure that the k meets the limits in Section g
4.2.2 at normal and credible abnormal conditions. The abnormal condition O
evaluation shall include:
a.
Precipitation of solid SNM to the most reactive credible extent; b.
Increasing the concentration of the SNM to the maximum credible extent due to effects such as evaporation; and c.
For arrays of units on concentration control, additional abnormal conditions to be evaluated (as applicable) include array size, unit spacing, and interspersed moderation effects.
4.2.5.2 Concentration Control - Powders and Pellets (Moderation Control)
The concentration of hydrogenous material within the SNM may be limited to a small percentage by weight of the SNM (moderation control) to prevent criticality, provided that the following conditions are met.
1.
The permitted concentration of hydrogera.as material shall be equal to or less than 50 percent of the critical concentration for the system in question;
(
l V
autuoutui amcara cart:
Septemoer 12, lwe pact uo.:
9-6 SPC ND 3330 947 (R 907/02)
-- - J
Siemens Power Corporation - Nuclear Division esp.2 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 PART I - LICENSE CONDITIONS
- ggy, 2.
The maximum reactivity of the system full of the materialin question, under the worst credible accident conditions, shall be limited by the guidelines given in Section 4.2.2; and 3.
The material shall be contained within a fireproof barrier or in a process area containing limited sources of hydrogenous material. In the absence of a fireproof barrier, special controls shall be used to prevent fires and to control the use of moderators in fire fighting in such process areas.
4.2.5.3 Concentration Control - Moderators Around/Between Fuel Rods and Bundles The concentration of hydrogenous material between and within discrete units of SNM such as fuel rods and bundles may be limited to a small percentage by volume of the space between the units SNM (moderation control) to prevent criticality, provided that:
l 1.
The maximum credible concentration of hydrogenous material shall be
[_}
equal to or less than 50 percent of the critical concentration for the system
(/
in question; 2.
The maximum reactivity of the system full of the materialin question, under the worst credible accident conditions, shall be limited by the guidelines given in Section 4.2.2; 3.
The fuel rods and bundles are stored or processed in such a manner that water is free draining if it is present; and 4.
Controls on the presence and use of other moderators in the area are evaluated and appropriately implemented.
4.2.6 Limits on Multi-Unit Arrays The spacing between units within an array shall be limited by mechanical means such that one of the following requirements are met 1.
The k of the array under the maximum credible accident conditions shall be g
limited by the guidelines given in Section 4.2.2; or 2.
For multi-unit arrays where k is not used as a basis, the number of units in the g
array shall not exceed 50 percent of the calculated critical number.
O autwauten amcarou can:
septemoer 1z, T w4 enas.o.:
4-9 SPC-ND3330 947 (R-UO 7/92)
Siemens Power Corporation - Nuclear Division EMF-2
]J SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 u-PART I-LICENSE CONDmONS ggy.
The mechanical design of equipment or storage arrays in which deformation or rearrangement could resu!t in the loss of a controlled parameter shall be reviewed by a person competent in mechanical engineering.
4.2.7 Criticality Safety Parameters 4.2.7.1 Criticality Data Critical parameters used to establish primary criticality safety limits shall be based on one or more of the following conditions (see Section 4.2.8 for sources of data currently acceptable to SPC):
1.
Criticality parameters obtained directly from experimental measurements; 2.
Criticality parameters derived from experimental measurements; or 3.
Cn!culations using methods validated in accordance with Section 4.3 of ANSI /ANS-8.1-1983 (reaffirmed in 1988).
O V
4.2.7.2 Enrichment Levels Design isotopic compositions shall be established and appropriate criticality safety controls implemented to assure conformance with the respective fissile element composition prior to initiating an operation.
Normally, equipment is designed to assure criticality safety by geometry control. Where batch control is utilized, enrichment level or other isotopic composition limits shall be clearly posted at the respective equipment or location.
4.2.7.3 Moderation Critical parameters derived from nuclear criticality safety analyses shall be based on optimum moderation, unless the requirements of Section 4.2.5 are applied, or other controls on moderation are established to ensure that the k,3 meets the limits in Section 4.2.2.
4.2.7.4 Reflection Critical parameters for units and arrays of units shall be based on full water reflection, unless other reflectors in the immediate vicinity could result in higher reactivities or other controls on reflection are established to ensure that the k,3 meets the limits in Section 4.2.2.
CN l
t i
V mtuoutuT Anetcarou cars:
septemoer 12,1994 paae uo.;
A-1 0 -
l sPC-ND 3330 947 (R-907/93
Siemens Power Corporation - Nuclear Division EMF-2 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 PART I-LICENSE CONDITIONS
- ggy, 4.2.7.5 Neutron Interaction Neutron interaction (exchange between individually suberitical units) shall be considered.
Consideration of the interaction between units or arrays of the SNM may be accomplished through the use of the solid angle method.
The solid angle method is applied according to the constraints in the
- Nuclear Safety Guide," TID-7016, Revision 2, except for the use of the nominally reflected solid angle acceptance criteria. The nominally reflected solid angle acceptance criteria are used to limit the allowable solid angle for arrangements of individually suberitical units provided that the following conditions have been met:
1.
Boundary conditions for the spacing between concrete walls and the array are as stated in Table 1 of Reference 21 of Section 4.2.8 except that a minimum separation of six inches shall be required; 2.
Concrete walls are less than or equal to seven inches in thickness; f
f O
3.
Separation distances given in Table 1 of Reference 21 are measured from t
the outermost vessel in the array to the closest wall; 4.
The array shall be limited in both number and size of vessels to arrays that I
are reasonable extrapolations of the conditions assumed in in Section 4.2.8 of Reference 21; and l
5.
All vessels within the array shall be suberitical when fully reflected by water and shall have a minimum edge-to-edge separation of 12 inches.
For arrays that violate any of the five conditions stated above, additional analyses shall be necessary to demonstrate the safety of the particular array in question or demonstrate the continued acceptability of using the nominally reflected solid angle acceptance criteria.
The methods in the preceding paragraph shall have been validated according to Regulatory Guide 3.14,' Validation of Calculational Methods for Nuclear Criticality Safet/."
O autwourut amcaron oart:
septemoer 12,1W4 pact wo.:
4-11
)
sPC-ND.3330 947 (R-U07/92)
SiemenS Power Corporation - Nuclear Division EMF-2 i
SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 i
i PART I-LICENSE CONDmONS f
- ggy, l
l, 4.2.8 Sources of Criticality Data and Analytical Technioues The sources of criticality data and analytical techniques currently used by SPC in performing criticality safety analyses are identified below:
f 1.
J. H. Chalmers, G. Walker, and J. Pugh, " Handbook of Criticality Data," UKAEA Handbook AHSB (S),1965.
2.
H. C. Paxton, J. T. Thomas D. Callihan, and E. B. Johnson, " Critical Dimensions 235U, 239Pu,and 233g,. TID-7028, Division of Technical l
of Systems Containing Information Extension, USAEC (1964).
3.
Subcommittee 8 of the American Standards Association Sectional Committee N6 and Project 8 of the American Nuclear Society Standards Committee, " Nuclear i
Safety Guide," TID-7016, Revision 2, Division of Technical Information Extension, l
USAEC (1978).
4.
H. K. Clark, " Critical and Safe Masses and Dimensions of Lattices of U and UO i
2 Rods in Water," DP-1014, Savannah River Laboratory (1966).
l 5.
H. K. Clark, " Maximum Safe Limits for Slightly Enriched Uranium and Uranium l
Oxide," Criticality Control of Fissile Materials, International Atomic Energy Agency, Vienna (1966), pp. 35-49.
l
?
6.
H. C. Paxton, " Criticality Control in Operations with Fissile Material," LA-3366, Los j
Alamos Scientific Laboratory (1972).
j l
7.
H. F. Henry, C. E. Newton, and J. R. Knight, " Extensions of Neutron Interaction i
Criteria," K-147'8, Union Carbide Corporation, Nuclear Division (1969).
8.
C. E. Newlon, AEC Research and Development Report," Minimum Critical Cylinder l
Olamaic:s of Hydrogen Moderated U (4.9) Systems," K-1629, Union Carbide i
Corporation, ORGDP, (1965).
I i
9.
Standards Committee N16 of the American National Standards institute," Nuclear l
Criticality Safety in Operations with Fissionable Materials Outside Reactors," ANSI l
N16.1 1969, American National Standards Institute, New York, NY (1969).
10.
R. D. Carter, G. R. Kiel, and K. R. Ridgway, " Criticality Handbook," Volumes I, ll, and Ill, ARH-600, Atlantic Richfield Hanford Company, (1968).
11.
W. Thomas, W. Weber, and W. Heinicke, "Handbuch zur Kritikalitat," (1970-73).
=cuo emmuccomous:
deptemoer 12,1we per uo.:
4-12 SPC-ND.3330.9C (R-1/07/92)
t
~
Siemens Power Corporation - Nuclear Division eu,.2 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 l
PART I - LICENSE CONDITIONS p gy, i
12.
L L Carter, C. R. Richey, and C. E. Hughey,"GAMTEC-II: A Code for Generating Consistent Multigroup Constants Utilized in Diffusion and Transport Theory l
Calculations," BNWL-35, (1965).
I 13.
H. Honeck, " THERMOS Code," BNL-5826, (1961).
l 14.
J. R. Ulley, " Computer Code HFN - Multi-Group, Multi-Region Neutron Diffusion i
Theory in One Space Dimension," HW-71545, (1961).
l l
15.
B. G. Carlson and W. Guber, "DTF Users Manual," UNC Phy/ Math-3321, Vol.1, l
(1963).
16.
J. E. Suich and H. C. Honeck, "The Hammer System," DP-1064, Savannah River Laboratory (1967).
l 17.
P. J. Hemmings, "The Gem Code," AHSB (S) R 105, United Kingdom Atomic Energy Authority (1967).
l l
18.
H. F. Henry, " Studies in Nuclear Safety," K-1380, (1957).
19.
C. L Brown, " Nuclear Criticality Safety Analysis, Uranium Fuels Plant," Jersey Nuclear Company, BNW/JN-29, (1971).
20.
" Determination of H/U Ratios in UO Water and ADU-Water Mixtures," JN-71-2, 2
(1971).
21.
C. L Brown, et al., " Validation of Boundary Conditions for Assuming Nominal Reflection in Solid Angle interaction Method (As Applied in Exxon Fuel Fabrication Plants)," BNW/XN-184, (1975).
22.
" SCALE:
A Modular Code System for Performing Standardized Computer Analyses for Ucensing Evaluation," NUREG/CR-0200.
O ausare amourm oor.
september 12, iw4 paos wo.:
4-15 SPC-ND3330 947 (R-UOW92)
~
i l
SiemenS Power Corporation - Nuclear Division EM F.2 l
SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227,IaiC DOCKET NO. 70-1257 i
PART l-LICENSE CONDITIONS REv.
7 TABLE l-4.1 SAFETY FACTORS FOR HOMOGENEOUS SINGLE UNITS SAFETY SAFETY PARAMETER CRITICAL PARAMETER FACTOR Safe Mass, M, Critical Mass, M 0.45 g
Safe Mass, MsL (Doubling of Msl is Critical Mass, M 0.75 g
excluded by construction)
Safe Spherical Volume, V, Critical Spherical Volume, V s 5f 0.75 g
V > 5#
0.80 k
Safe Infinite Cylinder, D, Critical Infinite Cylinder, D 0.85 u
Safe infinite Slab, S, Critical Infinite Stab, Sg S < 3 cm 0.75 g
S > 3 cm 0.85 g
Safe Concentration, C, Critical Concentration, C 0.50 p
TABLE l-4.2 SAFETY FACTORS FOR HETEROGENEOUS SINGLE UNITS SAFETY SAFETY PARAMETER CRITICAL PARAMETER FACTOR Safe Mass, M, Critical Mass, M 0.45 g
Safe Mass, M,t (Doubling of M,t Critical Mass, M 0.70 g
is excluded by construction)
Safe Spherical Volume, V, Critical Spherical Volume, 0.75 Vg Safe infinite Cylinder, D, Critical infinite Cylinder, D 0.85 g
Safe infinite Stab, S, Critical infinite Stab, S 0.55 l
g fh G
autwoucwt amc4Tcm oart:
September 12,1994 PAGE NO.:
4-14 SPC-ND:3330 947 (R-1/07/92)
Siemens Power Corporation - Nuclear Division Eur.2
(") SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 Lj PART I-LICENSE CONDITIONS
- ggy, CHAPTER 5 ENVIRONMENTAL PROTECTION 5.1 Effluent Controls 5.1.1 Gaseous Effluent Controls SPC shall maintain and use gaseous effluent treatment systems to maintain releases of radioactive material in gaseous effluents to unrestricted areas as far below the limits specified in 10 CFR 20.1302 as reasonably achievable. The major radionulides that can be emitted in gaseous effluents are U-234, U-235, and U-238. The chemcial forms of those nuclides include UF, UO F, UO (NO )2, UO, and U 0 -
s 22 2
3 2
38 Continuous isokinetic sampling shall be provided on all exhaust air stacks servicing areas in which uncontained radioactive materials are used, processed, or otherwise handled.
These samples shall be analyzed for radioactive material content on a weekly basis.
Samples of gaseous effluents potentially containing uranium shall be analyzed for gross alpha activity. The action levels described in Table I-5.1 shall be adequate to assure that
[o1 the effluent treatment systems are functioning properly and that immediate steps are KJ taken to rectify any observed deficiencies as soon as practicable.
In the event that the radioactivity in total gaseous effluents exceeds 50 pCi per calendar quarter, a report shall be submitted to the NRC within 30 days, identifying the cause and corrective actions to be taken by SPC to reduce release rates (Table I-5.2).
l if parameters important to dose assessment to the public (such as the distance to the nearest resident) change relative to gaseous effluents from the plant, a report shall be j
submitted to the NRC within 30 days, identifying the changes in the parameters, and providing an estimate of the resultant change in dose commitment.
Based on the average of the first 11 months and an estimate of the 12th,if the calculated dose to any member of the public in any consecutive 12-month period is about to exceed the limits specified in 40 CFR 190.10, SPC shall take immediate steps to reduce emissions so as to comply with 40 CFR 190.10.
Figure 1.1 in SPC's " Emergency Plan and Procedures", EMF-32, identifies the locations of all exahust stacks.
5.1.1.1 ILEPA Filtration Air from contaminated areas and process equipment or enclosures, where uncontained m
uranium compounds are handled, shall be passed through one stage of fire-resistant (JI HEPA filtration which meets Military Specification MIL-F-51068 prior to being exhausted x
September 12,1994 5-1 "E "*
$PCMDM30 947 (R-1107/92)
1 Siemens Power Corporation - Nuclear Division sur2 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 PART I - LICENSE CONDmONS REv.
to the atmosphere. Filtration is not required for air exhausted from rooms or facilities where radioactive materials are encapsulated or otherwise contained, and the outside surfaces of the containers have been surveyed and released.
Only HEPA filters which are certified by the manufacturer to be at least 99.97 percent i
efficient for the removal of 0.3 micron particles shall be used. The adequacy of final HEPA filter installations shall be determined by in-place testing to assure that they are at least 99.95 percent efficient for the removal of 0.8 micron particles prior to initiating operations with radioactive materials in the following instances:
i 1.
Startup of a new facility; t
2.
Following maintenance work on the filter bank and/or replacement filters; or 3.
Following actuation of the sprinkler deluge system if determined necessary by Plant Engineering following a visual / operational inspection.
5.1.1.2 Final HEPA Filter Protection Air exhausted from process equipment containing corrosive fumes may be liquid scrubbed, heated and/or diluted prior to final HEPA filtration. Final HEPA filters potentially exposed to corrosive fumes shall be visually inspected at least quarterly.
Final HEPA filters in the fuel fabrication buildings shall be protected from damage in the event of fire by one of the following design features:
1.
A liquid scrubber in the exhaust air stream upstream of the filters; or 2.
An automatic (actuated by rate-of-rise / heat detectors located in the exhaust air ducts) fog deluge system in the exhaust air stream upstream of the filters.
Final HEPA filter installations shall be equipped with continuous pressure differential measuring / indicating devices, the readings of which shall be recorded at least monthly.
The pressure differential across final HEPA filters shall not exceed four inches of water gage.
5.1.2 Llauld Effluent Controls SPC shall maintain and use liquid effluent treatment systems to maintain releases of radioactive material in liquid effluents to unrestricted areas as far below the limits specified O
in 10 CFR 20.1302 as reasonably achievable antt in accordance with 10 CFR 20.2003.
aucuoutm amcuos out:
September 12,1994 5-2 SPC-ND3330 947 (R 1/07/92)
Siemens Power Corporation - Nuclear Division EM F-2 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 PART I-LICENSE CONDITIONS g y, Procedures shall be adequate to assure that the effluent treatment systems function properly, and that immediate steps are taken to rectify any observed deficiencies as soon as practicable. See Table I-5.3 for action levels and required actions regarding liquid effluents.
Released liquid wastes are combined and discharged to the SPC-City lift station where the total combined liquid effluent from the plant is pumped to the Richland Municipal Sewerage System. The combined liquid effluent shall be continuously sampled at the SPC effluent station and the flow measured at the lift station. The composited samples shall be analyzed for uranium and regulated chemicals. Any increase in the chemical or uranium content of the composited samples statistically above those limits described in 10 CFR 20.2003 or the State of Washington Uquid Waste Discharge Permit shall be cause for an investigation and appropriate corrective action.
5.1.2.1 Sanitary Wastes Sanitary wastes shall be discharged to the sanitary sewer system which joins other liquid wastes prior to being discharged to the SPC-City lift station.
5.1.2.2 Process Coolina Waste Water Process cooling water shall be isolated from actual process atmosphere by double i
physical barriers. Process cooling water shall be discharged from various facilities (with the exception of the ELO Building) via building sewer systems separate from both sanitary and process chemical waste sewers. Process cooling water may be disposed of by discharge to the municipal sewerage system or discharge to the SPC Process Chemical Waste Lagoon System.
5.1.2.3 Process Chemical / Radioactive Wastes All process radioactive liquid wastes shall be routed to the lagoon system. Some lagoon solutions require further treatment prior to discard to the sewer. These solutions shall be treated, as necessary, for chemical / radioactivity removal prior to release to the sanitary sewer system.
The release of chemical wastes to the sanitary sewer system is controlled by local authorities via a permit system. The licensee shall notify NRC for informational purposes of any occurrences which, by permit, require reporting to the authorities.
=tututut amcaron oatt:
September 12,1994 5-3 I
sPc ND:3330 947 (R-907/92)
l Siemens Power Corporation - Nuclear Division eur.2 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 l
1 PART I-LICENSE CONDITIONS t
- ggy, 5.1.3 Process Chemical Lanoon Management System The lagoons shall be sealed on the bottom and all sides with an impervious liner to I
prevent the migration of lagoon contents to the adjacent subsurface soil or groundwater.
The liner consists of a double layer of impervious material, separated by a layer of sand or other material used to maintain spacing between liners. A system of sampling tubes shall be installed between the liners to provide sampling capability to permit detection of leaks in the upper liner, j
{
Routine monitoring of the integrity of the upper liners shall be accomplished by drawing a vacuum on each group of the "between-liners" sampling heads at least monthly, unless l
weather conditions (e.g. ice and/or snow) make it too hazardous to personnel to perform j
such sampling. In the event that a significant amount of liquid is pumped from any i
sampling head (s), the liquid shall be analyzed for fluoride and uranium content. If uranium and fluoride are present above previously measured levels, an investigation shall be initiated which shall include:
1.
Additional between-liner sampling; 2.
Lagoon solution sampling for comparison of the content of the sample to that of the lagoon; 3.
Checking the integrity of the lower liner of affected lagoons; for Lagoons 1 and 3 by activating sampling lines located between the lower liners and the original Petromat liners of these two lagoons; or checking the integrity of the lower liner of Lagoon number 4 by sampling the three dry wells associated with the three
- French Drains" located under the lower liner of this lagoon; and 4.
Making use of the lagoon test well system.
The between-liner sampling system is the first line of defense for detecting liner leaks, and sampling shall be scheduled monthly. The beneath-the-bottom liner leak detection system shall be activated at anytime the upper liner is determined to be leaking to confirm the integrity of the lower liner. Test wells are provided around the lagoon system as a backup to provide the capability to detect leaks penetrating both liner layers.
In the event that a leak in an upper liner is confirmed, the liner shall be repaired. A report of the leak, including results of the investigation and corrective actions taken, shall be forwarded, in writing, to the Administrator, NRC-Region V within 90 days of detection of I
the leak.
4 September 12,1994 5-4 SPC-ND3330.947 (Raio7/92) u- -
i SiemenS Power Corporation - Nuclear Division eur.2 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 PART I-LICENSE CONDITIONS
- ggy, 5.2 Environmental Monitorina SPC shall conduct a routine environmental surveillance program in relation to the operation of the Engineering and Manufacturing Facility. Surface and groundwater samples shall be collected frorn strategic locations in the environment and analyzed for pertinent chemicals and uranium.
5.2.1 Surface Samplina Sampling stations shall be established both on-site and off-site near points of expected maximum concentrations. The schedules for the various sampling stations are identified below. See Figure I-5.1 for the location of the stations.
Sample Station Sample Tvoe Samplina Freauency Analysis 1
1 Soil Quarterly Uranium 2
Soil Quarterly Uranium 3
Air Monthly Fluoride O
4 Air Monthly Fluoride
[
1 5
Forage Monthly Fluoride 1
6 Forage Monthly Fluoride 5.2.2 Groundwater Samplina Section 5.1.3 describes the between-liners sampling of the lagoons as well as the actions taken to confirm and repair possible leaks. The groundwater sampling program is described below. See Figure I-5.2 for the locations of the sample stations.
Presence of liquid Monthly Lagoon interliner sampling system Grab 2
Gross Alpha / Beta,
Quarterly GM Wells 1,5,6,7, and 8 and Grab chloride, No -N, TW Wells 6,7, and 21 3
NH -H, and pH 3
1 During the growing season only (April-October).
2 The analytical method shall be capable of detecting 5 picoeuries/ liter alpha and 15 picocuries/ liter beta.
September 12,1994 5-5 SPC-ND'3330 947 (R-UD7/92)
Siemens Power Corporation - Nuclear Division eur.2
[] SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO PART I - LICENSE CONDITIONS
- ggy, t
t 5.2.3 Sanitary Sewer Sludoe Samplino The release of radioactive material to the sanitary sewer system shall be controlled and I
monitored as described in Section 5.1.2. At the Richland sewage treatment facility, the slutige it removed from the process, de-watered to a semi-dry solid, and trucked to a sanitary landfill on a daily basis. Samples of sludge shall be taken monthly by SPC and analyzed for uranium. The analyses shall be converted to picocuries of uranium per gram of sludge as transferred to the landfill. If a running average of the analyses over a six-j month period exceeds 25 picoeuries per gram or any single confirmed result equals or exceeds 30 picoeuries per gram, an investigation shall be required, and a plan of action instituted. The action plan, as a minimum, shall require a reduction of discards to the sewer system until the sewer sludges contain less than 25 picocuries uranium per gram.
Any confirmed monthly sludge sample result of 25 picoeuries per gram or higher shall be brought to the attention of Chief, Fuel Cycle Safety Branch, NRC.
t O
V i
l i
AMEON AMMm DAE NE NO" September 12,1994 5-6 SPC-ND.3330.947 (T4-147S2) l
Siemens Power Corporation - Nuclear Division sup.2
(~'
SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 PART I - LICENSE CONDmONS
- agy, Figure I-5.2 Lagoon Test Well Locations Hom Rapids Road
\\
1 i
GM-8 GM-7 GM-5,
. GM-6
_..q l UGCCN S AI1.'
I l
i I
LAGCCN 1 '
Il I?
l 3
il j
il j
i guGCCN 2 l UGCCH $51
=r '
g i
- TW-21 r -*- - -'
uacCN c:
j TW-6 e
e s,-i G
TW-7 i
i uGCCH s i
i r
i 4
i i
I GM-1 s
AMENDMENT APPLCATION DATE:
PAGE NO,:
September 12,1994 5-11 SPC-ND.3330 947 (R 1/07/92)
Siemens Power Corporation - Nuclear Division EM F.2 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 J
PART I - LICENSE CONDITIONS REv.
6.4.2 Solid Radioactive Waste Uranium-contaminated solid waste is segregated into noncombustibis and combustible types, and is stored in designated areas within the controlled access ama. Containers used for this purpose shall be adequately sealed and apprcpriately labeled orior to being stored. In the event that such containers of waste are stored outdoors for extended periods of time, their physical integrity shall be visually inspected, and the accumulation shall be surveyed for external radioactive material contamination at least quarterly, and records of such inspections and surveys shall be maintained.
Combustible waste may be processed by incineration through the Solid Waste Uranium Recovery (SWUR) facility to obtain volume reduction with the ash being stored for the recovery of uranium. Ash uranium recovery may be pursued by the future installation of ash processing facilities; alternatively the uranium recovery may be performed by an offsite processor. When incineration at SWUR is not utilized, combustibles may be compacted, offsite or onsite, prior to burial at a licensed Low Level Radioactive Waste (LLRW) disposal site.
O Noncombustible waste is stored prior to shipmcat to a permanent waste disposal site.
)
G/
In the future, specialized facilities may be construcied onsite to allow for the radiological decontamination of non-combustible items, thereby icducing the volume of non-l combustibles for disposal at the LLRW disposal site. Waste packaged for disposal shall not be allowed to remain in storage for extended periods, but shall be scheduled for j
disposal on a current basis depending upon generation rate and cost-effective shipment sizes.
Certain containerized LLRW, not amenable to processing at SWUR (e.g., non-combustible HEPA filters or cedain chemically contaminated combustible wastes), may contain uranium in economically recoverable quantities. Processing of these containerized wastes for uranium recovery may be pursued via future construction of onsite processing facilities by SPC or via processing by a commercial vendor, onsite or offsite, 6.5 UF Cylinders e
New UF cylinders purchased by SPC shall conform to ANSI N14.1, which includes s
certification by the vendor that the cylinders comply with all fabrication, test, and cleanliness requirements specified therein. Periodic inspection and testing of cylinders shall be performed following heel removal. The heel removal procedures shall specifically exclude the use of hydrocarbons.
Cylinders of UF shall be received, unloaded, and stored within barricaded pads.
s Evacuated UF cylinders (containing heels) shall also be stored at these locations. As 6
(g) needed for processing, cylinders of UF are transferred to either an elevated dock or a g
V barricaded pad adjacent to the UO Building. Elevated or barricaded storage of bare UF 2
g autmout=T amcaTom cart:
Se tember 12,1994 6-3 SPC-ND.3330.947 (R-1/07/92)
t i
Siemens Power Corporation - Nuclear Division eur.2
} SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257
(
)
t PART I-LICENSE CONDmONS REV.
}
cylinders is designed to guard the cylinders against vehicular damage. Bare UF6 i
cylinders shall be stored in cradles providing spacing and stability.
Prior to shipping bare cylinders containing heels, the valves shall be covered and sealed.
When the cylinders are shipped in overpacks, the valves are not covered and sealed, but j
1 the overpack shall be sealed.
O t
i l
f O
AMENDMENT AMCATON DATE:
$gptgmbgr 12, j gg4 PAGE NO.:
g SPC-ND:3330 947 (R-vo7/92)
i Siemens Power Corporation - Nuclear Division sup.2 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 l
PART 11 - SAFETY DEMONSTRATION
- ggy, CHAPTER 9 GENERAL INFORMATION 9.1 CorDorate Information The names, addresses, and citizenship of the current principal officers of SPC are listed in Table 11-9.1. All of the common stock of SPC is owned by Siemens Corporation, a Delaware corporation with headquarters in New York City. Siemens Corporation is, in turn, wholly-owned by Siemens AG headquartered in Munich, Federal Republic of I-Germany.
The top organizational structure of the Nuclear Division of SPC is depicted in Figure 11-9.1.
l The Senior Vice President and General Manager reports to the President who reports to the Board of Directors and directs all Company functions. The handling, processing, and storage of special nuclear material, to which this license applies, is under the direction of the Plant Manager, Richland Plant, Nuclear Division, and Vice President, Engineering, Nuclear Division.
There are five departments within the Engineering and Manufacturing Facility which are g/
involved in the handling and/or processing of special nuclear material. These are Plant p
Operations, Materials and Scheduling, Manufacturing Engineering, Quality Control, and Research and Product Development.
Additional descriptive information on these organizations was presented in Chapter 2.
9.2 Financial Qualification A copy of the Annual Report for 1991 of Siemens AG, which describes the financial condition of the corporation with which SPC is affiliated, is appended at the end of this chapter.
r 9.3 Summary of Operatina Oblective and Process 9.3.1 General The objective of activities covered by this license application is the production of low enriched uranium fuel for light water reactors. This is accomplished by beginning with UF as a feedstock and incorporates processes ancillary to the main process including l
g liquid waste treatment and disposal, scrap recovery and recycle, and combustible waste incineration and disposal. Plant capacity is nominally 700 metric tons per year. A very j
brief description of the mainline process is given below.
ho AMMENT AMCATON DAW September 12,1994 p
PAGE NO.:
l SPC ND:3330 f,47 (R 1107/92)
+
SiemenS Power Corporation - Nuclear Division sur.2 V] SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET N
/
PART ll - SAFETY DEMONSTRATION ggy.
9.3.3 Uranium Scrare Reprocessina Discrepant UO is recycled into the process by oxidizing to U O and then blending into 2
3 g UO2 Powder or by dissolving UO in nitric acid to form uranyl nitrate which is pumped 2
into either conversion line and precipitated using the ADU process.
Mixed impure scrap (floor sweepings, grinder sludge, laboratory samples, etc.) is accumulated in a designated storage area and reprocessed on a campaign basis. The process for uranium scrap reprocessing is the classic nitric acid dissolution, solvent extraction-stripping, ammonium diuranate precipitation, calcination technique.
9.3.4 Process Chanaes Since Last License Renewal 9.3.4.1 Laaoon SA lon Exchance UX) System This process consists of pumping waste solutions from Lagoon SA through an existing primary sand filter to remove most of the suspended solids. The primary filter is located at the lagoon pump near Lagoon SA. Next the solution goes through a secondary sand
(]
filter with finer media to trap any remaining solids that might clog the ion exchange G/
column.
The solution then goes through an IX column to reduce the uranium concentration of the solution before discharge to the city sewer.
9.3.4.2 UO Dissolver 2
This process change is described in Chapter 15, paragraph 15.1.8.1.4.
9.4 Site Description Richland, along with Pasco and Kennewick, form the Tri-Cities metropolitan area in the
{
Columbia Basin area in the Southeastern portion of the State of Washington with a population of approximately 160,000. The SPC site lies just inside the northern boundary of the City of Richland on 320 acres. The fenced exclusion area is basically rectangular shaped and lies on the northern portion of the site. The center of the plant is approximately 930 feet south of the Horn Rapids Road which forms the northern boundary of the site.
The prevailing wind at the site is from the southwest, along the Yakima River corridor.
Secondary wind directions are from the northwest and southeast. The average annual precipitation in the Richland area is 6.4 inches, which is typical of a desert climate. The temperature falls below freezing approximately 100 days per year, with an inch of snowfall on the average of twice each month during December and January.
Maximum
')
(O temperatures of 100 F or more are recorded during July and August.
\\
acuoucut Amentom onic:
September 12,1994 9-3 SPC-ND.3330 947 (A.1/07/92) l
i r
Siemens Power Corporation - Nuclear Division eur.2 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 PART 11 - SAFETY DEMONSTRATION I
REv.
Figure 11-9.2 l
!A 1
i c
e
[
,!!i.
55 i
1
$5 h.*
cn sc
~
JI 2'3 g;
1 a
ty
,j g =:
gg e
e.
1 jg
.g lE
$U $
\\i my
-. - ~
)){
E"
?
(! eh' Si!
