ML19296F506
| ML19296F506 | |
| Person / Time | |
|---|---|
| Issue date: | 07/23/1980 |
| From: | Harold Denton Office of Nuclear Reactor Regulation |
| To: | Ahearne J, Gilinsky V, Hendrie J NRC COMMISSION (OCM) |
| Shared Package | |
| ML19250J298 | List: |
| References | |
| FOIA-82-93 NUDOCS 8010210583 | |
| Download: ML19296F506 (40) | |
Text
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JUL 2 3 E0 MENORANDUM FOR: Chairman Ahearne Commissioner Gilinsky Commissioner Hendrie Commissioner Bradford William J. Dircks De$ Willim y THRU:
Acting Executive Director for OperabM FROM:
Harold R. Denton, Director Office of Nuclear Reactor Regulation
SUBJECT:
COMPLIANCE OF NRC LICENSES WITH NRC REGULATIONS, REGULATCRY GUIDES, BRANCH TECHNICAL POSITIONS, AND LICENSEE COMMITMENTS On June 13, 1980 I transmitted a memorandum to you on the above subject.
The purpose of that memorandum was to advise you of the costs to the NRC of having to certify in detail how structures, systems and components of a nuclear reactor facility comply with each current safety-related regulation, regulatory guide and branch technical position. My memorandum discussed the difficulties of, and consequent resource needs associated with, comprehensive documentation of cmpliance with every aspect of each require-ment of the Commission's regulations governing nuclear reactor design and operation. On the other hand, my memo touched only lightly on the staff's overall assessment that our review provides adequate assurance that licensed facilities conform to the Commission's regulations. This dichotomy has raised a number of questions concerning our discharge of regulatory respon-sibilities with respect to assuring compliance with the Commission's regulations. As a result, I believe some amplification with respect to these questions is warranted.
As pointed out in my memorandum, our review is based on the Standard Review Plan (SRP). Each section of the SRP contains acceptance criteria which reflect the requirements in the Commission's regulations.
In those instances where the regulation is specific (e.g.,10 CFR 50.46 and Appendix K) the acceptance criteria in the applicable sections of the SRP reference the regulation and therefore the review does focus on explicit conformance with the regulation. In those instances where the regulation is stated in broad terms (e.g., General Design Criteria) the acceptance criteria in the applicable section of the SRP reference regulatory guides and branch technical postions generally related to the regulations and therefore the a.
review does focus on explicit confomance with these guides and positions.
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Thus, our review emphasizes an in depth evaluation of the principal systems
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a The Commissioners Background of the SRP The Standard 'ieview Plan was developed in the early 1970's at the time of substantial expansion of the technical review staff and a shift in review activities to a far more penetrating in-depth review of principal structures and systeas important to safety. Bis change in review included a substan-tial increase in independent technical assessment in such areas as design and performance of emergency core cooling systems, design and performance of containment and other accident mitigating safeguards, design and per-formance of radwaste systems, geology and seismology, and structural design techniques.
Although still an audit review in the sense that not every system and not every " nut and bolt" is explicitly evaluated, the expansion of technical review in the 1970's provided a far more comprehensive assessment by AEC-NRC of critical systems than had been conducted earlier by AEC.
As a result of expansion in staff and in depth of review, it was felt important to provide, for this newly expanded staff, clear guidance as to what was expected of the review that they were to conduct. The SRP was the written expression by experienced staff reviewers of the factors to be considered in properly reviewing particular syttems.
It is important to recognize that, although there is no explicit correlation between the SRP and the regulations, the experience upon which the drafters of the various portions of the SRP drew was their prior experience in revising appli-cations. For these prior reviews tne only criteria for judgment were those of the regulations as amplified by the GDC.
Imbedded in that judgment of " adequacy" by virtue of the experience of the reviewers involved in drafting the SRP is an assessment of conformance to the requirements of the GDC and other Commission regulations. Unfortunately for posterity and for the types of questions posed recently, but under-standable in light of the time and manpower pressures that existed in the early 1970's when the SRP was being developed, the chain of reasoning of the reviewers who drafted the SRP was not preserved. To develop an explicit technical basis relating the SRP to the Commission regulations is the activity of high manpower cost indicated in my memorandum of June 13, 1980.
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Congruence of the SRP and the Regulations i
i Each SRP is organized into four sections: areas of review; acceptance criteria; review procedures; and evaluation findings. The acceptance criteria contain a statement of the purpose of the review and the technical basis for detemining acceptability. The technical basis consists of specific criteric which typically include reference tc Part 50 and 100, and particularily the General Design Criteria of Appendix A, regulatory guides, codes and standards, and branch technical positions.
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The Commissioners While, as mentioned before, there has been no fully disciplined attempt to relate every rule and regulation to every applicable standard review plan or Regulatory Guide, we have made a post facto study of these interrelationst.ips and found, for example, that all but one General Design Criterion is specifically referenced in the SRPs. The only GDC not referenced explicitly in the SRP is GDC 51 " Fracture Prevention of Containment Pressure Boundary." Howt:yer, SRP Section 3.8.2, which should have included such a reference, does specify that the materials of steel contairnents or steel portions of steel and concrete containments be reviewed for compliance with Article NE-2000 of Section NE of the ASME Boiler and Pressure Vessel Code. Compliance with this section of the Code will in general provide assurance that the basic requirements of GDC 51 are met.
To provide some additional perspective as to the interrelationships between the GDC, the other regulations, the SRP's, and Regulatory Guides, a prelim-inary survey was made of these documents. Enclosure 1 to this memorandum provides a listing of:
1.
SRP sections where specific GDCs are referenced, listed by GDC (Table 1);
2.
Reg. Guides where specific GDCs are referenced, listed by GDC (Table 2);
3.
A cross-list of GDCs and regulations, listed by Regulatory Guide (Table 3); and 4.
A cross-list of regulations, GDC, and Regulatory Guides, listed by SRP section (Table 4).
In general, the degree of congruence between a regulation and the SRP for any particular set of structures, systens or components reflects the specificity of the regulation itself and any implementing guidance, the -
experience of the staff, and the influence and interests of the Corsaission, the ACRS, the Boards, the public, and the industry.
Therefore, although there are identification and documentation concerns as expressed in my June 13, 1980 memorandu:a, I am nonetheless confident that the link exists in fact so that when a license review is properly carried out in accordance with the SRP, such a review adequately assures conformance with the Commission's regulations.
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NRC FORM 318 (9-76) N RCM O240 D U S. GOV ERNMENT PRINTING OFFICE: 1979 289 363
The Commissioners SRP and Licensing Process Applicants for pemits and licenses have had fomal guidance on how to prepare their Safety Analysis Reports since June 1966, when a " Guide to tae Organization and Contents of Safety Analysis Reports was issued by the AEC.
h Revision 1 to this guide was issued in October 1972, as the " Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants."
The Standard Fomat was revised again in September 1975 (Revision T) and in November 1978 (Revision 3).
Since 1972 the staff guidance has required that applicants explicitly describe confomance with the General Design Criteria, (Appendix A to 10 CFR 50). Revision 3 (November 1978) was the first Standard Fomat to specify that appitcants should address conformance with Regulatory Guides, but most applicants have done this since about 1973. Most SARs since 1972-73 contain sections whic5 address the extent to which each GDC and, each applicable Regulatory Guide issued up to a time some months prior to the SAR submittal date is met.
Although justification for deviations from the Standard Review Plan has not yet been required to be explicitly documented, reviewers are expected to use the SRP as a guide in their review of all applications.
In cases of facilities suostantially constructed before the SRP was promulgated, the specific recommendations of the SRP or referenced regulatory guides may not have been followed or be needed. However, a design satisfying the basic safety requirements of the Commission's regulations is nevertheless required.
The absence of explicit documentation justifying the deviation makes the "conformance trail" that much more troublesome.
My confidence that our overall review assures compliance with Commission regulations is supported by the staff performance in hearing cases in which the issue of compliance with particular provisions of the Commission's regulations has been challenged. In such cases, testimony in addition to the sunraary statements in our SER is often needed to explain how our review has lead us to the conclusion that the system in question will perfom safely and in accordance with the applicable Commission's regula-tions. Although this may entall substantial additional explanation than that provided by the SER, the issue of ccmpliance has been generally adjudicated favorably to the staff.
Finally, it is important to recognize that although the staff's review of an application is partially an " audit" review, the applicant for a license is obligated to assure compliance with applicable regulatory requirements. It is the applicant.who bears the burden of proof on the
, issue. For issues in controversy the applicant bears this burden in the hearing process; for matters not in controversy before Licensing Beards, OFFICE k SURNAME DATEk.
