ML20147C777
| ML20147C777 | |
| Person / Time | |
|---|---|
| Site: | 05000584 |
| Issue date: | 12/07/1978 |
| From: | Heltemes C Office of Nuclear Reactor Regulation |
| To: | Gogolick C GIBBS & HILL, INC. (SUBS. OF DRAVO CORP.) |
| References | |
| NUDOCS 7812180403 | |
| Download: ML20147C777 (41) | |
Text
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, n. . - s o jf UNITED STATES t*$% NUCLEAR REGULATORY COMMISSION
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.3 WAShlNGToN, D. C. 2C555 ss .2 a s, ;5
..Ay DEC 7iM Docket No.: STN 50-584 Mr. Charles Gogolick GIBBSSAR Project flanager Gibbs & Hill, Inc.
393 Seventh Avenue flew York, New York 10001
Dear Mr. Gogolick:
SUBJECT:
REVISED GIBBSSAR REVIEW SCHEDULE Enclosed is the revised GIBBSSAR review schedule (in network format) which was approved on November 2,1978. This schedule revision was discussed preliminarily with you, prior to final approval, at a meeting here on October 23, 1978 with Mr. Harold Denton. Further details and implementation of this revised schedule was discussed with Mr. Robert Prieto and other G&H GIBBSSAR project personnel at another meeting on November 7,1978 immediately following final approval . It was agreed at the October 23 meeting that a listing of matters defined for the PDA extensions would be pro;ided to G&H GIBBSSAR project personnel for their consideration in advance of completion of our review, at a means of expediting the remaining GIBBSSAR review prccess and improving upon the revised GIBBSSAR schedule if possible. This listing is enclosed. It should be noted that this listing is a general one and specific items of applicability will have to be determined. If you can identify any problems with the revised schedule which have not already been discussed, or if you have specific questions regarcing the applicability of the PDA extension matters, i.e. the Category I, II, III or IV items, on the enclosed listing to GISBSSAR, please contact us as soon as possible to arrange for discussion and expeditious resolutica of such problems / questions. Sincerely, I C.J.H e s , Chief Sta ardization EM nch Division of Project Management Office of Nuclear Reactor Regulation
Enclosures:
- 1. Revised GIBBSSAR Review Schedule
- 2. PDA Extension Review Matters 7812180fo3 s
t Mr. Charles Gogolick CEC 7 13/8 cc w/encis: fir. Frederick W. Gettler, Vice President Power Engineering Gibbs & Hill, Inc. 393 Seventh Avenue New York, flew York 10001
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"PDA EXTENSION REVIEW MATTERS"-
- , The staff's detailed extension review will j be conducted according to the following guidelines:
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. a) Category 'I Matters- Shff review will determine whether you have cTearly celineated the extent to which the design already j
1 l conforms to these matters. There should be no changes to the design resulting from the staff's review of Category I mattars. , b) Category II Matters-Sh(f review will define the extent to whicn the design conforms, or provides an acceptable
- ' alternative, to these matters. For those cases where the-
. i design is not in substantial conformance with these matters or acceptable alternatives :are not provided you should 1 2 1 demonstrate' why conformance is not necessary. The oiltcome l of the staff review may result in additional requirements. , l . f c) Catiegory III Matters-staff review will determine the extent to - ..; which the design confonns to these matters or whether acceptable ..I . alternatives are provided. If the design does not conform to ; the stated Category III requirements or no acceptable alterna-tive has been provided, staff-approved revisions to the design . will be required. b
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l - . l d) Category IV Matters - Category IV matters are those which have not been reviewedTy the RRRC, but which the Director, NRR, . deems to have sufficient safety attributes to warrant their 1-
being addressed during the PDA _
review. These matters will be treated identically to the Category II matters. . , 4 i l' l b
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.i h' ENCLOSURE A 31 ,
t ' CATEGORY I MATTERS APPROVED 84 RRRC AND ISSUED l1 - FRC'4 MARCH 1974 THROUGH AUGUST 1978- - 1 - -. - ~
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i' EFFECTIVE .
! DATE DOCUMENT NO. REVISION TITLE
. u-: .n i . u:> c...,n . v. s. r 5:an u.. p . .
1/31/73 RG 1.7 2 Control of Combustible Gas Concantrar
- In'C6ntainm'ent* FoTlowing a L6:s-of-
'E ' ~
i Coolant Accident ~ i 9/1/78 RG 1.9 1 Selection, Design, and Qualific ticn jf for Diesel-Generator Units Uced as Onsite Electric Power Systems at
;i Nuclear Power Plants
] 1/9/76 RG 1.20 2 Comprehensive Vibration Assessman:
Prcgran for Reactor Internals Curing Preoperational and Initial Startup Testing
. 11/29/77 RG 1.28 1 Quality Assurance Program Require.mont: l (Design and Construction) 'l ,
l 6/20/78 .RG 1.29 ~~ 3 Seismic Design Classification i
,: 7/20/76 RG 1.31 2 Control of Ferrite Content in Stainles.
Steel Weld Metal
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1/14/77' RG 1.32 2 Criteria for Safety-Related Electric
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Power Systems for Nuclear Pcwer Plan:: 0/21/76 RG 1.33 1 Quality Assurance Program Requir: ment: l (Operation)
, 8/15/75 RG 1.35 '2 Inservice Inspection of Ungrouted Tendons in Pre-stressed Concrete Containment Structures 'i '
(1) 5/77 RG 1.33 2 Quality Assurance Requirements fcr Packaging, Shipping, Receiving, Star:n and Handling of Items for Water-Ccolcc l Nuclear Pcwer Plants 7/12/77 RG 1.39 2 Housekeeping Requirements for Wcter-
, Cooled Nuclear Power Plants .