I$
lli-i--
??
i i11 i
zw Q
fl
,(
I i%'
11
['"~J
- - -,1
,?
j
' !}] 3 } lI D' t_ g? l so l W j e l l <[ . /25 o[ 7x l} y s u e /, isf l 5 t j} _ / l n \\- t y 1 }: I f I [ il l r ~ p .p-
- li (1
{b N' \\ F [ 9 i3 N ',II. _ I autacwT amcarm oarc: September 12,1994 9-8 spc-wo saso w <n-voma j
I g3$m T Eo, OQVo'ml63 ' Zco5m, ~mg _ pO oOO mH ZO' n Lu ? r mmO5r Eorm>m gyymu%r-COmzmm z . m>YMa ! e D a< ) A(qeNc y %4 e= W %%sg s .N. Q4 g!, I-
- u
_Wg__*, "(k - L N=~ 1a I_ s o l ,ku = o hms io o 5. o n t bv / t ?m I I ray sf &l, oti o, ]~\\ ' _ - pt c r l a o F C .5t g ]l1 re n cL wri r c ) u ,. 4 ot , o Pc .o a o n,m. a o o i,., ,i, nu
- r. ts i...
en ,eo.. ma ? . C n. ~ m. ieM p o. r . on i. Sd ~ n s e ~- d a , f-n j}. ag ~. n ai / ]e er r e Ae .W c n a e n i di ng o::,[ 5 fan E h [- lc 7 NggH ~ 7'~- s T fg. \\ 4, O % e1 a i f. m.., av e m A.- [._. ~ , $4 4m*.*m/ 0".k 7 N m r f + +% 4 / / l / f,. h4},e. / p i. g .,5z! 5E9eB iE F m *n v b fmlm~.og i g y IS t 'p .8i" i gM, I
8 Siemens Power Corporation - Nuclear Division s u s.2 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 v PART 11 - SAFETY DEMONSTRATION g y, c 10.1.2 U.0 Buildina Description 2 The UO Building is SPC's main operating facility and houses the majority of the fuel 2 fabrication activities. Since 1970 when the original building was constructed, there have l been numerous additions, modifications and expansions. The building, as it currently l exists in 1994, is described in this section. The footprint of the building is shown on Figure 11 - 10.3. The basic facility consists of a 2-story,133 ft wide x 367 ft long building. Equipment room [ and administrative areas are located on the second floor at both the north and south ends of the building. An 80% x 112 ft,2-story wing on the northeast corner houses Unes l 1 and 2 UF conversion systems as well as other chemical operations. The outside walls g of the UO Building are 6-in thick, precast, reinforced concrete slabs attached and sealed 2 to reinforced concrete columns. The roof is a modified bitumen roofing system. The inner side of the precast wall panels, except in the UF conversion wing, are insulated s and have a fire-rated gypsum board interior applied on furring strips. The gypsum panel joints are taped and sealed and the interior surfaces are suitably painted. For ease in decontamination, the inner side of the wall panels in the UF conversion wing have the O s concrete finished smooth and painted. The compressive strength of all concrete exceeds 3000 psi and the 6-in thick concrete floor slabs were designed for 250 psf. The portion of the building south of column 19 was designed structurally for Zone i earthquake loadings (0.25 g) where the remainder of the building was designed for Zone 11 earthquake loadings (0.17 g). The interior partitions throughout the building are poured in place concrete, concrete block or fire-rated gypsum board. The analytical laboratory addition consists of a two-story 35 feet wide by 70 feet long building attached to the west side of the north UO Building. The outside walls are CMU j 2 concrete block. The roof is constructed of steel deck plates supported on steel trusses t and covered by a vapor barrier insulation plus a twenty year, built-up, asphalt roof. The inner sides of the concrete block walls are insulated and have a fire-rated gypsum board interior applied on furring strips. The gypsum panel joints are taped and sealed and the interior surfaces are suitably painted for ease of decontamination. The compressive strength of the poured concrete is 3000 psi and the slab floors were designed for 250 psf and 100 psf for six-and four-inch thick slabs, respectively. The wind load design is for 70 mph (sustained) and the seismic design is zone 28. The U 0 Facility, located immediately south of the UF conversion wing, is a 2-story 3s s structural steel facility. The building is approximately 29 x 50 x 24 ft high. The exterior walls and roof are 24-gauge sheet metal, insulated and finished on the interior with fire-rated gypsum panels. Care was used in sealing the sheet metal and gypsum panels in order to maintain the required pressure differentials and airflows in the metal structure., t All interior walls used for process control are fire-rated gypsum board on metal studs. All 3 ] interior surfaces are taped and sealed and suitably painted. The first and second story ^ " " * " ' * ^ " "
- September 12,1994 10-3 sPC-ND.3330.947 (R-U07r92)
Siemens Power Corporation - Nuclear Division eur-2 (G SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 d PART 11 - SAFETY DEMONSTRATION g y, j floor slabs are 3000 psi concrete. The Powder Storage Facility, located south of the Une 1 UF conversion area, is a high bay,50 x 50 x 24 ft high, pre-engineered metal building. s The exterior walls and roof are 24-gauge sheet metal, insulated and sealed. The floor slab was designed for 250 psf and made of 3000 psi concrete. The Tube Cleaning Room, located on the south end of the UO Building is a pre-engineered metal building, 2 40 x 24 x 16 ft high. The exterior walls and roof are 24-gauge sheet metal, insulated and finished on the interior with fire-rated gypsum panels. Interior partitions are fire-rated gypsum board on metal and/or wood studs. The gypsum panel joints are taped and sealed, and the interior surfaces are suitably painted. The floor slab was designed for 250 psf and made of 3000 psi concrete. Figures ll - 10.4, ll - 10.6, and 11 - 10.7 show the major equipment on both floors of the UF conversion wing. The two conversion l'.nes occupy Rooms 104A and 131 on the first 6 i floor and Rooms 204A and 210 on the second floor. SPC currently uses the Ammonium Diuranate (ADU) process for the majority of its conversion work. This process consists of the following major steps: vaporization, hydrolysis, precipitation, centrifugation, drying i and calcination. A proprietary, dry conversion system is also located in Room 104. The Miscellaneous Uranium Recovery System (MURS) is located in Room 104 and is h described in other sections of this Application. A UNH facility, which is used to dissolve b U 0 or UO in nitric acid to form uranyl nitrate, is located in Room 101 A. The UO 3s 2 2 powder lot bFending, compaction and granulation is accomplished in Rooms 104 and 128; whereas lube blending is accomplished in Room 180. The Scrap Recovery Facility and pellet dissolver are located in Room 240. The equipment locations and floor plans of the 2-story U 0 and the Powder Storage 3s Facilities are shown on Figure II - 10.8. The lower floor of the U 0 Facility is used 3s predominantly as a maintenance shop for work on contaminated equipment. A small area i of the lower floor adjacent to the powder storage area is partitioned off and used to load drums with powder from the second floor. The second floor of the facility is used to convert scrap UO to U 0. The U 0 is then screened and loaded into drums for 2 38 3s storage or transport for further processing. The Powder Storage Facility is used to store sealed buckets or drums of dry materials, including UO, U O, scrap, etc., awaiting 2 a3 further processing. ) Figure il-10.9 is a plan view of Room 100 which is a high bay area and contains ceramic process equipment. The process flow starts at the north end of the room where the UO2 powder is directed by hoppers to the pellet presses. The green pellets are processed through the sintering furnaces. The sintered pellets are then fed to the grinders, the wash station, dry station, and then through the pellet inspection stations. Once the pellets are accepted, they are placed in pellet storage cabinets until needed for further processing. The Powder Characterization Facility is located in Room 100A at the south end of Room 100. The Powder Characterization Facility contains the necessary equipment to conduct p) laboratory scale tests on each UO powder lot. ( 2 v AMENNENT A%CATON DATE: " ^ '
- September 12,1994 10-4 SPC ND.3330.947 (R-1/07/92)
Siemens Power Corporation - Nuclear Division sup.2 f) SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 v PART 11 - SAFETY DEMONSTRATION REv. Fuel rod tubing is processed through the Tube Cleaning Facility (Figure 11 - 10.10) prior to use in the manufacturing stream. The tubing is cleaned and inspected before it is transferred to Room 182 tor first-end welding. Figure 11 - 10.10 also shows the equipment arrangement of the south Equipment Room and the administrative area on the second floor. Figure 11-10.11 shows the space and equipment devoted to the rod fabrication phase of the manufacturing process. Room 182 is a high bay area, where Room 189 has headroom of approximately 9 ft. Pellets are removed from the storage cabinets in Room 100 and processed through the outgas furnaces in Room 182. After the pellets are outgassed, they are loaded into first-end welded tubes in the rod loading glove boxes. After rod loading, the second-end weid is made. The rods are then processed through the helium leak check, rod assay tester and rod x-ray equipment which is all located in Room 189. After the nondestructive testing in Room 189, some types of BWR fuel rods are transferred to Room 110 (the Etch Room) shown on Figure 11-10.12. The rods are cleaned in cleaning solutions and etched to remove any surface contaminants. Offgas 7 ( from the etching equipment is exhausted through a scrubber and HEPA filters. After L etching, the rods are transferred to Room 107 where they are processed through high-pressure, high-temperature autoclaves. The autoclaves are located in a reinforced concrete pit approximately 20 ft deep. After autoclaving, the rods are transferred through various inspection stations to the final rod storage area in Room 193. PWR and nonetched BWR fuel rods, on the other hand, are transferred directly from nondestructive testing in Room 189 to the final inspection table in Room 107 and then to the final rod storage area in Room 193. The bundle assembly, cleaning, inspection and shipping container loading operations are all done in Room 193 (shown on Figure 11 - 10.13). Rods are moved from the rod storage area to the rod insertion tables. The fuel bundle is assembled on the bundle assembly table and is then lifted to the vertical position on the assembly table and moved onto the bundle inspection granite column. After inspection, the bundle is cleaned, dried, and moved to the storage area or to the shipping container loading area where it is prepared for shipment. Immediately west of Room 193 is Room 195 where cage fabrication work is accomplished. Figure 11-10.14 shows the low bay area of the UO2 Building that is devoted almost exclusively to the various Quality Control and inspection functions that are necessary in the manufacture of nuclear fuel. These areas include chemical and physical testing laboratories, metallurgical laboratories, and a limited number of offices. The north end Equipment Room and administrative areas are shown on Figures II-10.15 and ll - 10.16. J ^ " " * " " " ^ " ^
- September 12,1994 10-5 i
SPC-ND:3330 947 (R-v07/92)
Siemens Power Corporation - Nuclear Division eur.2 (G] SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET N PART 11 - SAFETY DEMONSTRATION REv. 10.1.3 ELO Buildino Description The current ELO Facility consists of an original building constructed in 1974, and an 2 expansion constructed in 1982. The original facility consists of approximately 6000 ft on 2 the ground floor and 3000 ft in the basement. The addition, which effectively doubled the building size, utilized the same basic type of construction. The 1982 addition 2 2 consisted of approximately 6000 ft on the ground floor and 4000 ft in the basement.The entire ELO Facility is currently operated as a single facility and treated as such in this Application. The floor plan for the ELO Facility is shown on Figure II-10.17. The basement areas are constructed of poured-in-place concrete walls and floor slabs. The ground floor construction (approximately 70 x 172 x 12 ft high) is of concrete block on poured-in-place skirt walls. The inner side of all exterior walls are insulated and covered with fire-rated gypsum board. The gypsum panel joints are taped and sealed, and the interior surfaces are suitably painted. The interior partitions throughout the building, used for process control, are either concrete block or fire-rated gypsum board. The roof is made of steel deck plates supported on steel trusses, covered with insulation, and a 20-year, built-up [V roof system. The compressive strength of all concrete exceeds 3000 psi and the concrete floor slabs were designed for 250 psf. The basement area contains a chemical engineering development area, an engineering machine shop, the Gadolinia Scrap Recovery Facility, and other engineering test facilities. The first or ground floor contains an instrument laboratory, metallography laboratory, wet chemical laboratories and several engineering offices. The uranium enrichment in the various laboratories is less than 20 wt% U-235 in solid form, and not more than 5 wt% U-235 in liquid form. Activities include all operational steps of fuel manufacturing and process testing. 10.1.4 Contaminated Clothino Laundry The location of the Laundry Facility is shown on Figure ll - 10.1. The building is a combination of a pre-engineered metal building (20 X 30 X 12 feet high) and a concrete block structure (16 X 20 X 11 feet high) with a 6'9" x 17'9" airlock which provides for access, yet maintains the proper pressure differentials for contamination control. The exterior walls are 26-gauge sheet metal and concrete block and the roof is 24-gauge sheet metal, and built up metal insulated and finished on the interior. The interior of the roof is a suspended metal ceiling grid with acoustical lay-in panels 5/8-in thick. The interior of the metal walls are finished with fire-rated gypsum panels. Interior partitions are fire-rated gypsum board on metal studs. Care was used in sealing the sheet metal and n gypsum panels in order to maintain the required pressure differentials and airflows in the Iv) mcacut amcatcmioart: September 12,1994 10-6 SPC-ND.3330 947 (R-907S2)
Siemens Power Corporation - Nuclear Division sus.2 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 PART 11 - SAFETY DEMONSTRATION ggy. I i metal structure. All interior surfaces are taped and sealed and suitably painted. The floor was made of 3000 psi concrete and designed for 250 psf. Contamiaated clothing is cleaned in a water wash system, surveyed for contamination, i and returned to use. The effluent is directed to retention tanks where it is sampled prior i to being emptied either to the site lagoon system or the City sewer. l 10.1.5 Fuels Storace Warehouse The location of the Fuels Storage Warehouse is shown on Figure 11 - 10.1. The building i is a pre-engineered metal building 60 x 40 x 16 ft eave height. The exterior walls are 26-gauge sheet metal and the roof is 24-gauge sheet metal. The interior of the walls and roof are fully sealed with closure strips and insulated. The floor slab was made of 3000 psi concrete and designed for 250 psf. The warehouse is used for the storage of packaged special nuclear material in various compounds and forms. i 10.1.6 UNH Drum Storaos Warehouse t The location of the UNH Drum Storage Warehouse is shown on Figure II - 10.1. The building is a pre-engineered metal building 100 x 50 x 16 ft eave height. The exterior walls and roof are 26-gauge sheet metal. The interior of the walls and roof is fully sealed with closure strips and insulated. The floor slab is made of 3000 psi concrete and designed for 250 psf. The floor slab has a 6 inch curb and is coated to provide secondary containment and facilitate decontamination of possible spills or leaks. l The warehouse is used for the single tiered storage of UNH drums of less than 140 gU/t and less than 5% wt. U-235. 10.1.7 Radioactive Materials Warehouse i The location of the Radioactive Materials Warehouse is shown on Figure 11 - 10.1. The \\ building is a pre-engineered metal building 50 x 140 x 16 ft_ eave height. The exterior walls are 26-gauge sheet metal and the roof is 241auge sheet metal. The interior of the walls and roof are fully sealed with closure strips and insulated. The floor slab was made of 3000 psi concrete and designed for 250 psf. The warehouse is used for the storage of packaged special nuclear material in various ompounds and forms. 10.1.7.1 Ancillary Radioactive Material Storace l
- * ' " ^ * * *
- September 12,1994 10-7 l
SPc-ND-3330 947 (R-1107/92)
i I Siemens Power Corporation - Nuclear Division EMF-2 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 PART 11 - SAFETY DEMONSTRATION y y, On occasion, additional storage for uranium oxide product material is required. During those occasions, pre-engineered enclosures such as enclosed sea containers or trailers are used. These temporary warehouses are weather-tight and are positioned such that criticality alarm coverage is provided by existing detector units. I 10.1.8 Product Development Test Facl_lity (PDTF) The location of the PDTF is shown on Figure 11 - 10.1. The building is a high bay,29-ft eave height, pre-engineered metal building 40 x 40 ft. The exterior walls are 26-gauge sheet metal and the roof is 24-gauge sheet metal. The interior of the walls and roof are fully sealed with closure strips and insulated. A concrete pit (15 x 8 x 18 ft deep) inside the building provides space for various test vessels and piping systems. All concrete i used in the construction of the building had 3000 psi compressive strength and the floor slab was designed for 250 psf. A 3-ton bridge crane and hoist services the entire interior of the building. Adjacent to, and constructed as part of the main building, is a 40 x 20 x 12 ft high addition made of the same material as the main building which houses the _ process boiler. The PDTF houses an hydraulic test loop that is used to determine heat transfer / performance of nuclear fuel assemblies during spray and reflood cooling following a postulated loss-of-coolant accident (LOCA). l 10.1.9 UF. Receivina and Storace Facility The location of the UF Receiving and Storage Facility is shown on Figure ll-10.1. The s complete facility consists of: 1) an open-air,5-ton bridge crane and hoist supported by a steel structure which operates over an area of 165 x 25 ft; 2) a 10 x 10 x 10 feet high 2 sheet metal scale house; and 3) an asphalt area of approximately 8500 ft used for storage of UF cylinders. s The open-air bridge crane is used to load and unload UF cylinders on and off semi-6 2 trucks. The crane structure provides an under-crane storage area of about 4000 ft where 3 the UF cylinders can be moved using the 5-ton crane and hoist. The truck access bay s 2 and a total of approximately 1700 ft is provided with a sheet metal rain cover. The scale house provides a weatherproof structure for the certified scales used to weigh full and empty UF cylinders. An additional asphalt storage area of approximately 9000 ft2 g, s provided near the crane structure where UF cylinders can be handled with a forklift. The 6 entire facility provides load, unload and storage capabilities for up to 180 UF 30-inch s - cylinders. i, 1-September 12,1994 10-8 SPC-ND 3330.947 (A*0742)
- l
) 1 i
= _ - SiemenS Power Corporation - Nuclear Division EMF-2 [ SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 PART ll-SAFETY DEMONSTRATION
- ggy, 10.2 Utilities and Support Systems In the day-to-day operation of a nuclear fuels fabrication plant, many utilities and support systems are required. Utilities are required for the manufacture of the fuel and for the protection and safety of the product as well as the employees. A brief description of the following utilities and/or services is provided in this Ucense Application.
1. Electric Power 2. Compressed Air 3. Water 4. Sanitary Sewer System l 5. Gas and Chemical Storage l 6. Communications and Annunciations 7. Breathing Air 10.2.1 Electrical Power The electrical power for the SPC plant site is provided by the City of Richland industrial C$ system. In addition, SPC has installed a standby emergency system to supply certain plant power needs in case the municipal system fails. Both systems are described in this section. 10.2.1.1 Utility (Normal) Supolv The normal supply is a 12.5 kV underground cable which connects to a Richland substation bus dedicated to north Richland industrial loads. The bus is directly supplied through two breakers and a transformer from a 115 kV Bonneville power loop. The 115 kV loop is, in turn, supplied by Columbia River and Snake River hydroelectric generation. The normal 12.5 kV cable forms an underground loop within SPC's plant site and each plant building supply is tapped into the loop through a 3-phase,12.5 kV grounded-wye to 480/277V grounded-wye transformation. i 10.2.1.2 Standbv (Emeroencv) Supolv Standby emergency power is provided by three diesel-fueled 225 KW generators and one ' propone fueled 12.5 KW generator. All generators are turbine driven and are 3 phase, grounded-wye connected to supply regulated 480 Volt,60 Hz power. Upon a loss of ) normal power, the ignition systems and battery-driven starters for these turbines are i automatically energized and automatic transfer switches cause the generators to supply - power to emergency loads. C4 September 12,1994 10-9 meatm eucum ous: 1 SPC-ND:3330.947 (R 007/92) ~ -. _ _. _. _ _. _. _ _ _ _j
l 1 Siernens Power Corporation - Nuclear Division EM F.2 ] SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 PART ll-SAFETY DEMONSTRATION REv. The three 225 KW generators are, respectively, located in the North and South Equipment Rooms of the UO Building and in a small building immediately east of the north end of the UO Building.2This third generator feeds the East Equipment Room. The generators 2 in the North and South Equipment Rooms have elevated day tanks which provide fuel at I a positive head to a direct-connected fuel pump. Additional fuelis stored in three 3000 gallon underground tanks, one for each generator. In the case of a generator failure, the - north and south generators can be connected by manual switching to supply both ' emergency buses. In this case, some nonessential loads must be switched off. The 12.5 KW generator, located west of the Specialty Fuels Building, supplies power to the security radio and emergency lighting in the Central Guard Station. l 10.2.1.3.QQ Buildina 2 Normal 480/277V power is brought into the UO Building through separate loop-tapped 2 transformers located at the north, south and east side of the building. The two north-end transformer secondaries supply a 2000-amp,3-bus panelboard (inside the building), and a 4000-amp,3-bus panelboard (located in a switch cubicle outside the building) which distribute 480/277V, 3-phase power to the north half of the building through local distribution panelboards. An additional 400-amp, 480/277V, 3-bus panelboard dedicated to critical loads and - identified as " North Emergency Panelboard" is supplied through an automatic transfer switch which, on loss of normal power, switches the critical loads to a standby generator. l The loads currently assigned to North Emergency Panelboard "E" are: l i 1. .QQ Buildina (North) l 2 K-9 Exhaust Fans I = K-10 (POG) = Instrument Air Compressor and Dryer j Standby HVAC Pneumatic Control Air Compressor .l = Air Sampling Vacuum Pump ^ = Recirculating-Air Radiation Detector = Selected Personnel Survey Instrumentation = Criticality Detector and Warning System a Emergency Ughting i Annunciator Power = Office Building #1 = HVAC Control Power = O Emergency Generator Battery Power September 12,1994 10-10 SPC-ND3330.947 (R 1/07/92) l
SiemenS Power Corporation - Nuclear Division sur.2 [] SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 V PART 11 - SAFETY DEMONSTRATION
- pgy, i
Day Tank Diesel Fuel Pump Exhaust Duct Deluge System = Sintering Furnace Unes 1,2,3 and 4 Control Power Combustible Gas Detector Safety Shower Ughts Calciner Rotation Communication System (Office Building #5) 2. SF Buildina Process Drain Retention Tank Sampling Pump Emergency Ughting K-6 Exhaust Fan Standby HVAC Pneumatic Control Air Cornpressor l Deluge System HVAC Control Power Safety Shower Ughts Annunciator O Criticality Detector and Warning System O Recirculation-Air Radiation Detector Battery Charger Security Alarm Gas Analyzers (0, H Combustible Gas) 2 Shutdown Exhauster WUR) A south end transformer secondary supplies 4000-amp,480/277V,3-phase ir: coming switchgear which distributes 480/277V,3-phase power to the south half of the building through distribution panels. A 600-amp,3-phase,480/277V bus designated as " Emergency Motor Control Center E-1" (MCC-E-1) is supplied by an automatic trancier switch, which upon loss of normal power, will instigate an automatic transfer of the MCC-E-1 loads to the oncoming turbine generator. The loads currently assigned to the south emergency system MCC-E-1 are: 1. UQ Buildino (South) 2 K-3 Exhaust Fan Compressor Recirculation Pump K-22 i Air Sampling Vacuum Pump Emergency Ughting p Instrument Air Compressor (Joy #3) Turbine Generator Fuel Pump amourwt amcatow oart: September 12,1994 10-11 SPC-ND.3330S47 (R UO7S2)
Siemens Power Corporation - Nuclear Division EMF-2 f) SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 C/ PART 11 - SAFETY DEMONSTRATION ggy. I Air Dryer and Compressor Cooling Water Control Exhaust Duct Deluge System Closed Loop Cooling System Turbine Generator Battery Charger a Autoclave Sump Pump Combustible Gas Detector l The east end transformer secondary supplies 300-amp, 480/277V, 3-phase incoming switchgear which distributes 480/277V, 3-phase power to the east side (Une 2 conversion) of the building through distribution panels. A 600-amp,3-phase,480/277V bus designated as
- Emergency Motor Control Center NE" (MCC-NE) is supplied by a 600-amp normal feeder from the incoming switchgear by way of an automatic transfer switch. Loss of normal power willinstigate an automatic transfer of the MCC-NE loads to the oncoming emergency generator.
The loads currently assigned to the east emergency generator system MCC-NE are: 1. UQ Buildina (East) 2 Exhaust Fan K-31 Exhaust Fan K-32 (POG) Air Sampling Vacuum Pump Emergency Ughtire Calciner Rotation a Calcinor Rotary Airlock Une 2 Switchgear (emergency control power) Uninterruptible Power Supply (UPS) for MICON System Vaporization and POG Scrubber Une 2 Instrument Air Dryer Deluge System Emergency Generator Battery Power HVAC Control Panels Combustible Gas Detector Une 2 Vaporization Room Header Heater Zones 1,4,5 and 6 Ammonia Recovery Emergency Power Panel 10.2.1.4 SF Buildina Normal power is brought into the SF Building from a 12.5 kV tap on the plant loop, transformed to 3-phase,480/277V, grounded-wye. It is supplied to three circuit breakers (350, 450, and 900 AMPS) through a 1600 AMP, ground-gault protected main circuit breaker, outside the building. me outia mcuou oct: September 12,1994 10-12 SPc-ND3330.947 (R 1/07/92)
Siemens Power Corporation - Nuclear Division Eur.2 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 PART 11 - SAFETY DEMONSTRATION REv. Critical loads are supplied from a distribution panelboard which is connected by underground cable to the UO North Emergency Panelboard. 2 10.2.2 Compressed Air System 1 The Plant's compressed air system is actually four systems: normal plant air, instrument air, control air and emergency air. The nuclear fuel manufacturing portion of the plant's normal compressed air needs are supplied by three compressors located in the Compressor Building shown on Figure 11-10.1. These compressors supply air to the UO, SF, ELO Buildings, and the ARF. 2 Other facilities such as the Maintenance Building, Machine Shop and PDTF have small compressor units to serve their individual needs. Two of the compressors in the Compressor Building are 500 CFM units and the third is a 413 CFM unit which provides a total capacity of 1413 CFM for the manufacturing portion of the plant. All three units are classified as " oil-free". Normal plant demand is satisfied by running any two of the units with the third on standby. The system pressure is maintained at 115 psig. Normal compressed air is supplied to air dryers located in the North, South, and East Equipment Rooms of the UO Building, and in the ARF. The dryers dry the air to a j g dewpoint of -40"F, and then it is distributed as instrument air. I instrument air is reduced in pressure to 20 psig to be utilized as HVAC and process i equipment control air. l l In the event of an electrical power outage, the plant's emergency air system will function. The emergency system consists of the 413 CFM unit in the Compressor Building, a 42 CFM unit in the north equipment room of the UO Building, and a 34 CFM unit in the 2 south Equipment Room of the UO Building. All three of these units are connected to the 2 plant's emergency power system. The 415 CFM compressor provides air to the plant j until the various systems can be safely shut c'own. The smaller units in the equipment i rooms are devoted exclusively to the HVAC systems to provide sufficient operating and control air to allow the system to be shut down safely or to continue to operate in the j emergency mode. j 10.2.3 Water The water supply to SPC is furnished by the City of Richland. The source of water for Richland is the Columbia River. Wells, which were the earlier source of water for Richland l before construction of the present water filtration plant on the river, have been kept in 2 O operable condition as a secondary backup source. Primary backup is furnished by a AMENDedENT APPLCATION DATE: PAGE NO.: ) sPc-ND 3330 947 (R 1/07/92)
Siemens Power Corporation - Nuclear Division EMF.2 (} SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 C PART ll - SAFETY DEMONSTRATION
- ggy, 15,000,000-gallon equalizing reservoir on a bluff south of Richland at an elevation of 545 feet.
Water is supplied to the plant via two lines from the North Richland water grid into the plant's water loop. The City of Richland's Water Department has assured SPC that each main line would supply in excess of 1500 gpm at 40 psig for a total capacity in excess of 3000 gpm through both lines. Actual flow measurements taken during the entire 1991 year show that the maximum water required by the plant was approximately 280 gpm with a yearly average of 189 gpm. The plant's total water usage averages 42 gpm of sanitary water and 147 gpm of process water. Approximately 20 gpm (maximum usage) of the process water is further refined as deionized water for various product and laboratory uses. Since both sanitary and process water needs are taken from the same inlet waterline, vacuum breakers are installed to provide assurance that process water cannot backflow into either the city system or into the plant sanitary water system. Figure 11-10.18 shows the main waterlines into and around the SPC facility. (d Due to the redundant backup systems for Richland's water supply, the probability of complete loss of water supply is extremely small. Since the water is supplied to the fuels plant by a complete loop with a number of sectionalizing and isolation valves, the probability of a water outage due to a main pipe breakage is also considered extremely remote. Even if a complete water outage did occur, the only possible damage to the plant would be to the sintering furnaces and would not present a radiological or containment problem. i 10.2.4 Sewer System Sanitary wastes from the office buildings, as well as from the production buildings, flow to a sanitary drainage system leading directly to the Richland Municipal Sower System. Changeroom showers are not used for personnel decontamination, but are for employee convenience only. Employees are required to conduct a self-survey as they enter the changeroom from a radiation area. l Water is used for cooling purposes throughout the manufacturing facilities in various amounts, depending on the specific process, equipment and operation in progress. Process cooling water is isolated from actual process atmospheres by double physical barriers and discharged to the sanitary sewer system. Liquid chemical wastes, process cooling wastes, and sanitary wastes are combined and ,q flow through the flowmeter-sampling station where a composite sample is automatically (O taken of all flows to the Municipal Sewer System. The composite sample is analyzed for =catm amcatow oart: September 12,1994 10-14 SPC-ND:3330 947 (R*0F92) l
Siernens Power Corporation - Nuclear Division EM F.2 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 PART ll - SAFETY DEMONSTRATION REv. l uranium daily Monday-Friday. Figure Il-10.18a shows the main pipelines of the sewer system. 10.2.5 Gas and Chemical Storace The sintering furnaces and the calciners in the UF conversion areas are provided with g a reducing atmosphere consisting of a hydrogen-nitrogen mixture derived from liquid nitrogen and cracking anhydrous ammonia in dissociators. Pressure tanks for anhydrous ammonia and liquid nitrogen are outside of the buildings. Outside cylinder storage is provided for gases such as helium, argon, oxygen, acetylene, hydrogen and propane. l l Liquid chemicals used in the process, such as nitric acid and sodium hydroxide solution, f are stored in outdoor tanks near the point at which they are consumed. The chemical storage tanks are diked such that spillage or leakage is contained. The dikes do not drain to sewer; however, they may be pumped to sewer if so dispositioned. l J Solvents are used principally for cleaning components. Other than quantities required for normal use, solvents are stored in flammable liquid storage cabinets. { b 10.2.6 Communications and Annunciations Full telephone service is provided for every building and facility on the SPC site. Most of the telephones in the various offices are individual, direct-line telephones whereas most of the telephones in the production area are serviced by an operator-attended switchboard. Annunciation, both visual and audible, of off-standard conditions in the UO and SF 2 Buildings is provided in the lobby area of each building. Typical conditions that may be annunciated include: O V measwr uncarou mTc: NE@ September 12,1994 10-15 SPc-ND.3330.947 (R-1/07/92)
Siemens Power Corporation - Nuclear Division sup.2 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 PART 11 - SAFETY DEMONSTRATION
- ggy, Condition UO Building SF Building 2
Fire Alarm X X Emergency Power Mode X Various HVAC Malfunctions X X Criticality Alarm X X Autoclave Sump Pumps X X Low Water Pressure X X High Radiation, HVAC X X Low Fire Main Pressure X Cooling Tower Malfunction X Any indication of an off-standard condition in the UO or SF Buildings is also signaled at 2 annunciator panels in the Central Guard Station which is manned at all times. n ( i Cycling klaxon howlers are provided to indicate the occurrence of a criticality event. Single strike gongs (two strokes per second) are provided to indicate fire. A public address system is installed for special messages or instructions. All three sets of audible signal devices are provided in sufficient numbers to be heard in all parts of the buildings and nearby outside areas. 10.2.7 Breathina Air Breathing air requirements are satisfied by the use of a breathing air purifier tied into the plant compressed air system. This device is equipped with a catalyst cartridge for removal of carbon monoxide from the system. The purifier is certified for Class D breathing air per ANSI and OSHA. Breathing air is continuously monitored for O and CO content. O below 19.5% by 2 2 volume or CO above 10 ppm will automatically isolate and de-pressurize the main receiver tank, disabling the system, and sound an audible alarm. in addition to the main breathing air system, there is a backup breathing air system. The 3 system uses two 220 ft air bottles and is located outside Line 1 conversion area. It serves the vaporization room, powder prep area, hot oil dryer area and Une 2 vaporization room. .m " ^ " " ' ' aucuoen amcaron oars: September 12,1994 10-16 SPC-ND;3330 947 (R-UO7/92)
i Siemens Power Corporation - Nuclear Division sur.2 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 701257 PART ll - SAFETY DEMONSTRATION f
- ggy, 10.3 Heatina. Verd!lation and Air Conditionina (HVAC) 10.3.1 Criteria The HVAC systems have been designed and are operated in conformance with Part I, i
Chapter 5 of this Ucense. HVAC systems in those facilities which are involved with uranium handling and processing have the following basic characteristics: 1. The HVAC system is refrigeration-cooled and electrically-heated. 2. Air enters the process areas near the top of the rooms and exits near the floor. 3. Partial air recirculation may be used to reduce thermal loads. 4. When air is reused, it moves from areas of lower potential airborne contamination to areas of higher potential airborne contamination. 5. Provisions are made for continuous monitoring of the airflow in air V recirculating systems, and an indication of airborne concentration exceeding 8 MPC-hr ahead of the second filter, automatically shuts off i supply air fans, and switches the system to a once-through exhaust mode. 6. Recirculated air from areas of no contamination is passed through roughing filters prior to return to the work spaces. l 7. Recirculated air from areas of low potential contamination is passed through HEPA filters prior to return to the work spaces. No air is recirculated from process equipment hoods or enclosures. I 8. Exhaust air from production areas with a potential for contamination is passed through HEPA filtration with an in-place filter bank efficiency of greater the.n 99.95% for a 0.8 micron particle or di octyi sabacate (DOS) test. Exhaust air from process equipment enclosures or process offgas (POG) systems is double HEPA filtered prior to exiting the facility. 9. At least 7 air changes per hour occur in uranium processing areas. 10. The process areas are compartmentalized. 11. The HVAC system has been designed so that supply fans cannot be q started until the proper exhaust duct negative pressure has been obtained. h i AMENDMENT ARCATON DATE: NE NO" September 12,1994 10-17 SPc-ND.3330 947 (R-vo7/92) l
As. Siemens Power Corporation - Nuclear Division sur-2 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 PART 11 - SAFETY DEMONSTRATION
- pgy, 12.
