ML20050D057
| ML20050D057 | |
| Person / Time | |
|---|---|
| Site: | Zion File:ZionSolutions icon.png |
| Issue date: | 03/16/1982 |
| From: | Varga S Office of Nuclear Reactor Regulation |
| To: | Delgeorge L COMMONWEALTH EDISON CO. |
| References | |
| NUDOCS 8204120017 | |
| Download: ML20050D057 (5) | |
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Docket File i NRC PDR Local PDR MAR 161982 ORB 1 File D. Eisenhut J. Heltemes C. Parrish Docket Nos. 50-295 D. Wigginton and 50-304 OELD ety I&E (1)
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- l Mr. Louis 0. De1 George ACRS (10) c T
Director of Nuclear Licensing
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Comonwealth Edison Company
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Post Office Box 767
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Chicago, Illinois 60690
Dear Mr. DelGeorge:
On January 29, 1982 in a Comonwealth Edison Company (Ceco) letter to the NRC, guidance was requested to aid in the design of a low rad waste storage facility at the Zion Station site. This guidance in response to your specific questions is enclosed.
Sincerely, Steven A. Varga, Chief Operating Reactors Branch No. 1 Division of Licensing
Enclosure:
As stated cc: See next page l
l 8204120017 820316 DR ADOCK 05000295 PDR ORB 1 DID RP
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j Mr. Louis 0. De1 George I'
Commonwealth Edison Company cc: Robert J. Vollen, Esquire
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109 North Dearborn Street Chicago, Illinois 60602 I
Dr. Cecil Lue-Hing Director of Research and Development Metropolitan Sanitary District of Greater Chicago 1
100 East Erie Street Chicago, Illinois 60611 Zion-Benton Public Library District 2600 Emmaus Avenue Zion, Illinois 60099 Mr. Phillip P. Steptoe Ishan, Lincoln and Beale Counselors at Law One First National Plaza 42nd Floo-Chicago, Illinois 60603 Susan N. Sekuler, Esquire Assistant Attorney General Environmental Control Division 188 West Randolph Street, Suite 2315 Chicago, Illinois 60601 i
U. S. Nuclear Regulatory Commission Resident inspectors Office 105 Shiloh Blvd.
Zion, Illinois 60099 James P. Keppler Regional Administrator - Region III U. S. Nuclear Regulatory Commission 799 Roosevelt Street Glen Ellyn, Illinois 60137 n
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CECO Question 1: The recommended site boundary dose rate is 1 mrem /yr.
Is this considered an absolute maximum?
If not, what technical support would be required to raise the
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acceptable dose rate?
Is there a threshold value 4
above which installation of the facility would be i
considered an unreviewed safety question?
NRC Response:
The value of 1 mrem /yr is not intended as a limit.
In the context of the guidance document, it was quoted only l
as an incremental dose which, when added to other doses contributed by uranium fuel cycle facilities, would not likely impact on the ability of any nuclear power plant to meet the limit of 40 CFR 190. Regardless of the magnitude of the site boundary dose rate, it is incum-i bent upon the licensee to document in the evaluation conducted pursuant to 10 CFR 50.59 that neither the limits of 10 CFR Part 20.105 for exposure rates in unrestricted areas nor the limits of 40 CFR 190 are exceeded. Further-more, the 10 CFR 50.59 evaluation must also document that routine releases of radioactive material from the storage facility when added to other releases from the plant will not cause the dose design objectives of Appendix I to 10 CFR Part 50 to be exceeded. No site boundary dose rate " threshold value" has been established above which installation of the facility would be con-l sidered an unreviewed safety question. Guidance on when a facility modification such as installation of l
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an expanded low level radioactive waste storage facility should be considered an unreviewed safety question is j
I found in the enclosed IE Circular No. 80-18, "10 CFR Part 50.59 Safety Evaluations for Changes to Radioactive Waste Treatment Systems", issued August 22, 1980.
l CECO Question 2: As a minimum, a quarterly container inspection program is recommended. What percentage of the containers must be inspected at each inspection interval?
Is the inspection interval dependent upon container stack height (i.e., load on the base container)?
Is there a limit on a stack height?
NRC Response:
No specific guidance on percentage of containers to be inspected was provided because it is the staff's opinion that this number is a function of a number of parameters including 1) waste form, 2) container design, 3) environ-ment external to the container, 4) length of time in i;
storage, 5) quality control applied to container speci-fications and waste processing, etc. The staff believes that the load on the base container is certainly one of the parameters, together with those listed above, that should be evaluated in determining both an appropriate inspection interval and the number of containers to be inspected at each interval.
In any case, the staff l
certainly did not intend that extensive container handling (unstacking and restacking) would be necessary to conduct an adequate container inspection program.