NRC FORM 318 (9-76) NRCM 0240 C U.S. GOVERNMENT PRINTING OFFICE: 1979 289 369
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The Cocaissioners the applicant bears this burden in the staff's review process. As a part of an application for an operating license, Section 3.1 of a typical FSAR recites compliance with all GDCs. iihile a useful summary of conformance, the remaining thousands of pages of an application are needed to adequately evidence specific compliance. The staff's audit review process tests these assertions. This review results in changes in principal safety systems.
In general, these changes have not been limited to those necessary to comply in a minimal fashion with the language of the applicable regulation, but in general go beyond to assure the use of " good" design.
CONCLUSION The problem of documentation of conformance with the Commission's regulations is a vexing, manpower intensive effort to which the staff, due to tica and manpower limitations, has been forced to give inadequate attention. By good management effort, I hope to improve this situation an, -
gradually eliminate it. But to do so by an intense effort will be costo. This was the thrust of my June 13, 1980 memorandum. However, the defects in documen-tation should not be misconstrued as evidence of defects in the review process. Using a audit process, it is simply not possible for the NRC to state, based on its cwn knowledge, that every rule and regulation has been met for every applicable action by the applicant. However, considering the certifications made by the applicant, the degree to which guidance has been provided to the industry and the public regarding acceptable ways te meet the rules and regulations, the emphasis that the staff places in its reviews on areas of particular controversy and importance to safety, and the fact that both the ACRS review and the hearing process throw additional spotlights on the areas of safety significance relating to the regulations, gives confidence that our review process results in a reasonable basis for judgments as to whether the - julations have been met.
cfgint S@ed W H. R. Dant:s Hnrold R. Denton, Director 1* \\
Office of Nuclear Reactor Regul tion
Enclosure:
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MEMCRANCUI4 FCR: C:1 airman Ahearne Ccamissioner Gilinsky Ccemissioner Hendrie Commissioner Bradford William J. Dircks CIre$ ',yl!tyn 7 m.,,
THRU:
Acting Executive Director for Operation,s FROM:
Harold R. Centon, Director Office of Nuclear Reactor Regulation CCMPLINiCE CF NRC LICENSES WITH NRC REGULATICNS, REGULATCRY SUSJECT:
GUIDES, BRANCH TECHu! CAL POSITICNS, AND LICENSEE CCMMITMENTS Cn' June 13, 1980, I transmitted a mer.orandcm to you on the above subject.
The purpose of that memorandum was to advise ycu of the costs to *.he NRC of having to certify in detail how-structures, systems and ccaponents of a nuclear reactor facility comply with each current safaty-related regulation, regulatory guide and branch technical position. My memorandum discussed the difficulties of, and censequent rescurce needs associated with, comprehensive dcct.~.nantation of canpliance with every aspect of each require-ment of the Commission's regulations governing nuclear reactor design and On the other hand, my memo touched only lightly cn the staff's operation.
overall assessmant that our review provides adequate assurance that licensed facilities conform to the Co=aission's regulations. This dichotcmy has raised a number of questions concerning cur discharge of regulatory respon-sibilities with respect to assuring compliance with the Cec.issicn's regul ations. As a result, I believe sece amplification with respect to these questions is warranted.
As pointed cut in my memorandum, cur review is based on the Standard Review Plan (SRP). Each section of the SRP centains aaceptance criteria In those which reflect the requircents in the Camaission's regulations.
instances where the regulation is specific (e.g.,10 CFR 50.46 and Appendix K) the acceptance criteria in the applicable secticns of the SRP reference the regulation and therefore the review dces focus on explicit conformance In those instances where the regulaticn is stated in with t:.e regulation.
bread terms (e.g., General Cesign Criteria) the acceptance critaria in the applicabla section of the SRP reference regulatory guides and branch technical postions generally related to the regulations and therefore the review dces focus on explicit conformance with these guides and positions.
Thus, cur review emphasizes an in depth cvaluation of the ;rincipal systans and structures important to safety against detailed criteria, rather than focusing en an explicit accounting of ccupliance *f th broad principles.
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g 3ackcround of the SRP The Standard Review Plan was develcped in the early 1970's at the time of substantial expansion of the technical review staff and a shift in review activities to a far more penetrating in-depth review of principal structures and systems important to safety. This change in review included a substan-tial increase in independent technical assessment in such areas as design and performance of emergency core cooling systems, design and perforr.ance of containment and other accident mitiga'.ing safeguards, design and per-formance of radwaste systeas., geology and scismology, and structurai design techniques.
Although still an audit review in the sense that not every system and not every " nut and bolt" is explicitly evaluated, the expansion of technical review in the 1970's provided a far more coi:prehensive assessmant by AEC-NRC of cr'tical systems than had been cr 1 ducted earlier by1EC.
As a result of expansion in staff and in depth of revieu, it was felt important to provide, for this newly expanded staff, clear guidance as to The SRP was the
>< hat was expected of the review that they were to conduct.
written expression by experienced staff reviewers of the factors to be censidered in prc;:erly reviewing particular systems.
It is important to recognize that, although there is no explicit correlation between the SRP and the regulations, the experience upon which the drafters of the varicus pcrtions of the SRP drew was their prior experience in reviewing appif-cations. For these p.-ice reviews the cnly criteria for judg. ant ware those of the regulations as amplified by the GCC.
Imbedded in that judgment of " adequacy" by virtue of the experience of tha myiewars involved in drafting the SRP is an assessment of conformance to the requirements of the GCC and other Commissicn regulations. Unfortunately for posterity and for the types of questions posed recently, but under-standable in light of the time and manpcwar pressures that existed in the early 1970's when the SRP was t,cing developed, the chain of reasoning of the reviewars who drafted the SRP was not preserved. To develop an explicit technical basis relating the 2RP to the Ccmmission regulations is the activity of high manpower cost indicated in my memorandum of June 13, 19d0.
Ccngruence of the SRP and the Requiations Each SRP is organized into four sections: areas of review; acceptance criteria; review procedures; and evaluation findings. The acceptance criteria contain a statenent of the purpose of the review and the technical basis for determining acceptability. The technical basis censists of specific criteria which typically include reference to Part 50 and 100, and particularily the General Design Criteria of Appendix A, regulatory guides, codes and standards, and branch technical positions.
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'ihile, as mentioned before, there has been no fully disciplined attrpt to relate every rule and regulation to every applicable standard reviaw plan or Regulatory Guide, we have made a post facto study of t'ese interrelationships and found, for example, that all but one General Oesign Criterien is specifically referenced in the SRFs. The only GCC not referenced explicitly in the SRP is GCC 51 " Fracture Prevention of Mcwever, SRP Section 3.3.2, which should Centaiment Pressure Scundary."
have included such a reference, does specify that the matarials of steel centairaents or steel porticns of steel and concrete contaiments 5e reviewad for ccmpliance with Articie NE-2000 of Section NE of the AS'E 3 oiler and Pressure Vessel Code. Compliance with this section of the Code will. in general provide assurance that the basic requireents of GDC 51 are met.
To provide sc=e additional perspective as to the interrelationships bet.ceen the GCC, the other regulations, the SRP's, and Regulatory Guides, a pre 11m-inary survey was made of these dccuments. Enclosure 1 to this memorandum provides a listing of:
1.
SRP sections where specific GDCs are referenced, listed by GCC (Table 1);
2.
Reg. Guides tilere specific GDCs are referenced, listed by GDC (Table 2);
3.
A cross-list of GCCs and regulations, listed by Regulatory Guide (Table 3); and 4.
A cross-list of regulations, GCC, and Regulatory Guides, listcd by SRP section (Table 4).
In general, the degree of congruence between a regulation and the SRP for any particular set of structures, systems or ccaponents reflects the specificity of the regulation itself and any implementing guidance, the -
experience of the staff, and the influence and interests of the Cccaission, the AC35, the Boards, the public, and the industry.
Therefore, although there are idcatification and decunantation concerns as expressed in my June ?3,1980 a:emorandua, I c.1 nonetheless confident that the link exists in fact so that when a license review is properly carried out in accordance with the SRP, such a review adequately assures confor ance with the Ccamission's regulations.
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J b.Ccmmi,ssioners S3P and Licensing Process Applicants for pemits and licenses have had formal guidance en hcw to prepare their Safety Analysis Reports since June 1966, 'nhen a " Guide to the Organi:ation and Contents of Safety Analysis Reports"was issued by the AEC.
Revision 1 to this guide was issued in October 1972, as the " Standard Format and Content of Safety Analysis Reports for Nuclear Power P7 ants."
The Standard Fomat was revised again in Septe.ber 1975 (Revision 2) and in
'tovember 1973 (Revision 3).
