ML20154S658
| ML20154S658 | |
| Person / Time | |
|---|---|
| Site: | Westinghouse |
| Issue date: | 10/21/1998 |
| From: | Robert Williams WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| References | |
| CON-NRC-98-051, CON-NRC-98-51 TAC-L31068, NUDOCS 9810280049 | |
| Download: ML20154S658 (6) | |
Text
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Westinghouse Commercial Nuclear orawer a Electric Corporation Fuel Division f3$'32Eo""'282' 8
NRC-98-051 October 21,1998 Director Office of Nuclear Material Safety and Safeguards U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555
Dear Sir:
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SUBJECT:
CHANGED PAGES; LICENSE NUMBER SNM-1107; DOCKET 70-1151; REVISED TO INCORPORATE RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (TAC NO. L31068)
Westinghouse Electric Company hereby submits (six copies of) changes to a proposed Revision 11.0 of pages iv and Chapters 1.0 and 3.0, for the Application for Renewal of a Special Nuclear Materials License for the Commercial Nuclear Fuel Division at the Columbia, South Carolina Fuel Fabrication Facility. An enclosure to this letter describes the changes.
If you have any questions, please contact me at (803) 776-2610, Extension 3393.
Sincerely, WESTINGHOUSE ELECTRIC COMPANY Robert A. Williams, Licensing Project Manager Westinghouse Columbia Plant Docket 70-1151 r " 9 O '. b License SNM-Il07
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cc:
U. S. Nuclear Regulatory Commission g\\
ATTN:
Mr. Charles Gaskin j
Licensing Section 1, Licensing Branch i
FCS&S Division, NMSS 11545 Rockville Pike Mail Stop T8D14 I
Rockville, MD 20852-2738 i
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Enclosures I
9810280049 981021 PDR ADOCK 07001151L C
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NRC-98-051 October 21,1998 DESCRIPTION OF CHANGES l
The enclosed changed pages of the License Application reflect NRC Staff's expressed satisfaction with responses to their Request for Additional Information related to the original l
April 3,1998 submittal of the Amendment Request for Revision 11.0 of SNM-1107. In i
addition, another " structured incident evaluation" (i.e., " Specific Cause Determination") has been added to Section 3.7.2 (page 3.18).
The pagination of this submittal still reflects (to a minor degree) the previously discussed anomaly resulting from Westinghouse confusion of " Revision Number" versus " Amendment Number." This has been manually corrected in all on-site copies of the License Application; and, in accordance with a prior commitment, will be electronically corrected (with a complete correct copy of the License Application submitted to NRC Staff) when Revisions 10 and 11 l
have been approved.
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i REGULATORY-SIGNIFICANT PROCEDURES - Those procedures that contain, in whole or in part, actions that are important to environmental protection, health, safety, and/or safeguards.
RESTRICTED AREA -- Areas such as the Manufacturing Building, or equivalent areas, to which access is restricted by physical or administrative methods and which is monitored on a scheduled basis by the site Security Function.
SAFE MASS [3.7.3(b.2) and (c.5)] critical mass for a particular process or vessel given the credible material geometry for that process / vessel, and the License Evaluation Bounding Assumptions for that material type (e.g., homogeneous UO ) and reflection, 2
Optimum moderation and material density are assumed.
j SAFETY-RELATED -- Relevant to systems crucial or important to safety; and, those systems that improve the margin of safety (e.g., in the context of maintenance).
SAFETY-SIGNIFICANT - Relevant to systems crucial or important to safety (e.g., in the context of quality assurance).
SIGNIFICANT SPREADABLE RADIOACTIVE CONTAMINATION - For the purpose of determining notification requirements when an employee receives medical treatment outside of facility contamination control areas, significant means greater than 9000 disintegrations per minute (approximately 0.1 Annual Limit of Intake) total contamination on an affected patient's skin and clothing (inside the plastic wrap, or equivalent means, utilized to prevent the spread of contammation during patient transport and preparation for treatment).
UNRESTRICTED AREA - An area, access to which is neither limited nor controlled.
WILL - Denotes a mandatory requirement to take a stated action or course.
Docket No.
70-1151 Initial Submittal Date:
30APR90 Page No.
1.12 l
License No. SNM-1107 Revision Submittal Date: 03APR 98 Revision No. 11.0 l
based reporting process for an area will be reviewed as a part of the formal audits of the area, as described in Paragraph 3.6.1(b) of this Chapter.
l 3.7.2 STRUCTURED INCIDENT EVALUATION 1
An incident review committee -- comprised of the Engineering Component Senior Manager, the Manufacturing Component Senior Manager, and the Regulatory Component Senior Manager - will determine if reported process upsets and/or procedure inadequacies are to undergo structured incident evaluation.
Structured incident evaluations will be maintained by a datapack process. Documentation of this process will provide for the following information:
Results of a Root Cause Analysis or Specific Cause Determination, led by an e
individual with formal training in conducting such an analysis, including recommendations; Status of corrective action (s) implementation; Regulatory assessment; e
Notification documentation; e
Training documentation; Plant-wide applicability assessment; and, e
Miscellaneous infonnation pertaining to the incident and/or the evaluation.