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(2) 11/29/77 RG',1.52 , 2 Desi,gn;-Testing, and Maintenance for
' ,,Ggineered Safety Featura Atmos:hard j .
1 C16anup System. Air Filtration /and
! Adsorption Units of Light Water C:01a.
I Nuclear Power Plants (3)3/22/77 RG 1.63 1 El.ectriq Penetiration Assembl.ies in Containment ~ Structures for Light-Water i Cooled Nuclear Pcwer Plants . l'
! 1/9/76 RG 1.64 2 Quality Assurance Requirements for tM Design of Nuclear Power Plants 6/20/78 RG 1.68 2 Initial Test Programs for Water-C:olc )
Reactor Power Plants t
, 9/26/75 RG 1.68.1 0 Preoperational and Initial Startup Testing of Feedwater and Ccndensata Systems for Boiling Water Reac:cr Power Plants -
! 11/15/77 ,RG 1.72 .", 1 Spray Pond Plastic Piping I ..:
(1) 3/78 ^ RG 1 84 - 12 Code Case Acceptability - AS".E Secti::. III Design and Fabrication (1) 3/78 RG 1.85 12 Code Case Acceptability - ASME Sectica III Materials , j 5/26/77 RG 1.90 . 1 . Inservice Inspection of Pre-strested j
. .. Concrete Containment Structures with Grouted Tendons
} 8/22/75 RG 1.92 1 Combining Modal Responses and Spatial
.. Components in Seismic Respense Analysi.
I 2/6/76 RG 1.94 1 Quality Assurance Requirements for l
. Installation, Inspection, and Testing l
. of StrJctural Concrete and Stractural. ,
Steel during the Constructier. Phase " j of__ Nuclear Power Plants .- ;
- l. 10/21/76' RG 1.95 . 1 Protection of Nuclear Power Plant l Control Room Operators Against an Accidental Chlorine Release .
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$ DATE DOCU"ENT NO. REVISION TITLE j '
s. I- RG l'.99 E'f fsefs of Residual , Elements on - (4). 1/14/77 - 1
- Prsdicted' Radiation Damage to-P.eactor ji ,(
Vessel Materials . 4 '
'! 6/14/77 RG 1.100 1 Seismic Qualification of Electric Equipment,for Nuclear Power Plants * *
- i. (1) 10/76 RG 1.103 1 Post-Tensioned Pre-stressing Systems lI for Concrete Reactor Vessels and Containments ,
. 1/28/77 RG l.106 1 '
Thermal Overload Protection for Electr: Motors on Motor-0perated Valves 10/21/76 RG 1.107 1 Qualifications for Cement Grouting for Pre-stressing Tendons in Centair.me . Structures (1)' 5/77- RG l.116 0-R Quality Assurance Requirements fer Installation, Inspection, and Testin; of Mechanical Equipment and Systems i 9/27/77 'RG 1. 113-' 1 Periodic Testing of Electric Power and Protection Systems (5) 5/11/77 RG 1.120 l~ Fire Protection Guidelines for Nuclear Power Plants 11/15/77 RG 1.122 1 Development of Floor Design Res;on:a
. ' Spectra for Seismic Design of Floor-Supported Equipment or Ccmponents (1) 7/77 RG 1.123 -
1 Quality Assurance Requirements for :
. Control of Procurement of' Items and l Services for Nuclear Power Plants i i .
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li !! ' 11 EFFECTIVE . TITLE-l's, DATE DOCUMENT N0s, REVISION
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l: 1/14/77 RG 1.126 0-
, An Acceptable-Mode 1 and.Related 2 1' < Statistical Methods for the Analysic of Fuel Densification .
6/20/78 RG 1.128 1 Insta.llation Desi,gn. and Install.aticn of Large Lead $torage Batteries for
Nuclear Power-Plants .
i 2/18/77' RG 1.129 0- Maintenance, Testing and Replacemen: l of Large Lead Storage Batteries fer 1 s Nuclear Power Plants (6) 5/26/77 RG l.131 0 Qualification. Tests of Electric Cams,
. Field Splices and Connections for Light Water Cooled Nuclear Power Plants i 5/11/77- RG 1.132
~
O Site Investigations for Foundations of 1 Nuclear' Power P1 ants 3/22/7,* RG 1.134 '. 0 ' Medical' Certification and. Monitoring.
. . of Personnel Requiring Operator Licencu j 7/12/77 RG 1.135 O. Normal. Water Level and Discharge at Nuclear Power Plants i 8/31/77 RG 1.136 0 Material for Concrete Containments I(7) 9/27/77 RG 1.137 0 Fuel Oil Systems for Standby Diesel Generators 9/27/77 NUREG-0102 0 Interfaces for Standard Designs l
(SRP 1.8) . 11/15/77 RG 1.138 0 Laboratory Investigation of Soils for Engineering Analysis and Design
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of Nuclear Power Plants 11/15/77 RG 1.XXX 0 Permanent Dewatering Systems 11/29/77 RG'l.140' . 0 Design, Testing, and Maintenance
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Criteria for Normal ~ Ventilation Exhaust System Air Filtration and Adsorption Units of LWR's -
.1/31/78 RG 1.142 0 Safety-Related Concrete Structures 3/14/78 RG 8.19 .
