ML20079H274
| ML20079H274 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 01/11/1984 |
| From: | Foster D GEORGIA POWER CO. |
| To: | James O'Reilly NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| References | |
| GN-303, NUDOCS 8401230384 | |
| Download: ML20079H274 (8) | |
Text
{{#Wiki_filter:I Georgia Power Company N 333 Piedmont Avenue Atlanta, Georgia 30308 Telephone 404 526-7726 Maihng Address: Post Office Box 454S AManta, Georg a 30302 Georgia Power D. O. Foster the snuttnyn electrc system Vice Pres: dent and General Manager Vogtle Project January 11, 1984 United States Nuclear Regulatory Commission File: X78G03-M48 Office of Inspection and Enforcement Log: GN-303 Region II-Suite 3100 101 Marietta Street Atlanta, Georgia 30303
Reference:
Vogtle Electric Generating Plant - Unit 1, 50-424; Westinghouse NLP Printed Circuit Cards GN-275, dated 11/3/83, GN-298, dated 12/27/83 Attention: Mr. James P. O'Reilly Gentlemen: Georgia Power Company has completed its investigation into the above referenced concern and has concluded that a reportable event could exist. Based upon NRC guidance in NUREG-0302, Revision 1 and other NRC correspondence, Georgia Power Company is reporting this concern under 10 CFR 50.55(e), since duplicate reporting of a 10 CFR Part 21 and 10 CFR 50.55(e) event is not required. Enclosed is a copy of the evaluation performed for this concern. This response contains no proprietary information and may be placed in the NRC Public Document Room upon receipt. You s ru j o . O. Foster D0F/REF/cc xc: U. S. Regulatory Commission Document Control Desk Washington, D. C. 20555 l R. J. Kelly D. E. Dutton J. A. Bailey L. T. Gucwa R. E. Conway W. F. Sanders
- 0. Batum M. Malcom G. F. Head R. H. Pinson H. H. Gregory, III G. Bockhold, Jr.
J. T. Beckham, Jr. B. M. Guthrie C. W. Hayes P. D. Rice D. N. MacLemore R. A. Thomas E. D. Groover J. L. Vota 8401230384 840123 ON 'd O P gDRADOCK 05000424 I PDR d kD i I
s EVALUATION FOR A SUBSTANTIAL SAFETY HAZARD EVALUATION FOR A SIGNIFICANT DEFICIENCY . Westinghouse NLP Printed Circuit Cards Initial Report: On June 1, 1983, Westinghouse (E. P. Rahe) informed the NRC (R. C. DeYoung) of two potential problems within the Westinghouse 7300 Process Protection System. One of these problems involved heat sink adhesive failures of printed circuit cards. (The other ( ctncern involved the failure of NTC relay cards and'was reported by Georgia Power Company to the NRC on September 6, 1983.) W stinghouse did not list the Vogtle Electric Generating Plant in their list of plants having potentially defective NLP cards. Subsequent investigations ' identified that three defective NLP cards were located in the balance of plant (BOP) process cabinet. Georgia Power Company notified the NRC of this potential concern on October 4, 1983. Background Information: H:at sink adhesive failures were identifed at several utilities and were reported to W2stinghouse as field deficiencies All reported field failures have occurred only on the loop power supply (NLP) cards. Westinghouse has determined that NLP printed circuit cards shipped from Westinghouse Industry Electronic Division (WIZD) between August 1, 1980, and September 1, 1982, were equipped with a thermal heat sink assembly on the inverter transistors that is subject to potential failure in the adhesive bond in the thermal link assembly. The adhesive bond is between an insulating washer and the thermal link. Failure of the bond can cause the heat sink plate to separate from the thermal links and fall off of the printed circuit board. The plate is made of conductive metal and under certain circumstances could cause shorting of the low level eignals if it became vadged between cards in the card frame. Westinghouse has not de-fined the adhesive failure mechanism or the expected number of hours of system operation bsfore failure. For Plant Vogtle, the potentially defective heat sink assembly can be rscognized by the hex nuts visible on the top. side of the assembly. The engineering -aview identified the following defective cards in the balance of plant process cabinet iSv4QSPP1 (C-1) for Unit 1 only. 1. PQY-3000 Provides power supply to instrumentation loop P-3000 which controls the atmospheric power operated relief valve PV-3000. 2. QY-2791B Provides an isolated signal for the Train A diesel generator power output indication and computer input. 3. FQY-77744 -- Provides power supply to instrumentation loop F-7774 which provides the automatic start signal to a reactor makeup pump. An analysis of potential interactions was conducted to determine if there was a potential impact on plant safety. The analysis assumed the preexistence of a short circuit resulting from potential physical interactions with the NLP card conductive plate concurrent with the most limiting single active failure following the onset of an event (transient or accident condition) requiring a response from the impacted card.