Since 1972 the staff guidance has required that applicants explicitly describe confomance with the General Design Criteria, (Appendix A to 10 CFR 50). Revision 3 (Movember 1973) was the first Standard Fomat to specify that applicants should address confomnce with Regulatory Guides, Most SARs since but most applicants have done this since about 1973.
contain sections which address the extent to which each GCC 1972-73 and, each applicable Regulatory Guide issuci up to a time some months prior to the SAR subaittal date is met.
Althcugh justification for deviations frca the Standard Revied Plan has not yet been required to be explicitly docum2nted, revie..ars are expected to use the SRP as a guide in their review of all applications.
In cases of facilities substantially censtructed before the SRP was prceulgated, the specific recommendations of the SRP or referenced regulatory guides may not However, a design sacisfying the basic have been follcwed or be needed.
safety requirements of the Cccmission's regulations is nevertheless required.
The absence of explicit documentation justifying the deviation makes the "ccafamance trail" that much more trcublesome.
? y confidence that our everall review assures empliance with Ccamission regulations is supported by the staff performance in hearing cases in hich the issue of cc=pliance with particular previstens of the Commissicn's regulaticas has been challenged.
In such cases, testimony in addition n
to the summary statements in our SER is often needed to explain hcw cur review has lead us to the conclusion that the system in question will perform safely and in accordance with the applicable Caissicn's requia-Althcugh this may entail substantial additicnal explanation than tions.
that provided by the SER, the issue of c::npliance has bcen generally adjudicated favorably to the staff.
Finally, it is important to recognize that althcugh the staff's review of an application is partidly an " audit" review, the applicant for a license is obligated to assure ccepliance with applicable regulatory It is the applicant.who bears the burden of prcof On the requirements.
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. The Carnissioners As a part the appiteant bears this burden in the staff's review process.of an recites couplianca with all GD0s. khile a useful st= mary of confontance, the remaining thousands of pages of an application are needed to adequately The staff's audit review process tests these evidence specific ccapliance.
This review results in changes in principal safety systems.
assertions.
In general, these changes have not been attea to those necessary to ccmply in a minimal fashion with the language of the applicable regulaticn, but in general go beyond to assure the use of "gcca" design.
CONCLUSICN The probicm of documentation of confornance with the Ccmmi.,;on's regulations is a vexing, manpower intansive effort ta which the staff, due to tica 3y and manpcher limitations, has been forced to give inadequate attention.
- cod management effort, I hope to improve this situatica and to gradually This was 3ut to do so by an intense effort will be costly.
eliminate it.
the thrust of my June 13, 1980 memorandum. However, the defects in documan-tation :hould not be af sconstrued as evidence of defects in tne review process. Using a audit precess, it is simply not possible fo Howaver, censidering the met for every applicable action by the applicant.
certifications made by the applicant, the degree to which juidance has been provided to the industry and the public regarding acceptable ways to meet the rules and regulations, the emphasis that the staff places in its raviews on areas of particular centroversy and importance to safety, a.TJ the fact that both the ACRS review and the hearing prccess thr w additional spotlights on the areas of safety significance relating to the regulaticns, gives confidence that our review process results in a reasonable basis for judsments as to whether the regulations have been met.
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ENCLOSURE 1 PRELIMINARY SURVEY OF INTERRELATIONSHIPS BETWEEN THE REGULATIONS, SRPS AND REGULATORY GUIDES
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TABIZ 1 STANDARD RE'/IEU PLAN SECTIONS '.(HICP REFERENCE GENERAL DESIGN CRITERIA GDC SRP Sections 1
3.22/3.9.2/3.9.3/3.9.5/3.11/4.5.2/5.2.3/5.3.2/6.3/7.1/7.2
- 7. 3/7. 4/7. 5 /7. 6 / 8.1/8. 2/8. 3 /1 L. 3. 6 2.4.3/3.2.1/3.3.1/3.3.2/3.4.1/3.4.2/3.5.1/3.5.3/3.8.1/3.8.2 2
3.8.3/3.8.4/3.8.5/3.9.2/3.9.3/3.9.4/3.9.5/3.10/3.11/5 4.6 5.4.7/6.3/7.1/7.2/7.3/7.4/7.5/7.6/8.1/8.2/8.3/9.1.1/9.2.1 9.2.6/9.3.1/9.3.3/9.3.4/9.3.5/9.4.1/9.4.2/9.4.3/9.4.5/9.5.4/
9.5.5/9.5.6/9.5.7/9.5.8/10.3/10.4.7/10.4.9 3
7.1/7.2/7.3/7.4/7.5/7.6/8.1/8.2/8.3/9.5.1 3.5.1/3.E.2/3.5.3/3.6.1/3.6.2/3.8.1/3.8.2/3.8.3/3.8.4/3.8.5 4
3.9.2/3.1.3/3.9.4/3.9.5/3.11/5.4.1/5.4.6/5.4.7/6.3/6.7/7.1
- 7. 2/7. 3/7. 4/7. 5/7. 6 /8.1/8. 2/ 8. 3/ 9.1.1/ 9. 2.1/9. 2. 6/3. 3.1/
9.3.3/9.3.4/9.3.5/9.4.1/9.4.2/9.4.3/9.4.4/9.4.5/9.5.4/9.5.5 9.5.5/9.5.7/9.5.8/10.2.3/10.3/10.4.7/10.4.9 5
5.4.7/6.3/7.1/7.2/7.3/7.4/7.5/7.6/8.1/8.2/8.3/P.1.1/9.2.1 9.2.6/9.3.1/9.3.4/9.3.5/9.4.1/9.4.2/9.4.3/9.4.4/9.4.5/9.5.1 9.5.4/9.5.5/9.5.7/9.5.8/10.4.7/10.4.9 10 3.9.5/4.2/4.3/4.4/7.1/7.2/7.3/7.4/7.5/7.6 11 4.3 12 4.2/7.1/7.2/7.7 13 4.3/5.2.3/7.1/7.2/7.3/7.4/7.5/7.6/7.7/8.1/8.2/8.3/15.4.7 i4 3.9.1/3.9.2/4.5.2/5.2.3/5.4.2 15 3.9.1/3. 9. 2 / 3. 9. 4/ 5..?. 2 / 5. 4. 2/7.1/ 7. 2/7. 6 /7. 7 /9. 5. 5 16 3.8.1/3.8.2/6.1.1/6.2.1 17 6.3/8.1/8.2/8.3/9.5.4 9
18 8.1/8.2/8.3 19 5.4.7/6.4/7.1/7.2/7.3/7.4/7.6/7.7/9.4.1/10.4.9 20 2.9.4/4.6/6.3/7.'1/7.2/7.3/7.4/7.5/7.6/15.4.2/15.4.3 21 4.6/7.1/7.2/7.3/7.4/7.5/7.6/8.1/8.2/3.3 22 7.1/7.2/7.3/7.4/7.5/7.5/8.1/8.2/8.3 23 3.11/7.1/7.2/7.3/7.4/7.5/7.6 24 7.1/7.2/7.3/7. 4/7.6/7. 7/7. 5
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. 25 4.3/4.6/7.1/7.2/15.4.2/15.4.3 26 3.9.4/4.5.1/4.6/7.1/7.2/7.4/7.7/9.3.4/9.3.5 27 3.9.4/4.3/4.6/6.3/7.1/7.2/7.4/7.5/7.7/9.3.4/9.3.5 28 4.3/4.6/7.1/7.2/7.5/7.6/7.7 29 3.9.4/7.1/7.2/7.3/7.4/7.5/7.6/7.7/9.3.4 30 3.9.4/5.2.5 31 3.9.4/5.2.2/5.3.1/5.4.2 32 3.9.4/5.2.4/5.4.2 33 7.1/7.4/7.5/7.6/8.1/8.2/8.3/9.3.4
[.
34 5.4.6/5.4.7/6.1.1/7.1/7.3/7.4/7.2/8.1/8.2/8.3/10.3 35 6.1.1/6.3/7.1/7.2/7.3/7.5/7.6/8.1/8.2/8.3 36 6.3/6.6 37 6.3/7.1/7.2/7.3/7.5/7.6 38 J.2.1/6.2.2/7.1/7.3/7.5/7.6/8.1/8.2/8.3 39 6.2.1/6.2.2/6.6 40 3.9.6/6.2.1/6.2.2/7.1/7.3/7.5/7.6/15:6:5 41 6.1.1/6.2.5/6.5.2/7.1/7.3/7.5/7.6/8.1/8.2/8.3 42 6.2.1/6.2.5/6.5.2/6.5.6.