0 Occupational Radiation Dose As:ess-ment at LWR's - Design Stage Man-Rem Estimates
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! ,.- / ir . . 3 I' - g; . I EFFECTIVE % .. OATE DOCUMENT N0y. REVISION - c.w. c . TITLE . RSB 5-2 d' ' Rea t[r C'cohant Sy' stem 0verprbssdre (8) 3/14/78 ' Protection -
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( 1) . Indicates that the category I assigned by RRRC for the previous revisica c.' l this document was retained. Review by the RRRC for reassignment of the category is not required for document revisions which do not result in an j, j; increase in requirements. Revision 1 of this regulatory guide was assigned as a Category Il matter
~
; ( 2)
- ', effective January 9, 1976. It is the intent of the RRRC that revision 1
;! remain a Category Il mat However, revision 2 may be used in lieu of l
revision 1 if so desired (er.by applicants. - lI ( 3) Assigned as .a Category II matter by the RRRCs for those applications not - previously reviewed to revision 0. l 1
, (4) Category I for paragraph C.3 only. . Paragraphs C.1, C.2, and'C.4 are assigned by the RRRC as Category III matters.
( 5) In specifying category I .for this regulatory guide, the RRRC recognizes
- that the staff is utilizing Appendix A to BTP ASB 9.5-1 on operating reactors and all CP and OL. applications now under review.
( 6) In specifying the category I for this regulatory guide, the RRRC recogni: 03
; that the fire protection aspects are covered by Appendix A to BTP ASS 9.5-i -
which i,s.a Category II matter
;' '(7) Category I for all CP or PDA applications docketed after the implementatica date shown in the published guide. Certain provisions of the guida are also assigned by the RRRC as Category II and Category III matters.
(8) Category I for operating licenses issued prior to March 14, 1978. Assignd by the RRRC as a Category III matter for all other applications. . i
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.3 H . 1 13 ENCLOSURE B 4 11 CATEGORY 11 MATTERS APPROVED BY RRRC l
[D . FRCM MARCH 1974 THROUGH AUGUS.71973 . -
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l {' i - EFFECTIVE . .. .
. 2 .. . '
'j g DATE DOCUMENT REVISION YITLE'" , I t
# Ultimate Heat Sir.k for Nuclear Pcwer 11/12/75 RG 1.27 2
. Pl.. ants I
1/9/76 RG 1.52 . 1 Design, Testing, and Maintenance i: j. Criteria. for Engineered-Safety-Featura Atmosphere. Cleanup System Air Filtra-
' i; {!
tion and Adsorption Units of Light-W t: j, - Coole? Nuclear Power Plants
,i
- . (1) 8/77 RG l.59 2 s Design Basis Floods for Nucleir Power Plants
,5 ( -
(2)'. 3/22/77 RG 1.63 1 Electric Penetration Assemblies in
!' j, . Containment Structures for Light-Water- l Cooled Nuclear Power Plants i (3) 5/16/78- RG 1.68.2' .
.1 Initial Startup Test Program to Ce:ron-
" strate Remote Shutdcwn Capability for Water-Cooled Nuclear Pcwer F1 ants i
i 11/15/77 RG 1.91 1. Evaluation of Fxplosions Pestulated
.; to Occur on Transportation'Rcutes Near 1
Nuclear Power Plant Sites t* 1/28/77 RG 1.97 - 1 Instrumentation for Light-Water-C cled
! Nuclear Power Plants to Assess Plant Conditions During and Follcwing an
. Accident
'l 11/12/75 RG 1.102 1 Flood Protection for Nuclear Pcwer Plan':
0 9/15/76 RG 1.105 1 - Instrument Setpoints 6/14/77 RG 1.108 1 Periodic Testing of Diesel Generater Units Used as Onsite Electric Preer-Systems at Nuclear Power Plants t 4 e e 4 m
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- DATE DOCUMENT REVISION
-- - ' TITLE -l 3/22/77 RG 1.115 1 Protection Against. Low-Trajector./ l Turbine Missiles- -
,12/20/77 RG 1.117 1 Tornady Design. Classification -
8/31/77 RG 1.124 1 Service Limits and Loading Combin icrs j for Class 1 Linear Type Componen. i;;per 1 7/77 RG 1.130 0' Design Limits and Loading Combir.2:ians l for Class'l Plate- and Shell-Typ. ' Component Supports , 1 Fue) Oil Systems for Standby Diesel
.; (4) 9/29/77 RG 1.137 1 0 l Generators 1
- l 8/18/76 RG 8.8 2 Information Relevant to Ensuring th'c ,
Occupational Radiation Exposures i: l c .
. Nuclear Power Stations will be cs L s i i ,
. . . . As is Reasonably Achievable (Nuci: r
! ~-
. Power Reactors)
~
i 8/18/76 BTP A55' Guidelines for Fire Protection fc- l 9.5-1 Nuclear Power Plants Under Revieu cad " l Construction l
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. 4/13/77 BTP Materia 1' Selection and Processing l
! MTEB
~
5-7 Guidelines for BWR Coolant Pree-" ^ 1 Boundary Piping j (5)' 1/31/78' SRP 5.4.7 1 Residual Heat Removal System j
; i
.} (6) 1/31/78 RG 1.141 0 Containment Isolation Provisions T:e !
