ML20037A511
| ML20037A511 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 11/29/1976 |
| From: | Gossick L NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) |
| To: | Guyer T HOUSE OF REP. |
| Shared Package | |
| ML20037A518 | List: |
| References | |
| NUDOCS 8003050903 | |
| Download: ML20037A511 (18) | |
Text
-
's c
..-w c[4 Distributic I
~
(
i 4
9 Docket File f " IE (3) l NRC PDR Gecv NL1 ";;Hity (3).
local PDR H. Shapar NRR Reading D. Crutchfield LWR 4 Reading H. Denton B. C. Rusche P. Heineman E. G.' Case V. Stello NDY 2 J E;'S R. S. Boyd gg {3 Cacer teo. 50-346 R. C. DeYoung D. B. Vassallo
~S. Varga M. D. LynJ1 M. Service Tr.e Lcr. oracle Dnnyson Guyer united house or 8epresentatives G* Ertter (01053)
P. GroH naanirw3 on, D. C.
20515 t
L. Dreher
..p' '- - e c. Aig h ~e m~,
Dear Congrea man Guyer:
J. Yore, ASLB tour letter datro ccrocer 21, 1976, to ar. necanera.I. nm.ca,
~oirector of tue of fice e,f'auclear F.aterial sardty sed Jafeguaruc, haa aeen rererred to tnic orfice -for ' reply.
In your letter, yac requect tnatwe consiaer tne matters contairvo in a'lerter fro.
Linaa carr ard tnat we provide you sitit relevait infor.Lticn re<garuing '
rne saroty ceasures incoraorates into nuclear powr pinnes.
f.3. C;.rr inoicarr4 in iter letter tnat ner interest in nucle.;. cafety :stter; arow tro a recent sancuncement regarcing tJin propewa acaition ci-two nuclear units at ene site of tac :.;2Viu-cesae-nuclear i. owr plac.t
- w. ten is premntly ceing constructeu.
I oulu lirS to aMresa eacn or rne ap>cific 'ecncerns i, entifies vy r.n. Carr out tir.:t I vcul, 1 ice to cr iafI f aincuaa ta.5 Cavis-b u -
f ac il i ty.
Inis nuclear ; owc plant is locat u in norta central m ia cn tua soutusestern enera of La::e Erie in Ottawa County, Onio, x.ct
.I miles east of Tejeao, Ooio. me facilitg in seing constracr% urs t Coastruction Per:ait.%. cme-oO inscea by tne nuclear Be3ulator /
Canission.(d50) on Marco 24,,1971, a.ne sita.consista.or awur suu acres,.2., cat of vaica is flat cnd ic unuse.r.ar sa l ano.
'ne facil ity, iwntifice as Davis-3acce 1, irrorporates a tur-loop prassur (:N varer naclear ' reactor' ror its ste sa suoply syste a wnich teeas tne teraine generatc,r.
. m o net electrical output i;. swer 9uo aesasarts (n,#).
v;vic-desse I will ce jointly oea ay tw 21a.o Liaca cc.:cany ane rne Clevelard 41octric Ill uaiaat ing C%;anj.
A cor...etailea cc crir.eien of tne seire-Lane fr.cility :.as tac siro is contaim.i in tr.e vinal afory.mly tc Fesorr '(PSR) amittN oa :. area 31, IkH, 7/ tos p>rairtees fuentifla a:xw.
- t. copy c: to MM caJ r.i ) ot;sr accusara ur taint:9 tu.evf L-ene 1.rc avail;.nle rer ir.;g-ter im ny Im.cr.,
of t..e %;.olie at ti.o Iva cu w P::Asc Liurery, J1) 1..!=a atreur, rcr t cl intcn, Cnio, r u at raa m C'u Paulic cocc e t t w.. i wateu at i /17 a Strmr, a....,..asaing tm, v.
0 9
O D<I g
1 1
u vEA1y I
~
'0 T g soososo903 4
=.
?
w -
w 7
i w
y
~r'
-o
( -
- .y g.l..
2 ine ficnorr.ble 'Iennyson Guyer -
- c. Carr fuentifies eignt concerns in.ner letter and inquires wnotner scx::e aiditional action regarcing safWy measures can ce tar.cn.
Since i::y response to her concerns and her c;uestion on safety !<easures is extensive, I nave written it ad an enclosure to tais letter.-
tnat tne ciscussion of tne-I-Ex1feve.
your request.
- issues in tne enclosure is responsive to I nave enclonec a copy or yo'Ar letter requesting information nr.J a copy of your constituent's letter.
Sincerely,
($!gned) Lee V. Cossisk Lee V. Gossich Executive Direciar for Operations i-ucl;mrr-a:
1.