43 3.9.6/6.2.3/6.2.5/6.5.2/7.1/7.3/7.5/7.6 44 5.4.7/6.1.1/7.1/7.3/7.5/7.6/8.1/8.2/8.3/9.2.1/9.2.6 9.5.5/10.4.7/10.4.9 45 5.4.7/9.2.1/9.2.6/9.5.5/10.4.7/10.4.9 46 3.9.6/5.4.7/7.1/7.3/7.5/7.6/9.2.1/9.2.6/9.5.5/10.4.7/10.4.9 50 3.8.1/3.8.2/6.2.2/6.2.5/7.1/7.3/7.5/7.6 52 6.2.6 53 6.2.6
. 54 6.2.1/6.2.4/6.2.6/6.3/6.7/7.1/7.3/7.5/7.6 5.4.6/5.4.7/6.2.4/7.1/7.3/7.5/7.6/15.6.2/15.6.5 55 5.4.6/5.4.7/6.2.1/6.2.4/6.3/7.1/7.3/7.5/7.6
~
56 57 5.4.6/5.4.7/6.2.4/7.1/7.3/7.5/7.6 60 10.4.1/10.4.3/11.5 61 9.1.1 3
62 9.1.1 63 9.1.1/11.5 64 10.4.1/10.4.3/11.5 O
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TABIZ 2 REGULATORY GUIDES WHICH REFERENCE GENERAL DESIGN CRITERIA 4
GDC Reo Guides 1
1.68,1.69,1.70,1.71,1.72,1.79,1.80,1.81,1.84,1.85,1.87 1.103,1.104,1,105,1.106,1.107,1.128,1.133,1.136,1.10, 1.15,1.16,1.18,1.19,1.20,1.26,1.31,1.36,1.41,1.43,1.44, 1.50,1.55,1.66,1.67,1.65 1.29,1.48,1.57,1.59,1.60,1.61,1.76,1.92,1.102,1.117,1.122, 2
1.124,1.129,1.130,1.135,1.142 3
1.75,1.120 4
1.14,1.44,1.46,1.78,1.91.,1.95,1.96,1.106,1.115,1.142 5
1.104 13 1.56,1.97,1.133 14 1.31,1.36,1.56,1.83,1,121 15 1.56,1.83,1.121 17 1.6,1.32,1.75,1.93,1.108,1.137 18 1.32,1.108,1.118 19 1.78,1.95,1.97,1.114 20 1.22 21 1.75,1.118 23 1.77 30 1.45,1.65,1.84,1.85 31 1.36,1.65,1.83,1.99,1,106 32 1.83 35 1.2,1.7,1.82 36 1.82 37 1.82 38 1.82 39 1.82 40 1.82
'41 1.1,1.52
. 42 1.52 1.52 43 44 1.27 45 1.127 50 1.63,1.80,1.81 53 1.35,1.90 54 1.96,1,141 55 1 11,1,141 56 1.11,1.141 57 1.141 60 1.21,1.140 61 1.13,1.98,1.104,1.140 64 1.97 Appendix B 1.128,1.30,1.31,1.33,1.34,1.37,1.38,1.39,1.40,1.46 1.54,1.58,1.64,1.73,1.74,1.88,1.94,1.100,1.116,1.123 50.55A 1.26,1.46,1.53,1.62,1.75,1.118
-me emp.
TABI2 3 GENERAL DESIGN CRITERIA AND REGULATIONS REFERENCED IN REGULATORY GUIDES Number Reg. Guide GDC Reculation 1.1 Net Positive Suction Head for Emergency 41 Core Cooling and Containment Heat Removal System Pu=ps 1.2 Thermal Shock to Reactor Pressure 35 Yessels 1.3 Assumptions Used for Evaluating the Potential Radiological Consequences 'of a Loss of Coolant Accident for Boiling Water Reactors.
1.4 Assumptions Used for Evaluating the Potential Radiological Consequences of a loss of Coolant Accident for Pressurized Water Reactors 1.5 Assumptions Used for Evaluating the Potential Radiological Consequences of a Steam Line Break Accident for Boiling
~
Water Reactors 1.6 Independence Between Redundant Standb'y ~ '
17 (Onsite) Power Sources and Between Their Distribution Systems 1.7 Control of Combustible Gas Concentrations 35 in Centainment Following a Loss of Coolant Accident Supplement to Safety Guide 7, Backfitting Considerations 1.8 Personnel Selection and Training 1.9 Selection, Design, and Qualification of Diesel-Generator Units Used as Onsite Electric Power Systems at Nuclear Power Plants (Comme.ats requested by 1/26/79) l.10 Mechanical (Cadweld) Splices in Reinforcing 1 Bars of Category I Concreto Structures 1.11 Instrument Lines Penetrating Primary 55, 56 Reactor Containment Supplement to Safety Guide 11. Backfitting Considerations 1.12 Instrumentation for Earthquakes
. 1.13 Spent Fuel Storage Facility Design Basis 61 1.14 Reactor Co61 ant Pump Flywheel Integrity 4
1.15 Testing of Reinforcing Bars for Category 1
1 Concrete Structures 1.16 Reporting of Operating Information --
1 Appendix A Technical Specifications 1.17 Protection of Nuclear Plants Against Industrial Sabotage 1.18 Structural Acceptance Test for Concrete 1
Primary Reactor Containments 1.19 Nondestructive Examinatien of Primary 1
Containment Liner Welds 1.20 Comprehensive Vibration Assessment 1
Program for Reactor Internals During Preoperational and Initial Startup Testing 1.21 Measuring, Evtluating and Reporting Radio-60 activity in Solid Wastes and Release of Radioactivity in Liquid and Gaseous Effluents frcm Light Water-Cooled Nuclear Pcwer Plants 1.22 Periodic Testing of Protection System 20 Actuatien Functions 1.23 Cnsite Meteorological Programs 1.24 Assumptiens Used for Evaluating the Potential Radiological Consequences of a Pressurized Water Re'e'or Gas Storage Tank Failure 1.25 Assumptiens Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Soiling and Pressurized Water Reactors 1.26 Quality Group Classifications and Standards 1 50, 55A for Water Steam-and Radio-Waste-Containing Camponents of Nuclear Power Plants
3 1.27 Ultimate Heat Sink for Nuclear Pcwer 44 Plants 1.28 Quality Assuranct Program Requirerents App B (Design and' Construction) 1.29 Seismic Design Classificatien 2
1.30 Quality Assurance Requirements for '
App B the Instal ~ation, Inspection, and Testing of Instrumentation and Electric Equipment 1
1.31 Control of Ferrite Content in Stainless 1, 14 App B Steel Weld Metal 1.32 Criteria for Safety-Related Electric 17, 18 Power Systems for Nuclear Power Plants 1.33 Quality Assurance Program Requirements App B (Operation) 1.34 Control of Electroslag Weld Propertie's
~
App B 1.35 Inservice Inspection of Ungrouted Tendens 53 in Prestressed Concrete Containment Structures 1.36 Nonmetallic Thermal Insulation for 1, 14, 31 Austenitic Stainless Steel 1.37 Quality Assurance Requirements for App B Cleaning of Fluid Systems and Associated C mponents of Water-Cooled Nuclear Pcwer Plants 1.38 Quality Assurance Requirements for App 3 Packaging, Shipping, Receiving, Storage, and Handling of Items for Water-Cooled Nuclear Pcwer Plants 1.'39 Housekeeping Requirements for '4ater-Cooled App 3 Nuclear Power Plants 1.40 Qualification Tests of Continuous-Duty
. App B Motors Installed Inside the Containment of Water-Cooled Nuclear Pcwer Plants 4
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. 1.41 Preoperational Testing of Redundant 1
Onsite Electric Pcwer Systems to Verify Proper Load Group Assignments 1.42 Interim Licensing Policy on As-Low-As-Practicable for Gaseous Radiciodine Releases from Light-Water-Cooled Nuclear Power Reactors 1.43 Centrol Stainless Steel Weld Cladding of 1
Low-Alloy Steel Components 1.44 Control of the Use of Sebsitized Stainless 1, 4 Stec1 1.45 Reactor Coolant Pressure Soundary Leakage 30 Detecticn Systems 1.46 Protection Against Pipe Whip Inside 4
Containment 1.47 Bypassed and Inoperable Status Indication -
50.55A, for Nuclear Power Plant Safety Systems App 3 1.48 Design Limits and Leading Combinations for 2
Seismic Category I fluid System Components 1.49 Power Levels of Nuclear Power Plants 1.50 Control of Preheat Temperature for 1
Welding of Lcw-Alloy Steel 1.51 Inservice Inspection of ASME Code Class 2 and 3 Nuclear Pcwer Plant Compenents 1.52 Design, Testing, and Maintenance Criteria 41, 42, 43 for Post-accident Engineered-Safety-Feature Atmosphere Cleanup System Air Filtration and Absorption Units of Light-Wat -Cooled Nuclear Pcwer Plants 50.55A 1.53 Applicati. of the Single-Failure Criterien to Nuclear Pcwer Plant Protection Systems App B 1.54 Quality Assurance Requirements for
~
Protective Coatings Appliced to Water-Cooled Nuclear Power Plants 1.55 Concrete Placement in Category 1 Structures 1
5-1.56 Maintenance of Water Purity in Boiling 13, 14, 15 Water Reactors.