, Fluid Systems l 1
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i (1). Indicates that the category II assigned by RRRC,for the previous revinica ljj of this document das retained. Review by.-t-ha RRRC for retssignment af ci .... - _the category is net requir.ed . for . document..fev,isions * ' " which db not ; , i result in-an increate in requirements. , 4 (2)~ Arsigned as a Category I matter for those applications previo'usly reviewcd to revision 0. Category II for all other applications. ...a , . . , - . - . . (3) . category II for operating reactors. Assigned by the RRRC as a Category I!! matter for all other applications. - (4) Category II for paragraph C.l for all CP's or PDA's under review wh'ase [i SER's have not been issued prior to the implementation date shown in i; the published guide. Paragraph C.2 for.all operating reactors, OL i; applications, and CP and PDA applications under review whose~SER's cre
** ' completed prior to the implementation date shown in the published guide.
Certain provisions of this guide are also assigned by the RRRC as Catc; cry j , III matters. .
'l j (5) Category II for operating reactors and all other applications for which
- the issuance of the OL is expected prior to January 1,1979. Assigned 4 by the RRRC as a Category III' matter for all other applications.
i (6) Category II for operating. reactors and OL reviews. Assigned by the 332C as a Category III. matter for all other applications. j
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a e. e e , e i, , ENCLOSURE-- C l! . CATEGORY III MATTERS APPROVED BY RRRC
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FROM MARCH 1974 THROUGH I.UGUST 1978
.-s..-' ... .
ll EFFECTIVE - -
/I7 . : , . I., .. .
(l DATE DOCUMENT REVISION TITCE . 5/16/78 RG 1.56 1 Maintenance of Water Purity in Coiling Wat er. .Re actprs,,, (1) 5/16/78 RG 1.68.2 1 Initial Startup Test Program to C:::n-i strate Remote Shutdown Capability f:r Water-Cooled Nuclear Fewer Plants
- I Effects of Residual Elements on Prrdictr
, (2) 1/14/77 RG 1.99 1 Radiation Damage to Reactor Yessei Materials
' (3) 3/77 RG 1.101 1 Emergency Planning for Nuclear Power Plants 11/76 RG 1.114 1 Guidance on Being Operator at the Controls of a Nuclear Power Plant 5/11/76 RG 1.121 ' -0 Bases fcr Plugging Degraded PWR Steam i
;7; Generator Tubes l 11/29/77 RG 1.127 1 Inspection of Water-Control Structurcs
; Associated with Nuclear Power Plan:s (4) 9/27/77 RG 1.137 0 Fuel Oil Systems for Standby Diesei
+
Generators
~
{ (5) ,l/31/78 5RP 5.4.7 1 Residual Heat Removal System i (6) 1/31/78 RG 1.141 0 Containment Isolation Provisions fer
- Fluid Systems I 3/14/78 RSB 5-2 0 Reactor Coolant System Overpressuriza-
! (7) tion Protection J .
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; .. (2) . Paragraph.C.3 is a-Category.I' matter. ., . i (3) Indicates that the Category'III assigned by RRRC for the' previous Review by the RRRC for revision reassignment of this of document the category was retained is not requ ,, ired'for document revisic*ns
[ which do not result in an increase in requirements.
! (4) Category III for paragraph C.2 for all CP and PDA applications endar
.it review whose SER's have not been issued prior to the implementation
'date shown in the published guide. Certain provisions of this guide
- .' are also assigned by the RRRC as Category II matters.- ,
1
- l. (5) (ategory III for CP or PDA applications docketed prior to January 1, 1978, and for which OL issuance is hpected cfter January 1, if 79, j all Category II for all other applications. .
1 (6) Assigned by RRRC as a Category II matter for operating reactors and
, OL applications.
(7) Assigned by RRRC a a Category I matter for CL's issued prior to
.Maren 14,1978, and Category III for all other applications. ,
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! NRR CATEGORY IV MATTERS -
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! A. Regulatory Guides not categorized . .. , , , ,, .. ,
- ' Issue l Date Number Revision Title 4/74 , 1.12 1 Instrumentation for Earthqua'Kes ;
i 12/75 1.13 1 Spent Fuel Storage Facility Cesign Basis
~
8/75 1.14 1 Reactor Coolant Pump Fly.< heel Integrity
. 1/75 1.75 1 Physical Independence of Electric Systems 4/74 1.76 -
0 Design Basis Tornado for Nuclear Pcwer
~-
Plants 9/75 - 1.'79 1 Preoperational Testing cf Emergency Core Cooling Systems fer Pressuri sd Water Reactors Preoperational Testing of Instrumen*, 6/74 1.80 O Air Systems 6/f4 1,82 0 Sumps for Emergency Core Cooling cni
~
- Containment Spray Systems 7/75 1.83 1 Inservice Inspection of Pressuri:cd Water Reacter Steam Generator Tu s 11/74 1.89 0 Qualification of Class 1E Equipment
- for Nuclear Power Plants 12/74 1.93 0 Availability of Electric Power Scu'rces 2/76 1.104
- 0 Overhead Crane Handling Systems for
, Nuclear Power Plants S
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< : - Impl ementa-Appl i c'a5Th'_'I- . ~
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'~ tion Date Branch "SRP Section - .
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- 1. 1.1/24/75 tiTEB 5.4.2.1 BTP MTEB-5-3, . Monitoring i l 'i of Secondary Side Water j - .. , Chemistry in PWR Steam.