Supply and exhaust fans are interlocked so that the supply fan is shut off upon exhaust fan failure. 1 13. Uranium processing areas are maintained at a negative pressure to atmosphere and adjacent clean areas. 14. Continuous isokinetic air samplers are installed in the exhaust systems downstream of the final HEPA filter. 15. Rate-of-rise / heat detectors are installed in exhaust ducts and set to trip filter bank fog deluge systems at temperatures of 140*-190 F. 16. Visual indicators for reading the pressure drop across all HEPA filters are permanently installed. 17. Connections are provided for in-place DOS efficiency filter testing of all final HEPA filters. 10.3.2 SF Buildina HVAC Systems The SF Building has four independent HVAC systems. The SWUR Facility located in Room 173 is basically a once-through HVAC system with double HEPA filtration (K48 supply and K49 exhaust) serving Room 173 and the adjacent airlock. The remaining produciton facilities, laboratories, changerooms, and office sections is served by a combination (K5 supply and K6 exhaust) once-through and recirculation supply and exhaust system with double HEPA filtration. The incinerator itself is served by the K50 system consisting of a once-through HVAC system with double HEPA filtration. The incinerator shrcud cooling system is served by the K55 system consisting of a once-through system with double HEPA filtration. Simplified schematics of these HVAC systems are furnished in Figures 11-10.19 (SWUR K48/K49 and K50 HVAC systems), ll-10.20 (SF production area K5 and K6 HVAC systems), and ll-10.27a (SWUR incinerator shroud cooling system). 10.3.2.1 SWUR HVAC System The general features of the SWUR HVAC systems (Figure 11-10.19) are a once-through, i ceiling-to-floor airflow supply air system and a double HEPA filtered exhaust system. i 10.3.2.1.1 K48 Air Supply System ) 3 The K48 air supply system supplies about 6600 ft / min of 100% outside air to the SWUR i i p processing area and adjoining airlock. Each room has a minimum of 7 air changes per hour. Downstream of the fan are duct air monitors for measuring airflow quantities. NE N* September 12,1994 10-18 sPC-ND.3330 947 (P-110742)
Siemens Power Corporation - Nuclear Division eup.2 ^N SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 (J PART 11 - SAFETY DEMONSTRATION g y, Airflows are directional from ceiling supply to near-floor exhaust air grills or process hood inlets and always from areas of low contamination potential to areas of higher contam-ination potential. 10.3.2.1.2 K49 Air Exhaust System 3 Air supplied to Room 173 and the airlock, plus infiltration (approximately 7500 ft / min), is exhausted through the K49 exhaust system. The double HEPA filter arrangement in this system consists of the final HEPA filter bank plus individual or smaller filter banks of profilters and HEPA filters located in the exhaust ducts of areas or process equipment serviced. The K49 system exhaust air passes from the final filter bank through a duct air monitor (measures airflow quantities), the main exhaust fan, and is discharged from a stack located on the SF Building roof. The stack extends 20 ft above the roof of the SF Building, or 50 ft above ground elevation. The K49 exhaust system has one full capacity fan which is connected to normal power. 3 All final HEPA filters are in-place tested and assured to be 99.95% (minimum) efficient for 0.8 micron DOS cold aerosol. 10.3.2.1.3 Systems Controls The HVAC systems are controlled with temperature, pressure and flow sensor actuating valving and damper positions to hold temperature, pressures and pressure differentials constant in the various building areas. The K48 supply air system is interlocked with the K49 exhaust system to prevent operation of the K48 supply air system without the K49 exhaust system operating. Pressure sensors are provided for damper control to maintain a negative pressure (minimum -0.05-in water gauge) in the process areas relative to atmosphere. Automatic audio and visual alarms are activated when supply or exhaust system upsets occur. Pressure differentialindicating devices and airflow quantity meters are located in the controlled zones and/or on the main HVAC panel to provide system and zone operating conditions. Air samplers are located upstream and downstream of the final HEPA filters. 10.3.2.1.4 Deluce System A deluge system of fog spray nozzles is installed in the exhaust ducts a short distance upstream of the final filter banks. If rate-of-rise / heat detectors indicate a temperature of 140* 190"F, the deluge spray is automatically activated. The deluge system will protect (nw) the final banks from burning debris and heated air that would damage the mastic used AutNDMENT APPLCATON DATE: PAGE NO.: g sPC-ND.3330 947 (R-1/07/92)
l Siemens Power Corporation - Nuclear Division aus.2 T SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 l PART 11 - SAFETY DEMONSTRATION
- ggy, to cement the filter media to the filter frame. Should the deluge system be activated, differential pressure readings across the final filters shall be taken and an in-place DOS test made at the earliest opportunity.
10.3.2.1.5 Final Filter Bank that, in turn, is fastened and The final filter bank is encased in a sheet metal housing / min at one-inch water gauge sealed to a concrete slab. HEPA filters rated at 1000 ft pressure drop are mounted in welded steel structural frames. The HEPA filter medium is 100% moisture-resistant fiberglass, pleated over corrugated separators, and sealed in fire-resistant plywood frames. The individual filters are certified to remove 99.97% of 0.3 micron particles and meet or exceed Military Specification MIL-F-51079. 10.3.2.2 SWUR Incinerator Shroud Coolino System The SWUR incinerator is equipped with a cooling shroud to reduce heat losses to the SF Building Room 173 operating area. The K-55 system (Figure ll-10.27a) consists of filtered supply systems, pressure shrouds, a HEPA filtered exhaust and a stack. O. 10.3.2.2.1 K-55 Air Supply System 1 3 The K55 air supply system provides about 1200 ft / min of filtered outside air to the incinerator shroud. The air supply is assured by providing automatic switchover to the standby fan if the on-line fan fails. The system is also on the emergency power supply system. An inlet filter radiant heater frost protection system is provided. 10.3.2.2.2 incinerator Shrouds Air from the K-55 supply system is ducted under pressure to incinerator shrouding system. The main shroud consists of welded plate flow channels around the incinerator walls and between chambers. The north shroud cools the front face of the secondary chamber. The system is operated under pressure with respect to the incinerator and operating area to minimize the potential for uranium contamination of the cooling air. 10.3.2.2.3 K-55 Exhaust System The K-55 exhaust system consists of parallel HEPA filter units (one online and one on standby) and an exhaust stack. The filter units are self-contained metal framed high temperature nipple-connected HEPA filters. The HEPA filters are DOS tested and certified prior to installation. Filtered exhaust air is discharged from a 14 inch diameter stack extending 12 foot above the SF Building. An isokenetic air sampler is provided to monitor potential stack discharges of particulate uranium. autwoutuiamcatow oare:
- E*
September 12,1994 10-20 SPC-ND.3330.947 (R-iro7/92)
Siemens Power Corporation - Nuclear Division sup.2 O SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 V PART ll - SAFETY DEMONSTRATION gy, 10.3.2.3 SWUR Incinerator HVAC System The SWUR incinerator itself is served by the K50 supply / exhaust system. 10.3.2.3.1 K50 Air SuppIV System 3 The K50 air supply system provides about 900 ft / min of filtered outside air to the Incinerator. 10.3.2.3.2 K50 incinerator Exhaust System 3 The incerator exhaust gas from the afterburner chamber (approximately 950 ft / min) is exhausted through the K50 exhaust system. The exhaust system components are described in Section 10.4.6.3 of this document. The exhaust stack extends 14 feet above the roof of the SF Building, or 42 feet above ground elevation. All final HEPA filters were in-place tested and assured to be 99.95% (minimum) efficient for 0.8 micron DOP/ DOS cold aerosol. 10.3.2.4 SF Buildina Production Area HVAC System The general features of the SF Building (K5 and K6) HVAC system are a partially recirculated air, down-draft airflow, and a double HEPA filtered exhaust system. A simplified schematic diagram of the SF Building HVAC system is shown in Figure 11-10.20. 10.3.2.4.1 K5 Air Supply System The K5 system supplies air to the service areas of the building and the fabrication areas not serviced by the SWUR Facility systems. The K5 system provides approximately 3 3 26,000 ft / min of air. Approximately 5000 ft / min is recirculated from the service / office 3 areas, and about 3000 ft / min from the NAF area to reduce thermal loads in normal operation. The design is such that air moves from clean areas to areas of contamination potential. Recirculated air from the NAF area is passed through double HEPA filtration, the final bank of which shall have an installed tested efficiency of 99.95% for 0.8 micron DOS particles. Provision is made in the K5 recirculation system for continuous radiation monitoring of recirculated air prior to the second HEPA filter. An indication of an airborne concentration exceeding 8 MPC-hr automatically shuts down the K5 supply fan and-places the KS system on a once-through basis with the air exhausting through the final exhaust filters and stack. Retum-air from the service and office areas passes through a roughing filter and mixes with NAF recirculated air prior to passing through the recirculation HEPA filter bank. =cacuumcuom ons: September 12,1994 10-21 i SPc-ND 3330 947 (R4/07/92)
_ _ __ ~ _ _. _. i i i Siemens Power Corporation - Nuclear Division eup.2 ' SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 PART ll - SAFETY DEMONSTRATION REv. l l In all process areas, the airflow is directional. Supply air is introduced from diffusers in the upper portion of the rooms and exhaust air (including recirculated air) is drawn i through HEPA filters in housings located near floor level. Air is also exhausted through hoods. There are no overhead exhaust outlets. I To minimize the potential of contamination spreads, the fabrication area is compartmentalized. Each area is provided with floor level exhaust outlets (through HEPA filters) such that potential contamination would be localized to the room in which it might originate. 10.3.2.4.2 K6 Air Exhaust System l The SF Building production facilities, laboratories, changerooms and related areas are. served by the K6 exhaust system. The filter arrangement in this system consists of two 'l parallel final HEPA filter banks plus individual or smaller filter banks of prefilters and HEPA filters located in the exhaust ducts of areas or process equipment serviced. l l The final filter arrangement consiste of two separate and parallel filter housings O discharging to a common duct leading to the stack. One filter housing contains a bank of HEPA filters preceded by a bank of 35% ASHRAE roughing filters. Air exhausted from l the changerooms, NAF area, and a portion of the rod loading and Analytical Laboratory j is passed through this series of filtration stages prior to discharge. Also, following j prefiltration through individual combined HEPA and roughing filters, air exhausted from -l the remaining portion of the Analytical Laboratory, radiation monitoring offices, and i storage area is introduced to this filter housing between the banks of HEPA and roughing i filters and is drawn through the final bank of HEPA filters and discharged. The second of the final filter housings contains one bank of HEPA filters. Air from the j remainder of the process areas is drawn through upstream, individual combined HEPA - l and roughing filters and then through this final HEPA filter and discharged. Room air exhausted from the NAF area which is profittered through individual combined HEPA and roughing filters and normally recirculated shall be diverted upon detection of 8 MPC-hr of radioactive material by the in-line radiation detector to this filter housing and will be drawn through the final bank of HEPA filters and discharged. i Normally when part of the air is being recirculated through the K5 supply system, the K6 l 3 system exhausts approximately 16,000 ft / min to atmosphere; if the entire HVAC systems 3 ~ were on a once-through operation, approximately 26,000 ft / min would be discharged. The K6 exhaust air passes directly from the final filter banks to the exhaust fans and is discharged from a stack extending 50 ft above ground elevation. The K6 exhaust system has two full capacity fans which are used alternately to provide standby capability. Both t O fans are connected to normal and emergency power. P E " aucwoucut apeticatow oart: %teder 1E 1m 10-22 SPC ND-3330S47 (R-UO7S2) i -, = - - - -
Siemens Power Corporation - Nuclear Division eup.2 Q SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 (G PART ll - SAFETY DEMONSTRATION
- ggy, 10.3.2.4.3 System Controls The HVAC system is controlled with temperature, pressure and flow sensors actuating valving and damper positions to hold temperature, pressures, and pressure differentials constant in the various building areas. The individual systems are interlocked so that one of the KS exhaust fans must be started before the K5 supply system can be started. In the event of power f ailure, K6 fans are automatically transferred to an emergency electrical bus.
As long as the supply and exhaust plena pressures remain in the established ranges, the building pressure differentials will be maintained. All main duct HEPA filters are provided with differential pressure indicators. Sudden changes in differentia! pressure across a final filter bank indicating either excessive filter plugging or a filter rupture automatically alarms on the main annunciator panel. 10.3.2.4.4 Deluae System A deluge system of fog spray nozzles is installed in the exhaust ducts a short distance ( ) upstream of the final filter banks. Should the deluge system be activated, differential v pressure readings across the final filters shall be taken and an in-place DOS test made at tiie earliest opportunity. 10.3.2.4.5 Final Filter Banks The final filter banks are encased in sheet metal housings that, in turn, are fastened and 3 sealed to a concrete stab. HEPA filters rated at 1000 ft / min at one-inch water gauge pressure drop are mounted on special clean-room frames. Individual filter frames that are normally used for clean-room systems are welded together to form integral frames of the required sizes and shapes. The HEPA filter medium is 100% moisture-resistant fiberglass, pleated over corrugated separators and sealed in fire-resistant plywood frames. The individual filters are certified to remove 99.97% of 0.3 micron particles and meet or exceed Military Specification MIL-F-51079. 10.3.3 UO Buildina (North) HVAC Systems 2 The UO Building (north) has four air supply systems and four exhaust systems. The g supply air systems consist of the following: i The K1 system supplies air to the service areas of the building with a q portion of this area being recirculated and a portion being once-through. 'Q aw woutwTa m carowoart: "E " " September 12,1994 10-23 SPC-ND.3330 947 (R-UO7S2)
- l
Siemens Power Corporation - Nuclear Division eur.2 [' SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 PART ll - SAFET( DEMONSTRA MON
- ggy, The K2 system supplies air to the fabrication areas of the building with a portion of this air being recirculated and a portion being exhausted.
The K23 system is a once-through system which supplies the uranium i scrap and lube-blend areas. The K36 system is also a once-through system which supplies the UF -UO Une 2 conversion areas. g 2 The exhaust air systems consist of the following: The K3 building exhaust system provides room and process hood exhaust for portions of the service and fabrication areas which are not recirculated air. The K9 system exhausts air from the etch room and etch tanks; The K10 is a POG system which serves process equipment in the Une 1 conversion and the uranium scrap reprocessing areas. The K31 exhaust system, which is described in and primarily serves Une 2 conversion, also serves as the non-offgas exhaust for the UF -UO s 2 conversion, lube blend and uranium scrap reprocessing areas. i All final filter banks in the four exhaust systems are tested and assured to have an installed efficiency of 99.95% for 0.8 micron particles. Simplified schematics of these HVAC systems are furnished in Figures 11-10.21 and 11-10 22. 10.3.3.1 K1 and K2 Air SuDDiv Systems The K1 system, which supplies air to the service areas of the building, and the K2 system, which supplies air to the fabrication areas, both use partial recirculation. Recirculated air in the K1 system is only rough filtered while recirculated air in the K2 system is passed through double HEPA filters, of which the second has an installed efficiency of 99.95% for 0.8 micron particles. Provision is made in the K2 recirculation system for continuous radiation monitoring of recirculated air prior to the second filter. An indication of an airborne concentration exceeding 8 MPC-hr automatically places the K2 system on a once-through basis with the air exhausting through the final filter to the stack. The K1 supply system provides approximately 28,600 ft / min of air which is divided 3 m among various laboratories, offices, changerooms, HVAC equipment room, and i September 12,1994 10-24 1 SPC-ND.3330 947 (R-1/07/92) i
Siemens Power Corporation - Nuclear Division sur.2 /'"' SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 d PART 11 - SAFETY DEMONSTRATION
- ggy, production service areas. Approximately one-third of the system's airis recirculated within the K1 system, primarily from the second floor office areas, HVAC equipment room, and building lobby.
3 The K2 supply system provides approximately 45,000 ft / min of air to the pellet pgeparation and storage areas, autoclave area, and the etch facility. Approximately 8000 ft / min of the air in the K2 qstem is recirculated within the system from locations where radioactive contamination !s expected to be zero (or very low) such as the autoclave area, supply room, production offices, and gauge and scale calibration rooms. In addition, 3 another 4000 ft / min of air from clean service areas supplied by the K1 system is recirculated through the K2 system. In most of the process areas, the airflow is directional. Supply air is introduced from diffusers in the upper part of the room, and exhaust is taken into filter boxes near the floor around the periphery of the room. 10.3.3.2 K23 and K36 Air Supp!v Systems g 3 (v) The K23 system supplies approximately 12,000 ft / min of 100% outside air to the lube-blend and uranium scrap reprocessing areas, while the K36 system supplies 3 approximately 12,000 ft / min of 100% outside air to the UF -UO Une 1 conversion area. e 2 Airflows are directional from ceiling to near-floor exhaust air grills or process hood inlets, and always from areas of low-contamination potential to areas of higher contamination potential. 10.3.3.3 K3 Exhaust System 3 l The majority of the building air (approximately 44,000 ft / min) is exhausted through the K3 system. The doub'e filter arrangement in this system consists of the final filter bank i and individual or smalier filter banks of prefilters and HEPA filters in the exhaust ducts of l the areas or equipmer,t serviced. The K3 system exhaust air passes directly from the final filter bank and is discharged from Building, or 50 ft above a stack extending 22 ft above the highest portion of the UO2 ground elevation. The K3 exhaust system has two full-capacity fans which are used afternately to provide standby capability. Both fans are connected to normal and emergency power. 10.3.3.4 K9 Exhaust System i The K9 exhaust system provides exhaust service for the corrosive fumes from the fuel rod o etch and stop bath tanks. The exhaust is passed through a scrubber, dryer, and double HEPA filters and is discharged through a fiberglass-polyester stack 37 ft high. The WENDMENT A%CATON DATE: Se tember 12,1994 10-25 SPC-ND.3330 947 (R vo7/92) 4
i Siemens Power Corporation - Nuclear Division eup.2 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 PART 11 - SAFETY DEMONSTRATION REv. I ductwork materialin the K9 exhaust system is steel-coated inside with acid-resistant vinyl resin upstream of the scrubbers and stainless steel main ducts and PVC branch ducts downstream. i The K9 exhaust system has two full-capacity fans which are used alternately to provide standby capability. Both fans are connected to normal and emergency power. The K9 i 3 system exhausts approximately 6000 ft / min continuously. 10.3.3.5 K10 Exhaust System { 8 The K10 exhaust system provides approximately 1200 ft / min vent service for tanks and vessels in the UF Process, as well as offgas exhaust for the calciner, UF vaporizer room j s s and vaporizer chests. As in the etch process, fumes from these vessels are corrosive. The exhaust is passed through a scrubber, a dryer and double HEPA filters, and is discharged through a separate stack located on the roof of the UO Building. This stack { 2 extends 25 ft above the highest portion of the UO Building. The ductwork materialin the l 2 K10 exhaust system is stainless steel for ducts greater than seven inches in diameter, and i polyvinyl chloride for ducts less than seven inches in diameter. l The single K10 exhaust fan is connected to emergency power. Should the fan become inoperative for any reason (including loss of power), or the system pressure decay to a I predetermined set point, the UFs Process will shut down automatically. Air from the i general UF area will continue to be filtered and exhausted through the K31 system. ] s The scrubber is provided to remove ammonia and fluorides from the offgases, the dryer j to remove entrained liquids, and the filters to remove potential uranium. In addition to the j main scrubber in the K10 system, there is a second scrubber in the exhaust line from the calciner. j 10.3.3.6 System Controls The HVAC systems are controlled with temperature, pressure and flow sensors actuating l vaMng and damper positions to hold temperatures, pressures, and pressure differentials constant in the various building areas. In the event of power failure, K3, K9 and K31 i exhaust fans are automatically transferred to an emergency electrical bus. l 1 The power to the supply fans is interlocked so that exhaust fan failure, or failure of ' instrument or control air, stops the K1 or K2 supply fans. A similar upset in the K31 exhaust system will stop the K23 and K36 supply fans. In the event of power failure, the 4 supply fans will stop and air will windmill through the supply fans due to the continuance of the exhaust fans on emergency power. A rise in exhaust pressure in the K10 system higher than -4 inches water gauge automatically shuts down the UF Process. s AMENDMENT APPLCATION DATE: PAGE N0a September 12,1994 10-26 SPC-ND3330S47 (R-UO7S2) -\\
Siemens Power Corporation - Nuclear Division sup.2 i O SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 V PART ll-SAFETY DEMONSTRATION REv. The differential pressure between the zone containing the building's lobby, stairwell and j changerooms on the first floor, the offices and lunchroom on the second floor, and the zone comprising the rest of the building will be maintained at a minimum of 0.05-inches water gauge, with the former being at atmospheric pressure, or slightly above, and the latter below atmospheric pressure at all times. l i All main duct filters (K3, K9 and K10 systems) are provided with differential pressure l Indicators. Sudden changes in differential pressures across the final K3 or K9 filter banks indicating either a filter rupture or excessive plugging automatically (except the K10 i exhaust fan) activate alarms. 10.3.3.7 Deluoe System A deluge system of fog spray nozzles is installed in the main exhaust duct (K3) a short distance upstream of the final filter bank. If rate-of-rise / heat detectors indicate a i predetermined temperature rise, the deluge system is automatically activated. Should the deluge system be activated by any circumstance, differential pressure readings across the filter shall be taken and an in-place DOS test made at the earliest opportunity. The j K9 and K10 systems have liquid scrubbers in-line and ahead of the filters; therefore, they d are not equipped with deluge systems. j 10.3.3.8 Final Filter Banks The final filter bank for the K3 system is encased in a sheet metal housing that, in turn, 3 is fastened and sealed to a concrete slab. HEPA filters rated at 1000 ft / min at one-inch i water gauge pressure drop are mounted on special clean-room frames. The HEPA filter j medium is 100% moisture-resistant fiberglass, pleated over corrugated separators and i sealed in fire-resistant plywood frames. The individual filters are certified to remove i 99.97% of 0.3 micron particles and meet or exceed Military Specification MIL-F-51079. Following the scrubber, the exhaust air of the K9 system enters a filter plenum which houses both the primary and final HEPA filters. All construction features of the final filter bank of the K9 system are the same as those for the K3 system. The K10 system filters are single filters in series and are housed in separate housings. in the K10 exhaust system, there are two separate but identical legs. Each leg contains a scrubber, followed by a dryer, and the double HEPA filters. After the second filter in each leg, the legs are joined together in a common 12-in duct which penetrates the roof of the building to the fan and stack. The K10 filters are mounted on steel frames instead 1 of concrete slabs. All other construction features are the same as those for the other systems. o AMENDMENT APPUCATON DATE: PAGE NO.: September 12,1994_ 10-27 SPC-ND:3330 947 (R-UO7/92)
SiemenS Power Corporation - Nuclear Division eur.2 ("X SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 V PART 11 - SAFETY DEMONSTRATION REv. 10.3.4 Ul0 Buildina HVAC Systems (South Addition) _2 Although this building addition is physically attached to the original UO Building, it has 2 its own separate HVAC system (K20 supply, K21 exhaust, and controls). An airlock connecting the two facilities maintains separation of the two HVAC systems. A simplified schematic diagram of the HVAC system for this addition to the UO Building is provided 2 in Figure 11-10.23. 10.3.4.1 K20 Air Supply System The K20 air supply system serves all areas of this building. The K20 system also provides a small amount of air to the tube cleaning room which is a small structure physically attached to the UO Building (south). The tube cleaning room (no contamination 2 3 potent!al) has its own recirculation HVAC system and receives approximatcly 1000 ft / min from K20 for a non-process exhaust fan. Approximately 50% of the supply air is recirculated during normal operations. The design is such that air is moved from clean areas to areas of contamination potential. As such, return air from the office areas, HVAC equipment room, bundle assembly, cage inspection, and spacer fabrication areas is [J i mixed with incoming air, rough-filtered and distributed throughout the building. 3 The K20 supply system provides approximately 46,000 ft / min of air which is divided 3 among various areas within the building addition. About 9000 ft / min of the supply air serves the office area, HVAC equipment room, and tube cleaning facility while 20,000 ft / min serves the clean fabrication and assembly / shipping areas on the west half of the 3 building; all of which is redistributed (after rough filtering and mixing with incoming air) 3 to other areas within the building. Approximately 17,000 ft / min of supply air is provided for the rod fabrication / loading / welding areas and dischargsd through the HEPA filtered K21 building exhaust system. Airflow is directional in all production areas. Supply air is introduced from diffusers in the upper part of the room and exhaust air is removed via process equipment enclosures or grills located near the floor around the periphery of the rooms. 10.3.4.2 K21 Exhaust Systen The K21 exhaust system serves the entire building addition. This exhaust system provides double HEPA filtration with the first stages of filtration located near process equipment enclosures or the individual exhaust grills in the various areas and the final stage in a common main exhaust filter bank. The final filter bank is tested in-place and assured to be 99.95% (minimum) efficient for removal of 0.8 micron-size particles, cold-generated DOS aerosol. ,m('] The K21 system exhaust air is discharged from a stack extending 22 ft above the highest portion of the building, or 50 ft above ground elevation. The K21 exhaust system has a AMENDeAENT APPLCATON DATE: PAGE NO.: September 12,1994 10-28 sPC-NO.3330 947 (R 90742)
Siemens Power Corporation - Nuclear Division sup.2 [] SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 v PART II - SAFETY DEMONSTRATION REv. primary full-capacity exhaust fan and an alternate to provide backup capability. Both fans are connected only to normal power, 10.3.4.3 Systems Controls The HVAC systems are controlled with temperature, pressure and flow sensors actuating valving and damper positions to hold temperature, pressures, and pressure differentials constant in the various building areas. The building ventilation systems are interlocked so that one of the K21 exhaust fans must be operating before the K20 supply fan can be started. Duct airflow monitoring devices have been installed in the main exhaust duct as well as in major supply and exhaust ducts to assist in measuring airflow for maintaining prescribed air change rates in process areas and calibrating representativo duct air samplers. The bundle assembly and cage fabrication areas are maintained essentially at atmospheric pressure, or slightly negative. The rod loading / fabrication area is maintained (n) at a minimum of -0.05 inch water gauge in respect to the bundle assembly area, and 0.02 U to 0.05 inch water gauge with respect to the north UO Building pellet preparation area 2 which is maintained at approximately -0.10 inch water gauge with respect to atmosphere. As long as the supply fan (K20) and exhaust fan (K21) plena pressures are maintained within indicated ranges, the building pressure differentials will be maintained. A differential pressure switch is provided across the main K21 exhaust HEPA filter bank to activate alarms on excessive filter buildup or plugging. 10.3.4.4 Deluae System j A fog deluge spray nozzle is installed in the main K21 exhaust duct a short distance upstream of the final HEPA filter bank. If rate-of-rise / heat detectors indicate a predetermined temperature rise, the deluge spray and alarms are automatically activated. A steel spark-arrester screen is provided between the deluge nozzle and final filter bank. Should the deluge system be activated, differential pressure readings across the filter shall be taken and an in-place DOS test made at the earliest opportunity. 10.3.4.5 Final HEPA Filter Bank The final HEPA filter bank for the K21 system is encased in a sheet metal housing that, 3 In turn, is fastened and sealed to a concrete stab. HEPA filters rated at 1000 ft / min at one-inch water gauge pressure drop are mounted in welded steel structural frames. p The HEPA filter medium is 100% moisture-resistant fiberglass, pleated over corrugated (j separators and sealed in fire-resistant plywood frames. The individual filters are certified AMENDMENT APPLCATON DATE: PAGE NO.: September 12,1994 10-29 i sPC-ND.3330 947 (R 1/07/92) i
Siemens Power Corporation - Nuclear Division e u r.2 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 PART 11 - SAFETY DEMONSTRATION REv. to remove 99.97% of 0.3 micron particles and meet or exceed Military Specification MIL-F-51079. 10.3.5 UO Buildina Line 2 Conversion HVAC Systems 2 The general features of the UO Building Une 2 conversion HVAC systems are a 2 once-through ceiling-to-floor airflow supply air (K30) system, a double HEPA filtered (K31) building exhaust system, and a process offgas (K32) exhaust system with scrubbers, dryers, and double HEPA filters. A simplified schematic of the K30 air supply system, K31 building exhaust system, and K32 POG system is shown in Figure 11-10.22. 10.3.5.1 K30 Air Supply System 3 The K30 air supply system supplies approximately 28,000 ft / min of 100% outside air to the Une 2 conversion area, UNH facility, vaporization room, control room, and the facility's equipment room. Airflows are directional from ceiling to near-floor exhaust air grills or [b) process hood inlets, and always away from areas of low contamination potential to areas of higher contamination potential. 10.3.5.2 K31 Air Exhaust System 3 The K31 air exhaust system removes approximately 28,000 ft / min of air supplied from the 3 process areas served by the K30 air supply system, and 20,000 ft / min frorn areas supplied by the K23 and K36 (Figure 11-10.22 and Section 10.3.3) air supply systems. The 3 combined airflow, including infiltration, totals about 50,000 ft / min being exhausted through the K31 system. The double filter arrangement in this system consists of the final HEPA filter bank plus individual or smaller filter banks of profilters and HEPA filters in the exhaust ducts of the areas or equipment serviced. Ductwork in the K31 system is made of galvanized steel. The K31 system exhaust air passes directly from the final filter bank to the exhaust fans and is discharged from a stack extending 22 ft above the. highest portion of the UO2 Building, or 50 ft above ground elevation. The K31 exhaust system has two full capacity fans which are used alternately to provide standby capability. Both fans are connected to normal and emergency power. 10.3.5.3 K32 Exhaust System 3 The K32 exhaust system provides approximately 1000 ft / min vent service for equipment, tanks and vessels in the UF and dry conversion processes as well as offgas exhaust for p; the calciner and exhaust service for the UF vaporizer chests. As in the etch process, 6 tsV 6 fumes from these vessels are corrosive and an exhaust system separate from the main AMENDMENT APPLCATON DATE: PAGE NO.: September 12,1994 10-30 sPC ND3330 947 (R-1107/92)
Siemens Power Corporation - Nuclear Division sup.2 (O3 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET PART 11 - SAFETY DEMONSTRATION
- ggy, K31 system is required. The exhaust is passed through a scrubber, a dryer, and double HEPA filters and is discharged through a separate stack located on the roof of the UO2 Building. This stack extends 25 ft above the highest portion of the UO Building. The 2
ductwork material in the K32 exhaust system is stainless steel for ducts greater than seven inches in diameter, and polyvinyl chloride for ducts less than seven inches in diameter. The K32 system has two full-capacity fans which are used alternately to provide standby capability. Both fans are connected to normal and emergency power. A scrubber is provided to remove ammonia and fluorides from the offgases, a dryer to remove entrained liquids, and HEPA filters to remove potential uranium. In addition to the main scrubber in the K32 system, there is a second scrubber in the exhaust line from the calciner. 10.3.5.4 System Controls The HVAC cystems are controlled with temperature, pressure and flow sensors actuating valving and damper positions to hold temperatures, pressures, and pressure differentials constant in the various building areas. The K31 exhaust fan power is interlocked so that the fans are stopped upon excessive negative pressure (more negative than -14 inches water gauge) and instrument or control air failure. In the event of power failure, the K31 fans are automatically transferred to an emergency electrical bus. The power to the supply fans is interlocked so that a rise in K31 exhaust pressure higher than -8 inches water gauge (exhaust fan failure) or failure of instrument or control air stops the K30 supply fan. A similar upset in the K31 exhaust system will stop the K23 4 and K36 supply fans. In the event of power failure, the supply fans will stop and air will windmill through the supply fans due to the continuance of the exhaust fans on emergency power. A rise in exhaust pressure in the K32 system (fan failure) higher than -4 inches water gauge automatically shuts down the UF process. s The differential pressure between the conversion area zone and the zone comprising the rest of the north UO Building will be maintained at a minimum of -0.05 inch water gauge, 2 and the rest of the north UO Building will be maintained at a minimum of -0.05 inch water 2 gauge below atmospheric pressure at all times. As long as the supply and exhaust plena pressures remain in the indicated ranges, building pressure differentials will be maintained. Sudden changes in differential pressure ar.oss the final K31 filter bank or the K32 filters indicating a filter plugging automatically activates alarms. AMENDMENT APPLCATION DATE; PAGE NO.: September 12,1994 1 0-31 SPC-ND.3330 947 (R-UO742)
Siemens Power Corporation - Nuclear Division sur.2 (b3 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET PART ll - SAFETY DEMONSTRATION
- ggy, 10.3.5.5 Deluoe System A deluge system of fog spray nozzles is installed in the main K31 exhaust duct upstrer.m of the final filter bank. If rate-of-rise / heat detectors indicate a predetermined temperature rise, the deluge spray is automatically activated. Should the deluge system be activated, differential pressure readings across the filter shall be taken and an in-place DOS test made at the earliest opportunity. The K32 system has liquid scrubbers in-line ahead of the filters; therefore, it is not equipped with a deluge system.