Discussica or tae Ccccern:
or~ Lirna Carr' Incceing 1tr fr'oc Concrees.m Jayer 2.
3.
Incocing.)rr.f rcn LimasCarr D
DJ JL v
~
o [ig m Y
3 3
LhR 4
~'
h'P. lr AD/LWP MDLynch/ red Ea;;ga DBVassallo/
11// 7/76 11 ?/76 fe11f4 /76
[
.D.. 0 ELDDkDP5
{ f :. R Q i D:NFIG EDO CA
.,, c s FeAH pShoyY i.E_C_a o
B h
11/1 /76_ _11/p/76
)1/M,4-
- CRu _ fe IXGos. i.c_.k._..__
l 11(476 11_/p//76 11/f7/76 76 om,
Forsa AEC.318 (Rev. 9 5)) AEC.A 0740 W u. s. novsawaernt one=Tme'orasce es74.sas.see
(
m
- ~
-e a,
,p i
t e
4 Docket to. 50-346 he Honorable Tennyson Guyer United States l'ouse of Representatives washington, D. C.
20515
Dear Congreav an Guyer:
1 Your letter cated Octooer 21, 1976, to Mr. Kennetn R. CNoran, Director cf tne office of. tiue3 ear ttaterial Safety and 2teguards,
':ms rseen referres to_ rnis offica for reply.
In your letter, you regaest that we considet the~ natters contained in a _ letter frca
..Linaa Carr ard tnat wa,provice you witn relevant internation' regsruing the safety iaeasures incorpcrated into nucler.r power plants. Ms. Carr inaicateo in ner letter tnat her interest in nuclear safety if.atters i
i arose froa a recent announcement regarding tne proposed acuition of two nuclear units at ta site of tne Davis-basse nuclear cwer pli.nt walen is presently ceing constracted.
i I woula 114e to accress eacn of tne n;:ecitic concerna ide ntiflea by ns. Carr cut first I would like to briefly discuss tne Davia-Je:a facil ity. bis nuclear power plant is locatN in nurtn central O!io on tria scutnwestern shore of Lake Cr le in Ottwa County, Unio, asout 21 miles cast of toleco, onio. D e facility is neity constructed cruer ConstrucMcn Fermit i40. CITR-80 icsuad oy tne Nuclear m ulatory Comission Dikc) on naren 21, 1971.
T.e site ecosists of acout M ceres, most of which is flat and is unard :aarsalanc.
me-facility, ioentifica as Davis-Lasse.1, incorporates a two-loop pressur izea water nuclear reactor for its steam sugoly system unien f%s tae turoine generator.
lIhe net electr,ical output is c?>3ut Wu mawatts (ima).
Davis-Basse 1 will x Jointly owneu oy tne TolNo Ecison Cccpany ana the Cleveland Electric Illtssinating Con.cany. A r. ore uetaile; cescr iption of the Davis-casse facility ux3 tne site is containad in tne Final l
Safety Analysis Report (FS,AR) sunaitted on Haren 31, 137.5, of tae peruittees laentitieo aoove. A co;.y of tae PJM and all otner occut.ents pertaining.to osvis-basse 1. ate avaiicole ror inspoetion ny ca w ra j
of the public at the Ica Rapp Pub)ic Licrary, 31J.coicon Street, lort Clinton, a.io, ana at tw 4GC's A:a3 ic oca.pnt eg i atogt 1717.I street,.i.ri., Wasnington, D. C.
DJ DJ -
T R!
1LUL o F FIC s W summaus h Dats)
Form AEC SIS (Rev.9.$3) AZCM 0240 W v. s. eovsnmu sm? pamvmo orrscs: sera.sas. nee
o DiCIDSURE DISCUSSION OF THE CONCERNS-REGARDING*THE SAFETY OF NUCLEAR POWER PLANTS A number of concerns regarding the safety of nuclear power plants are identified in Ms. Carr's letter.
These are:
(1) training of nuclear power plant personnel; (2) construction procedures for nuclear power plants; (3) inspection of nuclear power plants; (4) effectiveness of safety systems; (5) statements of three General Electric engineers regarding nuclear safety; (6) " accidents" in 1974; (7) dispersal of fission products following postulated accidents; and (8)_ the effect of nuclear power plants on public health. Ms. Cart then inquires whetner safety measures can be improved in view of her stated concerns about the safety of nuclear power plants. Each of the eight concerns stated above are briefly discussed in this enclosure and, where appro-priate, references are cited where a more detailed discussion can be found.
Finally, the safety measures required by the Nuclear Regulatory Ccmission (NRC) for nuclear power plants are discussed.
1.