- Genera' tors
! <! 2, 11/24/75 CSB 6.2.1 BTP CSB-6-1, Minimum j
! 6.2.1A Containment Pressure M::e1 4
l' 6.2.18 .for PWR ECCS Performance 6.2.1.2 , !. Evaluation . i 6.2.1.3
, 6.2.1.4-6.2.1.5 i 3. 11/24/75 CSB 6.2.5 BTP CSB-6-2, Control of j
, Combustible Gas Concentra-i tions in Containment Folicwin; a Loss-of-Coolant Accident
- i. . .-
! 4. 11/24/75 ,.5SB' 6.2.3 BTP CSB-6-3, Determinatien cf f - Bypass Leakage Path in Qual l Containment Plants
! 5. 11/24/75 CSB 6.2.4 BTP CSB-6-4, Containmen:
Purging During Normal Plant l , Operations - i l
- 6. 11/24/75 ASB 9.1.4 BTP ASB-9.1, Overhead Har.dling
. . . Systems for Nuclear Power Plants j 7. 11/24/75 ASB 10.4.9 STP ASB-10.1, Design Guideliner
. for Auxiliary Feedwater Sys:cm i '
Pump Drive and Power Supply
; ; Diversity for PWR's i i
- 8. 11/24/75 SEB 3.5.3 Procedures for Composite See:fcn i- Local Damage Prediction (~.7,?
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Section 3.5.3, par.11.1.0) 4 6 g
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I c . 1" . l Implementa- Applicable tion Date . SRP Section w Title l -Branch
, - ... . w. .-.*
- e *.
- 9. 11/24/75 - SEB - 3.'7.1 J L .,Deiel,opmebt of Design Tim History for So'il-Strue:urc Interaction Analysis (ir.? 1
.] . Section 3.7.1," par. II.2)
.. , Pro,cedure,s for Seismic.Syst:.n
. 10. 11/24/75 SEB , .3.7.2 Analysis (SRP Section 3.7.2 par.II) ...
j 11. 11/24/75 SEB 3.7.3 Procedures for Seismic Sub-i system Analysis (SRP Sectica 3.7 i
- par.II) l -12. 11/24/75' . SEB 3.8.1 Design and Construction of Concrete Containments) SP.P i -
Section 3.8.1, par. II) - l
,, 13. 11/24/75 SEB- 3.8.2 Design and Construction of Steel Containments (SRP Se::icn 3.8.2, par. II) ,
.14. . 11/24/75 :SEB 3.8.3 Structural Design Criteria fer' Categcry I Structures In:id:
Containment (SRP Section 3.J.3, par.II) .
- 15. 11/24/75 SEB 3.8.4 Structural Design Criteria. f -
Other Seismic Category I Orc:=
~ .' ' ' * .
. (SRP Section 3.8.4, par. II) l
- 16. 11/24/75 SEB 3.8.5 Structural Design Crfteria for '
'~
Foundations (SRP Secticn 3.0.5, par.II)
- 17. 11/24/75 SEB 3.7 Seismic Design Requirer.ents for ,
i 11.2 Radwaste Sysems and Their H:usint j' . 11.3 Structures (SRP Section 11.2, BT? 11.4 ETSB 11-1, par. B.y) i - 1
- j. ,
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. Implementa-- . Applicable -
l- tion Date Branch SRP Section Title
- i
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j 18. 11/24/75 SEBL '.. 3.3,.2 f.1..Ternado,.Lo'ad..E ffec.t Combi . J. nations (SRP Sect' ion .3.3.2,. ) par. II.2.d) -
. , ,,, _j
; 19. 11/24/75 SEB-. 3.4.2 .D i
~ T(ynamic SRP Sectic'n Efects of Wave 3'.*4.2, par. ActionII)
{
- 20. 10/01/75 ASB 10.4.7 Water Ham.er for Steam l Generators with Preheaters (S7? l Section 10.4.7 par. , I.2.b)
I
- 21. 11/24/75 AB 4.4 Thermal-Hydraulic Stability (fr?
Section 4.4, par. II.5) j 1
- 22. 11/24/75 RSB 5.2.5 Intersystem Leakage Detecti:n (SU
. Section 5.2.5 par. II.4) c-c R.2. ".
i 23. 11/24/75 RSB 3.2.2- Main. Steam Isolation Valve Lesage Control System (SRP Sectica D.3 par. III.3 and BTP RSB-3.2) C. Other Positions, Implementa- Applicable tion Date Branch SRP Section Titl e
- 1. 12/1/76 SEB 3.5.3 Ductility of Reinforced Cen:rtta
~
and-Steel Structural Elema.1:: Subjected to Impactive or Irpuhtye - )
. Loads .
'2. 8/01/76 SEB 3.7.1 Response Spectra in Vertical Direction
- 3. 4/01 /76 SEB 3.8.1 BWR Mark III Centainment Pcol l
3.8.2 Oynamics
- 4. 9/01 /76 SEB 3.8.4 Air Blast Loads si
, 5. -10/01/76 SEB 3.5.3 Tornado Missile Impact Passive Failures During Lent-
; 6. 6/01/77 RSB ,
6.3 i Term Cooling Folicwing LCCA
- l
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- Impl e. ment a- Applicable j-tion Date
- Branch SRP Section Title -
~ -
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!, 7. 9/01/77 , .335B 6.3 c.gC,ontrol Recta Pccition Indic:- . ? - -
- l. .