1.57 Design Limits and Loading Combinations 2
for Metal Primary Reactor Containment System Components 1.58 Qualification of Nuclear Power Plant-App B Inspection, Examinatien, and Testing Personnel Cesign Basis Floods for puclear Pcwer 2
1.59 Plants 1.60 Design Response Spectra for Seismic 2
Design of Nuclear Power Plants 1.61 Damping Values for Seismic Design 2
of Nuclear Pcwer Plants 1.62 Manual Initiation of Protective Actions 50.55A 1.63 Electric Penetration Assemblies in 50 Containment Structures for Light-Water-Cooled Nuclear Power Plants 1.64 Quality Assurance Requirements for App B the Design of Nuclear Power Plants 1.35 Materials and Inspection for Reactor 1, 30, 31 Vessel Closure Studs 1.66 Nondestructive Examinaticn of Tubular 1
Products 1.67 Installation of Overpressure Protective 1
Devices 1.68 Initial Test Programs for Water-Cooled 1
Nuclear Pcwer Plants 1.68.1 Preoperational and Initial Startup Testing of Feedwater and Condensate Systems for Soiling Water Reactor Pcwer Plants 1.68.2 Initial Startup Test Program to Demenstrate 1
Remote Shutdown Capability for Water-Cooled Nuclear Pewer Plants
. 1.69 Concrete Radiaticn Shields for Nuclear 1
Power Plants 1.70 Standard Format and Content of Safety 1
Analysis Riports for Nuclear Power Plants - LWR Edition 1.71 Welder Qualification for Areas of Li,mited 1
Accessibility 1.72 Spray Pond Piping Made From Fiberglass-
.1 Reinforced Thermosetting Resin 1
1.73 Qualification Tests of Electric Valve App B Operators Installed Inside the Containment of Nuclear Power Plants 1.74 Quality Assurance Terms and Definitions App B 1.75 Physical Independence of Electric Systems 3, 17, 21 50.55A 1.76 Design Basis Tornado for Nuclear Power 2
Plants 1.77 Assumptiens Used for Evaluating a 28 Centrol Rod Ejection Accident for Pressurized Water Reactors 1.78 Assumptions for Evaluating the Habit-4, 19 ability of a Nuclear Pcwer Plant Control Rocm Curing a Postulated Hazardous Chemical Release 1.79 Preoperational Testing of Emergency Core 1
Cooling Systens for Pressurized Water Reactors 1.80 Preoperational Testing of Instrument Air 1, 50 Systems 1.81 Shared Emergency and Shutdcwn Electric 1, 50 Systems for Multi-Unit Nuclear Power Plants 1.32 Sumps for Emergency Core Cooling and 35, 36, 37, 38 Containment Spray Systems 39, 40
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1.83 Inservice Inspection of Pressurized Water 14, 15, 31, 32 Reactor Steam Generator Tubes 1.84 Design and Fabrication Code Case 1, 30 Acceptabili-ty - ASME Section III Division 1 1.85 Materials Code Case Acceptability -
1, 30 ASME Section III Division 1 1.86 Termination of Operating Licenses for Nuclear Reactors
~
1.87 Guidance for Construction of Class 1 1
Components in Elevated-Temperature Reactors (Supplement to ASME Section III Code Classes 1592, 1593, 1594, 1595, and 1596) 1.S8 Collection, Storage, and Maintenance of App B Nuclear Power Plant Quality Assurance Records 1.89 Qualification of Class IE Equipment for Nuclear Power Plants 1.90 Inservice Inspection of Prestressed 53 Concrete Containment Structures With Grouted Tendons 1.91 Evaluations of Explosions Postulated to 4
C cur on Transportation Routes Near Nuclear Power Plants l.92 Ccmbining Modal Responses and Spatial 2
Compcnents in Seismic Response Analysis 1.93 Availability of Electric Power Sources '
17 1.94 Quality Assurance Requirerents for In-App 3 stallation, Inspecticn, and Testing of Structural Concrete and Structural Steel During the Construction Phase of Nuclear Power Plants 1.95 Protectica of Nuclear Pcwer Plant Control 4, 19 Room Operators Against an Accidental Chlorine Release 1.96 Design of Main Steam Isolation Valve 4, 54 Leakage Centrol Systems for Boilding Water Reactor tiuclear Power Plants
..----w 8-1.97 Instrumentation for Light-Water-Cooled 13, 19, 64 Nuclear Power Plants to Assess Plant Conditions During and Following an Accident 1.98 Assunptions Used for Evaluating the 61 Potential Radiological Consequences of a Radioactive Offgas System Failure in a Boiling Water Reactor 1.99 Effects of Residual Elements en Predicted 31 Radiation Damage to Reactor Vessel Materials App B 1.100 Seismic Qualification of Electric Equipment for Nuclear Power Plants
.l.,101 Emergency Planning for Nuclear Power Plants 1.102 Flood Protecticn for Nuclear Pcwer Plants 2
1.103 Post-Tensioned Prestressing Systems for 1
Concrete Reactor Vessels and Contain6ents' 1.104 Overhead Crane Handling Systems for Nuclear 1, 5, 61 Power Plants
~
1.105 Instrument Setpoints 1
1.106 Thermal Overicad Protection for Electric 1, 4, 31 Motors on Motor-Operated Valves 1.107 Qualifications for Cement Grouting for 1
Prestressing Tendons in Containment Structures 1.108 Periodic Testing of Diesel Generator 17, 18 Units Used as Onsite Electric Pcwer Systems at Nuclear Pcwer Plants Calculation of Annual Doses to Man Frcm Apo I 1.109 Routine Releases of Reactor Effluents for the Purpc e of Evaluating Compliance with-10 CFR Part 50, Appendix I App I 1.110 Cost-Benefit Analysis for Radwaste Systams for Light-Water-Cooled Nuclear Power Reactors App I 1.111 Methods for Estimating Atmospheric Trans-part and Dispersion of Gaseous Effluents in Routine Releases frem Light-Water-Cooled Reactors
- N w - em>
,we-
.g.
1.112 Calculation of Releases of Radioactive App I Materials in Gaseous and Liquid Effluents from Light-Water-Cooled Power Reactors 1.113 Estimating Aquatic Dispersion of Effluents App I from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I 1.114 Guidance on Being Operator at the Controls 19 of a Nuclear Power Plant 1.115 Protection Against Low Irajectory Turbine 4
Missiles 1.116 Quality Assurance Requirements for App B Installation, Inspection, and Testing of Mechanical Equipment and Syste.-s 1.117 Tornado Design Classification 2
18, 21 50.55A 1.118 Periodic Testing of Electric Power -
and Protection Systems 1.119 Surveillance Program for New Fuel Assembly Designs 1.120 Fire Protection Guidelines for Nuclear 3
Power Plants 1.121 Bases for Plugging Cegraded PWR Steam 14, 15 Generator Tubes 1.122 Development of F1cor Design Response 2
Spectra for Seismic Design of Floor-Supported Equipment or Components 1.123 Quality Assurance Requirements for App 3 Control of Procurement of Items and Services for Nuclear Power Plants 1.124 Service Limits and Loading Combinations 2
for Class I Linear Type Component Supports 1.125 Physical Models for Design and Operation App K of Hydraulic Structures and Systems for Nuclear Power Plants
. 1.126 An Acceptable Model and Related Statistical App K Hethods for the. Analysis of Fuel Densification 1.127 Inspection of Water-Control Structures 45 Associated with Nuclear Power Plants.