Training of Nuclear-Power-Personnel In Ms. Carr's letter, the training of personnel is described as
" poor." While she does not specify whether this allegation is with i
respect to the design and construction phase or the operating phase, tne latter will be discussed in this section of the enclosure.
Tne training of construction personnel will be discussed in the following sections of this enclosure.
Recognizing the need for trained, capable, and experienced operating personnel for nuclear power plants, the NRC has issued regulations requiring that applicants for operating licenses submit information concerning their organizational structure ar.3 operating personnel qualifications. This information is reviewed in detail by the NRC and must De found acceptable pr for to issuance of an operating license. Ib provide guidance to applicants on this matter, the NRC has endorsed in Regulatory Guide 1.8, " Personnel Selection and Training," (Attachment 1), the criteria for the selection and training of nuclear power plant personnel contained in ANSI N18.1-1971, " Selection and Training of Nuclear Power Plant Personnel." (See footnote 1)
The AdSI standard cited above is one of many such standards published by the American National Standards Institute.
Footnote 1.
Copies may be obtained from the American Nuclear Society, 244 East Ogden Avenue, Hinsdale, Illinois 60521.
7
. Since this ANSI standard on the selection and training of operating personnel is quite extensive, a sumary of its salient features is provided. We standard establishes the minimum qualifications and training for all functional levels of the operating organization, including managers, supervisors, professional-technical, and operator-technicians-repairmen.
A taole is included (Attachment 2) which summarizes the required experience, academic training, and the NRC operator licensing requirements for all functional levels of operating personnel. Wese requirements are designed to provide assurance tnat the nuclear power plant operating personnel: (1) will be capaole of safely and efficiently operating the facility; (2) will understand tne complexities of the plant design; (3) will be capaole of properly manipulating tne plant controls; and (4) will maintain and repair the plant equipment in an acceptable manner.
Before any operating personnel can manipulate the controls of an operating nuclear power plant, they must obtain a 1icense from the t&C which authorizes them to operate a specific nuclear power plant, m e requirements for obtaining an operator's license are contained in 10 CFR Part 55, " Operator's Licenses" (Attachment 3).
Section 55.11 of Part 55 states in part that an applicant for an operator's license will be approved if the NRC finds that the individual is in good healtn and has passed a written examination and an operating test to determine that the applicant has learned to operate a specific nuclear power plant.
The scope of the senior operator written examination ~ is described in Sections 55.21 and 55.22 of 10 CFR Part 55 and covers 21 different aspects of reactor operation, ' including: (1) fundamentals of reactor theory; (2) general design features; (3) general operating characteristics; (4) conditions and limitations in the facility license; and (5) fuel handling facilities. While this list is not.
complete, it does indicate that-the operating personnel are required to demonstrate a broad, in-depth knowledge of all aspects of an operating nuclear power plant. Additionally, the operators of a t
facility are required to demonstrate an understanding of 12 separate matters, including: (1) the required manipulation of the console.
controls; (2) tne use and function of tne facility's radiation' monitoring systems; and (3) the emergency plans for the' facility..
t
l
. In su.cary, the operating personnel of nuclear power plants are carefully selected, intensively trained in a broad range of nuclear power plant operations, and carefully retested periodically by the NRC to determine that they can safely operate the facility for vnich they are licensed.
2.
Construction Procedures'for Nuclear Power P1 ants Ms. Carr indicates in her letter that she considers the construc-tion of nuclear power plants to be faulty.
In response to this statement, a d scription of tne metnod by which construction procedures are selected and approved for nuclear power plants is provided.
In the' following section on the inspection process, a diseassion on how assurance is provided that these construction procedures are followed, is also provided.
In the two-stage licensing process for a nuclear power plant, the review conducted by the NRC staff at the construction permit stage is directly concerned with the acceptability of the "... proposed design of tne facility, including, but not limited to, the principal architectural and engineering criteria for the design..." (See footnote 2). The minimum requirements for the principal design criteria for water-cooled nuclear power plants are contained in Appendix A, " General Design Criteria for Nuclear Power Plants,"
of 10 CFR Part 50 ( Attachm3nt 4).
Some of tne more important and essential elements of the principal design criteria for a nuclear power plant are the proposed construc-tion procedures. Many of taese procedures are contained and codified in well-known design and construction coces which have been published and acopted by national societies and institutes. A few examples of these codes are:
(1) the ASME Boiler and Pressure Vessel Code published by the American Society of Mechanical Engineers; (2) the ACI codes published by tne Acerican Concrete Institute; (3) the AISC Code for Steel Structures published by the American Institute for Steel Construction; and (4) the codes and standards of the Institute of Electrical and Electronic Engineers. This listing illustrates that there are widely used standards and codes which can be adopted to provide assurances enat acceptable construction proceaures are followed during the construction phase of nuclear power plants..These national stancards have been used in many different types of construction projects for many years and have provMed acceptable protection of the public health and safety.