3.7.3 NOTIFICATION OF REGULATORY AGENCIES 1
Cognizant Regulatory Agencies will be promptly notified of major safety incidents in accordance with all requirements from 10 CFR Parts 20 and 70. In particular, as points of additional clarification, the NRC Operations Center will be notified of the following types of inc.idents, within the time limits prescribed (Note: The incident time limit
" clock" will start when a qualified Emergency Coordinator classifies an event, or when a cognizant Regulatory Function Engineer makes the initial " eyes-on" assessment of the safety of the errant condition, whichever comes first. Emergency Coordinators will be trained and authorized to make 1-Hour notifications Emergency Coordinators will be trained and authorized to make 4-Hour notifications, if they cannot contact a cognizant Regulatory Function Engineer, at first attempt to do so, to come and make an initial
" eyes-on" assessment):
(a) 1-Hour Notifications (NRC Operations Center) l l
(a.1) Any incident for which an Alert or Site Area Emergency has been declared, as prescribed by the Site, Emergency Plan described in Chapter 9.0 of this License Application.
i Docket No.
70-1151 Initial Submittal Date:
30APR90 Page No.
3.18 l
License No. SNM-1107 Revision Submittal Date: 03APR98 Revision No. 11.0 l
(a.2) Any incident involving Quality Level A systems, for which accident controls cannot be initiated, whether or not regulatory limits are exceeded.
(b)-
4-Hour Notifications (NRC Operations Center) l
-(b.1) Any incident involving Quality Level B systems, for which accident controls cannot be initiated, whether or not regulatory limits are exceeded.
(b.2) Any nuclear criticality safety incident, in an analyzed system, for which less than previously documented double contingency protection remains (multi-parameter control or single-parameter control) and:
(b.2a) Greater than a safe mass is involved and double contingency protection is not restored within four (4) hours; or, (b.2b) Greater than a safe mass is involved and controls are restored within four (4) l hours, but:
- i. Only single contingency protection is restored; or, l
ii. Double contingency protection is restored but multiple original controls under both contingencies were modified or replaced.
(b.3) Any determination that a criticality safety analysis or evaluation was deficient, or that a particular system was not previously analyzed; and, that less than two unlikely, independent, and concurrent changes in process conditions would be required before a criticality accident would be possible.
(b.4) Any unanticipated /unanalyzed nuclear criticality safety incident, or incident involving a previously unanalyzed system, for which the severity and remedy are not readily determined.
(c) 24-Hour Formal Notifications (NRC Operations Center) l (c.1) Any incident for which the work area is unavailable for normal use for 24-hours, following a loss of radioactivity contamination control, where the unavailability is due to the contamination.
(c.2) Any incident for which Quality Level A or B system safety equipment is not performing its intended function.
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(c.3) Any incident for which an employee, having significant spreadable radioactive contamination (see definition, Section 1.4), receives medical treatment outside of facility contamination control areas.
i Docket No.
70-1151 Initial Submittal Date:
30APR90 Page No.
3.19 l
License No. SNM-1107 Revision Submittal Date: 03APR98 Revision No. 11.0 l
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- (c.4) Any incident for which a fire or explosion damages nuclear fuel and the damage affects the integrity of the licensed material or breaches its container, when the l
material involved is greater than five times the lowest Annual Limit on Intake (ALI) specificd in Appendix B of 10CFR20.100-20.2401 for the material.
l (c.5) Any nuclear criticality safety incident, in an analyzed system, for which less than previously documented double contingency protection remains (multi-parameter j
control or single-parameter control) and:
(c.Sa) Less than a safe mass (see definition, Section 1.4) is involved; or l
l (c.5b) Greater than a safe mass is involved, but a sufficient number of the controls that l were lost are restored within four (4) hours such that double contingency l
protection is restored.
(d)
A procedure will be prepared, maintained, and followed - in accordance with Subsection 3.4.1 of this Chapter - that details the information to be included in a l
notification. In particular, each notification of a nuclear criticality safety incident will include (where applicable) the following information:
Whether the notification is the result of an event, or of a deficient nuclear criticality safety analysis (including the time period for which the deficiency l
existed);
The safety significance of the incident; l
Potential criticality pathways involved, including brief scenario (s) of how accidental criticality could occur; Controlled parameters -- mass, moderation, geometry, concentration, etc. -
e involved; Estimated amount, enrichment, and form of licensed material involved -
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including applicable process limits and the percent of worst-case critical mass of the material, in the configuration, involved; A description of the involved failures or deficiencies - including applicable l
nuclear criticality safety controls or control systems; and.
Corrective actions to restore safety systems, and when each was, or will be, l implemented.
l Docket No.
70-1151 Initial Submittal Date:
30APR90 Page No.
3.20 l
License No. SNM-1107 Revision Submittal Date: 03APR98 Revision No. 11.0 l