1.128 Installation Design and Installation of 1
Large Lead Storage Batteries for Nuclear Power Plants 1.129 Maintenance, Testing, and Replacement 2
of Large Lead Stcrage Bitteries for Nuclear Power Plants 1.100 Service Limits and Leading Ccebinations 2
for Class I Plant-and-Shell-Type Component Supports 1.131 Qualification Test of Electric Cables, Field Splices, and Connecticn; for Light-Water-Cooled Nuclear Power Plants -
1.132 Site Investigaticns for Foundations of Nuclear Power Plants 1.133 Loose-Part Detection Program for the 1, 13 Primary System of Light-Water-Ccoled Reactors 1.134 Medical Certification and Monitoring of Personnel Requiring Oper.ator Licenses 1.135 Normal Water Level and Discharge at 2
Nuclear Pcwer Plants 1.136 Material for Concrete Containments 1
1.137 Fuel-011 Systems for Standby 17 Diesel Generators 1.138 Laboratory Investigations of Soils for Engineering Analysis and Design of Nuclear Power Plants 1.139 Guidance for Residual Heat Removal 1.140 Design, Testing, and Maintenance 60, 61 Criteria for Normal Ventilation Exhaust System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants wm
. 1.141 Containment Isolation Provisions for 54, 55, 56, Fluid Systems 57 1.142 Safety-Related Concrete Structures for 2, 4 Nuclear Power Plants (Other Than Reactor Vessels and Centainments) 1.143 Design Guidance for Radioactive Waste Management Systecs, Structures, and Competents Installed in Light-Water-Coo'*. Nuclear Power Plants 1
=
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.. YABl.E'4~ '
REGilLATIOilh GE'lERAL DESIGil CRITERI A Afl0 REGill. ATOR,Y_GillDES REfTjlEjlCED lil SECTIO!!S Of T!IE STAtlDARD REVIEW PLAtl.
SRP Section Title Regulation GDC Reg. Guide 2.1.1 Site Location and Description Pt 100 2.1.2 Exclusion Area Authority and Control.. Pt 100 2.1.3 Population Distribution......... Pt 100 2. 2.1 -2. 2. 2. Locations and Routes Descriptions....
1.78, 1.91, 1.95 2.2.3 Evaluation of Potential Accidents.... Pt 100 2.3.1 Regional Climatology...........
1.27, 1.76 l..
2.3.2 Local Meteorology 1.21, 1.23 i
2.3.3 Onsite Meteorological Heasurements Programs a
2.3.4 Short Tena (Accident) Diffusion Estimates.Pt 100 1.3, 1.4, 1.5, 1.23, 1.24, 1.25 1.27, 1.77 2.3.5 Long Tenn (Routine) Diffusion Estimates.
1.21, 1.23, 1.42, 1.111 2.4.1 ilydrologic Description..........
2.4.2 Floods..................
1.29, 1.59, 1.102, 1.135 2.4.3 Probable Maximun Flood (PMF) on Streams and Rivers................
2 1.29, 1.59, 1.102, 1.135 2.4.4 Potential Dam Failures (Seismically Induced)................ Pt 100-App.A 1.29, 1.59, 1.102, 1.135 2.4.5 Probable Maximum Surge and Seiche 1.29, 1.59, 1.135 i
Flooding.
I 2.4.6 Probable Maximum Tsunaml flooding 1.29, 1.59, 1.102, 1.135 i
2.4.7 Ice Flooding...........
1.27, 1.29, 1.135
SRP Section Title Regulation G0C Reg. Gufde 2.4.8 Cooling Water Canals and Reservoirs..
1.27, 1.29, 1.59, 1.102, 1.135-s 1.27 2.4.9 Channel Diversions..,........
2.4.10 Flood Protection Requirenents 1.25,1.29,1.59,1.102 2.4.11 Low Water Considerations 1.27 2.4.12 Dispersion, Dilution, and Travel Times of
.cidental Releases of Liquid E f fl uen ts...............
Pt 20 1.26, 1.27, 1.29 2.4.13 Groundwater..............
Pt 20 2.4.14 Technical Specifications and Energency Operation Requirements 2.S.1 Basic Geologic and Seismic Information.
Pt 100-App. A 2.5.2 Vibratory Ground Motion........
Pt,100-App. A 1.60 2.5.3 Surface faulting............
Pt 100-App. A 2.5.4 Stability of Subsurface Materials Pt ISO-App.A 1.132 i
2.5.S Slope Stability Pt 100-App.A 3.2.1 Seismic Classification.........
Pt 100-App.A 2
1.29 3.2.2 System Quality Group Classification App.B. 50.SSA i 1.26, 1.29, 1.48, 1,51 3.3.1 Wind Loadings.............
2 3.3.2 Tornado Loadings 2
3.4.1 Flood Protection 2
1.102 3.4.2 Analysis Procedures..........
2 i
SRP Section Ti tle Regulation GDC Reg. Guide 3.5.1.1 Internally Generated Missiles
~
(Outside Containment).......
4 1.13, 1.27, 1.115 3.5.1.2 Internally Generated Missiles (Inside Contaisiment)
A3HG or ACI 4
3.5.1.3 Turbine Missiles..........
Pt 100 4
1.115,1.1]7 3.5.1.4 Hissiles Generated by Natural Phenomena.......,.....
2, 4 1.76 3.5.1.5 Site Proximity Missiles (Except Aircraf t).........
Pt 100 1.76, 1,91 3.5.1.6 Aircraf t llazards Pt 100 3.5.2 Jtructures, Systems, and Components to be Protected from Externally Generated Hissiles.........
4 1.13, 1.27, 1.115, 1.117 3.5.3 Barrier Design Procedures 2.4 3.6.1 Plant sesign for Protection Against Postulated Piping failures in Fluid Systems Outside Containment.
4 1.29, 1.46 3.6.2 Determination of Break Locations and Dynamic Effects Associated with the Postulated Rupture of Piping....
4 1.46 3.7.1 Seismic Input...........
Pt 100-App.A 1.60, 1.61 i
3.7.2 Seismic System Analysis......
3.7.3 Seismic Subsystem Analysis 3.7.4 Seismic Instrumentation Program..
Pt 100-App. A 1.12 i
SRP Section Title Regulation GDC Reg. Guide 3.8.1 Concrete Containnent........
ACI 2, 4, 16, 1.10, 1.15, 1.18, 1.19, 1.35 -
50 1.55 3.8.2 Steel Containment ASME 2, 4, 16, 1.57 50 3.8.3 Concrete and Structural Steel Internal Structures of Steel or Concrete Conta innents............
ACI, ASMG 2, 4 1.10, ?.15, 1.55, 1.57 3.8.4 Other Cate90ry I Structures ACI 2, 4 3.8.5 Foundations.............
2, 4 3.9.1 Special Topics for Hechanical Components.
ASME, App.B 14, 15 1.68 3.9.2 Dynamic Testin9 and Analysis of Systems, Components, and Equipment.
ASME 1, 2, 4, 1.20, 1.67, 1.68 14, 15 I
3.9.3 ASME Code Class ~1, 2, and 3 Components, Component Supports, and Core l
Support Structures.........
1, 2, 4 1.48, 1.67 3.9.4 Control Rod Drive Systeam......
ASME 2, 4, 15, 1.26,1.29,1.48 20, 26, 27, 29, 30, 31, 32 3.9.5 Reactor Pressure Vessel Internals..
1, 2, 4, 10 t
3.9.6 Inservice Testin9 of Pumps and Valves 50.5SA 37, 40, 43, 46 3.10 Seismic Qualification of Category i Instrumentation and Electrical Equipment.........
Code 2
1.89, 1.100 l
4
. +,
SRP Section Title Regula tion GDC Reg. Guide 3.11 Environmental Design of Mechanical
~
and Electrical Equipment......
50.55A, 1, 2, 4, 23 1.31, 1.32, 1.40, 1.53, 1.63 App.8 1.73, 1.89 3.11.5 Chemical and Radiological Environmental Estimates.......
4.2 Fuel System Design.........
10 4.3 N ucl ea r De s i gn...........
10, 11, 12, 13, 20, 25e 27, 28 4.4 Thermal and llydraulic Design....
10 1.68 4.5.1 Control Rod System Structural Haterials.......'.......
26 1.31, 1.37, 1.44, 1.85 4.5.2 Reactor Internals Materials.....
ASME 1, 14 1.31, 1.44 4.6 20, 21, 23 25, 26, 27, 28 5.2.1.3 Con:pliance with 10 CFR 5 50.55a..
50.55A 1.84, 1.85 App.B 5.2.1.4 Applicable Code Cases........
5.2.2 Overpressure Protection......
15 1.68 5.2.3 Reactor Coolant Pressure Boundary i
Haterials.............
ASME.
1, 13, 14, 1.26, 1.31, 1.34, 1.37, 1.43, App.G 31 1.44, 1.45, 1.56, 1.71, 1.85 5.2.4 RCPB Inservice Inspection & Testing ASME 32 5.2.5 RCPL Leakage Detection.......
50.55A 30 1.45 i
~
r SRP Section T_i tA Regulation GDC Reg. Guide 5.3.1 Reactor Vessel Haterials......
ASME, App.G, 31 1.31, 1.34, 1.43, 1.50, 1.65 App.ll.
5.3.2 Pressure-Temperature Limits ASME, App.G.
App.ll.
5.3.3 Reactor Vessel Integrity......
ASME, App.G.
1.33, 1.37,, 1.38, 1.39 S.4.1.1 Pump Flywheel Integrity (PUR) 4 1.14 S.4.2.1 Steam Generator Haterials......