10.3.5.6 Final HEPA Filter Bank The final HEPA filter bank for the K31 system is encased in a poured and sealed concrete 3 structure. HEPA filters rated at 1000 ft / min at one-inch water gauge pressure drop are mounted in welded steel structural frames. Visual indicators for reading the pressure drop across the filters are permanently installed and means are provided for in-place DOS testing. The HEPA filter medium is 100% moisture-resistant fiberglass, pleated over corrugated ( separators and sealed in fire-resistant plywood frames. The individual filters are certified C to remove 99.97% of 0.3 micron particles and meet or exceed Military Specification MIL-F-51079. 10.3.6 U O Facility HVAC Systems 3 g The U 0 Facility and the tank canyon area are served by two individual air supply 33 systems (K38 and K39), but share a common double HEPA filtered exhaust system (K37) except for the tank canyon area which is single HEPA filtered. A simplified schematic of the K38 and K39 air supply systems and the K37 building exhaust system is shown in Figure 11-10.24. 10.3.6.1 K38 Air Supply System 3 The K38 air supply system provides about 4900 ft / min of 100% outside air to the tank canyon area and the ion exchange room. Airflows are directional from ceiling to near-floor exhaust air grills or process hood inlets and always away from areas of low contamination potential to areas of higher contamination potential. 10.3.6.2 K39 Air Supply System 3 The K39 air Fupply system provides about 12,700 ft / min of 100% outside air to the U O as receiving and processing areas, a hot maintenance shop, the powder storage area, and n an airlock. Airflows are directional from ceiling to near-floor exhaust air grills or process (} hood inlets and always away from areas of low contamination potential to areas of higher contamination potential. AMENDMENT APPLCATION DATE: PAGE No.: September 12,1994 10-32 SPC-ND.3330.947 (R.U07/92)
SiemenS Power Corporation - Nuclear Division sup.2 O SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 b PART 11 - SAFETY DEMONSTRATION gy 10.3.6.3 K37 Air Exhaust System 3 The K37 air exhaust system removes approximately 5500 ft / min of air supplied from the 3 process area served by the K38 air supply system and 12,700 ft / min of air supplied by the K39 air supply system. This combined airflow, including infiltration, totals about 3 19,300 ft / min being exhausted through the K37 system. The double filter arrangement in this system consists of the final HEPA filter bank plus individual or smaller filter banks of prefilters and HEPA filters in the exhaust ducts of the areas or equipment serviced. Room exhaust air from the tank canyon area is rough-filtered locally and mixed with K37 exhaust air from other areas which is passed through the K37 final HEPA filter bank. Ductwork in the K37 system is made of galvanized steel. The K37 system exhaust air passes directly from the final filter bank to the exhaust fan and is discharged from a stack extending 24 ft above the highest portion of the UO2 Building, or 50 ft above ground elevation. The K37 exhaust system has one full-capacity fan which is connected to normal power. All final HEPA filters are in-place tested and assured to be 99.95% min! mum efficient for 0.8 micron DOS cold aerosol. ,O 10.3.6.4 System Controls V The HVAC systems are controlled with temperature, pressure and flow sensors actuating valving and damper positions to hold temperatures, pressures, and pressure differentials constant in the various building areas. The building ventilation systems are interlocked so that the K37 exhaust fan must be started before the supply systems can be started. The K37 exhaust fan power is interlocked so that the fans are stopped upon excessive negative pressure (more negative than -12 inches water gauge) and instrument or control air failure. The power to the supply fans is interlocked so that a rise in K37 exhaust pressure higher than -4 inches water gauge (exhaust fan failure) or failure of instruments or control air stops the supply fans. The differential pressure between the possibly contaminated Maintenance Shop zone and the zone comprising the rest of the U 0 Building will be maintained at a minimum of 3 3 -0.05 inch water gauge, and the rest of the U 0 Building will be maintained at a minimum 33 of -0.051nch water gauge below atmospheric pressure at all times. The tank canyon and ion exchange areas will be maintained at a minimum of -0.05 inch water gauge below atmospheric pressure at all times. As long as the supply and exhaust plena pressures remain in the indicated ranges, building pressure differentials will be maintained. 10.3.6.5 Deluce System A deluge system of fog spray nozzles is installed in the main K37 exhaust duct a short (p distance upstream of the final filter bank. If rate-of-rise / heat detectors indicate a U) predetermined temperature rise, the deluge spray is automatically activated. Should the AMENDMENT APPLCATON DATE: PAGE No.: September 12,1994 10-33 SPC-ND:3330.947 (R-UO7/92)
Siemens Power Corporation - Nuclear Division Eur.2 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 PART ll-SAFETY DEMONETRATION REv. deluge system be activated, differential pressure readings across the filter shall be taken and an in-place DOS test made at the earliest opportunity. 10.3.6.6 Final HEPA Filter Bank The final HEPA filter bank for the K37 system is encased in a sheet metal housing that, 3 in turn, is fastened and sealed to a concrete stab. HEPA filters rated at 1000 ft / min at one-inch water gauge pressure drop are mounted in welded steel structural frames. l The HEPA filter medium is 100% moisture-resistant fiberglass, pleated over corrugated l separators and sealed in fire-resistant plywood frames. The individual filters are certified l to remove 99.97% of 0.3 micron particles and meet or exceed Military Specification MIL-F-51079. 10.3.7 UO2 Buildina Analvtical Laboratory HVAC Systems The Analytical Laboratory area of the UO2 Building is served by an air supply system (K57) and a double HEPA filtered exhaust system (K58). A simplified schematic of the b K57 air supply system and the K58 exhaust system is shown in Figure 11-10 V 10.3.7.1 K57 Air Supoly System 3 The K57 air supply system provides about 15,000 ft / min of 100% outside air to the laboratory areas. Altflows are directional from ceiling to near-floor exhaust air grills or process hood inlets and always away from areas of low contamination potential to areas of higher contamination potential. 10.3.7.2 K58 Air Exhaust System 8 The K58 air exhaust system removes approximately 16,000 ft / min of air supplied from the analytical laboratory area served by the K57 air supply system. The double HEPA filter arrangement in this system consists of the final two stage HEPA filter bank. Exhaust air from the laboratory area is passed through spray mist water web-type scrubbers (one for the air from the existing laboratory and one for the laboratory addition) and through the - K58 final HEPA filter bank. Ductwork in the K58 system is made of stainless steel upstream of the scrubbers and galvanized steel downstream of the scrubbers. The K58 system exhaust air passes directly from the final filter bank to the exhaust fan and is discharged from a stack extending 25 ft above the highest portion of the UO2 Building, or 50 ft above ground elevation. The K58 exhaust system has one full-capacity fan which is connected to normal power. All final HEPA filters are in-place tested and assured to be 99.95% minimum efficient for 0.8 micron DOS cold aerosol. m AMENOMENT APPLCATON DATE: PAGE NO.: September 12,1994 10-34 $PC-ND:3330 947 (R+07S2)
Siemens Power Corporation - Nuclear Division EuF.2 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 PART ll - SAFETY DEMONSTRATION
- ggy, i
10.3.7.3 System Controls l The HVAC system is controlled by a direct digital control (DDC) system which controls temperature, pressure and flow sensors actuating valving and damper positions to hold temperatures, pressures, and pressure differentials constant in the various areas. The building ventilation systems are interlocked so that the K58 exhaust fan must be started ' before the supply systems can be started. The K58 exhaust fan power is interlocked so that the fan is stopped in the event of excessive negative pressure (more negative than -12 inches water gauge) or instrument l or control air failure. The power to the supply fans is interlocked so that a rise in K58 exhaust pressure higher than -4 inches water gauge (exhaust fan failure) or failure of i instrument or control air stops the supply fans. The laboratory is maintained at a minimum of -0.05 inch water gauge below atmospheric pressure at all times. 10.3.7.4 Delune System i A deluge system of fog spray nozzles is installed in the main K58 exhaust duct a short O distance upstream of the final filter bank. If rate-of-rise / heat detectors indicate a b predetermined temperature rise, the deluge spray is automatically activated. Should the ) deluge system be activated, differential pressure readings across the filter shall be taken i and an in-place DOS test made at the earliest opportunity. j 10.3.7.5 Two Staae HEPA Filter Bank l The two stage HEPA filter bank for the K58 system is encased in a sheet metal housing 3 that,in turn,is fastened and sealed to a concrete slab. HEPA filters rated at 1000 ft / min at one-inch water gauge pressure drop are mounted in welded steel structural frames. ' The HEPA filter medium is 100% moisture-resistant fiberglass, pleated over corrugated separators and sealed in fire-resistant plywood frames. The individual filters are certified - to remove 99.97% of 0.3 micron particles and meet or exceed Military Specification MIL-F-51079. I 10.3.8 ELO Buildina HVAC Systems The ELO Building has two independent HVAC systems. The north side of the building is basically an office / service area recirculating-type (K26) supply system with an unfiltered exhaust. The south half of the building is a research and development area with . combination once-through and recirculation supply (K24), and a double HEPA filtered (K25) exhaust system. The north and south sides of the building are isolated with a w structural wall and access to the south side is gained via an airlock. A simplified l schematic diagram of these HVAC systems is shown in Figure 11-10.25. AMENDMENT APPL.CATON DATE: PAGE NO.: September 12,1994 10-35 SPC-ND;3330.947 (R-1107/92)
.~ Siemens Power Corporation - Nuclear Division EMF-2 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257. PART ll-SAFETY DEMONSTRATION
- ggy,
\\ 10.3.8.1 K26 Air Supply System ~ 3 The K26 air supply system provides about 5000 ft / min of air to the office / service area 1 3 (north half) of the ELO Building, and 1400 ft / min to the office portion of the adjacent ELO 3 addition. Except for about E07 ft / min makeup air, the K26 unit is a recirculating supply system. The K26 supply system is activated with a simple ON/OFF switch, operates at atmospheric pressure and has no electrical control interties with the south portion of the i building. j l 10.3.8.2 K24 Air Supolv System 3 The K24 air supply system provides about 10,000 ft / min of air to the south half of the ELO Building. The ground floor area consists of a mezzanine utility area and dark room; the basement includes a machine shop, test development area, and a development 3 laboratory. Approximately 7000 ft / min of air from the mezzanine, machine shop and test development area are recirculated through a single HEPA filter bank and returned back to the building. Air supplied to the development laboratory is exhausted through double HEPA filters. ) Recirculated air is passed through a roughing filter and a single HEPA filter with an installed efficiency of 99.95% for 0.8 micron DOS cold-generated aerosol. Provision is made in the K24 recirculation system for continuous alpha radiation monitoring of recirculated air upstream of the HEPA filter bank. The alpha air monitor is set to alarm i and annunciate when alpha activity exceeds 8 MPC-hr. 10.3.8.3 K25 Air Exhaust System [ 3 The K25 exhaust system provides about 2500 ft / min of exhaust air from the development i 3 laboratory and 500 ft / min of air from the test development areas for a total of about 3000 3 ft / min. The double HEPA filter arrangement in this system consists of a final HEPA filter bank plus individual profilters and HEPA filters in the exhaust ducts of the areas or j equipment serviced. Ductwork in the K25 system is made of galvanized steel. The K25 system exhaust air passes directly from the final filter bank to the exhaust fan and is discharged from a stack extending 8 ft above the building, or 20 ft above ground elevation. The K25 system has a single full-capacity exhaust fan connected to normal power. 10.3.8.4 System Controls i I The HVAC systems are controlled with temperature, pressure and flow sensors actuating vaMng and damper positions to hold temperatures, pressures, and pressure differentials O constant in the various building areas. The building ventilation systems are interlocked so that the K25 exhaust fan must be operating before the K24 supply fan can be started. i AMENOMENT APPUCATON DATE: PAGE NO.: September 12,1994 10-36 SPC-ND3330.947 (R-1/07/92)
u Siemens Power Corporation - Nuclear Division eup.2 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO.' 70-1257 PART ll - SAFETY DEMONSTRATION REv. Pressure sensors are pmvided for valve and damper control to maintain the development laboratory negative pressees at -0.10 inch water gauge. The K25 exhaust fan power is j interlocked so that the fan is swpped upon loss of negative pressure (less negative than -8 inches water gauge), instrument or control air failure, radiation or heat detection, or loss of differential pressure across the K24 supply fan. 1 The power to the K24 supply fan is interlocked so that a loss of exhaust duct negative i pressure (less than -8 inches water gauge), loss of K24 supply fan differential pressure, radiation or heat detection in recirculation or exhaust ductwork, or loss of instrument or control air stops the supply fan. The mezzanine, Machine Shop and test development areas are maintained at -0.05 inch 3water gauge in respect to atmosphere and the north half of the building. The development laboratory is maintained at between -0.05 and -0.10 inch water gauge in respect to atmosphere and negative to the adjacent Machine Shop area at all times. As long as the K24 supply fan and K25 exhaust fan plena pressures are maintained with!n indicated ranges, the building pressure differentials will be maintained. 10.3.8.5 Final HEPA Filter Bank The final HEPA filter bank for the K25 system is a sheet metal frame and housing that is 3 fastened to a concrete slab. HEPA filters rated at 1000 ft / min at one-inch water gauge pressure drop are mounted in steel frames. Visual indicators for reading the pressure drop across the filters are permanently installed, and means are provided for in-place DOS testing. The HEPA filter medium is 100% moisture resistant fiberglass, pleated over corrugated separators and sealed in fire-resistant plywood frames. The individual filters are certified to remove 99.97% of 0.3 micron particles and meet or exceed Military Specification MIL-F-51079. 10.3.9 ELO Addition HVAC Systems Although this building addition is physically attached to the original ELO Building, it has its own separate HVAC system: K45 (supply) and K46 (exhaust). The building is physically divided with the north portion serving as an office area and the south portion housing engineering test operations, an instrument laboratory, several metallography laboratories and various chemical laboratories. The north office portion of the building is served by the original ELO Building office (K26) supply system (see Figure 11-10.25) and - is separated from the laboratory portion by a series of airlocks. The south portion of the building is served by the K45 air supply and K46 exhaust systems. The general features of the ELO Building addition HVAC systems are a once-through ceiling-to-floor airflow (K45) supply air system and a double HEPA filtered (K46) building AMENDMENT APPLCATON DATE: PAGE NO.: September 12,1994 10-37 SPC-ND.3330 947 (R UO7/92)
Siemens Power Corporation - Nuclear Division EMF 2 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 PART 11 - SAFETY DEMONSTRATION REv. exhaust system. A simplified schematic of the ELO Building Addition HVAC system is shown in Figure ll-10.26. r 10.3.9.1 K45 Air Supply System l 3 The K45 air supply system provides about 16,000 ft / min of 100% outside air to the south portion of the ELO addition. Airflows are directional from ceiling to near-floor exhaust air 'l grills or process hood inlets, and always away from areas of low contamination potential to areas of higher contamination potential. 10.3.9.2 K46 Air Exhaust System The K46 air exhaust system removes approximately 18,000 ft / min of air supplied from the f 3 process areas served by the K45 air supply system. The double filter arrangement in this system consists of the final HEPA filter bank plus individual or smaller filter banks of prefiiters and HEPA filters in the exhaust ducts of the areas or equipment serviced. Ductwork in the K46 system is made of galvanized steel. 9 The K46 system exhaust air passes directly from the final filter bank to the exhaust fans and is discharged from a stack extending 23 ft above the main portion of the UO Building, or 35 ft above ground elevation. The K46 exhaust system has a single fu! capacity fan connected to normal power. 4 10.3.9.3 Systems Controls The HVAC systems are controlled with temperature, pressure and flow sensor actuating valving and damper positions to hold temperatures, pressures, and pressure differentials j constant in the various building areas. The building ventilation systems are interlocked i so that the K49 exhaust system must be operating before the K45 supply system can be i started. Pressure sensors are provided for damper control to maintain a negative pressure (minimum -0.05 inch water gauge) in the laboratory areas relative to the office area and to atmosphere, and to maintain an exhaust duct negative pressure between -4 and -12 inches water gauge. Any loss of instrument air prevents the supply and exhaust systems from operating. Both the K45 supply fan and K46 exhaust fan are interlocked so that a loss of exhaust duct negative pressure above -4 inches water gauge (toward i zero) will shut down the supply fan or an increase of exhaust duct negative pressure l below -12 inches water gauge will shut down the K45 supply fan and the K46 exhaust fan. ~ i Automatic audio and visual alarms are activated when any supply or exhaust system i upset occur. Pressure differential indicating devices and airflow quantity meters are located in the controlled zones and/or on the main HVAC panel to provide system and zone operating conditions. Ci =cw:mut appucato= onre oeptemoer 1z,1we nos e ivisu ] SPC ND.3330 947 (R-UO7/92)
- {
l .. ~
i Siemens Power Corporation - Nuclear Division EMF-2 O SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 V PART 11 - SAFETY DEMONSTRATION
- ggy, 10.3.9.4 Deluae System A deluge system of fog spray nozzles is installed in the exhaust duct in the K46 system l
a short distance upstream of the final filter banks. If heat detectors indicate a temperature l of 140-190*F, the deluge spray is automatically activated. Should the deluge system be activated by any circumstance, differential pressure readings across the final filters will be taken and in-place DOP/ DOS test made at the earliest opportunity, t 10.3.9.5 Final Filter Bank The-final filter banks in both the K46 and K56 systems are encased in a poured and 3 sealed concrete structure. HEPA filters rated at 1000 ft / min at 1-inch H O pressure drop 2 are mounted in we'ded, steel structural frames. The HEPA filter medium is 100% moisture-resistant fiberglass, pleated over corrugated separators and sealed in fire-resistant plywood frames. The individual filters are certified to remvoe 99.97% of 0.3 micron particles and meet or exceed Military Specification MIL-F-51079. 10.3.10 Contaminated Clothina Laundry HVAC System The general features of the contaminated clothing laundry HVAC system are a once-through ceiling-to-floor airflow supply system (K41) system and a double HEPA filtered exhaust system (K42). A simplified schematic diagram of the air supply system and the air exhaust system is shown in Figure 11-10.27. l 10.3.10.1 K41 Air Supolv System 3 The K41 air supply system supplies about 1900 ft / min of 100% outside air to the laundry l room which is divided into a cleaning area and adjoining sorting area. Airflows are directional from ceiling to near-floor exhaust air grills or a hood inlet, and always away from areas of low contamination potential to areas of higher contamination potential. 10.3.10.2 K42 Air Exhaust System Air supplied to the cleaning and sorting areas, plus infiltration and dryer exhaust 3 (approximately 4200 ft / min),is exhausted through the K42 exhaust system. The double HEPA filter arrangement in this system consists of the final HEPA filter bank and upstream primary HEPA filter bank plus individual prefilters located in the exhaust ducts of the two areas serviced. 3 The K42 system exhaust air (approximately 6300 ft / min.) passes from the two stage filter bank through the main exhaust fan, a duct air monitor (measuring airflow quantities) and p is discharged from a stack extending 25 ft above ground on the southwest side of the building. The K42 exhaust system has one full-capacity fan which is connected to normal Mb auc=outm a m oanonoart; septemoer 12,1ww4 PAGE NO.: SPC-ND:3330D47 (R-1/07/92)
Siemens Power Corporation - Nuclear Division aus.2 f3 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 LI PART 11 - SAFETY DEMONSTRATION g y, i i power. All final HEPA filters are in-place tested and assured to be 99.95% minimum efficient for 0.8 micron DOS cold aerosol. I 10.3.10.3 Systems Control The HVAC systems are controlled with temperaturo, pressure and flow sensor actuating valving and damper positions to hold temperatures, pressures, and pressure differentials constant in the various building areas. The K41 supply air system is interlocked with the K42 exhaust system to prevent operation of the K41 supply air system without the K42 exhaust system operating. Pressure sensors are provided to maintain a minimum negative differential pressure of 0.05 inch water gauge in the cleaning area relative to the atmosphere. The K41 supply fan is interlocked so that a loss of exhaust duct negative pressure above -3 inches water gauge (exhaust fan failure) or a signal from the exhaust duct heat detector will shut down the K41 air supply system. Automatic visual alarms are activated when any supply or exhaust system upset occurs. Pressure differential indicating devices and airflow quantity meters are located on the L main HVAC panel to provide system and zone operating conditions. 10.3.10.4 HEPA Filter Bank The final HEPA filters are enclosed in a sheet metal housing that, in turn, is mounted on structural steel legs fastened to a concrete slab. The HEPA filters are rated at 1000 3 ft / min at one-inch water gauge pressure drop and are mounted in welded steel frames. Continuous air samplers are installed downstream of the filter bank. Visualindicators for reading the pressure drop across the filters are permanently installed, and means are provided for in-place DOS testing. The HEPA filter medium is 100% moisture-resistant fiberglass, pleated over corrugated separators and sealed in fire-resistant plywood frames. The individual filters are certified to remove 99.97% of 0.3 micron particles and meet or exceed Military Specification MIL-F-51079. 10.4 Hadioactive Waste Handlina The facilities and processes which are involved in the handling of radioactive and chemical wastes produced by SPC are described in the following sections. IV auc~au m a m cu o~oc o September 1z,1994 paac wo.: l u-4u SPC-ND.3330 947 (Ra/07/92)
I l Siemens Power Corporation - Nuclear Division EMF-2 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 PART ll-SAFETY DEMONSTRATION
- ggy, 10.4.1 Lanoon System Description The lagoons provide containment for all uranium and chemically-contaminated liquid wastes generated at SPC.
Natural evaporation, controlled waste addition, waste discharge to the municipal sewer and water additions are used to control the volume of liquid stored in the lagoons. Inter-lagoon transfers are periodically made for both uranium accountability and volume control purposes. Sampling between lagoon liners is conducted monthly (unless prevented by freezing weather) to determine if leaks have occurred. There are six liquid waste storage lagoons, one solids leach pit, and one sand storage pit located along the east boundary of SPC (see Figure 11-10.1). The sand storage pit is located immeciately west oi Lagoon 5. Tree solids Isach p.t IJ lOcolGd imD&difOl'/ V' Ort of Lagoon 2. The dimension and capacities of the lagoons are: Lagoon Dimensions Est. Capacity 10' Gal 1 240' x 200' x 3' deep 1.4 i 2 240' x 100' x 3' deep 0.7 3 240' x 350' x 5'6" deep 3.5 4 240' x 290' x 6' deep 2."' 5A 240' x 175' x 7'6" deep 1.6 i 5B 240' x 175' x 7'6" deep 1.6 Leach Pit 40' x 54' x 8'6" deep 0.06 Sand Pit 39' x 300' x 6' deep 0.3 The lagoons and the solids leach pit have a " sandwich" construction; they each have two liners made of impervious material, separ&ted by 6 inches of sand. On Lagoons 1,2 and 3, the " sandwich" rests on an asphalt-type surface known as Petromat. In the sand layer between the impervious liners is an array of sample heads. The sand storage pit is single-lined and, therefore, does not have between-liner sampling capabilities. Tubing from each of the heads is routed to the berms on both the east and west side of the lagoons where small pumps can be periodically connected. Some additional sampler heads are located between the original Petromat liner and the lower liner. Lagoon 4 is equipped with three " dry wells" below the bottom liner. Samplers are located in each dry well. All sampler heads are pumped each month using small air-driven pumps (unless prevented by freezing weather) for leak detection purposes. C: = cam ameno= oam septemoer u, iw4 pass uo.: 10-gi $PC-ND.3330 947 (R 1/07/92)
Siemens Power Corporation - Nuclear Division eur.2 O SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 k.) PART ll - SAFETY DEMONSTRATION REv. Lagoons 1 and 2 are used as receivers of ammonia-bearing solutions from the conversion ) area with a low-level of uranium and is the feed lagoon for the ammonia recovery (AR) process. They have impervious floating covers to contain the ammonia fumes. Lagoon 3 is a storage lagoon for high uranium content wastes and serves as the feed lagoon to the LUR Facility. Lagoon 4 is a storage lagoon for low uranium content waste from the LUR process. Lagoon 4 also serves as the feed lagoon for LUR waste to the AR Facility. Lagoon SA is a storage lagoon for waste streams from the AR Facility and miscellaneous low uranium, low ammonia chemical wastes. Disposal of waste from Lagoon SA to the city sewer is accomplished after treatment by lon exchange to remove residual uranium (see Section 10.4.3). The waste sewering rate is automatical;y contiolled by a microprocessor via a flow measuring and control system. The sewered chemical waste is volume proportionally sampled. The chemical waste is monitored for uranium and chemical content and sewering rate is controlled based on waste composition so as to remain within discard limits. t i V Lagoon SB is currently used to store high uranium content waste to be treated for uranium recovery. Once emptied, this lagoon will be used as a batching lagoon in conjunction with Lagoon SA to receive AR Facility waste for metered discharge to the city sewer. The sand pit is used as a storage pit for sand and sludge that has been removed from the liquid storage lagoons during cleanup over the years of operation.' The solids leach pit is used for decontamination of sand. The process is described in detail in Section 10.4.4. Periodically, liquid waste solutions are transferred from one lagoon to another for accountability, volume control or maintenance purposes. A permanently mounted pump with interconnecting piping enables solution to be pumped from any lagoon to any other lagoon and also allows recirculation within any lagoon. 10.4.2 Ammonia Recovery (AR) Description 2 The AR process is housed in a 1635 ft insulated steel structure located north of Lagoon 1 (see Figure 11-10.1). The structure is designed to withstand UBC Zone 11 seismic loading 2 and 20 lb/ft windward pressure. An equipment arrangement is shown on Figure 11-10.28. The Engineering Flow Diagram is presented on Figure 11-10.29. A brief description of the ,m major equipment pieces is given in Table 11-10.1. The building also houses the Lagoon ( SA IX Process (see Section 10.4.3). aut.oum amccow nam deptember 12,1994 PAGE NO.: 10-42 SPC ND.3330 947 (Rm0742)
Siemens Power Corporation - Nuclear Division eur.2 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 PART 11 - SAFETY DEMONSTRATION ggy. The unneutralized low uranium content liquid process waste from the feed lagoons (Lagoon 1 or Lagoon 2) is transferred to the feed tank which provides approximately one hour surge capacity. Process waste from the LUR operation can be added to the feed I tank and processed with the conversion process waste. Sodium hydroxide is added to the feed tank to replace the ammonium in the ammonium salts, and to keep the pH high to reduce corrosion. The feed tank is maintained under a slight vacuum by venting through a scrubber. Feed solution is pumped via a flow control system and an energy recovery heat exchanger to the ammonia stripper. The heat exchanger is provided to reduce energy requirements for the process by using the hot stripper bottoms to preheat the feed solution. The ammonia stripping column provides for removal of ammonia from the waste solution by countercurrent contact with steam and is designed to produce 20 to 30 wt% ammonia product solution and waste effluent at less than 100 ppm ammonia. The bottoms from the stripper are pumped to the designateo waste treatment batching Is.gcor. (SA or SD) or recycled back to the feed tank, or the feed lagoon, depending on its ammonia concentration, temperature, and pH. O The air purge system on the pneumatic instrument lines is designed to prevent backup of process fluids in the event of excessive system pressure. A 50 psig rupture disk is provided at the top of the column to prevent overpressurization of the system. The pressure relief vents to the atmosphere through the tower roof. The condensables from the stripper overheads are removed in a downdraft condenser and are routed to the distillate tank. An automated deionized water injection system is provided to improve ammonia removal efficiency and to control the ammonia hydroxide i concentration. A scrubber is provided to remove ammonia from the process vessel offgas. The scrubber bottoms are routed to the feed tank. The scrubbed offgas is vented to the atmosphere via the building exhaust fan and stack. The control system is an electronic digital microprocessor system. The console can access process information and display it on the CRT. In case of transmitter failure, the processors will adjust to a safe value that can keep the process from going too far out-of-control. Deviations from set points are alarmed when they surpass their predetermined limits. A steam boiler provides a maximum of 3000 lb/hr of steam to the AR process. Boiler pressure controlis maintained by SCR's driven from pressure instrumentation. Steam header pressure instrumentation is provided to shut down the process and alarm on low and high steam pressure. O = c a cia a m cat m oart: september 12, Tw4 enac wo.: 10-4a SPC ND:3330D47 (4v0W92)
Siemens Power Corporation - Nuclear Division eus.2 (' SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 i PART ll - SAFETY DEMONSTRATION ggy. A 28,000-gallon, aboveground sodium hydroxide storage tank is provided to supply the AR process. The tank is insulated and heated with a 5 kW heater to maintain the caustic temperature above 65'F. All exterior piping is heat-traced and insulated. A safety shower is provided at the load-in facilities. Operation of the NaOH storage system is monitored end controlled at the control room. A 28,000-gallon, aboveground storage tank is provided to store NH 0H product from the 4 AR process. Facilities are provided to load out excess ammonium hydroxide for sale offsite. The NH 0H used for recycle to the process is transferred to either of two 10,000-4 gallon, ammonium hydroxide storage tanks. A concrete spill containment structure is provided for the outside bulk sodium hydroxide and product ammonium hydroxide storage tanks. The structure is designed to contain chemicals from a ruptured storage tank. Tne builo'ing exhaust system conJists of a two-cpced fan, temperature contrels, an ammonia monitor and stack. The exhaust fan is normally run on low speed 3 (approximately 1700 ft / min). The fan is switched to high speed (approximately 5000 h ft / min) if the building exhaust temperature exceeds 110 F. The building exhaust air V ammonia monitoring system is set to alarm locally and in the Line 2 Control Room. The AR Facility fire detection and alarm system has been tied into the existing plant systems. Criticality in the AR system is not deemed credible due to the extremely low concentration of uranium, in addition, the feed tank and stripping column were designed for solids to flush through the system. Nevertheless, the system is inspected quarterly for signs of buildup of uranium solids. The plant criticality detection and alarm system also covers the AR Facility. 10.4.3 Laooon SA IX Process The Lagoon SA lon exchange system is located in an addition to the Ammonia Recovery Facility (ARF) Building. The addition is 25' x 27' x 20' high. It is a pre-engineered metal building located on a concrete slab. A plan view of the addition is shown in Figure 11-10.28. The equipment arrangement is shown in Figure 11-10.29b. The building is insulated and designed to withstand Uniform Building Code Zone il seismic loading and a 20 lb/sq. ft. wind. The floor is sealed and caulked to be leak tight. The floor slopes toward the sump area and has a sill or curb for leak containment except for the doorway which is protected by a trench sloped to drain to the main sump. The process equipment is located in the sump area which has a 4 inch recessed floor. The sump pumps to o Lagoon 3. Controlis automatic, but can be controlled manually. The sump is designed I to collect minor leaks of lagoon solutions containing uranium and sodium and ammonium v =cuoutwT amcaron occ: Septemoer 12, Iw4 nos so.: 10-44 EPC-ND.3330 947 (R-1/07/92)
Siemens Power Corporation - Nuclear Division eur.2 (G3 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET N PART ll - SAFETY DEMONSTRATION REv. to collect minor leaks of lagoon solutions containing uranium and sodium and ammonium sulfates, fluorides and nitrates and ion exchange reagents such as sulfuric acid, sodium hydroxide, and water. The sump and floor areas are inspected annually for leaks. Those areas are normally dry, leaks are repaired as soon as practical. Lagoon SA waste solutions scheduled for discharge to the city sewer may be routed through the ion exchange (IX) process located in the ARF Building to further reduce the uranium content. The Lagoon 5A IX Process consists of two sand filters and an ion exchange column and associated auxiliary equipment. The process flow is shown in Figure ll-10.29a. Lagoon SA solution is pumped from the lagoon through the primary filter located at the lagoon pump out area to remove small particles and suspended solids. The solution then is pumped through the polishing filter in the ARF Building and to the adjacent IX column at a typical rate of 20 gallons per minute. Passage through the column reduces the uranium concentration by a factor of about eight or more. The uranium held up on the column is eluted from the column and transferred to Lagoon 3 for later uranium recovery. Lagoon 4 solution may be used as an aiuting solution. Other reagents such as carbonale solutions may be uced. rho regeneration cycle of the resin uses sodium hydroxide and sulfuric acid which are available at the facility. These are discharged from the resin to Lagoon SA. The polishing filter is backflushed to Lagoon U
- 3. The Lagoon SA filter is backflushed to Lagoon SA. The media of the sand filters is expected to last indefinitely and the resin is expected to last at least five years if reeded, the resin can be disposed of by incineration at SWUR.