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, A similar position relative to stack height is also held by the staff, i.e., no specific guidance was provided since allowable stack height is dependent upon a number of site specific parameters. These parameters i
include design loading of the container, height of the t.
storage structure, handling technique, type and stability q
i of the waste form, percent of void space within the con-tainer, container geometry and method / stability of stacking (pyramid or one-on-one with grating between layers), industrial safety requirements, container inspection method, etc.
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SSINS No.: 6830 I
Accession No.:
8006190038 IEC 80-18 UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C.
20555 August 22, 1980 IE Circular No. 80-18: 10 CFR 50.59 SAFETY EVALUATIONS FOR CHANGES TO RADI0 ACTIVE WASTE TREATMENT SYSTEMS Discription of Circumstances:
Recent inspection efforts at operating power reactors have revealed numerous instances in which licensees have failed to perform adequate safety evaluations to support ~ changes made to the design and/or operation of facility radioactive waste treatment systems.
These safety evaluations are required by the regula-tions of 10 CFR 50.59 whenever changes are made in the facility as described
- I in the Safety Analysis Report (SAR).
The inadequacies of the evaluations have caused radiological safety hazards to occur unidentified and therefore to remain unevaluated and uncorrected.
In two particular cases, the inadequately evaluated system changes resulted in system failures that caused an uncontrolled release of radioactivity to the environment.
In each of these situations, a proper 10 CFR 50.59 safety evalua-tion should have identified and corrected deficiencies in the system modifica-tion and/or operation and would have prevented the inadvertent release of radioactivity.
NRC ' followup examination of the situation indicates that the inconsistency
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j and/or inadequacy of licensee safety evaluations may be widespread.
A wide 1
range of opinions seems to exist among licensees as to what constitutes an appropriate 10 CFR 50.59 safety evaluation, particularly for radwaste systems.
Therefore, the following discussion and/or guidance is provided for licensee I
use in preparing future 10 CFR 50.59 safety evaluations to support changes in the design and/or operation of the radioactive waste treatment systems of licensed facilities.
Although the contents of this guidance are specifically directed to the radioactive waste systems, the general principles and philosophy of the 10 CFR 50.59 safety evaluation guidance are also applicable to the facility design and operation as a whole; thus, the application of 10 CFR 50.59 should
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reflect a consistent approach.
Discussion:
The requirements of 10 'CFR 50.59 are composed of three essential parts.
First, paragraph (a)(1) is permissive in that it allows the licensee to make changes to the facility and its operation as described in the Safety Analysis Report without prior approval, provided that a change in Technical Specifica-tions is not involved or an "unreviewed safety question" does not exist.
Criteria for determining whether an "unreviewed safety question" exists are i
defined in paragraph (a)(2).
Second, paragraph (b) requires that records of changes made under the authority of paragraph (a)(1) be maintained.
These records are required to include a written safety evaluation that provides the i
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lEC 80-18 L
August 22, 1980 Page 2 of 3 basis for determining whether an "unreviewed safety question" exists.
Paragraph (b) also requires a report (at least annually) of such changes to the NRC.
Third, paragraph (c) requires that proposed changes in Technical Specifications be submitted to the NRC as an application for license amendment.
Likewise, proposed changes to the facility or procedures and the proposed conduct of tests that involve an "unreviewed safety question" are required to be submitted to the NRC as an application for license amendment.
Any proposed change to a system or procedures described in the SAR, either by text or drawings, should be reviewed by the licensee to determine whether it irtvolves an "unreviewed safety question." Maintenance activities that do not result in a change to a system (permanent or temporary), or that replace components with replacement parts procured with the same (or equivalent)
I purchase specification, do not require a written safety evaluation to meet 10 CFR 50.59 requirements.
However, a safety evaluation is required to meet the provisions of 10 CFR 50.59 and any change must be reported to the NRC as required by 10 CFR 50.59(b) if the following circumstances occur: (1) com-ponents described in the SAR are removed; (2) component functions are altered; (3). substitute components are utilized; or (4) changes remain following comple-tion of a mainte. nance activity.
Notice to Licensees:
1 For all cases requiring a written safety evaluation, the safety evaluation must set forth the bases and criteria used to determine that the proposed change does or does not involve an "unreviewed safety question." A simple statement of conclusion in itself is not sufficient.
However, depending upon the significance of the change, the safety evaluation may be brief.
The scope I
of the evaluation must be commensurate with the potential safety significance of the proposed change or test.
The depth of the evaluation must be sufficient to determine whether or not an "unreviewed safety question" is involved.
These evaluations and analyses should be reviewed and approved by an appro-priate level of management before the proposed change is made.
f An important part of the "unreviewed safety question" determination is the evaluation and analysis of the proposed change by the licensee to assure that (1) potential safety hazards are identified, and (2) corrective actions are taken to eliminate, mitigate, or control the hazards to an acceptable level.