Code 14, 15 1.37, 1.85 31 5.4.2.2 Steam Generator Inservice Inspection Code 1, 32..
1.83 5.4.6 Reactor Core Isolation Cooling Sys tem (BWR).....*........
2, 4, 34 1,29, 1.46, 1.68 55, 56, 57 i
S.4.7 Residual lleat Removal (RllR) System.
2, 4, 5, 1.22, 1.26, 1.29, 1.33, 1.46, 34, 44, SS, 1.68 j
56, 57, 19 45, 46 S.4.8 Reactor Water Cleanup System (BWR).
1.26, 1.29, 1.56 S.4.11 Pressurizer Relief Tank '......
1.26, 1.29 6.1.1 Engineered Safety features Metallic Code 16, 34, 35, 1.7, 1.31, 1.36, '.37, 1,44, niterials 38, 41, 44 1.50, 1.54, 1.85 6.1.2 Organic Materials.........
1.3, 1.4, 1.54 I
6.1.3 Post-Accident Chemistry......
6.2.1 Containment Functional Design...
50.46 16, 50, 38, 1.4, 1.26, 1,29, 1.97 Code (4, 56, 39, 40 i
SRP Section Title Regulation GDC Reg. Guide 6.2.2 Containment lleat Removal 38, 39, 40 1.1, 1.26, 1.29, 1.82, 1.97 Systems..............
30 6.2.3 Secondary Containment functional Design...............
App.J 43 1.26, 1.52 -1.96 6.2.4 Containment isolation Systems 54, 55, 1.3, 1.4, 1.11, 1.26, 1.29 56, 57 1.41 6.2.5 Combustible Gas Control in Con ta innen t............
50.46 41, 42, 1.7, 1.26, 1.29, 1.52 43, 50 6.2.6 Containment Leakage Testing....
App. J 52, 53, 54 6.3 Emergency Core Coolin9 System...
50.46 2,4,5,17 1.1, 1.11. 1.29, 1.46, 1,47 20,27,35, 1.52, 1.68, 1.79 36,37,54,56 6.4 liabitability Systems..:......
19 1.52, 1.78, 1.95 6.5.1 ESF Fil ter Sys tems.........
Pt 100 1.3, 1.4, 1.25, 1.52 6.5.2 Containment Spray as a Fission Product Cleanup System.......
41,42,43 1.3, 1.4 6.5.3 Fission Product Control Systems 1.3, 1.4 6.5.4 Ice Condenser as a fission Product Cleanup System...........
6.6 Inservice Inspection of Class 2 and 3 Components Code 36,39,42, 45 6.7 Main Steam Isolation Valve Leakage Control System (BWR) 2,4,54 1.26, 1.29, 1.96, 1.102, 1.117 I
. o SHP Section Ti tle Regulation GDC Reg. Cuide 7.1 Introduc tion............
50.55A 1-5,10,12,13 1.6.1.7,1.11,1.22,1.29 15,19,20-24,25, 1.30,1.32.1.47,1.53,1.62, 26-29,33-35,37,38 1.63,1.68.1.75,1.78,1.89 40,41,43,44,46, 1.96,1.97,1,100,1.105, 50,54-57 1.118,1.120,1.12,1.45, 1.67,1.80,).95 7.2 Reactor Trip System 50.65A 1-5,10,12,13,15, 1.11,1.22,1.24,1.30,1.47, 19,20-24,25,26,27 1.53,1.62.1.63.1.68.1.75, 28,29,35,37 1.89,1,100,1,105,1.118, 1.120 7.3 Engineered Safety Featura Systens 50.55A 1-5,10,13,19,20-1.7,1.11,1.22,1.29.1.30, 24,29,34,35,37, 1.47,1.53,1.62,1.63,1.68 38,1-40,41,43,44, 1.75,1.80,1.89,1.96,1,100, 50,54-57 1.105,1,106.1.118,1.120 7.4 Systems Required for Safe Shutdown.
50.55A 1-5,10,13,19,20-1.6.1.7.1.11,1.22,1.29, 24,26,27,29,33, 1.30,1.32,1.47,1.53,1.62, 34.
. 1.63.1.68.1.69,1.80,1.89, 1.100 7.5 Safety-Related Display Ins trunenta tion..........
50.55A 1-5,10,12,13,15, 1.6,1.7,1,11,1.22,1,29, 19,20-24,25,26-29, 1.30,1.32,1.47,1.53.1.63, 33-35,37,38,40,41, 1.68.1.75,1.89,1.97,1,100, 43,44,46,50,54-57 1.105,1,118,1.120 7.6 All Other Instrumentation Systems 1-5,10,13,15,19, 1.6.1.7,1.11,1,12,1.22,
- Requi red for Sa fe ty........
20-24,28,29,33,35, 1.29.1.30,1.32,1.45,1.47, 37,38,40,41,43,44, 1.53,1.62,1.63,1.67,1.68 g
46,50,54-57 1.78,1.79,1.80,1.95.1.100 1.105,1.106.1.118.1.120 7.7 Control Systems Hot Required for 12,13.15,19,24,26, 1.30,1.63,1.68.1.120 S a fe ty..............
27,28,29 Appendix Bratich Technical Positions (EICSB).
1.105,1.106.1.118,1.120 7-A i
' SItP. Sec tion Ti tle Regulation GDC Reg. Guide i
Appendix General Agenda Station Site Visits..
7-8 8.1 In troduc ti on.............
50.55A 1-5,13,17.18, 1.6.1.9.1.E9,1.30,1.32, 21,22,33-35, 1.40,1.41,1.47,1.53,1.63, 38,41,44 1.68.1.73,1.75,1.81,1.89, 1.93,1.100,1.106,1.108, 1.120,1.128,1.129 H.2 O f fsi te Power Sys tem.........
50.55A 1-5,13,17,18, 1.30,1.32.1.41,1.47,1.68 21,22,33-35, 1.93,1,118,1,120 38,41,44 8.3.1 A-C Power Sys tems (Onsi te)......
50.55A 1-5,13,17',18, 1.6.1.9.1.29,1.30,1.32, 21,33-35,38, 1.40,1.47,1.53,1.63.1.68, 41,44 1.73,1.75,1.81,1.89.1.93, 1.100,1,106.1.108.1.118,1.120 8.3.2 0-C Power Systems (Onsite)......
50.55A 1-5,13,17,18, 1.6.1.29,1.30,1.32,1.41, 21,22,33-35, 1.47,1.53,1.63,1.68,1.75, 38,41,44 1.81.1.89,1.93,1.100,1,118, 1.120,1.128,1.129 Table 8-1 Acceptance Criteria for Electric Power.
t 9.1.1 New fuel Storage............
2,4,5,61,62, 1.13,1.29,1,102.1.115,1.117 63 i
9.1.2 Spent Fuel Storage..........
2,4,5,61,62, 1.13,1.29,1.102,1.115,1.117 63 f
9.1.3 Spent fuel Pool Cooling and Cleanup i
System..
9.1.4 Fuel Handling System....
i 9.2.1 Station Service Water System (SWS)..
2,4,5,44, 1.26,1.29,1.102.1.117 f
45,46 i
e
SRP Section Title Regulation GDC Reg. Guide 9.2.2 Coolin9 Water System.........
9.2.3 Demineralized Water Make-up System (DWHS).............
9.2.4 Potable and Sanitary Water Systems..
9.2.5 Ul tiaa te !!ca t S i n k..........
9.2.6 Condensate Storage facilities 2,4,5,44,45 1.26,1.29,1.102.1.117 46 9.3.1 Compressed Air System (CAS)......
2,4,5 1.26,1.29 9.3.2 Process Sanpling System (PSC).....
1.21,1,26,1.29,1,143 9.3.3 Equipment and Floor Draina9e System (EFDS) 1......
2,4 1.29 9.3.4 Chemical and Volume Control System (PWR) (includin9, Boron Recovery System.)
2,4,5,26,27, 1.26,1.29,1.102.1.117 29,33 9.3.5 Standby Liquid Control Systan (SLCS).
2,4,21,26,27 1.26,1.27.1.102.1.117 9.4.1 Control Room Area Ventilation I
Sys tem (C RAVS)............
2,4,5,19 1 '6,1.29,1,117 9.4.2 Spent fuel Pool Area Ventilation System (SFPAVS)...........
2,4,5 1.13,1.26,1.117 9.4.3 Auxiliary and Radwaste Area Ventilation System (ARAVS)............
2,4,5 1.26,1.29,1.117 9.4.4 Turbine Area Ventilation System (TAVS).