Footnote 2.
Refer to Section 50.35(a)(1) of 10 CFR Part 50.
In addition to the national standards discussed above, the rmC has established additional guidance for applicants for 1icenses to construct and operate nuclear power plants. mis guidance is contained in the IEC Regulatory Guides (Attachment 5) which are issued to either supplement the national standards or to provide guidance for acceptable design criteria wnen the national standards are not sufficiently conservative for certain aspects of tne g
construction of nuclear power plants.
Where possiole, the IEC staff issues design criteria for specific systems or components of nuclear power plants. mese specific design criteria are contained in Branch Technical Positions and reflect the knowledge obtained from the review of previous appli-cations for licenses and from the operating experience of licensed nuclear power plants.
Finally, the staff establishes ac hoc design criteria for tne con-struction of nuclear power plants when the existing national standards, regulatory guides and Branch Technical Positions do not address unique design features of a proposed facility.
^
3.
Insoection of Nuclear Power-Plants One of the concerns of Ms. Carr is with respect to the adequacy of inspection of nuclear power plants. m is is a very important e4:ent in the safety of nuclear power plants since the adequacy of construction and the safety of operation of nuclear power plants depends directly on the quality of the inspection process.
Recognizing this importance, the IEC has issued regulations requiring the establishment of a "... Quality Assurance (QA) program to be applied to the design, fabrication, construction, and testing of tne structures, systems, and components of the facility)." (See footnote 3) The requirements of this QA program are contained in Appendix B, " Quality Assurance Criteria for Nuclear Power Plants ad Fuel Reprocessing Plants," to 10 CFR Part 50.
Tne IGC staff reviews both the proposed QA program of applicants A
and tne QA organization which will implement the QA program. Prior to authorizing construction of the safety related structures, systems or components of a nuclear power plant, the staff must find botn tne applicant's CA program and QA organization acceptaole. However, tne !EC staff's involvement does not end here. Tne Office of Inspection-and F.nforcement (OIE) periodically conducts both scheduled and unannounced field inspections of the implementation of the applicant's QA program and those of his contractors and suppliers. This field Footnote 3.
Section 50.34(a)(7) of 10 CFR Part 50.
- 5-inspection starts about nine months prior to the submittal of an application for a construction permit and continues throughout the construction phase, tne preoperstional test program and the operating lifetime of the facility. mese field inspections by OIE during Q
the construction phase are extensive and cover: (1) a review of the applicant's QA performance, including audits of the applicant's QA records and documentation; (2) a witnessing of the construction practices and an inspection of the facility at various stages of construction; and (3) a review of the qualifications and training of the construction personnel as.well as those of the quality assurance and quality control (QVQC) personnel. Se review of the qualifications and training of the OVQC personnel is conducted for all construction personnel at the site, including the specialized subcontractors, and at the manufacturing facilities of the vendors and suppliers.
Tne NRC QA review can best be visualized as a multiple tier process involving the entire QA program. To grossly simplify the process, QC personnel for each subcontractor and vendor review the gaality of the design art fabrication. In turn the QA personnel of the subcontractors and vendors review the adequacy of the QC review.
The applicant's QA personnel tnen review the adequacy of the QA review oy the subcontractors, vendors, and that of its primary e
contractor.
Finally, OIE reviews the implementation cf the entire QVQC program by all parties in the design and construction priases.
Chile any inspection process is potentially fallible due to the human element, the QA programs for nuclear power plants descrioed above minimizes the possibility that gross negligence and/or incompetence would cause nuclear power plants to be con-structed with significant deviations from the design criteria for these facilities.
There nave been a nurcer of nuclear power plants which have been designed or constructed in a manner which represented significant departures from tne design criteria fcr these plants. The installation of tne safety-related electrical cables in the Davis-Besse 1 facility is an example of this. However, tne faulty construction procedures in this instance were discovered by OIE personnel implementing the QVQC program in tne manner descriced above, at least two years prior to any decision to issue an operating license for this facility.
Recognizing tne inherent limitations of any inspection process, the NRC does not assume that all deficiencies in ettner the design or installation of tne safety-related structures, systems or components will be found by tne QA program. Accordingly, the NRC requires applicants to conduct an extensive preoperational test program for each system and component.
For example, diesel generators for the
a
. on-site power supply of a nuclear power plant must undergs hundreds of cold startup tests to demonstrate the adequacy of their ' design and installation and certain important safety-related structures are proof tested to demonstrate their adequacy.
While the foregoing discussion is primarily oriented towarcs the design and construction phase, a similar inspection program is conducted tnrougbaut the operating 1ifetime of a nuclear power plant, including visual inspections, design reviews and periodic testing.