All realistic failure modes and/or malfunctions must be considered and protec-tion provided commensurate with the potential consequences.
All applicable j
regulatory requirements, including Technical Specifications, must be complied f
with so that the proposed change shall not represent an "unreviewed safety question." Also, the margin of safety as defined in the bases of the Technical Specifications shall not be reduced by the proposed change.
For radioactive waste systems, the appropriate portions of 10 CFR 20, 30, 50, 71, and 100, the facility Technical Specifications, and 40 CFR 190 (Environ-mental Dose Standard) are applicable.
Additional specific criteria that should be reviewed prior to the modification of radioactive waste systems are presented below:
(1) System modifications should be evaluated against the seismic, quality group and quality assurance criterfa'in Regulatory Guide 1.143.
Design
IEC 80-18 August 22, 1980 Page 3 of 3 provisions for controlling releases of radioactive liquids, as presented in Regulatory Guide 1.143, should also be evaluated.
(2) Radiological controls should be evaluated against the criteria in Regulatory Guide 1.21 and Standard Review Plan Section 11.5, " Process and Effluent Radiological Monitoring and Sampling Systems."
l (3) Systems involving potentially explosive mixtures should be evaluated against the criteria in Standard Review Plan Section 11.3, " Gaseous Waste Management System," subsection II, item 6.
(4) System design and operation should be evaluated to assure that the radiological consequences of unexpected and uncontrolled releases of radioactivity that is stored or transferred in a waste system are a small fraction of the 10 CFR 100 guidelines; i.e., less than 0.5 rem whole body dose, 1.5 rem thyroid from gaseous releases, and less than the radionuclide concentrations of 10 CFR 20, Appendix B, Table II, Column 2 from liquid releases at the nearest water supplies.
(See Standard Review Plan
. Sections 15.7.1, 15.7.2, and 15.7.3 for more details.)
The evaluation must include an analysis encompassing the above criteria to the extent that the criteria are applicable to the proposed changes; i.e., if the modifications involve a change addressed by the above regulations and criteria, then the modifications must be evaluated in terms of these regulations and criteria.
In, conclusion, for any change in a facility radioactive waste system as
' described in the SAR, a safety evaluation is required in accordance with 10 CFR 50.59.
In this safety evaluation and the "unreviewed safety question" determination, the evaluation criteria in Items 1-4 above should be used.
If the proposed modification (design, operation, or test) represents a departure from this evaluation criteria, one of the following actions should be taken:
(1) The proposal should be modified to meet the intent of the criteria; I
(2) The evaluation / determination must present sufficien't analyses to l
demonstrate the acceptability of the departure; or, (3) Commission approval must be received prior to implementing the modification (i.e., an unreviewed safety issue may be involved).
No written response to this circular is required.
If. additional,information g
L regarding this subject is required, contact the Director of this office.
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IEC 80-18
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August 22, 1980 RECENTLY issue 0 IE CIRCULARS Circular Date of No.
Subject Issue Issued to e
L 80-17 Fuel Pin Damage Due to Water 7/23/80 All holders of PWR Jet from Baffle Plate Corner OLs and PWR cps 80-16 Operational Deficiencies In 6/27/80 All power reactor Rosemount Model 510DU Trip facilities with an Units And Model 1152 Pressure OL or a CP Transmitters 80-15 Loss of Reactor Coolant Pump 6/20/80 All power reactor Coo, ling and Natural Circula-facilities with an I
tion Cooldown OL or CP 80-l4 Radioactive Contamination of 6/24/80 All holders of power Plant Demineralized Water and research reactor j
System and Resultant Internal licenses (operating Contamination of Personnel and construction permits), and fuel cycle licensees 80:13 Grid Strap Damage in 5/18/80 All holders of reactor Westinghouse Fuel Assemblies OLs and cps 80-12 Valve-Shaft-To-Actuator Key 5/14/80 All holders of reactor May Fall Out of Place When OLs and cps l
Mounted Below Horizontal Axis 80-11 Emergency Diesel Generator 5/13/80 All holders of a power Lube Oil Cooler Failures reactor OL or CP l
80-10 Failure to Maintain 4/29/80 All holders of reactor p
Environmental Qualification OLs and cps l'
of Equipment i
80-09 Problems With Plant Internal 4/28/80 All holders of a power f
Communications Systems reactor,0L or CP t
80-08 BWR Technical Specification 4/18/80 All General Electric Inconsistency - RPS Response BWRs holding a power l
Time reactor OL 80-07 Problems with HPCI Turbine 4/3/80 All holders of a power Oil System reactor OL or CP
.