2,4,5 1.26,1.29,1,117 9.4.5 En91neered Safety Feature Ventilation 2,4,5 1.26,1.29,1,117 System (ESFVS) e
W Sl(P Section Title Regulation GDC Reg. Guide 9.5.1 Fire Protection System.......
3,6 1,70,1.101 9.5.2 Communications System (CS).....
9.5.3 Lighting Systems (LS)........
9.5.2 tommunications System (CS) 9.5.3 Lighting Systems (LS)........
9.5.4 Diesel Engine fuel Oil Storage and 2,4,5,17 1.26,1.29,1.68,1,102, Trans fer Sys tem...........
1.ll7,1.ia?
9.5.5 Diesel Generator Cooling Water 2,4,5,44,45, 1.26,1.29,1.68.1.102,1,117 System...............
46 9.5.6 Diesel Generator Starting System..
2,4,15 1.26,1.29.1.68 9.5.7 Diesel Engine Lubrication System..
2,4,5 1.26,1.29,1.68,1,102,l.117 9.5.8 Diesel Generator Combustion Air 2,4,5 1.,26.1.29.1.68,1,102,1,117 Intake and Exhaust System......
9.5.9 Main Steamline Isolation Valve Scaling System (BWR)........
10.2 Turbine Generator 1,68 1
10.2.3 Turbine Disc Integrity ASME 4
10.3 Main Steam Supply System (HSSS) 2,4,34 1.26,1.29,1,102,l.117 10.3.6 Steam and Feedwater System ASME 1
1.31,1.36,1.37,1.44,1.50 Haterials..............
1.71,1.85 10.4.1 Main Condensers (HC)........
60,64 1.68 10 4.2 Main Condenser Evacuation System 1.26 (HCES)..........'.....
SRP Section Title Regulation GDC Reg. Guide 10.4.3 Turbine Glan4 Sealing System (TGSS).
60,64 1.26 10.4.4 Turbine Bypass System (TBS).....
1.68 10.4.5 CirculatingWaterSystem(CWS) 10.4.6 Condensate Cleanup System (CCS)...
ASME 1.56 10.4.7 Condensate and Feedwater Systen (C&FS) 2,4,5,44,45,46 1.26,1,29,'1.68,1,102,1,117 10.4.8 Steam Generator Blowdown System (SGBS) 1.143 10.4.9 Auxiliary Feedwater System (AFS)...
2,4,5,19,44, 1.26,1.29,1.62,1,102,1,117 45,46 11.1 Source Terms..............
Pt 20 App.I 1.112 HUREG-0016-0017 11.2 Lio; id Waste Systems.........
Pt 20 App.I 1.21,1.60,1,143 11.3 Gaseous Waste Systems Pt 20 App. I 1.140,1.143 11.4 Solid Waste Systems.........
Pt 2O App. I 1.143 f
11.5 Process and Effluent Radiological Monitoring and Sampling Systems...
60,63,64 1.21,1.97 12.1 Assuring That Occupational Radiation i
Exposures are As Low as Practicable (ALAP) 12.2 Radiation Sources..........
12.3 Radiation Protective Design Features.
12.4 Dose Assessment...........
12.1 Assuring That Occupational Radiation Pt 20 1.8 (8.8, 8.10)
ExposuresareasLowasPracticable(ALAP) i
. +
SRP Section Title Regulation GDC Reg. Guide 12.2 Radiation Sources..........
12.3 Radiation Protective Design Features.
Pt 20 1.21,1.52,1.69 (8.8)
Pt 20 12.4 Dose Assessnent 12.5 llealth Physics Program Pt 20 1.8.1.16,1.39(8.2,8.7, 8.8,8.9,0.10) 13.1.1 Manageaienc and Technical Support Oroanization.............
50.34(a)(9) 1.8 (b)(7) 13.1.2 Operating Organization........
13.1.3 Qualifications of Nuclear Plant Personnel..............
13.2 Training Pt 55 1.8. 1.101 13.3 Emergency Planning..........
Pt 100 App.E 1,.97, 1.101 13.4 Review and Audit-13.5 Plant Procedures 50,54(1)(j) 1.33 (k)(1) 13.6 Industrial Security Pt 73 1.17 14 ~. 0 Initial Plant Test Programs (PSAR)..
14.1 Initial Plant Test Programs - FSAR..
1.68 14.2 1.68 15.0 Introduction l
15.1 Increase in lleat demoval by the Secondary System i
g c,
c= :2.
SRP Section Title Regulation GDC Reg. Guide 5, 5 15.1.1-15.1.4 Decrease in Feedwater Temperature.
Q Increase in Feedwater Flow, Increase in Steam Flow, and
-g Inadvertent Opening of a Steam g
Generator Relief or Safety Valve.
15.1.5 Spectruin of Steam System Piping Failures Inside and Outside of Con ta inment (PWR).........
Pt 100 15.2 Decrease in lleat Removal by the Secondary System 15.2.1 Steam Pressure Regulator Failure 1S.2.5 (Closed), Loss of External Load or Turbine Trip, Closure of Main Steam Isolation Valve (DWR), and Loss of Condenser Vacuum..~.......
15.2.6 Loss of Non-Emergency A-C Power to the Station Aux 111arles......
15.2.7 Loss of Normal Feedwater Flow...
15.2.8 Feedwater Systen Pipe Breaks Inside Pt 100 and Outside Containment (PWR)...
I 15.3 Decrease in Reactor Coolant Flow Rate....
7.......
15.3.1 Loss of Ferced Reactor Coolant Flow 15.3.2 Including Trip of Pump and Flow Controller Halfunctions......
I 15.3.3 Reactor Coolant Pump Shaft Seizure Pt 100 15.3.4 and Reactor Coolant Pump Shaf t Break 15.4 Reactivity and Power Distribution Anomalies.............
i
SRP Section Title Regulation GDC Reg. Guide 15.4.1 Uncontrolled Control Rod Assenbly With-drawal From a Subcritical or. Low. Power Startup Condition...........
15.4.2 Uncontrolled Control Sod Assembly Withdrawl at Power.........
20,25 15.4.3 Control Rod Hisoperation (System Halfunction or Operator Error...
20,25 15.4.4 Startup of an Inactive Loop or 1S.4.5 Recirculation Loop at an Incorrect Temperature, and Flow Controller Halfunction Causing an increase in DWR Reactor Coolant flow Rate...
15.4.6 Chemical and Volume Control System Halfunction That Results. in a Decrease in the boron Concentration in the Reactor Coolant (PWR)...
~
15.4.7 Inadvertent Loading and Operation Pt 100 13 of a fuel Assembly in an Improper Position 15.4.8 Spectrum of Rod Ejection Pt 100 1.5, 1,77 Accidents (PWR).........
i 15.4.9 Spectrun of Rod Drop Accidents Pt 100 1.77 (DWR)..............
15.5 Increase in Reactor Coolant Inventory............
15.5.1 Inadvertent Operation of ECCS and 1S.S.2 Chemical and Volume Control System Halfunction That increases Reactor
. Coo l a n t i n ven to ry........
i
9 SRP Section Title Regulation GDC Reg. Guide 15.6 Decrease in Reactor Coolant Inventory.
15.6.1 Inadvertent Opening of a PWR Pressurizer Safety / Relief Valve or a BWR Sa fe ty/ Rel i e f Val ve..........
15.6.2 Break in Instrument Line or Other Lines from Reactor Coolant Pressure Boundary That Penetrate Containment.......
55 1.11 15.6.3 Steam Generator Tube Failure Accident (PWR).............
Pt 100 15.6.4 Main Steam Line Greak Accident (BWR)..
Pt 100 1.5 15.6.5 Loss of Coolant Accidents Resulting from Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Dounda ry...........
50.46 40, 55 1.52 15.7 Radioactive' Release fran a' Subsystem or Component 15.7.1 Waste Gas System Failure........
Pt 100 15.7.2 Radioactive Liquid Waste System Leak or failure............
Pt 100 15.7.3 Postulated Radioactive Releases due to Liquid Tank Failures........
Pt 20' 15.7.4 Fuel llandling Accidents 1.25 l
15.7.5 Spent Fuel Cask Drop Accidents.....
Pt 100 1.25 f
15.8 Anticipated Transients Without Scram..
I 15.8 Anticipated Transients Without Scram..
Pt 100 1,3, j,4
F 1
SRP Section Title Regulation GDC Reg. Guide 16.0 Technical Soecifications.......
50.36 17.1 Quality Assurance During Design and 1.28,1.30.1.37,1.39,1.54, Construction.............
App. 0 1.58,1.64,1.74,1.88,1.94 17.2 Quality Assurance During the 1.8.1.28,1.30,1.37,1.39, Operations Phase App. B.Pt 55 1.54,1.58,,1.64,1.74,1.88, 1.94
&S I
l 0
4 9
1
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