In su. mary, the inspection process for nuclear power plants is an extensive, in-cepth, continuing program whose objective is to n.inhize numan error wnich might adversely affect tne design, construction and operation of these facilities.
Its success can only be gauged by the operating history of tne licensed operating plants. mis matter is discussed in Item 9 of this enclosure.
4.
Effectiveness of Safety Systerrs Ms. Carr alleges in her letter that nuclear power plants have unproven safety systems. The response tc tnis concern involves two principal considerations. Tne first of these is the design philosophy for safety-related systems while the second is concerned with experimental verification and testing.
Se preceding discussion in Item 2 descri'oed the review and acceptance of the design criteria for nuclear power plants. 21s review includes a determination by the tac of "...
the adequacy of structures, systems, and components provided for the prevention of accidents and the mitigation of the consequences-of accidents." (See footnote 4) ne review of the emergency core cooling systems (ECCS) of water cooled reactors is an example of one of these systems wnich are analyzed and evaluated by tne NRC.
Assuming a loss of coolant accident (IOCA), the proposed ECCS is analyzed in accordance with very specific requirements contained in Section 50.46 and Appendix K of 10 CFR Part 50. Rese design criteria for safety-related systems in general, and of the ECCS in particular, were developed after extensive design, review, and testing. As an example, the public hearing at wnich the acceptance criteria for ECCS were developed lasted for aoout 1 year and 'includeo extensive testimony and cross-examination by the NRC staff, tne electric utilities, tne reactor vendors and numerous intervenorp representing various public interest groups.
It should be noted that the final acceptance criteria which were adopted by'the NRC included a considerable amount of conservatism to provide relatively large safety factors in establishing the design of safety-related systems.
Footnote 4.
Section 50.34(a)(4)(ii; of 10 CFR Part 50.
D 4 S e NRC does not assume.that the adoption of conservative design criteria established for safety-related systems is the only metacd of providing assurance that safety systems will function as requitad in the event-of an accident.
Accordingly, a multipl4 ~ step test program is conducted, including the the extensive precperational test program and the periodic tests performed during tne plant operating lifetime as discussed in Item 3 of this enc 1csure.
Finally, an extensive ECCS testing program is continuing at the National Reactor Testing Site (NRTS) in Idaho to descnstrate the conservatism of the NRC acceptance criteria. mis tasc program will not only establish a proof of principle for ECCS, ic vill also permit a detailed verification of the analytical mocels for the fluid flow, mass transfer and heat transfer characteristics of water-cooled nuclear power plants.
In sumary, the conservative design ~ criteria, the extensive preopera-tional test program, the periodic testing, and the test program at NRTS all provide assurance that the safety systems will function properly in the event of an accident.
5.
Statement of Eree General-Electric-Engineers Regarding' Safety of Nuclear Power P1 ants Ms. Ca r refers to a statement by three nuclear engineers to the effect that "... a major atomic power plant disaster is 1ikely within twenty-four years." It is assumed that this statement is attributed to the three engineers who resigned from the General Electric Company (GE) on February 2, 1976. mese engineers are Mssrs. Dale G. faridenbaugh, Richard B. Hubbard, and Gregory C. Minor. These engineers subsequently testified before the Joint Comittee on Atomic Energy on February 18, 1976; the written version of their testimony consists of 72 pages.
Recognizing the responsible nature and considerable experience of these three GE engineers, the Joint Comittee conducted five days of hearings on tnis matter.
Se NRC likewise recognizeo the excellent qualifications of these three GE engineers and prepared a detailed response contained in
- r. 334 page report which was submitted to the Joint Comraittee on Maren 2,1976. Since tha testimony of the three GE engineers and that of the NRC is extensive, it shall not be sumarized. However, it should be noted that all of the testimony cited above is contained in the two-volume record of these hearings.
Se value of.the statements made by the three GE engineers can be oest evaluated by reviewing the various public investigations conducted regarding the charges which have been made relating to nuclear reactor safety.
i
,~.. -.,
- In the instance cited above, the Joint Comittee conducted an open hearing, with the media present, and published a hearing record containing about 1800 pages of testimony. All interested parties wno could make a significant conttibution were heard, including the three GE engineers, the NRC Comissioners, memoers of the Advisory Committee on Reactor Safety (ACRS), representatives of GE, representatives of the Consolidated Edison Company, Robert Pollard who is a former employee of the NRC, me sers of the NRC staff, and members of the NRC Licensing Board and the Appeal Board.
It may be concluded from the lengthy testimony, the large list of witnesses and the varied toples in = :cention, that the question of the safety of nuclear power plants had an excellent presentation before the Congress and, through the media, directly to the American puolic. m is legislative review of the charges, and overview by the Congress of the performance of the NRC, permits the Congress and the public to become informed on the subject of nuclear safety.