Typical uranium concentrations of Lagoon SA solutions fed to the IX column are 1-2 ppm and corresponding effluent concentrations are 0.1-0.2 ppm. During the loading cycle from 500,000 to 1,000,000 gallons of waste solution are passed through the column to load from 4-8 kilograms of uranium on the resin column. The basis for criticality safety is concentration control. The uranium concentration in any part of the system is maintained at less than 50% of the minimum critical concentration. The highest uranium concentration in the system is that of the loaded ion exchange resin. Typical uranium concentrations of the resin just before elution are less than 10% of the i minimum critical concentration. A Criticality Safety Analysis has shown that all parts of the system are suberitical even under abnormal conditions of the transfer of uranium bearing lagoon solids to the first sand filter or saturation of the ion exchange resin with uranium. The facility is covered by the existing criticality accident detection and alarm system. The system is monitored for uranium buildup by sampling the solid phase of Lagoon 5A semiannually and analyzing for uranium. The sand filters and resin bed are also sampled and analyzed for uranium semlannually. In addition, the resin is inherently safe since it A saturates with uranium at less than 140 gU/t. Buildup of uranium on the column is j monitored by process control and accountability samples. autwoutwT amcatow caTt; September 12,1994 PAGE NO.: 10-4b SPC-ND 3330.947 (R4/07/92) i
Siemens Power Corporation - Nuclear Division EMF-2 p SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 G PART ll - SAFETY DEMONSTRATION REv. For fire detection there are rate-of-rise / fixed temperature detectors in the ceiling. These detectors set off alarms locally, at the Central Guard Station, and the Richland City Fire Department. The ion exchange room has a hand-held fire extinguisher. There is a two-hour rated fire wall between the original building and the new addition and a similar fire wall between the outside storage area for sulfuric acid and the building housing the ion exchange system. There are no gaseous or particulate releases since all the radioactive materials are in liquid form in closed vessels or in large double lined lagoons. The release of radioactive materials from process equipment is prevented since all vessels and associated piping are designed to withstand a pressure of at least 100 psig versus the maximum pump discharge pressure of 38 psig. Radiation work shall be controlled through the Radiation Work Permit System. All operations shall be conducted within the ALARA concept. The environmental impact of the ion exchange process for treating Lagoon SA solution is judged to be insignificant. The equipment is located in a building within the restricted area which is committed to industrial use. The equipment is located in an area cpocifical!y designed to contain any spills or leaks. There are no gaseous effluents or ,m (V) increases in the quantities of radioactive waste generated. All underground transfer lines entering and leaving the new facility are completely double-encased (inner pipe surrounded by a sealed secondary plastic containment shell), with electronic leak detection systems that alarm locally and in the conversion Line 2 control room. The primary pipes are tested hydrostatically annually and the secondary pipes biennially. The process generates small quantities of additional chemical wastes, but this is more than offset by the positive environmental impact of reducing the quantities of uranium discharged to the Richland city landfill. 10.4.4 Lacoon Uranium Recovery (LUR) Facility Description The LUR Facility is provided to recover LEU from stored high uran!am content liquid chemical wastes (see Section 10.4.1). Following uranium reevnry, the waste is treated for ammonia removal (see Section 10.4.2), then disposed to th e municipal sewer. The LUR Facility is located adjacent to Lagoon 4 as shown on Figure ll-10.1. The equipment consists of six process vessels, one chemical makeup vessel and associated pumps, piping and filters. A brief description of the major equipment is given in Table ll-10.2 and the equipment layout is depicted in Figure 11-10.30. The equipment is not housed, but is partially covered. The process equipment is not freeze-protected and is, therefore, shut down and winterized during cold weather. A process flow diagram for the uranium recovery system is presented in Figure 11-10.31. O Approximately 5000 gallons of high uranium content waste is pumped into each of the two precipitators. The uranium is precipitated from solution by addition of a reductant AMENDMENT APPLCATIOed DATE: September 12,1994 PAGE NO; I O-4ti sPC-ND;3330 947 (R-1/07/92)
Siemens Power Corporation - Nuclear Division eup.2 ^ (O 's SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 l PART 11 - SAFETY DEMONSTRATION
- ggy, and allowed to settle. At the end of the settling period, the supernatant is decanted to lagoon storage. This procedure is repeated until the total precipitate accumulated in the precipitator vessel reaches the desired batch size. The uranium precipitate is slurried to the washer tank and water-washed for extraneous chemical removal. The washed precipitate is then slurried into a container, dissolved in aluminum nitrate solution,,
pumped through a centrifuge to a plastic 55-gallon drum, and placed in storage. This cycle takes a day so the capacity is approximately 10,000 gallons per day. The recovered uranium is purified for reuse through existing solvent extraction facilities. A vent system is in place for removal of NO, fumes generated during the dissolving of the precipitate. At this time, an upgrade to the vent system is being planned with a scrubber, HEPA filtration and air sampler. In addition to routine processing of lagoon solutions, the centrifuge at LUR is used for separating mop powder solids after processing in ELO. The basis for criticality control is mass control. The mass in any vessel is maintained at less than 45% of the minimum mass required for criticality. The typical mass in any [ vessel is usually less than 30% of the minimum critical mass. A minimum spacing of 3 V ft is maintained between all process vessels. 10.4.5 Solids Uranium Recovery Facility The lagoons contain a wide array of solids types, consisting largely of soluble inorganic salts containing ammonium and sodium fluorides, sulfates, and nitrates; silicates, precipitates of uranium, gadolinium, zirconium, calcium, aluminum, and other salts and metals; sand; silts and clays; and miscellaneous debris blown in by the wind. Some of the lagoon solids contain chemical constituents in economically recoverable quantities, most notably uranium and ammonia. Both the quality and composition of the lagoon solids vary from lagoon to lagoon, depending on the process waste streams that have been historically managed in each lagoon. Processing of lagoon solids removed in conjunction with past lagoon relining / repair operations has historically been accomplished in a solids uranium recovery facility, consisting of a sand trench, leach pit, and two vibrating screens. Solids placed into the sand trench were processed via a series of physical screening / sizing and washing operations which allowed relatively higher uranium fines and liquids to be returned to the lagoons (for future uranium removal) and low uranium insoluble sands to be placed into the leach pit. Contingent on meeting certain size and uranium concentration specifications (per Amendment 19 to Condition 8 of the U.S. Ecology's NRC License 16-19204-01), the washed and sized sand can be disposed of unpackaged and in bulk as n backfill material at the U.S. Ecology Hanford Low-Level Radioactive Waste Disposal (v) Facility. auceevem amcuou out; september 12,1934 PAGc No.: 10-47 J SPC-ND.3330 90 (R-1/07/92)
i Siemens Power Corporation - Nuclear Division Eur-2 l SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 l' PART ll-SAFETY DEMONSTRATION REv. SPC has no plans for further usage of the sand trench / leach pit system for the processing of lagoon solids. A new Solids Processing Facility (SPF) is currently being planned that will meet the processing requirements for the entire lagoon solids inventory. Preliminary i development work on the SPF calls for solids retrieval via dredging and/or manual { collection, followed by dissolution and a variety of solids separation steps. Clarified liquid from the process will be routed to the Lagoon Uranium Recovery (LUR) Facility for uranium reclamation. Residual insoluble solids (sands, silts, clays, fines) will be disposed of either as nonregulated solids, low-level radioactive waste (containerized or bulk packaged), or mixed waste, depending on the degree and effectiveness of solids treatment processes. Appropriate license revisions to accommodate the SPF will be submitted as necessary. j 10.4.6 Solid Waste Uranium Recovery (SWUR) Facilltv Description l The Solid Waste Uranium Recovery (SWUR) facility is designed to incinerate combustible uranium-contaminated wastes for volume reduction and uranium recovery. The j incinerator, though designed to meet hazardous waste incineration criteria, is not permitted as a hazardous waste incinerator. The incinerator and auxiliary systems are located in Room 173 of the Specialty Fuels Building (See Section 10.1.1). f 'The process is divided lato four systams: foed preparation, waste incineration, offgas I treatments, and ash handling. l t 10.4.6.1 Feed Preparation Solid wastes generated in the nuclear fuel fabrication activities at SPC are sorted at points of generation to separate dangerous wastes from nondangerous solid wastes. The dangerous wastes are packed in plastic-lined 55-gallon DOT 17H metal drums for onsite .i mixed waste storage. The nondangerous wastes are sorted into combustible and noncombustible waste fractions in the UO Building. The noncombustible wastes are 2 surveyed item-by-item for uranium content and packaged for compaction and/or disposal j at a licensed radioactive waste burial ground. The combustible wastes are packaged in plastic-lined 55-gallon DOT 17H metal' drums for storage prior to incineration. All drummed waste packages are labeled for contents as well as weighed and assayed for j uranium content prior to storage, disposal, or incineration. 4 The combustible waste packages are transferred from storage to the SWUR facility where the contents are again sorted as described above. This second sort provides additional opportunity to verify that no dangerous wastes are present in the feed stream. The 8 combustible wastes are then packaged into plastic-lined 3.5 ft cardboard boxes for feed to the incinerator. The incinerator feed boxes are weighed and assayed for uranium then j conveyed to the incinerator hydraulic ram feeder which is automatically controlled to feed the incinerator primary chamber through a guillotine door. The ram feeder has a sealed j am ousm amecoats: septemDer 12,1w4 PAGE NO.: 10-45 SPC-ND.3330 947 (R-1107/92)
- l
Siemens Power Corporation - Nuclear Division eur.2 (U3 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET N PART ll - SAFETY DEMONSTRATION
- ggy, 3
loading hopper and can handle up to one yd feed batches. The guillotine door and ram feeder are interlocked to prevent simultaneous opening of both doors. The feed hopper and ram face are automatically inerted using a nitrogen fire-suppression system if a fire is detected by a heat detector located on the feed hopper, 10.4.6.2 Waste incineration The safe batch incinerator has a nominal capacity of 90 kg/hr of uranium-combustible waste consisting of paper, plastics, wood, etc. The incinerator is a commercial controlled-air unit modified to minimize ash holdup and to facilitate good carbon burnout and system cleanout. The unit consists of primary and secondary chambers, each constructed of a carbon steel shell internally coated with a mastic material for acid gas corrosion protection and lined with both insulating and high density castable refractory selected to minimize permeation of uranium contamination and to provide good service life. Both chambers have propane-fueled burners and combustion air ports. The primary chamber is operated at 1400 F - 1800 F, and the combustion air flow is controlled to near g y stoichiometric requirements to promote quiescent burning with minimal particulate v entrainment. Combustion products and pyrolysates are passed to the secondary chamber where excess air and high temperature (approximately 2000 F) and a minimum tuercecor.d retidence timo combine in ecmpletely burn all combustible gases, including dioxins which may form from PVC incineration. Alarms and interlocks are provided to prevent feeding the incinerator if either the primary or secondary chamber temperature is too low or high; if there is low combustion air pressure; if high pressure occurs in the primary chamber; if high liquid level occurs in the quench column; or if low liquid level occurs in the packed column scrubber. An extremely high chamber temperature or propane bumer malfunction automatica!!y shuts down the incinerator. 10.4.6.3 Offoas Treatment The incinerator exhaust gas from the secondary chamber contains particulates, vapors, and gases (including acidic gases) which result from the combustion of the cellulose, rubber, and plastics present in the waste feed. Cooling of the gases and removal of the acidic gases and potentially radioactive particles is accomplished by system components which consist of a quench column, a high-energy venturi scrubber, a packed column, a 1 mist eliminator, a reheater and HEPA filtration modules. The quench column is divided into an upper contacting section and a lower sump section. In the contacting section, cooling and saturation of the incinerator exhaust gas occur simultaneously by the evaporation of scrub solution liquid. Excess solution collects in the 10-inch diameter sump section while the saturated gas is routed to the inlet of the (o) venturi scrubber. An automatic emergency quench system is provided to supply process AMENOMENT APPLCATON DATE: September 12,1994 PAGE NO.: 10-4W SPC-ND 3330.947 (R 1/07/92)
Siemens Power Corporation - Nuclear Division EM F-2 p SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 G PART 11 - SAFETY DEMONSTRATION ggy. water to the quench column if a high column outlet temperature or loss of normal power occurs. The variable throat, high-energy venturi scrubber, located between the quench column and the packed column, removes more than 99 wt% of the offgas particulates. Scrub solution is injected through a nozzle located upstream of the throat. An alarmed interlock is provided to prevent feeding the incinerator if the scrub solution flow is low. The packed column is designed to slightly cool the offgas and to remove the acidic gases from the gas phase by countercurrent contact with recycled scrub solution. The gas is discharged from the packed column through a polypropylene mist eliminator. Alarmed interlocks are provided to prevent feeding of the incinerator if a high packed column inlet temperature. The gas enters the electric reheater where it is warmed to a minimum of 15*C above the saturation temperature. This heating reduces the relative humidity of the gas to prevent wetting of HEPA filters. The reheater is controled by a silicon controlled rectifier which maintains the proper temperature difference Dom the reheater inlet to the filter outlet. Alarms and interlocks are provided to prevent feeding of the incinerator if a high heater differential temperature or high filter inlet temperature occurs. ( w/ The offgas module contains the prefilter and two banks of HEPA filters. The stainless steel module housing is designed and reinforced to withstand the 150-inch H O vacuum 2 to which it may be exoosed. DOS testing of the final HEPA filter bank for verification of i 99.95% removal efficiency for 0.8 micron particles can be accomplished. Particulates that have been scrubbed from the gas stream are removed from the scrub solution by the filters. The filter elements are made from combustible materials so that once they are expended, they can be burned in the incinerator. Minimum expected replacement period between element changeout is 8 hours. The scrub solution liquid is circulated by redundant pumps to insure a continuous stream of scrub solution liquid. Automatic switchover of pumps occurs upon on-line pump failure. Scrub solution liquid is cooled by a plate-type heat exchanger. Cooled liquid is used to improve acid gas absorption and to reduce the packed column discharge gas temperature. An alarmed interlock is provided to prevent feeding the incinerator if a high cooler outlet temperature occurs. Caustic addition to the system is automatically controlled by the pH controller in the packed column, scrub solution liquid line. Scrub solution Nacl concentration is maintained at 6% in order to prevent corrosion of metal components. Other major chemical constituents of the scrub solution liquid are Na CO (from CO absorption),and Na SO (from sulfur compounds present in the paper 2 3 2 2 4 f-m products). A total dissolved solids analyzer, which basically works on solution ( conductivity, is provided to control the blowdown stream. The blowdown stream is amoenmcaron tart: september 12,1994 PAGE NO.: I u-Du SPC-ND.3330 947 (R+07/92)
i Siemens Power Corporation - Nuclear Division EMF-2 O SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 a PART ll - SAFE'IY DEMONSTRATION
- ggy, proportionally sampled and discharged to SPC's surface impoundment system (see Section 10.4.1).
10.4.6.4 Ash Handlina Ash formed from the combustion of wastes is pushed along the hearth by incoming feed and by an internal ash plow. The ash passes through an ash gate into an ash cooling chamber located at the end of the hearth. The cooled ash is discharged periodically into 30-gallon DOT 17-H metal drums. Each ash drum is homogenized and sampled for uranium and U-235 isotopic content and its net weight is determined. The drums are then stored for future recovery of the contained uranium. 10.4.7 Plutonium-Contaminated Waste Storaae A waste storage facility is provided for storing Pu contaminated waste which remains from a previous mixed oxide fuel fabrication facility. The Pu concentration in the contaminated waste is greater than allowed for Class C waste and therefore no disposal site exists which is licensed to receive this waste. The facility is described below and depicted in Figure 11-10.2. ] The storage facility is located in Room 162 of the SF Building. The facility is a below-grade room (approximately 12 x 20 x 20 ft deep) constructed of reinforced concrete and covered by steel floor grating overlaid with steel plate. The room contains a sump for lic,uld collection which is monitored by a liquid level alarm. A sump pump is installed j which can be manually activated and which discharges to a waste retention tank south of the UO Building. ] 2 Drum storage is on steel grating to support the drums off the concrete floor and on a mezzanine also fabricated of steel grating. Ingress and egress for personnel and i equipment is from the top of the room. i The room is ventilated. Air is drawn down from the roof and exhausted near floor level through one st.ge of HEPA filtration into the SF Building exhaust system. The exhaust air is continuously sampled and monitored prior to the installed HEPA filter. The air sample is analyzed weekly. 10.5 Fire Protection 10.5.1 Buildina Codes and Standards All permanent buildings at the SPC Engineering and Manufacturing Facility were 1 ( constructed in accordance with the applicable sections of the following building codes 4 and standards. PaGE NO.: l V-bl aucuovon amcara mts: Septemoer 1z,1w4 sPC-ND.3330.947 (R it07/92)
Siemens Power Corporation - Nuclear Division eur.2 [] SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 U PART 11 - SAFETY DEMONSTRATION
- ggy, UBC (Seismic Zone ll)
Uniform Plumbing Code Uniform Mechanical Code Uniform Fire Code National Fire Codes (NFPA) National Electrical Code ANSI-C1 ASHRAE Standards Washington Administrative Code, Chapter 296-24 Washington Administrative Code, Chapter 296-44 Richland Municipal Code and Zoning Regulations Richland Municipal Ordinances Numbers: = 3777 (adopt. Building Code) 3877 (adopt. Plumbing Code) 3977 (adopt. Mechanical Code) 10.5.2 Fire Protection Llability inspections p} SPC has elected to self-insure with regard to property damage. American National iV insurers (ANI) schedules a fire protection audit of its' policy holders, among whom is SPC, approximately every year by an acknowledged fire protection consultant. Richland's Department of Fire and Emergency Services conducts annual fire protection inspections of SPC's Engineering and Manufacturing Facility. The most recent copies of these a.:ctits and inspedions Le appe:Jed (see Appendix A). = 10.5.3 Fire Protection Procram 10.5.3.1 Combustible Solid Waste Handlina and Storsae Outside metal waste containers are provided by the City of Richland for clean wastes. Contaminated combustible wastes are properly sorted into metal boxes or drums, sealed and stored on an outside pad for future uranium recovery or disposal per approved procedures. Combustible wastes generated inside the process and other buildings (either clean or contaminated) are collected in metal waste containers and emptied daily into the appropriate waste storage containers. 10.5.3.2 Flammable Llauid Storaae Flammable liquids are stored in approved safety containers or cabinets near the final-use location. Additional storage for flammable liquids is provided for in approved safety p cabinets in the warehouse complex. U m c a c c a m c c om o c c: deptemoer 1z,1994 mc wo.: l u-52 SPC ND.3330.947 (R-U07/92)
Siemens Power Corporation - Nuclear Division sur.2 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 v PART ll-SAFETY DEMONSTRATION g y. 10.5.3.3 Combustible Liould Storaos Combustible liquids are stored in approved metal containers near the final-use location. Additional storage for combustible liquids is provided for in fire-resistant metal warehouses located away from any radioactive material storage areas. 10.5.3.4 Fire Prevention ) The manifolds for supplying combustible gases to the facility, including backup hydrogen for the sintering furnaces, are located outside the main building structure. All combustible gas distribution piping meets applicable NFPA codes. Combustible gas burn-off devices and combustible gas detection equipment are used j where necessary w pc nt explosion and fires around sintering fumaces and ovens. The HEPA exhaust filters in the UO and SF Buildings are protected from high 2 temperatures and burning debris in the event of fire by automatic deluge systems in the exhaust plenums immediately upstream of the final filter bank. i U 10.5.3.5 Fire Detection and Alarm Rate-of-rise / fixed temperature heat detectors are used in the facility to detect fires. This fire alarm equipment is installed to provide automatic, as well as manual alarm signals in event of a fire. The system includes an annunciator in the Central Guard Station which indicates which zone in the system has actuated (see Figure 11-10.32). A signal.is.also_ --- - automaticallyiransmitted to the Richtand Fire Departmerit; The fire alarm is a single-strike gong (2 strokes /sec). The fire alarm system is inspected and tested in accordance with the applicable, preventive maintenance procedures. 10.5.3.6 Fire Defenses SPC's Engineering and Manufacturing Facility is located within the city limits of Richland and thus, is served by the Richland Fire Department. The Washington Surveying and Rating Bureau has graded the City as Class 3 in its last survey. The closest Richland fire station is located at the intersection of McMurray and Jadwin Avenue, about 5 road-miles from the plant. The Fire Department estimates running time to the plant to be approximately 6 minutes. The City has Mutual Assistance Agreements with surrounding communities, counties and the DOE (which has a fire station at the Hanford 300 Area located 2 miles northeast of the plant site). The DOE fire-fighting staff is well-trained in nuclear fire safety precautions and has available equipment for radioactive fire fighting. The Richland Fire Department receives annual training in radiological safety precautions from SPC personnel. AMENDMENT APPLICATION DATE: September 12,1994 PAGE NO.: WM sPC-ND.3330 947 (R-UO7/92)
Siernens Power Corporation - Nuclear Division EMF-2 7 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 j (U PART ll - SAFETY DEMONSTRATION REv. The plant site is fed water from the north Richland water grid through 10-inch diameter water pipes which enter the plant site from the north and south. The plant loop to the hydrants is an 8-inch diameter pipe. There are 13 fire hydrants on plant site (see Figure 11-10.18). There are Multipurpose ABC, Halon, Met-L-x, CO, BC Dry Chemical, Purple-K 2 Dry Chemical, and AFFF fire extinguishers provided throughout the facility at selected locations. These fire extinguishers are lnspected and tested in accordance with the applicable, preventive maintenance procedure. SPC's Plant Emergency Response Teams (PERT) receive annual training in or incipient fire-fighting techniques. The Richland Fire Department has the main responsibility for fighting fires on the plant site. 10.5.3.7 Responsibilities The Manager of Safety, Security, and Ucensing has the responsibility for inspecting and testing the plant fire extinguishers. The Manager of Plant Engineering has the responsibility for inspecting and testing the 9(V plant fire alarm system. 10.6 Criticality Accident Alarm System The criticality accident alarm system used in the SPC facilities employs neutron criticality detectors (NCDs), which are operated in two-out-of-n (where n = 3 to 6 per comparator panel) coincidence to minimize the possibility of spurious trips due to NCD malfunction or response to radiation other than neutrons characteristic of a criticality accident. The SPC criticality accident alarm system consists of NCDs, comparator units, associated control panels, an annunciator panel, and howlers. Each comparator unit is capable of monitoring the failure and trip signals of up to six NCDs. Each NCD consists of a BF or He tube (externally moderated) with associated pulse 3 3 amplifiers, trip circuits, performance audit circuits, and associated power supplies. Failure audit circuits check the operation of each BF tube, pulse amplifier, multivibrator, and low and high voltage supplies by requiring a min)imum background count rate of 100 counts per minute. This background count rate is provided by an internal radiation source in each BF tube. If the minimum background is not detected by the audit circuit, i 3 a failure (indicating malfunction) is signaled. Such failure signals do not interfere with operation of the criticality accident alarm system. n Because the trip circuitry is not audited, redundant trip circuits are provided in each NCD. Failure and trip signals from all NCDs are fed to comparator units. Redundant trip 6eplember 1Z,1W4 pmENm 10-:>4 AMENDMENT APPLCATON DATE: SPC-ND3330 947 (R-UO7/92)
SiernenS Power Corporation - Nuclear Division EMF-2 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 PART ll-SAFETY DEMONSTRATION REv. detection is provided in the comparator units. The comparator panels display all failure f and trip signals. When two or more NCDs connected to the same comparator unit are t tripped, the criticality accident alarm (howlers) is activated. i An annunciator unit monitors the status of the comparator units, and an audible alarm is actuated in the Central Guard Station in the event that the annunciator unit detects either i a failure or trip signal. j The minimum detectable criticality burst width for the SPC criticality accident alarm system is 50 microseconds. An overall system reliability test is conducted quarterly. The trip point of each NCD is set sufficiently high to minimize false alarms. The system is designed to trip the alarms within 0.5 seconds when exposed at 5% above the trip point. O O AMENDMENT APPLCATION DATE: DeplemDer 12, iih4 iG-55 PAGE No; SPC-ND.3330.947 (R-1/07/92)
i - Siernens Power Corporation - Nuclear Division eur.2 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 { j PART 11 - SAFETY DEMONSTRATION REv. l PART 11 - SAFETY DEMONSTRATION ' Table 11-10.1 Ammonia Recovery Major Equipment j 1. Strip Column 38' X 2' diameter l 22 tray i Outlet Demister 2. Feed Tank 11.5' X 8' diameter j 3,300 gallons j Agitated l 3. Scrubber _ Bottom - 10' X 18" diameter 8' packing j ) Top - 10' X 6" diameter 8' packing j 4. Distillate Tank 9' X 3' diameter j 400 gallons j 5. Ammonium Hydroxide Storage Tank 24' X 14' diameter i 28,000 l 6. Sodium Hydroxide Storage Tank 24' X 14' diameter i 28,000 gallons i 8 7. Cooling Tower 2.5 X 10 BTU /HR 8. Transformer 1000 kVA 9. Steam Boiler 945 Kw i l 1 i AMENDMENT APPLCATON DATE: DepI M I12,l m PAGE NO.: bbbe SPC-ND:3330 947 (R-U07/92) m h M-ei e 1'-*t-i e-e e-W <tew m e-ra. m -aw
f Siemens Power Corporation - Nuclear Division EMF-2 /l U' SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 PART 11 - SAFETY DEMONSTRATION REv. PART ll-SAFETY DEMONSTRATICN - Table 11-10.2 LUR Process Equipment TANKS T450 Precipitator #1 (6,000 gallon) T451 Precipitator #2 (6,000 gallon) T452 Precipitator Washer #1 (350 gallon) T453 Precipitator Washer #2 (350 gallon) T454 Filter Wash Tank #1 (70 gallon) T455 Filter Wash Tank #2 (70 gallon) T456 Precipitation Agent Mix Tank (50 gallon) T-660.1 Dissolver Vessel (100 gallon) T460.2 Dissolver Vessel (100 gallon) T-660.3 Dissolver Vessel (100 gallon) T-660.4 Dissolver Vessel (100 gallon) PUMPS P450 Precipitator #1 Slurry Pump P451 Precipitator #2 Slurry Pump P-652 Precipitator Washer Decant Pump P 653.1 Decant Product Pump (portable) 4 P-653.2 ANN Pump (portable) P454 Filter Wash Pump #1 P455 Filter Wash Pump #2 P456 Precipitation Agent Makeup Pump P457 Preciphation Decant Pump P458 Lagoon Waste Feed Pump P459 LUR Waste Transfer Pump FILTERS F-652 Washer Decant Finer (2540 micron) F-657.1 Precipitation Decant Filter (5-10 micron) F457.2 Precipitation Decant Filter (5-10 micron) F457.3 Precipitation Decant Filter (1-3 micron) F457.4 Precipitation Decant Fiber (13 micron) F458 Lagoon Feed Filter (200 mesh) AGITATORS A-650 Precipitator #1 Agitator A-651 Precipitator #2 Agitator A 652 Precipitator Washer #1 Agitator A453 Precipitator Washer #2 Agitator A-654 Filter Washer #1 Agitator A455 Filter Washer #2 Agitator A-656 Mix Tank Agitator \\ U l AMENDMENT APPLICATION DATE: beptemDer 12,1994 noENo.. W-66b SPC ND.3330.947 (R-it07/92)
SiemenS Power Corporation - Nuclear Division eur.2 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 PART 11 - SAFETY DEMONSTRATION REv. i LIST OF FIGURES i FIGURE ll-10.1 Plant Equipment Layout Site Plan ll-10.2 Plant Equipment Specialty Fuel Building ll-10.3 Plant Equipment Layout UO Building 2 ll-10.4 Plant Equipment Layout UO Building 2 ll-10.5 Plant Equipment Layout UO Building 2 11-1 0. 6 Plant Equipment Layout UO Building 2 11-1 0. 7 Plant Equipment Layout UO Building 2 11-1 0. 8 Plant Equipment Layout UO Building 2 ll-10.9 Plant Equipment Layout UO Building 2 ll-10.10 Plant Equipment Layout UO Building ll-10.11 Plant Equipment Layout UO Building ll-10.12 Plant Equipment Layout UO Building 11-10.13 Plant Equipment Layout UO Building O Y lll-10.15 Plant Equipment Layout UO Building l-10.14 2 Plant Equipment Layout UO Building 11-10.16 Plant Equipment Layout U Building 1 ll-10.17 Plant Equipment Layout EL Building ll 10.18 Horn Rapids Road Site Arrangement - Fire and Water Supply ll-10.18a Horn Rapids Road Site Arrangement - Sanitary Drain ll-10.19 Simplified Schematic HVAC System - SF Building SWUR Facility 11-10.20 Simplified Schematic HVAC System - SF Building Production Facility 11-10.21 Simplified Schematic HVAC System - Original UO Building g ll-10.22 Simplified Schematic HVAC System - UO Building Line 2 Conversion Area g 11-10.23 Simplified Schematic HVAC System - Main (South) Addition ll-10.24 Simplified Schem'atic HVAC System - U O Facility 3g 11-10.25 Simplified Schematic HVAC System - ELO Building 11-10.26 Simplified Schematic HVAC System - ELO Building Addition ll-10.27 Simplified Schematic HVAC System - Contaminated Clothing Laundry Facility ll-10.27a Simplified Schematic - SWUR Incinerator Shroud Cooling System ll-10.28 Plant Equipment Layout - ARF Building ll-10.29 Ammonia Recovery Waste Management Facility Engineering Flow Diagram ll-10.29a Flow Diagram Lagoon SA - Sewer lon Exchange Project ll-10.30 Plant Equipment Layout - LUR Facility ll-10.31 Engineering Flow Diagram - LUR Production Facility 11-10.32 Fire Alarm System i autwowtw7 amcarm oart: september 12,1994 PaGE NO.: lUDDC spc-ND200.947 (41/07/92) ] A
gyi.mo ame.se I e i 7 I e I s = _. _ _.. _. _. _.. _. _. _ _ -.. _. _.. _ _. _. _.,. I + 3 r+ i .l l ,. = ~ ~
- _. >
j .. s / / N q JcQ J t l = mr-
- AD gy
. -.,, g i i ...a a3; _L i x g / r __y.__.-.' \\.___." _.a_._._ yA=_. O lase.m,aset ag L aMKos.i t.00ne.., 66. C,soo F - I.,q]) M j = = I o L:-=- lt=: - m,. TJ - e; l 0 "= / .,a l. . "T-U kn,% ~ ~- u %..nd, l C
- c rg 1 f~~-(
~ i ~,- E i E"E'l y\\, I I .--l 5gp o V m i ,~. l.