e - Engineering Evaluation: A total of 12 potential interactions were identified and analyzed for impact on plant safety. Of these 12 interactions, 2 resulted in unacceptable conse-quences. In one case the potential interactions of an NLP card conductive plate could ! render the train A penetration area ESF filtration unit inoperable. Failure of the train A unit concurrent with a single failure of the train B unit following a LOCA results in the inability to maintain a negative pressure in the penetration area. This could result in the unfiltered release of radioactivity to the environment in excess of analyzed releases. Although a qualitative assessment of this scenario indicates that the total offsite doses, including the unfiltered release, are within the 10 CFR 100 guidelines, the scenario represents an unanalyzed event which is not considered in the present safety analysis. The other case involved the potential interaction of an NLP card conductive plate which renders the isolation valve on the auxiliary component cooling water (ACCW) return header from the reactor coolant pump (RCP) #2 thermal barrier inoperable. This valve functions to automatically isolate the ACCW return header in the event of a thermal barrier failure. A backup isolation valve is provided for re-dundancy. Failure of one valve due to the instrument loop interaction with the NLP card conductive plate concurrent with a single failure in the backup valve results in the inability to mitigate the consequences of a thermal ~ barrier failure and the transport of reactor coolant to the auxiliary building environment through the ACCW surge tank overflow. Although a qualitative assess-ment of this scenario indicates that the total offsite dose is a small fraction ofthe 10 CFR 100 guidelines, the scenario represents an unanalyzed event which is not considered in the present safety analysis. It should be pointed out that the potential interaction of these heat sinks with other systems is a low probability event. The system interaction study was conducted to identify possible failure modes. Review for Reportability NUREG-0302 Revision 1, " Remarks Presented at Public Regional Meetings to Discuss Regulations for Reporting of Defects and Noncompliance" was reviewed for guidance relating to what the NRC may consider to be a major reduction in the degree of protection provided to public health and safety. The following guidance was obtained from page 21.3(k)-1 concerning the definition of substantial safety hazard: Question 1: Please elaborate on the definition of " Substantial Safety Hazard" as used in Part 21. For instance, give examples of what the NRC would con-sider to be a " major reduction in the degree of protection provided to public health and safety." l l 0
. Response: Appendix A to NUREG-0090-7, Report to Congress on Abnormal Occurrences, June 1977 lists a number of events that may help to illustrate the NRC concept of " Substantial Safety Hazard." Specific illustrations of what we would consider to be " major reduction in the degree of protection provided to public health and safety" include: o Exposure in excess of 25 rems, whole body (10 CFR 20.403(a)(1)) o Exposure of an individual in an unrestricted area to more than 0.5 rem in-one~ calendar year (10 CFR 20.105(a)) o Release of radioactive material to an unrestricted area in excess of 500 times the limit of Appendix B, Table II, 10 CFR Part 20 (10 CFR 20.403(b)) o Exceeding a safety limit as defined in the facility technical specifications o A deficiency:which seriously compromised the ability of a confinement system to perform its designated function ~ 1 -Also, in the section concerning " Remarks by 13ut Office of Standards Development 1 to Public Regional Meetings on 10 CFA Part 21 by W. E. Campbell, Jr." page 6 - states: " Basic component" as applied to nuclear power reactors means a plant structure, - s, 'em, component or part thereof necessary to assure (1) integrity of the ' reactor coolant pressure boundary or (2) the capability to shut down the reactor and maintain it in a safe shut down condition or (3) the capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to those referred to in Section 100.11 of Title 10. (This definition also appears in'Part 10 CFR 21 as definition of - a basic' component). ~ NOTE: Attachment A is a restatement of Sec. 100.11. ' Attachment B is a restatement of Sec. 20.403 Attachment C is a restatement of Sec. 