In addition to the legislative review discussed above, all of the -
charges brought by the three GE engineers, and by Mr. Pollard, were brought before the ACRS for a detailed technical review. mis committee is composed of members who have considerable experience and expertise in the field of nuclear safety and are thereby well qualified to conduct an in-depth technical review of all the charges made by the four individuals cited above.
mere is also an additional forum whereby contentions regarding the safety of nuclear power plants can be thoroughly aired. m is forum is the public nearing (an administrative procedure), which must be held prior to issuing a license to construct a nuclear power plant.
(Refer to Item 8 of tnis enclosure for a discussion of this hearing process.)
In su:anary, there has been a thorough legislative, technical, and administrative review of the charges made by the three engineers cited by Ms. Carr. While the individuals involved have excellent qualifi-cations, they have not persuaded either the Congress, the ACRS, the licensing coards or the courts that their charges have sufficient merit to warrant either:
(1) a shutdown of the presently cperating nuclear power plants; or (2) a moratorium on the issuance of licenses to construct and operate additional nuclear plants.
. 6.
, Incidence of Abnormal Occorrences During 19_74 In her letter, Ms. Carr alleges tnat d... In 1974 there were 1,421 accidents at the operating nuclear power plants in the country."-
It is assumed that Ms. Carr is referring to what were previously identified as " abnormal occurrences." If she is, there were 1 566 abnormal occurrences reported to the NRC during 1974 by operating nuclear power plants.
In order to place this topic into proper perspective, it is neces-sary to understand what was meant by an abnormal occurrence in 1974, and what is meant by that term now. First, it is emphasized tnat an aonormal occurrence was not necessarily an accident. In fact, an aonormal occurrence was very rarely anything resembling an accident. Recognizing tnis fact, the NRC has changed the nane of an abnormal occurrence to a reportable occurrence. 'Ihis action was taken by the NRC pursuant to the implementation of Section 208 of the Energy Reorganization Act of 1974 which established the NRC.
Under Section 208, tne NRC is required to report to the Cor,gress every quarter, a listing of each abnormal occurrence defined in that Section as "...an unscheduled incident or event which the Commission determines is significant from the standpoint of public health or 57fety." Tne NRC, accordingly, now reviews in detail every reportable occurrence contained in the Licensee Event Reports (LER) to determine which of these will be categorized as abnormal occurrences as defined above and, therefore, reported to the Congress.
In July 1975, the NRC developed interim criteria (Attachcent 6) for evaluating the reportable occurrences contained in the LER's to determine whether they are significant from the standpoint of public health and safety. These interim criteria define as significant (and reportable), events involving an actual loss of protection, or a major reduction in the degree of protection, provided for the health and safety of the public.
For each facility licensed under 10 CFR Part 50, the Technical Specifications for this facility set forth requirements for reporting to the NRC in 1974 any information concerning " abnormal occurrences" (now identified as reportable events) that took place In general, these events were those that caused, or threatened to cause, a condition affecting safe operation of the facility. A l
l I
t
pr+cise definition of what constitutes an " abnormal occurrence" at a given nuclear power plant in 1974 was set forth in the Itchnical specifications for that facility and emphasized that the NRC's interest and concern in requiring reports of these events is associated with significant occurrences only. However, since the significance of a given event may not be clear, and because of the inherently subjective nature of assessing significance, the NRC required that tncse occurrences that may have significance with respect to safety, also ce reported. This led to increasingly stringent reporting requirements and consequently, to a rapid escalation in the number of repcets sutnitted by licensees.
Both in 1974 and in the present, the NRC collects and evaluates tnis cperational information concerning licensed nuclear facilities to assess safety, and to form the basis for comparing plant performance with the design objectives. These reportable occurrences are incidents or events that involve system, component or structural failure or malfunction, personnel error, design deficiencies, management defic,iencies, and other matters that are related to plant safety in various ways. Because of the multiple levels of protection, or " defense-in-depth," including the provision of redundant safety systems and components, such events do not, in general, affect i~
safety directly and do not have an actual impact or consequence on the health and safety of the public. However, information regarding these events is useful to the NRC and to the nuclear industry in their efforts to improve safety.
Therefore, these events are brought to the attention of the NRC through a variety of reporting reqairements or by NRC inspection, and appropriate enforcement and corrective measures are thereafter taken.
With the understanding of what constitutes a reportable event and an abnormal occurrence and what is the objective of the NRC in requiring the submittal of LERs, the " abnormal occurrer.ces" of 1974 will now be discussed. Of the 1566 reportable events in 1974, a very small number of these incidents had any direct significance to the public health and safety.