- a..,.,. _d. [y]_
l,.,n.; i vaq.M, / - ,,, ~. o ..n_./.s , nom. 1 a 1 r -- N.g= g._3.- r p'eg O L._m_-._._. -+ _. ', f ]l
- 1 p -- 5'- ~Pg,
],'r_,'la i ('~ ' 77'. o al qa 3 ; -=q1a -.~.,
- 4... m 3 y
y-g._ ri, I _= g q .i c _ op.l 16 _..'r. 7,,. TL ;r.':r. I ne, .I ] Yi$ i I p 1l /'. ,q - 3 ~ ' ' ~ c J r i t.. -p_7 Cl D I 0 ll.dgI y ->n .,a L..J ,,o..
- a~
= > > l r-7 o p._f. mm..,u. b,n U.l.= W I Il ]l i F JTE am n (
- e. %
L..J,., a... l p' eng=i ,R m Lm-n,, e ,s,_:.=4.LQ. _. m _ _ _. J,m5 "*" 5 l gu, j / ,.._._a._._-__s m l.3, Ii t i i 3-. y, f- ~._._._._._- [ L_.__.=.=...:.._ .__.2.- =_._.__. __m A { 0 l 7 l 6 l
4 1 3 l 2 l t 7
- g;,5, wo 2
3 s .o ,3 i, o 2, n n r. n x v a a 3o 3 3, n 3 n H ,L n. 3 2 3 3 2 1 3 3 3 i EMF-2, Rev Page 10-56 September 12,1994 c ANSTEC APERTURE ~ + l CARD /r I.i Il U ruruv.m.isic s.im.cc.u - ll' '. Abc Available on -l l \\.svt. 4 r Aperture Card I i.i !i i 1 .l s _t lL., i o. i i l. ( rm1 -lg .._._._.__._.y. i i j l! c [_L oo j. j .l. i--- E j ! .,c p [ ;-- - -- r' !
- i.., =.
s.
- c. %
i i i! .l 1 g ;' la.v :';,.g l.I 11r-.1 ..I l \\ jy_ I,;;;.; I.l a,'a ;. %.e I.I ~ J ~ o C -(.L _,{ " ~ g;g,s 7 g [ _l"s"r - r j I .l g g. U- ...". > E. J 'g d -- 'l L_:,_ ~ - _J___._l l u Ir ; ^i ljl 4, l I r.
- ,n.,i i ruiun g'tr *#1 y"'$,'t 5'**
.l l - 'cm. r., r.,. - ", I ~ - .i i--- L.. g c i l l g i n ll 1 1 u-l .a. D. ,w l l l1j Figure 11-10.1 ~<. l {_. l
- **;'5 1
l n - ~. ~ 5"**** I j l, 'jy2
- l..,
a-j.l l i g. _. kf- - {,_ V~ ._._._.-._./L.- _J. -~ - - - ' C~~ _ I_._.J B
- :==.
N O9Z'dbO h O 1 _.1 Siemens Po-er Cort.orotion - ND j Advanced Nuclear F uels Corpor ation .. _.._. J PLANT 4.di k EQUIPMENT L AYOUT 4 4.: SITE PL AN 4 r_g wrw. ...n...o a i -... i i. ,u .,,. ~..... I. l 1 DS-JE - 182 - 94 o i-.{ ,... ~. - w l t l 3 l 2 l i _j
ww f 8 l 7 l 6 l H P-6*.B 4~ l ELECT [ O ' ' ' ' ' ma J F-6s7-3 % b ~ 1 1 I _.h 110.t c at =
- k.J I
j/< i 1,.eso ,u.t.a, m,c, 7 .%-es2 i .. s., I r Tk 653 -' Q r ~ D r.cs7-i r.csv.2 7 O p3,,c3 P 6,2 TK-Es2 intr.g"!n~ua P "" ] "fu,..ss_2 qc m s v ~2 A in.ss4 Tx-sss ty c,,, o( e...s. 0 TY ss6 (Avu Dkm O O CHEpiC AL SHED D D C~- C E c-T ADt t = DLSK B SHrii-I r mm hOhrf[R5 A p% **'* B l 2l 7 l 6 l [_ w
{ c l 3 1 2 l 1 H EMF-2, Rev Page 10-88 September 12,1994 BC3 ANSTEC -_b APERTURE CARD r..s. Also Available on Aperture Card i ocsx r O* tv 4no-4 [ASSOLvt DRUW$ ) O t,0- 2 1x460-3 E O! N auct C 70 f6 0,f 7 ' LI^I*" 3 [ h-D c Figure 11-10.30 0 g i l i. _:_i Sjement Power Corporotion - ND i Acivonce d Nucleor f uels Corporation . y,.
- 7. o..
~ w.ooi PLANT ygv-E ~ EOUIPMENT LAYOUT 4-2:+- LUR FACILITY A 5-JE-183-94 bh-i i 1 3 1 2 1 _J
o th 8 rj$'i az .J d th 2;h 0" E, E c', i d f be 5 ~ 4 o,. et =. a~r1 g n .y i s W .w s u,n _1}3 rl.:.,l l n?- p t-r i l lU l ll ),.; i R m ! I-u h = ++ lI f, - I. . yj -)i t.
- =
i l. ,pb cu il j,o i !Ii {3 W i i h.h tr I i,.,1 v i._._.J J / .m i,., J-n it l /,.. r.- Iji m i
n!
1 m L .U;{ _i J_lt n'. -! 1._.____._.J_._ .J._3-fE4 i L._._._ i e -- n ,<1 ,lo i I i c l t g l-r*- {2' 1
- a
'. _. _. l 9
- .m l
[ l l**i i Il' 1 '* 'i j i g I L l.. L~ L-] q. R !l l l! j i !!! jj*. m '- a . es l !..] L..-._J. li i
- I i
s m j I Zl l \\ l q }l.;y. -.. l~(i n ci .:p r-,. 1 ( o .- ) l E !. L-- 5 inli! ~ 3 {I%, I, , !f j i i P,^ i 13: T l l .L u 3; I s d L:- J et + l - p- , rj., I f t s j( ijl i" L! W,. l ,,,, g t.u ab a - ri u-j j i. g i l.'b SI g,, M;-t-**! l 1 _J%_n D,f, ~' i i r i i a a. 'J.s{ ; c. p Q q l 3, $1 g-- g ? rr.i ~ s c., .II J 't 'I i l g g v1 e m, i c f.i). J"~-*~ '~~" D]r, c W!b}; ]' l', I$ jd! I > 5 b; 1 i i s i.
- y i
a LA ' .cA.J y!; c l k N' 'y,' i (?giil_ ( i e l , x;; sp -a >- pj i Lgr i %_ r- . _r 9 k J- - -. i ! i. i i i i k i l \\p \\ .3 g___ g r -,ry - r - m i. p-,lDQ, i ,r ' !l U,Ig'n ' hll.,C, i\\. M g) l,
- r []
y' 6l. -.)f / L I 7' i t i i, & I di. 4 El ini. -.3 ll }~ .P 1 ll 1 i ss..eyu_ - 00TL M!i / u '.i il l 6 h,7d l, l, 1 a s.t.c,r.= _a e,, [+- 3, n, s ',., - - - i y 6rt J, g,
- qn.
i m.1 _ r\\ l + i;} _i r-t (./ i = o i o j = l 3 i o
~ i o o v x
- .y 11 O> IO~
t'! a t: s ::s e 9 l.'3 b ~ E 9 di[t &(e-N? a n 2% w O ft [ t k h.I ". 6 & 1 _ 't .a. t;*e !!c r u 's $ $ !.J gi j te- " E,iii! pp III.. - u <. qm'g il r:,p't.
- ,,s emy N
I l 4 me e i I 1,g 3F. 3 L cm a[ I j, I[ .,*ge a w t 5 pl} s: D!f ' I Q! ij -+ f f ,f %s ,r a y H t l' j :-i}: 5 l-. H i e- !; I( ! x != t, d 1 'a,- [ l,/s)CC n i h g\\ =3 / l li L\\ is i }l ' V IV O.s 1:D i H-d "i ll! I NN _ il "rf NT6 ^ ~~ g t ) e;: fi I s)g!ge:o!C! != 1 ye 4; .m 2;w b 0 [ I,.----=-- .glg. g;, x f J eW 5 7 1; 3:5 N
- E y
t = a i o i o
Siemens Power Corporation - Nuclear Division EM F-2 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 PART ll - SAFETY DEMONSTRATION
- ggy, CHAPTER 12 RADIATION PROTECTION 12.1 Proaram SPC's goal is to maintain radiation exposure as far below the limits specified in 10 CFR 20 as is reasonably achievable by establishing and maintaining a radiation protection program which includes:
1. Written policy guides, standards and procedures; 2. Using, to the extent considered practical by SPC, procedures and engineering controls based on sound radiation principles to achieve occupational doses and doses to members of the public that are ALARA; 3. Personnel training; 4. Access controls; 5. Periodic reviews of the radiation protection program content and Os implementation, and reviews of exposure records; 6. Radioactive effluent and waste control; 7. Use of protective clothing and personal protective equipment (including respiratory protection; 8. Inspections and audits; and 9. Recordkeeping. 12.2 Postina and Labelina Radioactive material, airborne radioactivity, radiation, and high radiation areas as defined by 10 CFR 20 are Mentified, and their boundaries are visibly marked. Signs denoting these areas are placed su inat at least one sign is visible from any approach. 12.3 Internal and External Radiation - Personnel Monitorina A personnel monitoring program has been established. The program consists of the following: 1. Establishment of internal and external exposure limits, including internal /Q Company guidelines which are generally lower than NRC limits; Q AMENDMENT APPLCATON DATE: beptemler 12,1994 PAGE NO.: 12N SPC-ND.3330.947 (R-907/92)
i Siemens Power Corporation - Nuclear Division sup.2 l SPECIAL NUCLEAR MATERIAL LICENSE NO._SNM-1227, NRC DOCKET NO. 70-1257 n PART ll-SAFETY DEMONSTRATION
- ggy, 2.
Provision of dosimeters to measure personnel external exposure for personnel likely to exceed 10 percent of the NRC's external radiation limits, 3. Provision of a dose tracking system and bloassay program to estimate - internal exposure for personnel likely to exceed 10 percent of the NRC's internal radiation limits
- 4.
Maintenance and analysis of exposure records. The occupational exposure received by SPC employees and visitors shall not exceed the NRC limit. Company guidelines (in Chapter 2 of EMF-30,"Siemens Power Corporation Safety Manual") have been established at or below the NRC limits in order to prevent i overexposure. All employees are advised of the National Council of Radiation Protection and Measurements recommendation to keep radiation exposure to an embryo or fetus to the very lowest practicable level during the entire gestation period and of the NRC's dose limit for a fetus. SPC determines the current calendar year dose for all personnel who require monitoring ) O and attempts to obtain records of cumulative exposure. I All personnel likely to exceed 10 percent of the NRC's external radiation limit wear NVLAP j accredited dosimetry, with the exception of extremity monitoring. Finger rings are used 1 when extremity (finger) monitoring is required. Direct-reading dosimeters may be used when timely information is needed. Self-reading pocket dosimete_rs (pencils) may be used with x-ray operations or where external radiation exposures have the potential for exceeding Company guides in a matter of days, j i Radiation exposure dosimeters are analyzed quarterly for personnel whose external' j exposure is likely to exceed 500 mrem (deep dose) per calendar year. Finger rings are j processed quarterly, pencils read weekly, and readings are documented. TLD's assigned to personnel who are not likely to exceed 10% of the annual NRC limits are analyzed annually. All personnel likely to exceed 10 percent of the NRC's internal radiation limit participate in the internal dose tracking system and bloassay programs. Internal dose may be estimated from airborne, stay time, and respiratory protection records or from bioassay results or from a combination thereof. Internal and external exposures are appropriately combined for personnel who require both internal and external dose monitoring. Internal, external, and combined dosimetry results are evaluated by the Health Physics Component to determine that the exposures of personnel are within limits. Exposure data are also reviewed by the ALARA Committee AMENDMENT APPLICATION DATE: september 12,1W4 PAGE NO.: 12-2 SPC-ND:3330.947 (R UO7/92) c,-- + n 1
i Siemens Power Corporation - Nuclear Division EMF-2 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 I l PART ll - SAFETY DEMONSTRATION REv. j to determine the effectiveness of programs to maintain exposure as low as reasonably achievable. 12.4 Radiation Surveys The routine radiation survey program consists of the following: 1. A detailed survey of radiation levels is performed during the initial startup of each operation and storage area which involve radioactive material that is likely to significantly change the existing radiation levels. These surveys l provide the necessary information for determination if the area must be l classified as a radiation area and the necessary boundaries of such an j area. They also identify areas where ALARA modifications may be l necessary. 2. Monthly radiation surveys are performed in all areas where radioactive { materials are processed or stored and where personnel have access. These surveys identify areas where the radiation status has changed and f potential areas for ALARA modifications. l 3. Radiation surveys are performed on all incoming and outgoing shipments f of radioactive materials to assure that such shipments conform to j applicable regulations of the NRC, the USDOT, the U.S. Postal Service, the State of Washington Utilities and Transportation Commission, and the i international Atomic Energy Agency. Procedures are in place to assure j that necessary radiation surveys are performed on all receipts and i shipments of radioactive material. l 4 When proposed test or nonroutine production work involves radioactive material, radiation i surveys are performed prior to the start of such work both to confirm the levels of radiation present and to permit evaluation of methods to reduce exposure during the work. Surveys are also performed during the work to confirm that radiation levels have l not increased significantly. I 12.5 Radiation Safety Trainina t All new employees and contractors receive Radiologicaly Safety Orientation which is part j of a New Employee Orientation Class. l l All new employees and contractors, who will be working wiih SNM or enter into restricted, areas of the plant, receive an additional Radiation Workers Training Class. An examination is given at the end of the class to test the attendees
- comprehension of the!
j class material and content. AMENDMENT APPUCATION DATE: bWPlWIIIWW3 be 5559 PAGE NO.: IN~ 2 i SPC-ND3330.947 (R-U07/92) 1 l . ~ a
Siemen.c Power Corporation - Nuclear Division eug.2 { SPECIA. 6 JCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 PART 11 - SAFETY DEMONSTRATION
- ggy, Annually, all employees and contractors, who work with SNM or enter into restricted areas of the plant, receive an Annual Refresher Radiation Workers Training Class. An examination is also given at the end of this class to test the attendees' comprehension of the class material and content.
12.6 Reports and Records t AI! n.;' ". d records of the Radiation Protection Program required in Chapter 2 are maintan.. a Company files in accordance with 10 CFR 20 Subpart L Such reports and records are maintained for a minimum period of 5 years unless longer retention periods are specified. i Personnel who require monitoring for either internal or external radiation exposure, will { be provided a report M nir exposure. 12.7 Instruments i The criteria for selecting radiation measurement instruments for performing radiation and i g contamination surveys, sampling airborne radioactivity, and monitoring area radiation are j described in Table 1-3.2. The Manager of Safety, Security, and Ucensing is responsible for maintaining adequate quantities of radia" leasurement instruments and related equipment. The Manager of Plant Engineering is responsible for the maintenance and calibration of radiation safety instruments and equipment. The following general requirements apply to all such equipment and instruments: 1. All radiation detection instruments are inspected (and repaired when i necessary) and calibrated at least semiannually or tagged out; 2. Instruments are calibrated following any maintenance deemed likely to affect operation before they are put back into routine service; 3. Each on-line radiation detection instrument is checked for proper operation by Health and Safety Technicians or by electronic surveillance daily (Monday through Friday for a normal work week). When daily checks are j performed in a manner which qualifies as calibration, separate semi-annual calibrations are not required; I 4. Portable survey instruments are source-checked each shift they are used; AutwoutwT AmcAToN oATE: September 12,199~4 PAGE NOa I2M SPC-ND 3330 947 (R-1/07/92)
Siemens Power Corporation - Nuclear Division EMF-2 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 l PART 11 - SAFETY DEMONSTRATION REv. 5. AC-operated personnel contamination survey instruments are provided with individual check sources to allow personnel to source-check the instruments at random intervals; 6. Calibration sources are traceable to the National Institute of Standards Technology (NIST); and 7. Dose rate instruments are inspected (and repaired if necessary) and calibrated at least quarterly or tagged out. 12.8 Protective Clothina The types of protective clothing required for operating personnel in normal, maintenance and accident conditions are dependent upon the work assignments and the levels and types of contamination present and are specified in the applicable SPC Radiation Work l Procedures which govern those conditions. i For inspection activities, lab coats and cotton or rubber shoe covers may suffice, perhaps supplemented with cotton gloves and surgeon's cap. Most normal operating and maintenance assignments will require rubber shoe covers, full coveralls, and surgical-type rubber gloves or light-weight PVC gloves. For more unusual type operations or maintenance full cotton hoods and cotton boots may be required. Industrial safety equipment is also available (face shields, goggles, and acid suits). In addition, both half-i mask and full-face negative pressure respirators, as well as full-face positive pressure respirators, are available. The protective clothing requirements are specified in the applicable Radiation Work Procedures. 12.9 Administrative Control Levels (includina Effluent Control) The action levels, alarm set points, frequency of measurements and actions to be taken for the various radiation protection monitoring programs are described below. 12.9.1 Occupational Exposure (Internal and External) The bloassay program, including frequency of measurements for determining internal exposure, is described in Section 12.12. The routine urinalysis and lung count results are i reviewed by the Health Physics Component to determine any unusual trends or potential exposures. If the internal exposure of an individual exceeds action levels or appears uncertain, additional analyses and/or removal from further exposure are considered. The external exposure personnel monitoring program, including frequency of measurements, is described in Section 12.3. The quarterly dosimeter results are reviewed by Radiological Safety personnel to determine any unusual trends or exposures. If the l i 2 wououtuumcarom oart: SeptemtsrT21994 PAGE NO.: SPC-ND 3330 947 (R-UO7/9h
Siemens Power Corporation - Nuclear Division eup.2 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 L PART ll - SAFETY DEMONSTRATION
- ggy, extemal exposure status of an individual becomes uncertain, the individual is removed from further exposure and the Supervisor, Radiological Safety will determine the exposure status and advise supervision of any special controls or restrictions to be applied.
1 12.9.2 Airborne Effluents The gaseous effluent controls are described in Chapter 5. The action levels are listed in Table I-5.1. Samples are counted by a single channel analyzer utilizing a surface barrier detector. The lower level of detection varies with the stack being sampled and ranges from 1.4 x 10-15 pCi/ml to 3.5 x 10-14 for a 7-day sampling period. The lower level of detection for mixed fission and activation products is 7.2 x 10'14 Ci/ml for a 7-day sampling period. The action levels specified in Table I-5.1 include shutdown requirements to reduce emissions. In the event that.the calculated dose to any member of the public in any consecutive 12-month period could be anticipated to exceed the limits specified in 40 CFR 190.10, the Company takes immediate steps to reduce emissions to maintain compliance with 40 CFR 190.10. O Q1 The facility air sampling program is discussed in Section 12.13. 12.9.3 Liauld Activity (Effluent Monitors) Uguid effluent leaves the site at the south boundary and is discharged to the municipal sewer. This effluent is sampled continuously and a composite sample is analyzed each workday (Monday through Friday for a normal work week) for uranium and regulated chemicals. The amount of uranium is determined fluorometrically with a minimum detection level of 0.1 pgU/ml (0.1 ppm or 2.1 x 10 Ci/ml). Action levels are to investigate any sample result greater than 0.1 ppm and to shut down the processes which could discharge uranium upon a confirmed sample result greater than 1.0 ppm. Trends of the average concentration of uranium and the total uranium discharged are reviewed by the ALARA Committee. There is a potential for release of uranium to the groundwater by leakage from the process chemical waste storage lagoon system. The inter-liner leak detection system and associated action levels are described in Chapter 5. The groundwater test wells and sampling programs are also described in Chapter 5. Action levels for resampling and investigating are based on certain indicator chemicals, although all data are reviewed and any significant change investigated. ( pO September 12, T994 PAGE NO.: 12C autwoutwTa m catomoart: SPC-ND:3330 947 (R-1/07/92)
Siemens Power Corporation - Nuclear Division EMF-2 (') SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 V PART ll - SAFETY DEMONSTRATION
- ggy, 12.9.4 Fenceline TLD Prooram SPC maintains a fenceline TLD system. Comparisons with a Battelle Pacific Northwest Laboratory network of environment TLDs in near and distant areas are used to demonstrate that SPC's external radiation dose contribution to the public is insignificant.
12.10 Respiratory Protection The primary objectives of SPC's respiratory protection program are to limit the inhalation of airborne radioactive materials and harmful air contaminants, to comply with permissible exposure limits, and to protect employees in oxygen-deficient atmospheres. These objectives are normally accomplished by the application of engineered controls, including process, containment and ventilation equipment. When such controls are not feasible or are not effective to the extent necessary, the use of respiratory protective equipment may be appropriate. In general, however, the use of respirators is less desirable in providing respiratory protection than the use of engineered controls. The use of respiratory 1 protective equipment is subject to the following considerations regarding circumstances under which respiratory protection may be needed: Q Routine operations are planned activities that are generally repetitive and occur frequently. For such operations, potential sources of airborne radioactive materials and other harmful air contaminants or oxygen-deficient atmospheres shall be identified so that respiratory protection may be accomplished by process, containment, and ventilation measures and by pre-planning of work. The use of respirators as a substitute for practicable engineered controls in routine operations is inappropriate. Respirators may be considered for use, however, while engineering controls are being instituted or evaluated. Nonroutine operations are activities that are either non-repetitive or else occur so infrequently that adequate limitation of exposures by engineering controls is I impractical. To the extent that process, containment and ventilation controls are not reasonably feasible in nonroutine operations, the use of respirators is appropriate. Emeroencies are unplanned events characterized by risks sufficient to require immediate action to avoid or mitigate an abrupt or rapidly deteriorating situation. i Although emergencies are unplanned, preparations must be made for coping with potential emergencies. SPC's preparations include a program for providing respiratory protection for use in emergencies that are likely to entail respiratory hazards. The advance preparations for a particular potential emergency depend i on both its possibla consequences and the probability of its occurrence (see EMF-32, Emergency Plan, for further details on plans for dealing with O emergencies). l ' b awarm appuccow out: September 12,1994 PAGE NO2 I 2-7-~ SPc ND 3330 947 (R UO7/92)
i Siemens Power Corporation - Nuclear Division sup.2 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 PART 11 - SAFETY DEMONSTRATION REv. ? Most operations can be readily categorized as routine, nonroutine or emergency. In l dealing with situations which are not easy to categorize, sound judgment must be exercised in using engineered controls where feasible and by avoiding unwarranted use of respirators. ] The period of time respirators are worn continuously and the overall durations of use are each kept to a practical minimum. It is necessary to allow respirator users adequate relief 1 from wearing respirators at reasonable intervals and to limit total time of use. However, i it is difficult to realistically assign specific time limits on respirator use because of wide variations in job requirements and in the physical capacities and psychological attributes of individuals. Such factors must be taken into account in establishing a respirator j program. Provision is made for the respirator users to leave respirator required areas for relief in case of equipment malfunction, undue physical or psychological distress, procedural or communication failure, significant deterioration of operational conditions, l or any other condition that might require such relief. l SPC uses respiratory protection equipment that is or has been certified by NIOSH/MSHA. i Respirator users are medically approved to wear respirators, are appropriately trained and 1 are fit tested in the selected respirator that the wearer uses. The protection factor for the O respirator chosen is greater than the expected concentration divided by the effective DAC. 1 12.11 Occupational Exposure Analysis Occupational exposure analyses have been performed and documented in the form of an annual ALARA Report. The ALARA Report for 1991, is appended to this Chapter as Appendix A. It contains a history of external and internal exposure data in Tables ll and j V. These tables include estimates of dose based on airborne contamination, in-vivo (lung count) data and urinalysis. Trend evaluation is discussed in Section 3.0. Measures taken to furthe' minimize personnel dose are presented in Section 4.0. j r 1 12.12 Measures Taken to implement ALARA SPC has committed to maintaining a functioning ALARA Committee as a subgroup of the Health and Safety Council. The membership and activities of this Committee are outlined in Chapter 2. Company Policy in this area is outlined in Chapter 3. 12.13 Bioassey Program The bioassay program established by SPC is conducted to confirm the results of radioactive material contamination control and personnel protection programs. It also may be used to estimate internal exposures due to internally deposited radioactive material and/or intakes of soluble uranium for chemical toxicity assessments. O aucNoutNTam cuoNoaTE: September 12,1994 PAGE NO,: 12 C SPC-ND:3330 947 (R-907/92)
4 Siemens Power Corporation - Nuclear Division eur.2 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 PART 11 - SAFETY DEMONSTRATION REv. The frequencies and types of measurements are established on the basis of the exposure potential of the individual's work assignment and the physical and biological properties i of the radioactive material with which the individual works. l Radiation exposure due to internally deposited radioactive materials is kept as low as is reasonably achievable primarily by the implementation of engineered controls. Administrative controls and precautionary procedures are also employed to complement the engineering controls. Aspects of the bloassay program are as follows: 1 1. Employees normally working in contaminated areas containing transportable uranium compounds submit urine specimens every 28 days for uranium analysis. Scheduling is staggered to help provide continuous assurance of airborne contamination control. i 2. Routine lung counting frequencies are established for all operators and maintenance personnel assigned to work in areas where nontransportable compounds are processed. The minimum frequency for lung counts is N semiannual for personnel normally working in areas exceeding 10 percent of DAC the previous quarter. i 3. Bioassay investigations are undertaken when action levels are exceeded and elevated results are confirmed. The investigations are documented and discussed at Al. ARA and Health and Safety Council meetings. When appropriate, corrective actions are executed. 4. Non-routine bioassays, consisting of in vivo (lung) counts, urine samples, and/or fecal samples are requested when individuals are suspected of acute exposures exceeding 200 DAC-hours. The Health Physics Component evaluates the resu!ts of such assays. The exposures from these bioassays may be substituted for those from airborne measurements. In evaluating the bioassay results, the Health Physics Component uses internationally recognized models and may make use of site specific data. 12.14 Air Samplina and internal Exposure Prooram The air in all general areas where uncontained radioactive materials are handled, processed or are likely to exist is regularly sampled and the samples are analyzed for radioactivity. The air sampling system consists of a combination of equipment and instruments. Field-run lines (flexible or hard-piped) connect individual air samplers to PAGE NO.: lI-g AMENDMENT APPLICATION DATE: September 12,1994 SPC-ND:3330 947 (R 1/07/92) l i
Siemens Power Corporation - Nuclear Division eur.2 (9 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 ig PART 11 - SAFETY DEMONSTRATION
- ggy, overhead vacuum lines such that the samplers may be moved about and located at any desirable point without causing obstructions or creating industrial safety problems. The locations of air samplers are determined, in part, by the need to measure potential airborne contamination, representative airborne contamination at workstations, and background airborne contamination.
The frequency of exchanging and analyzing filter papers of the workstation sampling units is based on historical experience. Sampler filters may be exchanged and analyzed more frequently in the event of suspected elevated airborne contamination; to assess the effectiveness of radioactive material containment / confinement following equipment modification or maintenance; or to assess the air quality of sequential shift operations. Permanently-mounted air sampling equipment used to determine worker's breathing zone concentrations is evaluated for representativeness following significant process or equipment changes; at least every 12 months for work stations which averaged greater than 10% of DAC the previous year; and at least every 24 months for remaining work
- stations, o
Specialized air sampling or monitoring equipment such as continuous air monitors, portable high volume air samplers, and lapel air samplers, is available to supplement the normal air sampling system and for use in special studies. SPC pursues maintaining radiation exposures as far below the normal limits specified in 10 CFR 20.1201 and 1202 as reasonably achievable. Respiratory protective equipment is required for entry into areas where airborne radioactive materials are known to exist in excess of the occupational concentration stated in 10 CFR 20. 12.14.1 Particle Size Distribution Effects SPC may elect to alter Derived Air Concentrations (DACs) and Annual Limits of Intake (Alls) based on the results of particle size distribution measurements. The method of obtaining such measurements and applying the results is described below. Particle size distribution measurements will be taken using Andersen 1 ACFM Non-viable Ambient Particle Sizing Samplers with a pre-separator and 8 stages (Model #20-830). The particle size range for each stage in micrometers is: stage 0: 9-10; stage 1: 5.8-9; stage 2: 4.7-5.8; stage 3: 3.3-4.7, stage 4: 2.1-3.3, stage 5: 1.1-2.1; stage 6: 0.7-1.1; and stage 7: 0.4-0.7. A backup filter is for size range 0-0.4 microns. (Glass fiber filters have efficiencies of 0.997.) o a m a t e a m c u ou o ut: september 12,1994 PAGE NO.: N*10-l SPO-ND.3330 9/7 tR-UOM12)
i Siemens Power Corporation - Nuclear Division eup.2 i (] SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 V PART ll - SAFETY DEMONSTRATION
- ngy, For multimodal distributions, the method of analysis consists of estimating the fractional activity, the geometric mean, and geometric standard deviation of each subdistribution.