, - tion
. ~"in theof ECCS Manual (Handwheel)'Vaiva
~
.s II 8. 4/01/77 RSB 15.1.5 Long-Term RecoveFy frcm Steinlica Break: Operator Action to Preven-
* "* Overpressurization - -
, 9. 12/01/77 RSB 5.4.6 Pump Operability Requirements
!i 5.4.7 ,
!> . 6.3 . . . . ..
- 10. 3/28/78 RSB 3.5.1 Gravity Missiles, Vessel Seti
! Ring Missiles Inside Cent:frmant
- 11. 1/01/77 ' AB 4.4 Core Thermal-Hydraulic Analysis
. 12. 1/01/78 PSB 8.3 Degraded Grid Voltage Conditicns
! 13. 6/01/76 CSB 6.2.1.2 Asymmetric Loads on Cc=penents
.' . , . . Located Within Containment Sub-
..,_ compartments 14.' 9/01/77.. ;CSB 6.2.6 Containment Leak Testing Pr:gr:m
- 15. 1/01/77 CSB '
6.2.1.4 ' Containment Response Due to ??.in Steam Line Break and Failure cf
! . MSLIV to Close <
l 16. 11/01/77 ASB 3.6.1 Main Steam and Feedwater Pipe
'. 3.6.2 Failures ... .
- 17. 1/01/77 ASB 9.2.2 Design Requirements for Cooling
.- Water to Reactor Coolant Pu.mps i .- .
- 18. 8/01/76 ASB 10.4.7 Design Guidelines for Water Haxer in Steam Generators with Tco Feedring Design (BTP ASB-lC.2)
~
lb. 1/01/76 I'C3B 3.11- Environmental Control' Sy:ttms -for Safety-Related Equipment -
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. ENCLOSURE E . ~
DESCRIPTION OF OTHER POSITIONS IDENTIFIED AS NRR CATEGORY IV MATTERS IN ENCLOSURE D 1 l l I Numbering scheme corresponds to that used in Item C of l
. Enclosure D; e.g., the first "Other Position" identified I l
as a Category IV matter'in Item C of Enclosure D is designated I,V.C.1, etc. _ e s g( 1
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', ENCLOSURE E )
i); " j IV.C.1 DUCTILITY OF REINFORCED CONCRETE AND STEEL STRUCTURAL Ei.EMENTS ' (3.5.3) SUBJETED TO M2fiVE OR IMPULSIVE LOADS "
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' In the ev .luation of overall response of reinforced' concrete struc; ural elements (e.g., missile barriers, columns, slabs, etc.) subjec:cd :c .
l impactive or impulsive loads, such as . impacts due.to.cissiles, .u:.:s:icn
,' of non-linear response (i.e., ductility ratios greater than uni:y) cf ,l the structural elements is generally acceptable provided that the tcfety functions of the structural elements and those of safety-related nr.+ s j' and components supported or protected by the elements are maintained. ~
The folicwing surmari:es specific SEB interim positions for revim :nd acceptance of ductility ratios for reinforced concrete and steal , structural elements subjected to impactive and impulsive loads. 1 SPECIFIC POSITICNS -
- 1. REINFORCED CCNCRETE MEMBERS 1.1 For beams, slabs, and walls where flexure controls design, the permissible ductility ratio ( y ) under impactive and impulsive loads should be taken as ,
. p = 0.05 for p-p' > .tOS i
..op' . ....
y = 10 for p p' <
._ .005 where p and p'are the ratios of tensile and compressive
. einforcing as defined in ACI-318-71 Code.
1.2 If use of a ductility ratio greater than 10 (i.e., u> 100) 7 is required to demonstrate design adequacy of structural elements against impactive or impulsive loads, e.g., missile impact, such a asage should be identified in the plant SAR. Information justifying the use of this relatively high ductility ' - value shall be provided for SEB staff review. , , ,,.
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- L . t i . I i 1.3 For beam-columns, walls, and slabs carrying axial ccmpression ", l ' - l loads and . subject to impulsive or impactive leads producing flexure, the permissible ductility ratTo'% flexure shouJd.
~ -
4 l ' be'as follows. :
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- (a) khen compression controls the design, as defined by' an
' interaction diagram, the. permissible ductility ratio
. shall be 1.3. . .. .. .....,..,.,,,_,,,,,_;.. ,
(b) khen the compression loads do not exceed 0.l fc'Ag or one- , third of that which would produce balanced conditions, which- . ever is smaller, the permissible ductility ratio can be as
.given in Secticn 1.1. ,
(c) The pemissible dutility ratio shall vary linearly from 1.3 to that given in Section 1.1 for conditions between those specified in -(a) and (b). (See Fig 1.) 1.4 For structural elements resisting axial ccmpressive impulsive c'r
.impactive loads only, without flexure, the pemissible axial
! ductility ratio shal1 ~ be 1.3. . . a..'.. i i e: . . . . .
1.5 For shear carried bf concrete only y . 1.0 .- For' shear carried by concrete and stirrups or bent bars' .',
~ '
y = 1.3 [ .. For shear carrie'd entireb tirrups y = 3.0 . .
~
2.0 STRUCTURAL STEEL MEMBERS __ , 2.1 For flexure compression and shear a y = 10.0 .
- 2. 2 For coltans with slenderness ratio (1/r) equal to or less than 20 r
~
- p. = 1.3 .
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IV.C.2 RESPONSE SPECTRA IN THE VERTICAL DIRECTION l
.(3.7.1) .