In fact, the vast majority of these incidents were of no significance in and of themselves, but were required to be reported by the licensees so that any evolving pattern of similar occurrences at several or many plants may be detected, and remedial action taken, as soon as possible. Tne NRC reported in its first quarterly report to the Congress on abnormal occurrences, NUPEG 75/090, " Report to Congress on Abnormal Occurrences, January - June 1975,"only six reportable events which occurred in 1974 and wnich were subsequently categorizea oy the NRC as abnormal occurrences.
.. Tne NRC's responsibility to keep the puolic informed regarding nuclear safety does not begin and end with its reporting requirements to the Congress. Currently, License? Event Reports, containing every reportable occurrence, as well as press releases by NRC regional offices and by licensees, are placed in about 122 local NRC Public Document Rooms every 2 weeks, as well as in the Washington, D.C., NRC Puolic Document Room. Lastly, NRC issues each month a
" Grey Book," which includes all Licensee Event Reports and shut-down information.
In su.raary, the " accidents" to which Ms. Carr refers are the reporting of minor events by licensees in compliance with the Technical Specifications of the operating nuclear power plants and do not constitute accidents in the usual definition of this word. %ey have no significance in regards to public safety other than to allow tne NRC and tne nuclear industry to correct minor deficiencies in the design, operation, and maintenance of nuclear power plants.
7.
Dispersal of Fission Products-Following Postulated Accidents In her letter, Ms. Carr alleges that "The accidental destruction of a nuclear power plant's core could be deadly up to ninety miles away, depending on weather conditions." While the basis for Ms. Carr's statement is not known, it is ar3umed that she is referring to an estimate made by extrapolating the information contained in hch-nical Information Document 14844, dated March 23, 1962. It is important to note that the calculations described in this document are considered to yield conservative, upper-bound distances since the analytical model on which this document is based, assumes no engineered safety features.
However, there are specific design features which are an integral t
l part of nuclear power plants and whose design basis assumes that I
tnere is a release from the reactor pressure vessel of the fission products contained in the nuclear core. Tnis assumption is made on a deterministic basis (i.e., no rational mechanism is assumed to be required to obtain this release) so as to impose extremely conservative design conditions on the engineered safety measures wnicn are physically incorporated in the power plant to mitigate the consequences of any postulated accident. m is release assumption is in accordance with the requirements of section 100.11(a) of 10 CFR Part 100 (Attachment 7). How=ver, this assumption implies i
that there is a catplete failure of the safety systems which are specifically designed to prevent this release of fission products I
from tne reactor core. mis method of designing safety systems to l
witnstand postulated worst case accidents, then assuming a failure i
of tnese systems and designing physically separate backup systemt, whien are diverse in principal, is known as " defense-in-depth."
l
- Some of the engineered safety systems which are _ typically incor--
porated into the plant design and which mitigate the consequences, of the. postulated accident are the prir 'ry containment, the secondary containment,' containment sprays, and uarcoal ' filters. Prior to licensing a nuclear p wer plant,' the NRC staff is required to demonstrate that the individual doses received by the public -
at specified distances from the facility following the design
- basis accident (i.e., the fission product release frora the reactor pressure vessel) are within the guideline values contained -
in 10 CFR Part 100. Taese specified distances are identified as the radius of the exclusion area and. the radius of the Jow population zone. Typical values of tnese distances are about 1/2 mile for the exclusion area and about 3 to 5 miles for the low population zone.
Tnese distances vary with plant site and are dependent on the power level of a facility, the engineered safety features, and the pertinent meteorological conditions of the plant site.
In the event of an accident at a nuclear power plant, the emergency plans, might require an evacuation of the public witnin the low popu-lation zone and possioly beycad the low population zone, depending on the severity of the accident.
However, Ms. Carr's allegation that the release.of fission products would be deadly up to ninety miles away must be considered incredible.
For her statement to.
i De crediole, one must assume a ccaplete failure of all safety systems designed to protect the nuclear core and a failure of both the primary containment in combination with tne highly unlikely postulated design basis accident (DBA). For the design of the containment the postulated DBA is usually a complete break in the largest pipe of the primary coolant system.
In sumary, the pnysical systems incorporated into the design of nuclear power plants as part of the " defense-in-depth" concept provide assurance that the hazardous consegaences of any fission product release from an operating nuclear power plant will be limited to a short distance from tne plant. Further, the eraergency plans for a nuclear power plant provide assurance that the public can be safely evacuated from tais limited area if a severe accident does indeed occur.
d
_ ~.
. 8.