20.105 Based upon NRC guidance in the-above referenced reporrs and attachments, Georgia Power Company realizes that these events may not exceed Part 10 CFR 100 limics and may therefore not be considered reportable under Part 10 CFR 21 or 10 CFR 50.55(e). However, the~ assumed failure of one train of the ESF B train penetration area filtration unit and resulting failure of the A train from the failure of the NLP card does represent a potential system interaction that results in the inability ^ of a confinement system to perform its designated function. Thus this event can constitute a substantial safety hazard due to the potential to have unfiltered - plant releases to the atmosphere. Since the magnitude of the releases have not been absolutely quantified, Georgia Power Company considers this concern reportable since the health and safety of the public may be affected. 4 m n ,en, ~ ,-r,---, w g w-- ,..,-1-- ,e a-,---r+ wa w "- 1
8v ' -4_ Also, guidance in 10 CFR 50.55(e) requires the reporting of.any event that could adversely affect the safety of operations of the nuclear power plant at any time throughout the expected lifetime of the plant. The potential to have plant releases in excess of those analyzed in the FSAR does constitute an effect on the safety of operations since health and safety of the public may be affected. I e I I
b2 O ATTACHMENT A Restatement of Sec. 100.11 Sec. 100.11. Determination of exclusion area, low population zone, and population center distance - (a) As an aid in evaluating a proposed site, an applicant should assume a fission produce release from the core, the expected demonstrabla leak rate from the containment and the meteorological conditions pertinent to his site to derive an exclusion area, a low population zone and population center distance. For the purpose of this analysis, which shall set forth the basis for the numerical values used, the applicant should determine the following: (1) An exclusion area of such size that an individual located at any point on its boundary for two hours immediately following onset of the postulated fission product release would not receive a total radiation dose to the whole body in excess of 25 rem or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure. (2) A low population zone of such size that an individual located at any point on its outer boundary who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage) would not receive a total radiation dose to the whole body in. excess of 25 rem or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure. (3) A population center distance of at least one and one-third times the distance from the reactor to the outer boundary of the low population zone. In applying this guide, the boundary of the population center shall be determined upon consideration of population distribution. Political boundaries are not controll-ing in the application of this guide. Where very large cities are involved, a greater distance may be necessary because of total integrated population dose consideration. (b) For sites for multiple reactor facilities consideration should be given to the following: (1) If the reactors are independent to the extent that an accident in one reactor would not in1tiate an accident in another, the size of the exclusion area, low population zone and population center distance shall be fulfilled with respect to each reactor individually. The envelopes of the plan overlay of the areas so calculated shall then be taken as their respective boundaries. (2) If the reactors are interconnected to the extent that an accident in one reactor could affect the safety of operation of any other, the size of the exclusion area, low population zone and population center distance shall be based upon the assumption that all interconnected reactors emit their postulated fission product releases simultaneously. This requirement may be reduced in relation to the - degree of coupling between reactors, the probability of concomitant accidents and the probability that an individual would not be exposed to the radiation effects from simultaneous releases. The applicant would be expected to justify to the satisfaction of the Commission the basis for such a reduction in the source term. (3) The applicant is expected to show that the simultanous operation of multiple reactors at.a site will not result in total radioactive effluent releases beyond the allowable limits of applicable regulations. .__a