The predicted total distribution is compared with the measured distribution. The object is to minimize the sum of the squares of differences for each stage between the measured and estimated distributions. As an indication of fit, a chi-square statistic will be determined assuming n-2 degrees of freedom, where n is the number of stages. The chi-square is a surrogate statistic used to determine the goodness-of-fit. When a statistically good fit has been achieved, the fractional activities and AMADs will be used to determine the DACs, etc. in accordance with Appendix B to this chapter. The level of confidence required to demonstrate goodness-of-fit will be p greater than or equal to 0.8, where p is the probability of obtaining a value equal to or less than the chi-square statistic when the hypothesized distribution is true. If this level cannot be achieved, the data will be discarded and additional data taken. However,if additional data can not be fitted,i.e. p is also less than 0.8 then the Health Physics Component will apply a conservative analysis and will document the analysis. At least three particle size measurements will be taken for each grouping of locations. If SPC chooses to adjust DACs and ALis by particle size, particle size analysis will be O performed at least semi-annually in each group of locations for which particle size credit V is taken. After one year, the Health Physics Component may relax the frequency to once per calendar year if data for a group of locations does not differ significantly from previous measurements. Particle size will be reassessed following significant process changes deemed likely to change the particle size distribution. Using the results of particle size measurements and knowledge of the process, the Health Physics Component will decide whether specific operations or specific locations can be grouped together for characterization purposes. Particle size distributions for maintenance operations may be different than for routine operations. There is a practical difficulty in obtaining particle size data for maintenance operations; namely, several days to weeks of sampling time are generally required and maintenance operations seldom last long enough. In the absence of particle size data during such operations it is assumed particle size data during normal operations would still be the best data to use, if particle size data is available, and if maintenance operations yield significantly lower particle size distributions than for routine work, then for prolonged operations, i.e. longer than 24 hours, the DACs will be modified for maintenance operations. Should assignments be performed away from normal working areas, the Health Physics Component will decide whether to take credit for particle size. If credit is taken, the Health Physics Component will use conservative judgment is assigning a DAC adjustment factor and will document the analysis, i l O V wtam amcatou care September 2,1994 pace No.: 1W SPC-ND:3330 947 (R-1/07/92)
Siemens Power Corporation - Nuclear Division eur.2 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 PART 11 - SAFETY DEMONSTRATION ngy. 12.15 Surface Contamination Radioactive materials are contained and/or confined during processing, transfer and storage to maintain the intake of such materials by personnel as low as reasonably achievable. Operations involving readily dispersible forms of radioactive materials are accomplished within enclosures such as process equipment, glove boxes, glove port hoods, laboratory-type hoods, etc., which are exhausted to facility exhaust air systems. 12.15.1 Facility Surveys A detailed survey of contamination levels is performed for each new operation involving radioactive material and frequently thereafter, for a period of time dependent on demonstrated operational controls as reflected in survey results. Operations involved in the production of fuel are repetitive in nature, equipment and systems employed in development activities are normally similar to those used in the production facilities, and containment and confinement principles are employed consistently throughout SPC's facilities, thus making it possible to effectively establish routine and repetitive contamination surveys. The frequencies of routine surveys are determined by a ( combination of professional judgment and experience and are periodically reviewed by ( the Radiological Safety Component. In general, reduced frequencies are permitted when the stability of an operation, as demonstrated by the consistency of survey results and the relationship between observed values and operational controls, is established. Such reduced survey frequencies are approved by the Health Physics Component. The following frequency schedule is applied to the facility contamination survey program: 1. Contaminated radioactive materials areas - weekly 2. Noncontaminated radioactive materials areas - monthly 3. Intermediate areas - daily 4. Lunchrooms adjacent to radioactive materials / radiation areas - daily i Operational controls are considered adequate for uranium operations when the following conditions prevail: 1. The is no visible or smearable contamination in excess of 10,000 dpm/100cm2 on exposed or non-hooded surfaces in the process areas (not applicable to nonroutine tasks being conducted under special controls). 2. There is no smearable contamination on intermediate area floors greater than 500 dpm alpha /100 cm2 autwoucui amcatow o=TE: September 12,1994 paas wo.: 1242 - j SPC-ND.3330.947 (R-1/07/92) l
_. _ = i Siemens Power Corporation - Nuclear Division sup.2 L SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 l PART ll-SAFETY DEMONSTRATION
- ggy, l
l 3. There is no smearable contamination in clean areas greater than 200 dpm i alpha /100 cm8 I L Cleanup of uranium contamination in excess of the levels specified in conditions 1-3 above is initiated during the shift detected. 12.15.2 Release of Personnel. Materials. Ecutoment. Facilities and Shloments L Contamination surveys are performed on all personnel leaving contaminated areas, on all materials, equipment and facilities to be released from radiation protection requirements, and on allincoming and outgoing shipments of radioactive materials. l 1. Personnel. Contamination surveys a'e conducted according to the following schedule: a. All persons leaving contaminated areas are required to survey themselves for contamination with survey instruments. b. Personnel are not released to eat or leave the respective facility if { their personal clothing is contaminated in excess of the following limits, except with the approval of the Supervisor, Radiological Safety and the respective facility manager: Smearable and Fixed Uranium: 200N dpm alpha /100 cm*. c. For routine release, personnel skin surface contamination shall not exceed 200 dpm alpha /100 cm'. j 2. Materials. Eauipment and Facilities. Contamination surveys are performed by Health and Safety Technicians on all materials and equipment removed from contaminated areas, and on areas or facilities to be released from radiation protection requirements. Decontamination of facilities and equipment prior to release for unrestricted use or termination of license is in accordance with levels established in Chapter 3. 3. Shloments. Shipments of radioactive materials arriving at SPC are surveyed if required by 10 CFR 20.1906. All outgoing shipments of radioactive materials are packaged and surveyed in accordance with 10 CFR 71 and 49 CFR 173.443. 4 1 200 dpm (alpha) per 100 cm8 represents the practical lower detection level for most direct-reading contamination survey instruments. September f2,1994 PAGE NO.: 1M3-AMENDMENT APPLICATON DATE: l SPC-ND.3330.947 (R-14742) .. ~
Siemens Power Corporation - Nuclear Division aus.2 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 PART 11 - SAFETY DEMONSTRATION
- ggy, t
APPENDIX A CHAPTER 12 ALARA REPORT L i l O I 1 l l i O =cwoutwr wucatom ars: 5eptember i2, isve p.c;s uo.: SPC ND:3330 947 (R-197S2) l i
Siemens Power Corporation - Nuclear Division EMF-2 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 PART ll - SAFETY DEMONSTRATION REv. APPENDIX B i CHAPTER 12 ADJUSTMENT OF DAC/ALI BASED ON PARTICLE SIZE MEASUREMENTS l SPC may alter the DACs (Derived Air Concentrations) and ALis (Annual Limits of Intake) for class W and class Y compounds by taking credit for particle size distributions. Adjustments will be made according to the following formulas: AU (AMAQ, Ha (1 p), DAC (AMAQ AU (1 p) H (AMAQ DAC (1 p) i and i Hw(AMAQ DJAMAD D,(AMAQ DlAMAQ O. e N,o( 1 p) Dy1p) D,,( 1 p) D/1 p) Where H () is the committed dose equivalent; fnp' I, and f, are the relative i tb o g fractions of H due to deposition in the nasal passage, the trachea and bronchial g tree, and the pulmonary regions, respectively; and Dnp, D, and D, are the tb o deposition probabilities in the nasal passage, the trachea and bronchial tree, and t the pulmonary regions, respectively. When applied to class Y uranium compounds, such as UO and U O, and class 2 3g W uranium compounds, f =0=f and f = 1 according to ICRP 30. np tb p Therefore the following equation will be used: H,o(AMAD) DlAMAD) l H o( 1 P) U/1 P) ri If the particle size distribution is multimodal, the following equation will be used: h o seP ember 12, isse pacc uo.: i autwoucur apptcaron oars: 5%ND 3330.947 (R-1/07/92)
4 F Siemens Power Corporation - Nuclear Division sup.2 l SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 PART 11 - SAFETY DEMONSTRATION [
- ggy, APPENDIX B (Cont.)
H (AMAD) s. D,(AMAD(d,)) a. D,(AMAD(d,)) D,(AMAD(d,)) w =i +a +... + a,,. H ( 1 p) D,( 1 p) D,( 1 p) D,( 1 p) w where a; is the fraction of the activity associated with the distribution having AMAD(d). 3 When multiple measurements are taken for a specific location or group of samples, the grand average H (AMAD) will be determined as follows: g H (AMAD) D,(AMAD(d,,)) # ' D,(AMAD(d,,)) D,(AMAD(d,n)) \\ w ^ "~ ~ H,0( 1 p) D,( 1 p) D,( 1 p) D,( 1 p) r I %./ D,(AMAD(d,,)) a,. D,(AMAD(d )) D,(AMAD(d,m)) ag + a,,. +g +... + a,,,. s D,( 1 p) D,( 1 p) D,( 1 p) +... D,(AMAD(d )) o. D,(AMAD(d)) D,(AMAD(d)) p +a,. g +m g. j 0(1 ) 0(1 ) 0,( 1 ) p p where is the fraction of activity associated with the distribution having k AMAD(d ) of the J-th measurement. Where 3 a, = 1 and No. of measurements; a.g., an + a,, + a, a p s =3 + a,, + ast + a, + ass 3 C =catut amcarou cars: 5eptember 12, is94 PAGE NO.: SPC ND.3330 947 (R-1/07/92)
i Siemens Power Corporation - Nuclear Division EMF-2 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 PART ll - SAFEW DEMONSTRATION
- ggy, CHAPTER 13 ENVIRONMENTAL SAFETY - RADIOLOGICAL AND NONRADIOLOGICAL i
The SPC environmental surveillance program consists of periodic collection and analysis of samples from the plant gaseous and liquid effluent streams (including the leak detection system for the liquid chemical waste storage lagoons) and selected elements of the local environs. Information on environmental monitoring and releases is provided by the supplement to Applicant's Environmental Report EMF-14. 13.1 External Radiation The background radiation level in this part of Washington State is approximately 80 l mrem /yr. There are no plant radiation sources which increase external exposure at the plant boundary above background. m 'i 13.2 Gaseous Effluents Continuous isokinetic sampling is provided for each exhaust stack with the samples collected and measured at least weekly. The samplers are located downstream of the final HEPA filters. The total uranium discharged out of all stacks has averaged less than 25 microcuries per year during the last five years. This amount of uranium (less than 16 grams / year) has had essentially no impact on the cumulative off-site exposure due to uranium fuel cycle operations. Soil samples off-site have shown no increasing trend in uranium concentration and have remained well within acceptable levels. Using a worst case CHl/O value of 0.114E-04 with ICRP 30 methodology, a committed dose equivalent of approximately 21 micro rem may be estimated from gaseous discharges. The Fuel Services Building, which houses equipment contaminated with fission products from irradiated fuel exams, etc. at reactors, is regulated under the State of Washington Radioactive Materials Ucense WN-1062-1. SPC employs the same isokinetic sampling and HEPA filtering systems in this building as in its other buildings in which uncontaminated radioactive materials are handled or processed. 13.3 Llauld Effluents All liquid effluents which leave plant boundaries are joined together on plant, proportionally sampled and pumped into the City of Richland sewer system. These effluents consist of three types of liquid waste which are blended into a single sewer flow-fg sanitary wastes, noncontact cooling water, and liquid chemical waste. t l V autwowtwr amcarou oaTE: September 12,1994 13-1 i "E " " SPC-ND:3330 947 (R-907/92) i 1
SiemenS Power Corporation - Nuclear Division EMF-2 (d' SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO i PART 11 - SAFETY DEMONSTRATION
- ggy, The discharge of nonradioactive chemicals in liquid waste is regulated by Waste Discharge Permit No. ST3919 issued by the Washington State Department of Ecology.
The discharge of radionuclides is regulated by 10 CFR 20.2003. Uranium releases to the sewer have consistently remained below both the CFR quantity limit and concentration limit. A review of chemicals discharged shows that waste discharges have been consistently within the limits of the Waste Discharge Permit ST3919 with few exceptions. The exceptions have not been of significant concern to the Washington Department of Ecology. 13.4 Groundwater 13.4.1 Aoulfer Description The unconfined groundwater aquifer occurs at depths from 10 to 50 feet below the surface and is approximately 20 feet thick. Its direction of flow in the immediate vicinity of the SPC plant is south-southwest to north-northeast. The impermeable layer which defines the base of the unconfined aquifer is a sitt aquitard whose thickness ranges from (; four to 33 feet. v 13.4.2 Early Samplina Procram During the early years of plant operation the single lined (Petromat) lagoons experienced localized failures (tears, breaches) which allowed limited amounts of process wastes to enter the groundwater. The lagoon liner failures were detected in part by analyses of groundwater samples taken from monitoring wells adjacent to the process lagoons. These early, single lined lagoons were repaired when failure was detected and were subsequently upgraded with double liners with leak detection capability between the liners. Hypalon was chosen as the replacement lining material and was considered much superior to the Petromat. By 1983 allliquid - receiving lagoons had double hypalon liners at a minimum. In recent years certain of the lagoons have been upgraded further by the addition of a high density polyethylene (HDPE) liner (s). In 1982 a special groundwater study was conducted to determine the results of past j lagoon failures. Certain waste chemicals were detected in concentrations above i background in groundwater samples. A report of findings was supplied to the NRC and to the Washington State Department of Ecology. No action was required by either agency as a result of the findings and none was taken by SPC. Groundwater samples have continued to be taken and analyzed routinely as part of SPC's ongoing environmental protection / surveillance program. This program serves to ,q track the presence / movement of past contamination in addition to serving as a detection system for any new releases. NE NT MENOMENT A%CATON DATE: geptember12,1gg4 13-2 SPc-ND:3330 947 (R-90P92) l
i SiemenS Power Corporation - Nuclear Division eur.2 p SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 U PART 11 - SAFETY DEMONSTRATION nsv. CHAPTER 14 NUCLEAR CRITICALITY SAFETY 14.1 Administrative Practices l l Administrative and technical practices for ensuring criticality safety are detailed in documents EMF-30 (Safety Manual) and XN-NF-31 (Nuclear Criticality Physics Manual), ( respectively. These documents are prepared, approved, and implemented in accordance with the Approval and Responsibility Matrix (see Figure 1-2.3) in Chapter 2. The practices used to ensure criticality safety include the following: l Establishing a validated methodology for performing CSAs. e l Establishing the procedures to ensure that all equipment and operations are reviewed and approved for criticality safety before introduction of the SNM. Performing CSAs for the equipment and operations to be authorized for SNM C processing. The CSA includes an assessment of the worst credible accident condition. All modifications to existing equipment and operations with the SNM are analyzed for criticality safety impact before their implementation. Details on basic assumptions in CSAs are given in Section 14.3. The completed CSA is subjected to an independent second-party review to assure that the methods, accident conditions, models, cross-sections, etc., are adequately conservative, Establishing the specifications and procedures required to ensure that the limits o assumed in the CSA are conservatively met at credible accident conditions. Before introducing the SNM, inspecting equipment and procedures to ensure that the assumptions employed in the CSA are conservative. Ensuring that the limits and procedures are known and understood by the personnel dealing with the SNM using postings, specifications, operating procedures, and training. Auditing equipment, operations, and procedures to ensure continued compliance e with the limits established in the CSA and to ensure that all credible accident conditions have been adequately addressed. A AMENDMENT APPifCATION DATE. gg ggghg7gg,jgg4 PAGE No.: j 4.j SPc-ND.3330 947 (R-907/92)
SiemenS Power Corporation - Nuclear Division EMF-2 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257_ l PART ll - SAFETY DEMONSTRATION
- pey, 14.2 Preferred Approaches to Eauipment Desian and Criticality Control The preferred equipment design criterion is favorable geometry; i.e., the design is safe at the worst credible combination of geometry, enrichment, moderation, reflection, and interactions with other systems. When it is not practical to depend upon equipment designs where dimensions are limited to assure criticality safety, the reasons for not using i
favorable geometry shall be documented as part of the CSA and appropriate administrative controls are specified to ensure that no single credible violation of the specified limits will result in a criticality. t 14.2.1 Parameter Control There are three general approaches to controlling parameters that directly affect K, to assure compliance with the double contingency principle i.e., that no single credible accident, error, or process upset is capable of causing an accidental criticality. These are l listed in the order of preference along with a brief discussion of each as follows-1. Controlling two different parameters that directly affect K for each specific critical i g excursion scenario (diverse independent control). Examples of two different parameters that directly affect K, are mass and moderation. 2. Controlling a single parameter that directly affects K, for each specific critical excursion scenario using two independent and differently designed means of control (independent control). An example would be ensuring dry product from j a process by controlling the process within specified ranges or monitoring moisture with in-line instrumentation and then confirming dry product by laboratory analysis of product samples. 3. Controlling a single parameter that directly affects K, for each specific critical excursion scenario using two independent but like or identically designed means of control (redundant control). An example would be ensuring a safe concentration by using two independent density-measuring devices that are independently interlocked to two different fail-closed block valves that would prevent a transfer to unfavorable geometry vessels if the allowed. uranium concentration were exceeded. 14.2.2 Mechanism Control Beyond these general approaches to controlling parameters, each individual control mechanism falls into one of three categories, passive engineered control, active engineered control, or administrative control. These three categories are listed in order of preference with a brief discussion of each. It must be noted that in many cases a method of control cannot be exclusively categorized as either passive engineered or acwoutui Amcaron oart: September 12,1994 PAGE NO.: 14 2 SPC-ND.3330 947 (R-907/92)
SiemenS Power Corporation - Nuclear Division EMF-2 ] SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 (O PART 11 - SAFETY DEMONSTRATION REV. active engineered. In addition, engineered controls are never entirely independent of some administrative requirements. 1. Passive engineered control: A method of ensuring a limit by control over dimensions, fixed material composition, or absolute exclusion of undesired materials. Passive engineered controls are human dependent for: proper design, initial installation, verification while in service, and repairs. Geometry control is usually a passive engineered control. 2. Active engineered control: A method of control that relies upon a detector / feedback / control mechanism to regulate a process parameter. Active engineered controls are human dependent for: proper design, initial installation, calibration / testing while in service, and repairs. Active engineered mechanisms may be electrical, mechanical, hydraulic. or some combination of these or other devices. 3. Administrative control: Those controls that are procedurally implemented and, therefore, are highly dependent on the human element. () 14.2.3 Administrative Controls The administrative controls employed at SPC include: 14.2.3.1 Mass Limits and Controls Mass limits may be used as a control alone or in conjunction with controls on geometry, spacings, moderation, etc. 4 Mass-controlled systems such as 5-gallon powder containers are controlled to not more than one " safe" batch, defined as 45% of the minimum critical mass for a fully reflected spherical geometry. Additional controls imposed to assure safety at credible accident conditions are: 1. Certified dry (typically 2.0 wt% maximum equivalent moisture content) containers may be stored edge-to-edge in designated arrays in moderation-controlled areas; 2. Moderated materials (including all materials not certified dry) are stored in single tier planar arrays with at least 12 inches edge-edge spacings; and 3. Only designated containers may be used. f% AEf0W ANAM ATE: PA E NO.: September 12,1994 14-3 SPC-ND.3330.947 (R 1/07S2)
Siemens Power Corporation - Nuclear Division eur.2 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 l PART ll-SAFETY DEMONSTRATION
- ggy, i
Such arrays are typically demonstrated safe with at least two simultaneous accident conditions such as double batching, double stacking, edge-to-edge contact, and loss of moderation control. 14.2.3.2 Specina Limits and Controls I Units in-transit are typically specified to be maintained at least one foot edge-to-edge from .l other fissile units. Violation of this spacing requirement for any single unit must remain a safe condition. The arrangement of fixed units may not be changed without an Engineering Change Notice (ECN); this administrative control is part of the plant-wide l configuration control system. l 14.2.3.3 Moderation Limits and Controls Moderation limits are typically used in systems with low potential for accidental moderation and then only after the material has been certified as " dry." Control of .l moderation interspersed between fissile units (such as buckets) is typically not assumed. [ Water is the most common and most probable moderator, except for certain organic compounds such a zine stearate, AZB, etc., used in the pelletizing operation. The 1 approved additives are listed in a CSS, along with a factor used to convert the weight j percent additive to weight percent water equivalent. A single moderation limit for all powders / pellets in the plant is used except in special cases. Moderation limits are typically specified as up to 1.0 wt% water (as water) plus up to 1.0 wt% water equivalent as an approved additive. It is noted that 2.0 wt% water is not i acceptable (even with zero additive) and that 2.0 wt% water equivalent as approved additive is not acceptable (even with zero water). Certain cases (such as Nauta mixers) may not allow additives. The single credible accident condition evaluation shall assume at least double the maximum equivalent moisture content permitted in the material; i.e., for a system permitted to contain 2.0 wt% moisture equivalent (1.0 wt% water plus 1.0 wt% additive), the CSA shall demonstrate an acceptable k with at least 4.0 wt% moisture equivalent g in the material as an accident condition. Systems with moderation controls are subjected to a sensitivity analysis. The systems are 3 modeled at the maximum credible uranium density (typically 3.5 gm U/cm = 3.97 gm 3 UO /cm ). The H/U (typically, wt% water) of the SNM is the key variable for sensitivity 2 analyses. The specified moderation limit shall not be greater than half of the moderation required to produce a k of 0.95 assuming the worst credible combination of reflection g and interaction. Prior to being introduced into geometries that are unsafe at optimum moderation, SNM is held and sampled in a safe geometry vessel until it has been certified dry and released AMENDMENT APPLICATION DATE: $9ptembgf 12,1gg4 PAGE NO.: j4-4 SPC ND:3330.947 (R-UO7/92) r
SiemenS Power Corporation - Nuclear Division EMF-2 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 PART 11 - SAFETY DEMONSTRATION ngy. by supervision. Once certified, controls are imposed to preclude unsafe moderation levels. Certified dry powders are placed into drums fitted with neutron absorbers to preclude criticality if accidentally moderated. 14.2.3.4 Uranium for U-235) Concentration Limits and Controls Primary reliance may be placed on maintaining low uranium concentrations, provided that the probability of producing higher than specified concentrations is very low coupled with the requirement for sampling each container or tank of solution to verify acceptable concentrations before they enter unsafe geometries. The maximum acceptable concentration limit does not exceed 50 percent of the minimum critical value. I The concentration assumed for analysis is typically higher than the maximum credible l concentration. Concentration controls are ensured by process analysis (determination l that unsafe concentrations are unlikely even at upset conditions) and by analysis of the concentration before it is removed from a geometry-controlled vessel. Concentration I controlled containers are strictly controlled in designated areas. 14.2.3.5 Neutron Absorbers l 14.2.3.5.1 Fixed Poisons Fixed poisons are typically used as a secondary control. The materials currently used are B C granules in cores of cylindrical geometries, boric acid in welded tubes for dry powder 4 drums, and boral plates placed in finished bundle storage arrays. All poisons are subjected to destructive analysis or visual inspection to verify their presence. l Neutron absorbers are modeled assuming the most reactive credible conditions are met: 1. Lowest credible absorber concentration or less; and 2. Most reactive credible combination of fissile geometry, fissile moderation, and absorber geometry. The continued presence of the absorber is verified at least every five years. Neutron absorbers are typically used to ensure safety at one or more improbable accident l conditions rather than to ensure safety at normal conditions. AMENDMENT APPLCATON DATE: September 12,1994 PAGE NO.: j f.$ i SPC-ND3330.947 (R-1/07/92)
Siemens Power Corporation - Nuclear Division sup.2 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 PART 11 - SAFETY DEMONSTRATION
- ngy, 14.2.3.5.2 Soluble or Blended Poisons Currently soluble poisons are not used at SPC.
l ? Blended poisons are typically used as a secondary or tertiary control. The material currently used is Gd O23 blended into powders that are then pressed into pellets. Laboratory analysis confirms the presence of Gd O23 before credit is taken for its presence. Lot traceable follower cards that list the wt% gadolinia added to the material accompany the material through the process. Neutron absorbers are modeled assuming the most reactive credible conditions are met: 1. Lowest credible absorber concentration or less. 2. Most reactive credible combination of fissile geometry, fissile moderation, and absorber geometry. 14.3 Basic Analytical Assumptions and Technioues The most fundamental assumption used in criticality safety analysis is "if a parameter is not controlled, it is assumed to be optimum for criticality." 14.3.1 Accident Analysis The kg (or reactivity) of a system is strongly dependent on factors that must be conservatively addressed in the CSA. The Criticality Safety Component and second-party reviewer shall assure that both normal operation and credible accident conditions are evaluated in the CSA. A multi-disciplined team of experienced personnel from the Criticality Safety Component, Plant Engineering, and operating user shall jointly evaluate credible accident conditions. The technique (s) of analysis may include, but are not limited to, any of the following: e Hazard Barrier e Fault Tree e Component Failure Modes and Effects e What if i e Hazops. 14.3.1.1 Continoencies The following are examples of possible contingencies (single, independent failures) that are typically considered in an analysis- \\ l AMENDMENT A%CATON DATE: $0ptember 12,1994 PAGE NO.: 14-6
- l SPC-ND.3330 947 (R-907/92) i
SiemenS Power Corporation - Nuclear Division sup.2 O SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 b PART ll-SAFETY DEMONSTRATION REv. 1. Exceeding mass limits. Mass control restricts fissile material accumulations to 45 percent of a critical mass. Some of the possible contingencies are over-batching or double-batching, enrichment errors, accumulations of materialin enclosures, autoclaves, blowdown tanks, cleaning baths, waste systems, liquid systems, ventilation systems, etc. 2. Violation of geometry control. Geometry control restricts the size and shape of individual units and the shape of arrays. Some of the possible contingencies are corrosion, bulging of tanks due to chemically or mechanically induced pressure, bowing, bending, or sagging. 3. Violation of interaction controls. Interaction controls limit the minimum spacing between individual units or arrays of fissionable material. Some of the possible contingencies are operator error, accidental dislodgement of storage items or racks, possible equipment installation errors, and mechanical integrity of the system d6 sign. 4. Violation of raoderation controls. Moderation control excludes or limits the mass, Q configuration, and/or amount of moderating material that may be mixed with C/ fissionable material. Safe limits are based on optimum water moderation, unless moderation control can be assured. Some of the possible contingencies to be considered in cases where less than optimum moderation is assumed are water i supply left running due to operator error, inadvertent routing of water lines into an unmoderated area, hydrogenous liquid mistakenly taken to be non-hydrogenous, moderator inadvertently mixed with fissionable material, inadvertent dissolution, etc. 5. Violation of reflector conditions. Safe limits are usually based on full-water reflection. Consideration must also be given to the proximity of other reflectors and the potential reflection by material more effective than water. 6. Violation of neutron poison control. Poisons may be a fixed part of an array. The j contingencies to be considered are inadvertent omission during equipment changes, erosion, corrosion, or other geometry changes. 7. Equipment failures such as vessel deformation or structural collapse of racks, hangers, shelves, cabinets, or cubicles used for storage of fissile material, or loss of power, air pressure, etc. j 8. Water or oil flooding caused by leaky or ruptured lines (e.g., water cooling coils in a furnace or building water pipes), water addition from fire-fighting operations, tank overflows, siphoning, instrument malfunction, etc., and water backup in ('. i drains due to line clogs downstream. O AW NDMENT APPLCATON DATE: September 12,1994 PAGE NO.: j 4.f SPC-ND:3330 947 (R-1/07/92)
Siemens Power Corporation - Nuclear Division sup.2 [ SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 PART ll-SAFETY DEMONSTRATION
- ggy, l
9. Fire or explosion failures such as structural collapse of racks, hangers, shelves, f cabinets, cubicles, bird cages, and storage boxes; collapse of building structures; failure of storage containers; rearrangement of fissionable material as a result of fire fighting efforts; or water addition (see item 8). 10. Changes in density / concentration and/or moderation of the contained SNM, due to extremes of temperature and humidity, precipitation, etc. 11. SNM leaks and spills, accidental introduction or accumulation of the SNM into vessels or other locations that have not been adequately reviewed for criticality -
- safety, t
12. Human errors such as unauthorized procedures. 13. Potential for enhanced reactivity due to diameter, pitch, and interspersed l moderation effects in systems composed of normally close-packed rods or other i arrays without constraints on pitch. If accident conditions are not independent and of low probability, then combinations of } conditions are assumed. if combinations of accident conditions may lead to an i unacceptable k, then design changes and/or administrative controls are specified to l g minimize the probability and consequence of accidental deviation from the limits. If any single credible accident can lead to an unacceptable kg, design modifications or i administrative controls are implemented as required to convert this single accident into two or more independent accidents of low probability. 14.3.1.2 Parameters In addition to the independent failures listed above, the following parameters are also considered by the analyst: ] 1. Enrichment. The maximum credible enrichment (typically 5.0 percent) shall be considered along with the possibility for higher reactivities at lower enrichments. Reactivity typically increases with enrichment. However, it may be found, for example, that an array of 5-gallon buckets loaded with up to 42 kg of 3.0 percent - enriched material is more reactive than those loaded with up to 18 kg of 5.0 percent enriched material. This effect may also change with moderation within and between containers. 2. Form. Heterogeneous systems such as fuel assemblies immersed in water have higher reactivity potential than homogeneous systems such as a tank of uranyl nitrate solution. ^' " *^ " '" ^* September 12,1994 PAGE PdO.: 14 8 SPC-ND:3330.947 (R-1/07/92)
SiemenS Power Corporation - Nuclear Division EM F-2 p' SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 \\J PART ll - SAFETY DEMONSTRATION ngy. The composition of the fissile material affects reactivity. For example, the reactivity of a solution such as uranyl nitrate-nitric acid-water tends to decrease with increasing nitrogen content. Therefore, the minimum credible nitrogen content should be evaluated. Nuclides such as boron, cadmium, and gadolinium should always be neglected unless their continued presence is assured. The optimum credible value for the density and moderator content of the fissile material should be evaluated. 3. Geometry. For heterogeneous systems such as arrays of fuel rods, the optimum credible value for parameters such as fuel diameter and water-to-fuel volume ratio (Vw/Vf) should be assumed. For containment vessels, the most reactive credible geometry (including allowances for corrosion, tolerances, etc.) should be assumed. The possibility of deformation to more reactive geometry due to events such as pressurization shall be considered, Q] r The arrangement of units in a system should be evaluated to assure acceptable interactions at normal and credible accident conditions. Violation of the spacing requirement for portable in-transit containers is a credible accident condition. 4. Reflection and interspersed moderation. The neutrons escaping from the surface of a tank (for example) may collide with an adjacent concrete wall, water surrounding the tank (if flooded), or the water in the body of a person next to the tank. As a result of this collision, the neutron may be reflected back into the tank to potentially cause additional fissions. Without the collision in the " reflector" (concrete, water, etc.), the neutron would have escaped from the system (i.e., leaked), or for arrays of units, it may have traveled to a different fissile unit and caused additional fissions there. Such leakage and reflection are evaluated. Single units and arrays of units are typically evaluated with full water reflection (30 cm thick water layer) and/or with concrete reflection, if appropriate. Arrays of units are evaluated with various amounts of interspersed moderation (typically water of variable density) to determine the peak reactivity condition. 14.3.2 Criticality Analysis Techniaues 1 i The methods described below have been validated according to Regulatory Guide 3.14 " Validation of Calculational Methods for Nuclear Criticality Safety". j 14.3.2.1 Sinole Parameter Limits b/ AMENDMENT APPLCATION DATE: September 12,1994 PAGE NO.: 14-9 SPC-ND:3330 947 (R-1/07/92)
e Siemens Power Corporation - Nuclear Division sup.2 SPECIAL NUCLEAP. MATERIAL LICENSE NO. SNM-1227 NRC DOCKET NO. 70-1257 PART ll-SAFETY DEMONSTRATION
- gey, Single-unit safety for elementary geometries may be demonstrated using established critical dimension limits from any of the references listed in Chapter 4 and appropriate safety factors as described in Tables 1-4.1 and I-4.2.
14.3.2.2 Bucklina Conversion K values for elementary systems and compositions may be estimated using buckling g calculations with data from souregs such as References 1 and 2 of section 14.5. For buckling calculations k, = W 1 + x Bl. 2 The formulae for geometric buckling of standard geometries are listed in Table 11-14.1. TABLE ll-14.1 EXPRESSIONS FOR GEOMETRIC BUCKLING IN TERMS OF ACTUAL DIMENSIONS AND EXTRAPOLATION DISTANCES 2 Geometry Buckling B Sphere of radius r r ,e ,r + s, Cylinder of radius r and height h ' 2.405'a r sa r + 6, r h + 26, r Cuboid of dimensions a, b, and c r r 3: + + , a + 28, r b + 28, 1c + 26, Buckling equations may also be used to determine various equivalent geometries for any given material. This becomes usefulin some surface density calculations. 14.3.2.3 Surface Density If an array of well-spaced units containing fissile material was compressed to form a slab on the floor and that slab was, in turn, reflected by water, it would be desirable for suberiticality to be the result. The surface density model for calculating limiting conditions of arrays derives from analyzing of such a situation. September 12,1994 PAGE NO.: 141Q aucsousut amcaron oars: SPC-ND:3330.947 (R-1/07/92) -j
Siemens Power Corporation - Nuclear Division EM F-2 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 PART 11 - SAFETY DEMONSTRATION
- ggy, The surface density method views an array in terms of its fissile density as projected onto a bounding plane, e.g., the ground or a floor. Comparison of the resulting value to the density of a water-reflected critical slab of the same composition is then the basis for establishing material limits.