Effect of Noclear Power Plants on'Public Health Ms. Carr alleges that:
"It has been proven that there has been an increase of infant mortality, cancer, and other health problems near nuclear plants." This statement is not accurate. A number of individuals with a scientific background have made such charges both in the press and in published books. However, these individuals have been unable to prove their charges in forums expressly establisned for this purpose. These forums include public hearings held pursuant to the regulations under which each nuclear power plant must be licensed.
At these public hearings, sworn testimony is submitted with subsequent opportunity for cross examination and rebuttal. The decision of the licensing boards on the. safety _ of a proposed nuclear power plant may oe then subject to an extensive appeal process which can, and has, led to a judicial review by the U. S. Supreme Court. Considering the extensive legal process to dich a proposed licensing action is suojected,_ Ms. Carr's content. ion that these charges have been proven cannot be accepted.
However, in response to her statement on public health, a discussion of ~
tne NRC's reqairements governing routine releases of radioactivity from an operating nuclear power plant is provided. The fundamental concept underlying the NRC requirements in this matter is identified as "as low as reasonably achievable." These requirements are contained in Appendix I to 10 CFR Part 50 which was published after an extensive public hearing on the subject. Since these requirements are fairly complex, they are sumarized in the following manner. Essentially, the routine releases (i.e., non-accident relatea) of radioactivity from an operating nuclear power plant must be held to amounts wnich result in whole body radiation doses to.the public through all possible-pathways (e.g., air-grass-cow-milk), that are a small fraction (about two to four percent) of the normal background radiation in our environment. Most experts in the field of radiation assessment oelieve this small incremental radiation dosage to be negligible.
l Indeed, greater incremental radiation dosages can occur from receiving l
diagnostic medical x-rays.
In sumaary, the NRC has established requirements that limit the radia-l.
tion releases from oparating nuclear power plants to amounts which result in incremental radiation dosages to the public that are a small fraction of the radiation which exists in our normal environ-ment and in our normal daily functions.
i, l
-l t
i
9.
Safety Measures-for Nuclear Power Plants Established by the NRC E_ Protect Puolic Healtn Prior to discussing the safety measures for nuclear power plants which have Deen established by the NRC to protect the public health, it should be noted that the nuclear industry represented by the electric utilities who operate the nuclear power plants; the architect-engineers wno design and construct them; tne reactor vendors who manufacture the nuclear reactors; the many diverse suppliers; and the numerous technical consulting firms, are all dedicated to tne task of protecting the public health.
Indeed, the prime responsibility for doing so is vested in their hands.
Re role of the NRC in the licensing process is to estaclish require-ments for nuclear power plants which are then issued as regulations contained in Chapter 1, Nuclear Regulatory Commission, of Title 10, Energy, of the Code of Federal Regulations (CFR). The NRC also estaolishes guidance in the form of regulatory guides, branch tech-nical positions and ad hoc design requirements (refer to Item 2 of this enclosure). Be ultimate responsibility of the NRC with regard to safety is to determine whether a proposed or operating nuclear poWr plant can be constructed and operated in a manner wnich will not adversely affect the puolic health.
E ensure that this objective will be acnieved, the NRC: (1) establishes and reviews qualification and training requirements for all operating personnel (refer to Item 1 of this enclosure); (2) establishes and reviews design criteria (refer to Item 2 of this enclosure); (3) establishes guidance for, and reviews, the inspection process for the construction and operation of nuclear power plants (refer to Item 3 of this enclost;re); (4) estaclishes acceptance criteria for the ECCS designed to protect the reactor core, including reviewing the detailed analysis of the ECCS, and reviews all safety-related structures, i
systems and components (refer to Item 4 of this enclosure); (5) l monitors the operating history of_ operating nuclear power plants (refer to Item 6 of this enclosure); (6) estaclishes requirements for, and reviews, engineered safety features designed to mitigate the consequences of postulated accidents (refer to Item 7 of this enclosure);
and (7) establishes requirements for, and continuously monitors, the routine releases of radioactive materials from operating nuclear power l
plants (refer to Item 8 of tnis enclosure). hhile this list is neither i
comprehensive nor complete, it does serve to illustrate the nature and enaracter of the effort taken by the NRC to ensure that appropriate safety measures are proposed, implemented, and maintained for nuclear power plants.
I l
i
. The entire licensing and regulatory process conducted by the NRC is also subject to technical, administrative, judicial and legislative review to determine that the NRC does indeed perform its primary oojective of protecting the public interest (refer to Item 5 of this enclosure). ibe excellent operating history of the civilian nuclear power plants in which not one death is attributable to accidents associated with the nuclear cores' of civilian nuclear power plants, cemonstrates the effectiveness of the safety measures proposed and implemented by the nuclear industry, reviewed and regulated by the NRC and reviewed again by the ACRS, the courts and the Congress. -
O t
_