s,. ~ ATTACHMENT B RESTATEMENT OF 10 CFR 20 SEC. 20.403 Sec. 20.403. Notifications of incidents - (a) Imaediate notification. Each licensee shall immediately report any events involving byproduct, source or special nuclear material possessed by the licensee that may have caused or threatens to cause: .(1) ' Exposure of the whole body of any. individual to 25 rems or more of radiation; exposure of the skin of the whole body of any individual of 150 rems or more of radiaton; or exposure of the feet, ankles, hands or forearms of any individual to 375 rems or more of radiation; or (2) The release of radioactive material in concentrations which, if averaged over a period of 24 hours, would exceed 5,000 times the limits specified for such materials in Appendix B, Table II; or (3) A loss of one working week or more of the operation 'of any facilities affected; or (4)~ Damage to property in excess of $200,000. (b) Twenty-four hour notification. Each licensee shall within 24 hours of dis-covery of the event, report any event involving licensed material possessed by the licensee that may'have caused or threatens to cause: (1) Exposure of the whole body of any individual to 5 rems or more of radiation; exposure of the skin of the whole body of any individual to 30 rems or more of radiation; or exposure of the feet, ankles, hands, or forearms to 75 rems or more of radiation; or (2) The release of radioactive material in concentrations which, if aver-aged over a period of 24 hours, would exceed 500 times the limits specified for such materials'in Appendix B, Table II; or (3) A loss of one day or more of the operation of any facilities affected; or (4) Damage to property in excess of $2000. (c) Any report filed with the Commission pursuant to this section shall be pre-pared so that names of individuals who have received exposure to radiation will' be stated in a separate part of the report. ~(d)-Reports made by licensees in response to the requirements of this section must be made as follows: (1) Licensees that have an installed Erergency Notification System shall make the reports required by paragraphs (a) and (b) of this section to the NRC Operations Center in accordance with g 50.72 of this chapter. ~(2) All other licensees shall make the reports required by paragraphs (a) and (b) of'this section by telephone and by telegram, mailgram, or facsimile to the Administrator of the appropriate NRC Regional Office listed in Appendix D of this part. Note: Item (b)2 of Sec. 20.403 refers to Appendix B; Table II refers to concentrations in air and water above natural backgrounds. This table is not reproduced due to its length.
l ATTACHMENT C RESTATEMENT OF SECTION 20.105 Sec. 20.105. - Permissible levels of radiation in unrestricted areas. (a) There may be included in any application for a license or for - amendment of a license proposed limits upon levels of radiation in unrestricted areas resulting from the applicant's possession or use of radioactive material and other sources of radiation. Such applications should include information as to anticipated average radiation levels and anticipated occupancy times for each unrestricted area involved. - The Connaission will approve the proposed limits if the applicant demon-strates that the proposed limits are not.likely to cause any individual to receive a dose to the whole body in any period of one calendar year in excess of 0.5 rem. (b) Except as authorized by the Commiesion pursuant to paragraph (a) of this section, no lb.ensee shall possess, use or transfer licensed material in such a manner as. to create in any unrestricted area from radioactive material and other sources of radiation in his possession: (1)-Radiation levels which, if an individual were continuously present in the area, could result in his receiving a dose of excess of two milli-rems in any one hour, or (2) Radiation lefels which, if an individual were continuously present in the area, could result in his receiving a dose in excess of 4 100 millirems in any seven' consecutive days. (c) In addition to other requirements of this part, licensees engaged in uranium fuel. cycle operations subject tothe provisions of 40 C. F. R. Part 190, " Environmental Radiation Protection Standards for Nuclear Power . Operations," shall comply with that part. f'- i s 1 r.---, y -,x,-- -y-mn.. ,,-g-,w, -~~,- - -, ~ v w,- -e--,-- w -~~~-w-e-na,- =~--=w-- ~ * " - ~ + * ' -w ='"* *^~*}}