Subsequent to the issuance of Regulatory Guide 1.50, the report
'; " Statistical Studies of Vertical and Horizontal Earthquake Spe:tra" was issued in Jcnuary 1976 by NRC as NUREG-0003. One of the important conclusions of this' report is that the response spectrum for vertical motion .can be taken as 2/3 the response spectre 1 for horizontal motion over the entire range of frequencies in the We: tern
' Unit'ed Statese According to Regulatory Guide 1.60, the vertical response spectrum is equal .to the horizcntal respense spectrum between 3.5 cps and 33 cps. For the. Western United States only, consistsn:
with the latest available data in NUREG-0003, the option of taking the vertical design design, response spectrum as 2/3 the horizontal res;cnse spectrum over the entire range of frequencies will be accepted. For other locations, the vertical response spectrum will be the sa o as that given in Regulatory Guide 1.60. . s- -
^
IV,C.3 SWR MARK III CONTAINMENT POOL DYNAMICS . (3.8.1 . 3.a.2)
- 1. POOL SWELL <
- a. Bubble pressure', bulk swell and froth swell loads, drag ,
pressure and other pool swell loads should be treated as # abnormal pressure loads, Pa - Appropriate load combinations and load factors should be applied accordingly,
- b. The pool swell loads and accident pressure may be combined in accordance with their actual time histories of eccurrence.
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f.1 . , 5 2. SAFETY RELIEF VALVE (SRV) DISCHARGE t . i a. The S,RV' loads should be treated as live loads in all load 5 combinations 1.5Pa where a load factor of 1.25 should be . -
!, applied.to the appropriate SRV.-1 cads., _ .. ..,: ,, ,
- !. t - '.. . : ..~
!1 b. A single active failure causing on. 'SRV e 'd' iscRarge %uit ~
ll be considered in combination with the Design Basis ' Accident (DBA). -
- c. Appropriate multiple SRV di; charge'should.be considered in
. combination with the Small Break Accident (SBA) and Inter-j- .
mediate Break Accident (IBA). , I; d. Thermal loads due to SRV discharge should be treated as T 0 for normal operation and aT . f r accident conditions.
- e. The suppression pocl liner should be designed in accordance U.v;: 'with-the ASME Boiler and Pressure Vessel Code, Division 1
~'
2 Subsection ftE to resist the'SRV negative pressure, considering
;: strength, buckling and low cycle fatigue.
;: l' IV.C.4 AIR BLAST LOADS J (Pa, Ta, To as defined in ACI 3S9-740)
; (3.8.4 ) ,. ,
The following interim. position on air blast loadings on fluelear power Plant. Structures should be used as guidance.in evaluating analyses. '
. 1. An equ' i valent static pressure may be used for structural analysis purposes. The equivalent static pressure should be obtained frca '
the air blast reflected pressure _or the overpressure by multiplying l these pressures by a factor of two. Any proposed use of a dyn:mic
- load factor less than two should be treated on a case by case basis, i
Whether the reflected pressure or the overpressure is to'be use: f:r individual structural elements depends on whether an incident blas: wave could strike the surface of the element. *
- 2. No load factor need be'specified for the air blast loads, and the load combination.should be: .
'l. U=D+L+B where, U is the strength capacity of a section D is dead load . <
L is live load
- e
.. 8 is air blast load.
- 3. Elastic analysis for air blast is required for concrete structures a of new plants. For steel structural elements, and also for rein- -
forced concrete elements in existing plants, some inelastic resp:r.se may be permitted .with app opriate limits on ductility ratios.
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- j. 4. Air blast generated ground shock and air blast wind pressuFe may-i; be ignored. Air blast generated missi4es may be importent in ij sit' uations where explosions arbpoitulated to occur..in vessels -
i! which may fragment. , ,
- 2. . , . . , . -
.5. Overturning and sliding stability should be assessed by multiplying the structure's full projected area by the equivaleft static l . ' pressure and assuming only the blast. side of the structure is loaded. Justification for reducing (he average equivalent st'atic
; pressure on curved surfaces should be considered on a case by case
, basis. .
l 6. Internal supporting structures should also be analyzed fer the effects of air blast to determine their ability to carry loads applied directly to exterior panels and slabs. lioreover. in-vented structures, interior structures may require analysis enn if they do not support exterior' structures. ,
~
l 7. The equivalent static pressure should be considered as potenticily acting both inward and outward.
-{ . ,
. -IV.C.5 TORNADO MISSILE PROTECTION -
I (3.5.3) ,, ,
, j
- As.an interim.messure,the' minimum concrete wall and roof thickness for tornado missile protection will be as follows:
Wall Thickness Roof Inict.ne:: l Concrete Strength (psi) (inches) (inches) 1
~
3000 ~ 27 24 l Reg,1.cn I. 4000 24 21 w ~ . . 5000 21 18 i ' 3000 24 21
- Region II 4000 -
21 18 , 5000 19 - 16 3000 21 16 J Region III 4000 18 16 5000 16 14 These thicknesses are for protection against local effects only. Designers must establish independently the thickness requirements for overall strue:; response. Reinforcing steel should satisfy the provisions of Appendix C, - 349 (that is, .2% minimum, EWEF). The regions are described in Re ulatory Guide 1.76. -
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t . . I- - 11 . IV.C 6 , 3 (6.3) -PASSIVE ECCS FAILURES DURING L0tlG-TERM CO i ! - . . v .- - ~ , Passive failuris in the ECCS, having 1pa , l ii rates equal to or leds than the long-term cooling period following sidered. ,, e ::n-a po( r ng ( having in the plant design design. features "and " " bases as described
' " * ~ .
i I l The leak detection system should include detectors and alarm alert the operator of passive ECCS leaks in sufficient time ! priate diagnostic and corrective actions may be taken on a tim The diagnostic and corrective actions would include then ident .