The surface density model described by Reference 3 of Section 14.5 is a well-verified and recent formulation. It suggests an allowed surface density, Sigma-allowed, expressed as: Sigma-allowed = 0.54 Sigma-crit (1 - 1.37*f), where Sigma-crit is the surface density in grams per square centimeter of a critical water-reflected infinite slab. Sigma-crit can be obtained from references such as ARH-600(2) or DP1014(1). i f is the " fraction critical", i.e., the ratio of the mass of a unit in the array to the critical mass of an unreflected cphere of the same material. f = Mass U per unit / minimum critical bare sphere i A The value of f is determined as if each unit were infinitely tall. For multi-tier arrays, the sum of all tiers at a given planar location is used as the mass for a single unit. Since negative surface densities are not achievable, the fraction critical must not exceed 0.73 for this formulation to be applicable. As an example: Waste drums are limited to 100 grams U and a maximum enrichment 235 of 5.0 wt%. This corresponds to a net uranium mass of 2 kg. For a four-tier array, the total uranium mass per vertical unit is 8 kgARH-600(2) reports the minimum critical mass j for a bare sphere of 5.0 wt% enriched U to be approximately 60.6 kg and the surface 3 density of a critical water reflected infinite slab to be 12.5 g U/cm. Therefore, the fraction ) critical, f, is, f = 8/60.6 = 0.132 Therefore: Sigma-allowed = 0.54
- 12.5 * (1 - 1.37
- 0.132) = 5.529 Waste drums are limited to 2,000 g U and 5.0 wt% enrichment. Sigma-actual for a four-tier array is:
4
- 2,000/d2 = 8,000/3,266.12 = 2.449 Ob autwatwT amcarou care:
September 12,1994 PAGE NO.: 14-11 SPC-ND.3330.947 (R-1/07/92)
Siemens Power Corporation - Nuclear Division EMF.2 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 PART 11 - SAFETY DEMONSTRATION REv. i Where d2 = center-to-center spacing of the drums. The drum diameters are 57.15 cm i and no spacing between the drums is assumed. r The sigma-actual / sigma-allowed for a four-tier array is 0.443, 14.3.2.4 Solid Anale Neutron interaction in multi-unit arrays may be assessed in accordance with the Solid ) Angle Method. Neutron interaction (exchange between individually suberitical units) must be considered. Consideration of the interaction between units or arrays of the SNM may be accomplished through the use of the solid angle method. The solid angle method is applied according to the constraints in Reference 3 of Section 14.5, except for the use of the nominally reflected solid angle acceptance criteria. The nominally reflected solid angle acceptance criteria are used to limit the allowable i solid angle for arrangements of individually suberitical units provided that the following conditions are met: 1. - Boundary conditions for the spacing between concrete walls and the array are as stated in Table 1 of Reference 4 of Section 14.5, except that a minimum separation of six inches shall be required. 2. Concrete walls are less than or equal to seven inches in thickness; 3. Separation distances given in Table 1 of Reference 4 of Section 14.5 are measured from the outermost vessel in the array to the closest wall. 4. The array shall be limited in both number and size of vessels to arrays that are reasonable extrapolations of the conditions assumed in Reference 4 of Section 14.5. 5. All vessels within the array shall be suberitical when fully reflected by water - and shall have a minimum edge-to-edge separation of 12 inches. For arrays that violate any of the five conditions stated above, additional analyses shall be necessary to demonstrate the safety of the particular array in question, or demonstrate the continued acceptability of using the nominally reflected solid angle acceptance criteria. O PAGE NO.: 14-12~ i AMENDMENT APPLCATON DATE: September 12,1994 SPc-ND.3330.947 (R 1/07/92) ') I
Siemens Power Corporation - Nuclear Division eur.2 (N SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 PART 11 - SAFETY DEMONSTRATION REv. 14.3.3 QCalculations Usina Computer Codes Most k,3 calculations for non-elementary geometries or SNM compositions are performed using cross-sections and computer codes from the SCALEW system of codes. This system of codes includes CSAS, BONAMI, NITAWL, XSDRN-PM, ICE, and severa! cross-section libraries. These cross-sections and codes have been extensively benchniarked against data from critical experiments. This benchmarking has been performed by SPC and others around the world. Allimprovements/ corrections to the codes are maintained and distributed by the Radiation Shielding information Center (RSIC) as soon as practical. SPC incorporates corrections and improvements when prudent to do so. l { The KENO-V.a. a 3-D Monte Carlo computer code, is the most frequently used code for i estimating system ken-The cross-section libraries available through the SCALE package include Hansen-Roach 16-group,27-group NDF4,218-group NDF4 and 123-group GAM-THERMOS. (O) Calculations are routinely performed using the Hansen-Roach 16-group cross-section library with resonance corrections by BONAMI and NITAWL The resonance self-shielding effects are conservatively modeled. BONAMI and NITAWL are used to estimate the Dancoff factor and the effective moderation cross-section for resonance nuclides. A Monte Carlo code is also used to calculate Dancoff factors and effective moderation cross-sections for heterogeneous systems where an infinite lattice is not assumed as in BONAMI. Thus, more accurate average values for finite lattices and, if desired, for individual rod types such as edge, corner, and interior may be employed. Since the effect of these refinements is typically small, and since the results using infinite lattice parame-ters are often more conservative, this approach is considered optional. Replicate KENO-V.a calculations using NITAWL-processed cross-sections from the 27-group NDF4, 218-group NDF4 and 123-group GAM-THERMOS libraries may be and frequently are performed for comparison. When the SCALE system of codes is used to calculate k,g, key items checked by the analyst and second-party reviewer include: 1. Verifying that the appropriate resonance self-shielding corrections have been applied to the cross-sections used. 2. Verifying that the nuclides and the atom densities in the mixing table conservatively represent the media in the system. O ,w wo w w u m c u ou o us' September 12,1994 P AGE N " 14-13 i SPC-ND.3330 947 (R U0742)
SiemenS Power Corporation - Nuclear Division Eup.2 ~] SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 (O PART 11 - SAFETY DEMONSTRATION ngy. 3. Ensuring that dimensions and media are appropriate for the condition being l tested. To help verify the geometry model, computer plots of the KENO geometry model are routinely made. These plots are used to ensure that the overall model and the units in the model are appropriate. 4. Verifying that the solution is well converged. If significant trends in k between g generations are present or if the generation k distribution is abnormal, determine g if additional generations or a different start type is needed. 5. Examining the data on flux, fission density, and by-group absorption-leakage fission for anomalies. Assure that the most reactive region was adequately sampled and, if necessary, replicate the case with all or most first-generation neutrons starting in the most reactive region. Previously approved codes / cross-sections (GAMTEC, DTF, HFN, HAMMER, CCELL for r example) have been used for analyses and may be used in the future. Other codes such as CASMO, and MCNP may be employed after appropriate validation. p 14.3.4 Methods Validation SPC is currently using the SCALE-4.2 version of the SCALE package which is distributed by the RSIC. RSIC provides sample problems, including inputs and outputs, for test cases that have been run on RSIC computers so users of SCALE 4.2 can check each of the modules used and ensure they are running correctly on their machines. After the SCALE 4.2 package was installed and appeared to be functioning correctly, several of these sample problems were used to verify each of the criticality safety modules were providing expected results. These test cases exercised each of the modules in the criticality safety sequence. A comparison of the 25 KENO.Va test cases provided by RSIC and these same cases run on the SPC HP workstation SSLO1 is given in Table 11-14.2 below. All cases except #21, in which the SPC calculation is more conservative (higher) than the RSIC case, agree within normal statistical uncertainties. No anomalies were found in the output for Case
- 21. All test cases run on workstation SSLO1 were, therefore, verified to have acceptable results.
O autwowcut amcatow om: September 12,1994 PaGE NO2 14.j 4 SPc ND 3330 947 (R 1/07/92)
I Siemens Power Corporation - Nuclear Division i aus.2 O SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 .Q PART ll - SAFETY DEMONSTRATION REv. TABLE 11-14.2 COMPARE 25 KENO.Va RSIC TEST CASES WITH SPC HP WORK STATION SSLO1 RSIC k, SPC k i m Delta-k POOLED CASE ,,y (RSIC MINUS STANDARD ID STANDARD POOLED AVERAGE AVERAGE SPC) DNN DON DEVIATION gig,,) 1 1.0065 0.0047 1.00608 0.00446 4.20E-04 6.48E-03 0.06 2 1.0065 0.0047 1.00608 0.00446 4.20E-04 6.48E-03 0.06 3 1.0116 0.0053 1.0054 0.00517 6.20E-03 7.40E-03 0.84 4 1.0122 0.0054 1.01067 0.00528 1.53E-03 7.55E-03 0.20 5 1.0245 0.0036 1.02225 0.00422 2.25E 03 5.55E-03 0.41 6 0.7496 0.0037 0.74949 0.00382 1.10E-04 5.32E-03 0.02 7 1.0055 0.0038 1.00026 0.00409 5.24E 03 5.58E-03 0.94 8 0.9491 0.0041 0.94911 0.00412 1.00E-05 5.81E43 -0.00 9 2.2848 0.0078 2.28484 0.00777 -4.00E-05 1,10E-02 -0.00 10 1.0065 0.0047 1.00608 0.00446 4.20E-04 6.48E-03 0.06 11 1.0065 0.0047 1.00608 0.00446 4.20E-04 6.48E-03 0.06 12 1.0094 0.0055 1.00327 0.00518 6.13E43 7.56E-03 0.81 i 13 1.003 0.0043 1.00213 0.00358 8.70E-04 5.60E43 0.16 14 0.9976 0.0049 1.00324 0.00465 -5.64E-03 6.76E-03 -0.83 15 1.0095 0.0049 1.00087 0.0043 8 63E-03 6.52E-03 1.32 16 0.9885 0.0029 0.99212 0.00267 -3.aA 03 3.94E-03 -0.92 17 1.0101 0.0164 1.0089 0.01664 1.20E 03 2.34E-02 0.05 18 1.0146 0.0073 1.01892 0.00752 -4.32E-03 1.05E-02 -0.41 19 1.0117 0.0047 1.01518 0.00628 -3.48E-03 7.84E43 -0.44 20 1.0001 0.0063 1.00883 0.0057 -8.73E-03 8.50E-03 -1.03 PAGE NO.: 14 15-AMNNT ANATON DATE: September 12,1994 SPC.ND-3330 047 (R 1107/92)
) i SiemenS Power Corporation - Nuclear Division Eur-2 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 i PART ll - SAFETY DEMONSTRATION asy. TABLE 1114.2 (Cont'd.) COMPARE 25 KENO.Va RSIC TEST CASES WITH SPC HP WORK STATION SSL01 RSlCk, SPC km Delta-k POOLED CASE g (RSIC MINUS STANDARD ID STANDARD ST NOARD pgg AVERAGE SPC) DEVIATION DEVIATION DEVIATION g,p,) 21 0.9844 0.0036 1.00063 0.00316 -1.62E-02 4.79E-03 -3.39 22 1.0079 0.0048 1.00579 0.00444 2.11 E-03 6.54E-03 0.32 23 1.0048 0.0046 1.00618 0.00446 -1.38E-03 6.41E-03 -0.22 24 1.0044 0.0043 1.00617 0.00412 1.77E 03 5.96E-03 -0.30 25 1.0059 0.0043 0.99998 0.0039 5.92E-03 5.81E-03 1.02 O The computer codes currently used for calculating k have been validated as prescribed g in ANSI /ANS 8.1 to determine the calculational bias. Generally, neither the bias nor its uncertainty is constant and both should be expected to be functions of composition and other variables. The determination of calculational bias is dependent upon the system's characteristics. The analyst is responsible for determination of the appropriate bias correction for the system being evaluated. 14.3.5 Selected Validation Cases Used Selected validation cases used by SPC in the validation process are listed below. 14.3.5.1 PNL Critical Experiments (Reference 6) The Reference 6 of Section 14.5 experiments involve three flooded clusters of 4.31% enriched rods with variable spacings between the clusters and with various absorbers between the clusters. The case numbers referenced below were taken from Reference 6 of Section 14.5. The data used for these cases include the corrections made to 235 Reference 6 on August 14,1979. Among these corrections was a change in the U enrichment from 4.29 wt% to 4.31 wt%. Figure ll-14.1 shows the general layout of the critical mass experiments. Table ll-14.3 lists data pertinent to the critical mass experiments. %./ aww:wt=T amcatom cart: September 12,1994 PA E e j 4.j g SPC-ND,3330.947 (R-UO7/92) l
6 I Siemens Power Corporation - Nuclear Division eur.2 l C SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 ( PART 11 - SAFETY DEMONSTRATION
- ggy, i
i i ? ) A L f l i $0.BaMl LM nm l ['
- )2 mn tmma-+
305 snm imm) Y Y - - -el-l TB-o k ooo o u CARBON STLEL TANK oo / \\. r i a 0o000 a E <[ E l WATER LEVEL _ ~ u ooooo s C 1$2 m a imin) = = L l j E W W ML o o M8:50 8s6 35 mm a ooo eo g { 606l = M Al ANGLT l u g o o o o og g a j CI f f fac W E e Polices PLAR f, 111mmaD9 mm a 7 - = ~==5 l g g g sacocco ,, oo,,, ACRYLIC etAK$ =
== y 5 153:50 8:LM mm o d 6061
- 4 Al
,,'g lCHA8fNEL l OO 5 g f 90T!DM OF ML l 8R#5 25 4 mm Nim / RYL C R$ .,14 ACRTLIC PLAR g,, g +3lIl am imink --- OLS mm--. IND YlEW rim virw i i i r FIGURE 11-14.1 PNL CRITICAL EXPERIMENT LAYOUT (REFERENCE 6) l Autwout=T amcaton cart: September 12s 1994 PAGE NO.: j 4.j 7 I SPC-ND3330 947 (R-1107/92) i
Siemens Power Corporation - Nuclear Division eur.2 C SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 PART 11 - SAFETY DEMONSTRATION 1 REv. Cases 001,002, and 003 determine the critical size of one cluster. The critical size was interpolated based on experiments with integral numbers of rods per edge; the critical number had a fractional number of rods on one edge and either 8,9, or 10 rods on the other edge. These three cases were modeled using cell-weighted cross-sections. A suffix "x" on the case name was used to denote cases modeled with cell-weighted cross-sections. i TABLE ll-14.3 EXPERIMENTAL DATA ON CLUSTERS OF 4.29 WT% U235 ENRICHED UO RODS IN WATER ( ) 2 FUEL CLUSTERS NUMBE LENGTH x WIDTH CRITICAL SEPARAT!ON 2 R IN 25.40 mm P.TCH BETWEEN FUEL EXPERIMENT ARyY(2 (FUEL RODS) CLUSTERS (3) (Xc, mm) NUMBER 1 10 x 11.51 0.04 O 001 = 1 9 x 13.35 0.01 003 1 8 x 16.37 0.03 002 = 3 15 x 8 106.4 0.1 004 3 15 x 8 106.0 0.1 032(4) (1) Error limits shown are one standard deviation. (2) Clusters of fuel rods aligned in a single row. (3) Perpendicular distance between the cell boundaries of the fuel clusters. (4) Rerun of Experiment 004. Case 004 involved three 15x8 clusters with no absorber plates. Cases 007,008,013, and 014 involved three clusters with 304L steel absorber plates. Two plate thicknesses and different absorber spacings from the central cluster were tested. Cases 009,010R,011, and 012 are similar to the previous four except that the 304L steel contained either 1.05 or 1.62% boron. Case 031 involved three clusters with BORAL absorber plates. Cases 029 and 030 involved three clusters with Zircaloy-4 absorber plates. AMENDMENT APPLCATON DATE: September 12,1994
- GE No :
14-18 SPC-ND:3330 947 (R-1/07/92)
l i 1 Siemens Power Corporation - Nuclear Division eur.2 { SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 v j PART li-SAFETY DEMONSTRATION ngv. The calculation results with the 16,27,123, and 218 group cross-section libraries are included in Table 11-14.4. All cases used resonance parameters calculated by the CSAS modules. Most models included provisions for using different resonance parameters for interior and edge rods. The "latticecell" parameters were input for the interior rods and "more data" parameters equal to those resulting from the "latticecell" parameters were used for edge rods. The "x" suffix on the case name denotes cell-weighted cross-sections. Suffixes such as "a" are for explicitly-modeled rods. The results indicate that cell-weighted calculations tend to yield higher k-eff values than i explicit models. The results with 16-group results are accurate or conservative; other cross-sections tend to be nonconservative for the cases modeled. OO I n(J u awsourut amcaros cart: September 12,1994 PaGE NO.: 14.j 9 SPC-ND.3330.947 (R-UO792)
O O O l TABLE 11-14.4 m l { REFERENCE 6 CASES CALCULATION RESULTS WITH 16-GROUP CROSS-SECTIONS, $g3 l 5 27-GROUP CROSS-SECTIONS,218-GROUP CROSS-SECTIONS, AND 123-GROUP CROSS-SECTIONS 5@ i rm Ha Ubrary 27-Group NDF4 218-Group NDF4 Group GAM-TH RMOS O i r-4 h Case ID l k k k k g g g g mo 1 Avg. Std.Dev. Avg. Std.Dev. Avg. Std.Dev. Avg. Std.Dev. gQ >o l 1.00355 c001x 0.002490 1.005910 0.002640 0.997410 0.00268 0.996550 0.002690 m3 I e. l 1.00905 c002x 0.002570 0.998280 0.002740 0.99660 0.00235 1.001870 0.002530 gO l 1.00845 hh c003x 0.002520 1.002680 0.002340 0.996850 0.00259 0.99650 0.002670 l 1.00435 ZE c004 0.002650 0.998530 0.002660 0.988460 0.00276 0.990370 0.002750 wg l 1.00244 h c005a 0.002650 0.998820 0.002320 0.994690 0.00231 0.989980 0.002450 o oO l 1.00198 c005b 0.002520 h 3. W m 6 Z 9. c006a l 1.00188 0.002360 0.989960 0.00230 0.989430 0.0026 0.993350 0.002740 z gg c006b 0.99954 0.002370 c007a l 1.00425 0.002470 0.990630 0.002380 0.990670 0.00263 0.987780 0.002780 6 'z Z i y c007x 1.00788 0.002530 0.997890 0.002560 1.000310 0.00248 O O l 1.00148 h h 0.002310 0.998180 0.002590 0.989780 0.00279 0.987790 0.002410 c008a 9 E A c008x l 1.00109 0.002420 1.001770 0.002680 0.998320 0.00237 Q l1.00062 h 0.002330 0.996130 0.002450 0.993840 0.00274 0.992210 0.002430 h c009a a l 1.00481 h c010a 0.002390 0.994440 0.002350 0.993140 0.00238 0.992430 0.002530 y 3 s a SE a 5 2
i OO _I TABLE 11-14.4 (Cont'd.) ]F REFERENCE 6 CASES CALCULATION RESULTS WITH 16-GROUP CROSS-SECTIONS, O3 27-GROUP CROSS-SECTIONS,218-GROUP CROSS-SECTIONS, AND 123-GROUP CROSS-SECTIONS l 5 r$ 5 4 v3 Ha n Roach Library 27-Group NDF4 218-Group NDF4 GAM TH RMOS i O E mo M Case ID k k k k g g g g mO c011a 1.00356 0.002750 0.993820 0.002770 0.991950 0.00259 0.991490 0.002650 l KS ~ > I) c012a 1.00063 0.002440 0.991030 0.002610 0.988110 0.00245 0.997500 0.00250 MR m% c013a 1.00142 0.002400 0.997760 0.00240 0.994030 0.00242 0.993580 0.002550 J 55' n ra c013x 1.01149 0.002270 0.993060 0.0025 1.003010 0.00246 C i OZ l E i l y c014a 0.99991 0.002410 0.993520 0.002430 0.988660 0.00229 0.991370 0.002460 i u) zE 4 u) 5 c014x 1.00732 0.002590 0.997080 0.00240 1.000020 0.00263 mq v i 80 c029a 0.99956 0.002420 0.99366 0.00230 0.993850 0.00245 0.992810 0.002730 g g E I o) z c030a 1.00278 0.002570 0.992410 0.00259 0.990640 0.00239 0.992790 0.002420 o g6 i h L* c031a 0.99736 0.002390 -i ro h Avg. 1.00328 0.002470 0.996280 0.00250 0.993990 0.00252 0.993020 0.002580 g z z m Std.Dev. 0.00345 0.000120 0.004110 0.000150 0.004260 0.00015 0.003540 0.000130 O O h
- These cases were not calculated b
o xm i Z o P 6 y 1 e a E -e s 7 to
Siemens Power Corporation - Nuclear Division EMF-2 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 PART ll-SAFETY DEMONSTRATION REv. 14.3.5.2 B&W Critical Experiments (Reference 7) The Reference 7 of Section 14.5 experiments involve nine (3x3 array) partially-flooded clusters of 2.46% enriched rods with variable spacings between the clusters and with various absorbers between the clusters. A generallayout of these experiments is shown in Figure 11-14.2. Specific parameters are included in Tables 11-14.5 and 11-14.6. The case numbers referenced below were taken from Reference 7 of Section 14.5. All cases except 2321 have the clusters separated by one pin pitch (1.636 cm). The water contains about 700-770 ppm dissolved boron. Case 2321 has four pitches between the clusters and zero dissolved boron. The calculational results with 16 group cross-sections are listed in Table 11-14.7. All cases used resonance parameters calculated by the CSAs modules. O' 4 l l l j O I Q autsoutut apetcarou onTE: September 12,1994 PaGE NO.: 14-22 sPC-ND.3330 947 (R-vo7/92)
Siemens Power Corporation - Nuclear Division aus.2 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 PART 11 - SAFETY DEMONSTRATION
- ggy, t
i ? t a Fuel Rod Position 'I i ~ r V//////////////! W
- /////////A Top Cml(2.54 cm)
Moderator Level Bt Pm Fuel Rod 177.3 153.66 sm 150 em H I i. Bottom Cral(154 cm) .1 j OJ cm g rfffffffgyf, ,3
- gg ygfjg
~ =9 r f 5.04 cm Alumumm Ben Mais N [4 94,[/////// // /// //// / // // // /// / / ////f // /// Bottom of Core Tent i I r(,\\) FIGURE 11-14.2 B&W CRITICAL EXPERIMENT LAYOUTS i t aucNousNT appucaton 04Tc: September 12,1994 PAGE NO.: 14-23 SPC ND3330 947 (R 1107/92)
Siemens Power Corporation - Nuclear Division sur.2 t SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 f PART 11 - SAFETY DEMONSTRATION
- ggy, TABLE 11-14.5 PROPERTIES OF 2.46% ENRICHED UO FUEL RODS 2
PARAMETER AVERAGE o(x) o(x) o.d., em 1.206 0.002 0.0003 j Wall thickness, em 0.081 0.003 Wall material 6061-T6 Al Pellet diameter, em 1.030 0.001 0.0005 Total length, em 156.44 0.41 0.05 Active fuel length, em 153.34 0.88 0.02 Pellet length, em 1.914 0.008 0.001 I Weight of UO g/ rod 1305.5 39.7 1.0 p Weight of U/ weight UO, % 88.13 0.01 0.000 p Weight of U, g/ rod 1150.5 35.0 0.9 235 Weight of U , g/ rod 28.29 0.86 0.02 Enrichment, weight 25/ weight Uranium, % 2.459 0.002 0.001 I 3 Pellet density, g/cm 10.29 0.005 0.02 3 Bulk density, g/cm 10.22 0.36 0.01 2 3 4 EN,,o, (summation of impurities, cm /cm oxide) < 4 x 10 TABLE 1114.6 CERTIFIED CHEMICAL ANALYSIS OF B C 4 Total boron, wt% 77.8 Total carbon, wt% 20.8 Anhydrous B O, wt% 0.10 2 3 Boron plus carbon, wt% 98.6 Particle size,30 to 50 mesh, % 91.2 O AucNoWENT APPLICATION DATE: September 12,1994 PCE NO.: 14-24 SPC-ND3330 947 (R-UO7/92)
I Siemens Power Corporation - Nuclear Division eur.2 O SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET N PART ll - SAFETY DEMONSTRATION
- ggy, i
TABLE 11-14.7 i REFERENCE 7 CASES CALCULATION RESULTS WITH 16-GROUP CROSS-SECTIONS k l g CASEID STANDARD AVERAGE DEVIATION a-2265zb 1.01932 0.00196 l a-2266 1.01179 0.0019 l a-2268zb 1.01263 0.00194 l a-2271zb 1.01034 0.0018 l a-2321 a 1.00651 0.00225 l a-2321b 1.00535 0.00205 l a-2321c 1.0064 0.00217 l 1 Additional SPC validation documentation is contained in Reference 8 of Section 14.5. 14.3.6 SNM Properties 14.3.6.1 Homoneneous Systems i The most frequently encountered homogeneous fissile materials are: 1. UO Powder with various moisture contents; 2 2. U O powder with various moisture contents; 3 g 3. UF (solid, liquid, or gas) in shipping cylinders or in piping; s 4. UO F -HF-water solutions resulting from UF hydrolysis; 22 s 5. Ammonium Diuranate (ADU) as dry powder, aqueous slurry and pastes" with water contents between these extremes; and 0N' l AMENDMENT APPLCATION DATE: September 12,1994 PAGE NO.: 14-25 I SPC-ND.3330.947 (R-U07/92)
Siemens Power Corporation - Nuclear Division eur.2 ( SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 G PART 11 - SAFETY DEMONSTRATION
- ngy, 6.
Uranyl nitrate (UNH) in aqueous or organic solution. Uncertain compositions or mixtures are typically modeled as UO -water if deemed to be 2 conservative. Homogeneous mixtures are routinely modeled as UO -water. This is adequate to cover 2 the range of credible reactivities of UO, U 0, and UO F. Since the nitrogen in mixtures 2 3 s 22 such as UNH and ADU absorbs neutrons more than equal amounts of oxygen or fiuorine, the nitrogen content is typically set to zero or to a minimum credible value with no credit for the excess nitrate and/or ammonium that may be in the actual system. A given two-phase mixture such as UO -water is modeled as a single homogeneous 2 phase consistent with the methods used in SCALE lil and SCALE IV. Other UO -water 2 mixtures with other uranium densities may be similarly formulated and used in XSDRNPM/ KENO to calculate k,3. The mixture with the highest k,3 shall be taken as the optimum moderation condition. Data are available (e.g., Section Ill.B of ARH-600) to allow rapid estimation of the optimum conditions. Figures lil.B.10(5)-1 and Ill.B.10(5)-2 of ARH-600(2) contain kg, material V buckling, migration area, and extrapolation distance data for a wide range of uranium densities in the UO -water and uranyl nitrate-water systems. Based on these data and 2 confirmed by several actual cases analyzed. the optimum moderation condition assume ;! i 3 for preliminary estimates is typically 1.9-2.1 gn' U/cm. The actual optimum may differ slightly depending on the geometry and reflection conditions. j i it is noted that the uranium concentration to yield the minimum critical mass is different from that for peak reactivity. For systems without mass limits, the units shall be assumed to be full of the most reactive credible mixture. For systems with specified mass limits such as 5-gallon cans of UO2 Powder, the most reactive uranium density for a single or double-batched container is assumed. Typically, the most reactive geometry-mixture combination is to fill the container with a saturated UO -water mixture of about 0.82 gm U/cc (single batch) or about 1.64 gm U/cc (double 2 batch). If higher uranium densities were used, the fissile geometry would be less reactive for the modeled contained mass. 14.3.6.2 Heteroaeneous Systems Typical heterogeneous systems are: l {Al G AMENDMENT APPLCATION DATE: September 12,1994 PAGE NO2 j f.26 SPC-ND:3330.947 (R-1/07/92)
) Siemens Power Corporation - Nuclear Division eur.2 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 J PART ll - SAFETY DEMONSTRATION REv. 1. Unclad UO pellets under trays which are routinely stacked; 2 2. Normally close-packed fuel rods on geometry control or with controls on number of rods together; and 3. Fuel assemblies. Most operations with heterogeneous systems are not normally moderated. The only two normally moderated heterogeneous systems currently used are rod etching and rod autoclaving. For moderator densities less than 100% (full density water), the optimum moderation condition is strongly dependent on the system geometry. All systems must have an acceptable k when fully flooded and fully reflected. Certain g cases, typically rod storage cabinets with tiers of " slabs" of rods, may exceed the kg limits if optimum moderation and close-fitted reflection are assumed. The range of moderation levels with unacceptable k 's is typically small and the optimum water g density is typically low,1-10 vol% water, for example. 4 h While such conditions are deemed not credible, additional controls are implementec.o / make the probability even more remote: 1. The array is fully enclosed and water must be capable of freely draining from the system; 2. The array is designated as a moderation-controlled area. The low probability of unacceptable moderation at conditions such as fire fighting are verified; and i 3. Strict controls on storage of hydrogenous materials such as paper and plastics are imposed. Even if low density interspersed moderation was accidentally introduced into these cabinets, they would typically remain suberitical because: 1. The rods within the slabs would be grossly undermoderated compared to the optimum-pitch rods modeled; 2. Typical cabinets are free standing with significant reflection only by the floor. Therefore, neutron leakage would be much greater than the fully reflected case. Attaining a fully reflected exterior with low density moderator inside would be difficult. The spacing between tiers of slabs would greatly reduce the kg if filled with full density water; and O AMENOMENT A%CATON DAR: September 12,1994 PAGE NO.: j f.27 SPC-NO3330.947 (RM/07/92) ~
y ? i SiemenS Power Corporation - Nuclear Division eup.2 [ SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 PART 11 - SAFETY DEMONSTRATION
- nev,
-i i 3. The cabinets typically contain steel walls, support members, and trays which are conservatively not modeled but, in fact, neutron absorptions in this steel would tend to lower the peak k to acceptable values for all moderation levels. g 14.4 Structural intearity Policy and Review Proaram All changes to systems with potential criticality safety impact require an Engineering Change Notice (ECN) which requires approval by qualified persons responsible for engineering design, operations, and safety (including criticality safety), i 14.5 References 1. H. K. Clark, " Critical and Safe Masses and Dimensions of Lattices of U and UO2 Rods in Water," DP-1014, Savannah River Laboratory (1966). i 2. R. D. Carter, G. R. Kiel, and K. R. Ridgway, " Criticality Handbook," Volumes I,11, { and Ill, ARH-600, Atlantic Richfield Hanford Company, (196S). /~' 3. J.T. Thomas, ED., " Nuclear Safety GuidernD-7016/ Revision 2," NUREG/CR-0095 and ORNUNUREG/CSD-6, U.S. Nuclear Regulatory Commission (1978). l 4. C. L Brown, et. al., " Validation of Boundary Conditions for Assuming Nominal Reflection in Solid Angle interaction Method (as Applied in Exxon Fuel Fabrication i Plants)," BNW/XN-184, (1975). 1 5. " SCALE: A Modular Code System for Performing Standardized Computer Analyses for Ucensing Evaluation," NUREG/CR-0200. 6. NUREG/CR-0073: " Critical Separation Between Suberitical Clusters of 4.29 wt% Enriched UO Rods in Water with Fixed Neutron Poisons." 2 7. BAW 1484-7: " Critical Experiments Supporting Close Proximity Water Storage of Power Reactor Fuel." l 8. EMF-94-175 " Validation and Verification of KENO.Va". fD O AMEfCMENT A%WO@ ATE: September 12,1994 PAGE NO.: j4 2$ $PC-ND.3330,947 (R-vo7/92) L
Siemens Power Corporation - Nuclear Division Eur-2 SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1227, NRC DOCKET NO. 70-1257 PART 11 - SAFETY DEMONSTRATION REv. covered by a vapor barrier insulation, plus a 20-year, builtup, low melt asphalt roof with a final coating of gravel. The inner side of the precast wall panels are insulated and I covered with fire rated gypsum board. The gypsum panel joints are taped and sealed and the interior surfaces are suitable painted. The compressive strength of all concrete exceeds 3000 psi and the 6-in thick concrete floor slabs were designed for 250 psf. The fire loading of the building is kept to a minimum through monthly inspections described in Paragraph 2.6.4 of Chapter 2 of Part 1. Fire extinguishers are strategically positioned throughout the building and inspected monthly. Fixed / rate-of-rise temperature sensors throughout the building provide fire alarm l capability. l 15.2.1.4 Environmental Safety l 15.2.1.4.1 Containment / Confinement - Material processed in the SF Building is contained or confined in open or closed primary containers for UO Of 2 U 0 Powders and UO Pellets in the NAF pellet fabrication area; hoods 38 2 ( for powder and pellets in the OC laboratory; and hoods, boxes, drums and the incinerator for SWUR. l The concrete floors in the ELO Building are sealed to be liquid tight and there are no floor drains. Uguid effluents which could contain uranium or other hazardous material are treated to reduce such materials to levels within regulatory limits prior to discharge. 15.2.1.4.2 Heatina. Ventilation and Air Conditionina (HVAC) -The SF Building has four independent HVAC systems. The SWUR Facility located in Room 173 l Is basically a once-through HVAC system with double HEPA filtration (K48 supply and K49 exhaust) serving Room 173 and the adjacent airlock. The remainder of the production facilities, laboratories, changerooms, and office sections is served by a combination (K5 supply and K6 exhaust) once-through and recirculation supply and exhaust system with double HEPA filtration. The incinerator itself is served by the K50 system consisting of a once-through HVAC system with double HEPA filtration. The incinerator shroud cooling system is served by the K55 system consisting of a once-through system with double HEPA filtration. Simplified schematics of these HVAC systems are furnished in Figures ll-10.19 (SWUR K48/K49 and K50 HVAC systems),11-10.20 (SF production L area K5 and K6 HVAC systems), and ll-10.27a (SWUR incinerator shroud cooling system). l =cacut amcatom onit: September 12,1994 15-30 SPC-ND.3330 947 (R-UO7/92) L}}