, isolation subsystem isof the faulted ECCS line before the performance degraded.
end i include: The design bases of the leak detection system sncu (1) f Identification and justification of the maximum leak rate; (2) i Maximum allowable time for operator action and justificatio e :r;
* (3)
Demostration that the leak dettetien system is sensitive ugh to eno
' initiate and alarm on a timely basis, i.e. , with sufficient s: tima le the leak can create undesireable consequ dundant equipment. ne b f.r:
ng :f a-(4) The minimum time to be considered is 30 minut ECCS train and that the leak can
, and be isolated; aultad D I
(5) alarms and a leak detection system that :- me'nts be imposed.of IEEE-279, except that the si.ngle failure criteri IV.C.7 - . . . (6.3) _ CONTROL ROOM POSITIOff INDICATION OF M Regulatory Guide 1.47 specifies automatic positioneach indicatien bypass three or deliberately conditions are met: ., induced inoperable condition if th (1) designed to perform an automatic safety functi g .
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l; (2) The bypass or inoperable condition cangeasonably be expected - to occur. gore frequently than oncq-per year. . ll h, \ (3) The bypass' or in6perable.conditiE i's ' expel:t'ed 'to . occur when 1he l system is normally required to operate. '
- Q.L
~
'i Revision one of the Standard Review Plan in Section 6.3 requires
] conformance with Regulatory Guide 1.47 with.the intent beir.g that any manual (handwheel) valve which could jeopardize the i
j g operation of the ECCS, if inadvertently left in the wrong positi:n, h must have position indication in the control rocm. In the pC; stre < , o reviews it-is important to confirm that standard.. designs ir.:iu:e in: 2 0 .
. design feature. Most standard desi.gns do but this matte'r was proha .
;\ not specifically addressed in scme of the first PDA reviews.
4- ]
-l IV.C.8 LONG-TERM RECOVERY FROM STEAM LINE BREAX - OPERATOR ACTICN TO
., (15.1.5) PREVEili OVERPRESSURIZATION (FWR) a A steam line break causes cooldown of the primary system, shrinkage of 4
f 4 RCS inventory and depletion of pressurizer fluid. Subsequent to plant trip ~, ECCS actuation, and main. steam system isola tion, the RCS inven-
' tory increases and expands, refilling the pressurizer, Without operat:r 1 action, replenishment of RCS inventory by the ECCS and expansion :,t icw temperature could repressurize the reactor to an unacceptable pressure-temperature regi'on thereby cerrpremising reactor vessel integrity. A.ai-
, I yses-are required.to show that following a main steam line break that
; (i) no additional fuel failures result frca the accident, and (ii) the j pressures following the initiation of the break will not compromise tha
.t integrity of the reactor coolant pressure boundary giving due considera-j tion to the changes in coolant and material temperatures. The analy:es
- should be based on the assumption that operator action will not be taken l
until ten minutes after initiation of the ECCS. IV.C.9 PUMP OPERABILIT/ REOUIREMENTS , (5.4.6 1 5.4.7' In some reviews, the staff has found reasonable doubt that some types of 1 6.3) engineered safety feature pumps would continue to perform their safety function in the long term following an accident. In such instances there , has been follewup, including pump redesign in some cases, to assure
.; that long term performance could be met. The following kinds of infor- ,
1
? mation may be sought on a case-by-case basis where such doubt arises. l
, a. Describe the tests ' performed to demonstrate that the pumps are capable of operating for extended periods under post-LOCA conditions. -
including the effects of debris. Discuss the dacage to pump seals Aaused by debris over an extended period of operation.
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- 1, b. Provide detailed diagrams of all water cooled seals and compo'-
nents in the pumps.
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i c. Provide a description of fhe composition of the pump shaft l seals and the shafts. Provide an evaluatra.: of lost..of shaft seals.
' ~
- d. Discuss how debris and post-LOCA An'/i'rerdent'51 ' conditions wer"e factored into the specifications and det,ign of the pump.
i 3 " ' ~ i IV.C.10 GRAVITY MISSILES. VESSEL SEAL RIllG MISSILES INSIDE CONTAlfiMENT ' (3.5.1) Safety related systems should be protected against loss of functica due to 1 internal missiles from sources such as those associated with pre::urizac g ccmronents and rotating equipment. Such sources would include but nct be i limited to retaining bolts, centrol rod drive assemblics, the vessal saal ring, valve bonnets, and valve stems. A description of the methe.. used
- to afford protection against such potential missiles, including the bases therefor, should be provided (e.g. , preferential orientation of the poten-tial missile sources, missile barriers, physical separation of redundant 1
sa fety systems and ccmponents) . An analysis of the effects of such potan- { tial missiles on safety related systems, including metastably supr:r:ed i
- equipment
;-t...., m which en -could w re.fall -.upon impingement, sho,uld also be proviced.
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a e i i I IV.C.ll CORE THERMAL-HYDRAULIC ANALYSES ' (4.4) l j In evalu'a ting'the thermal-hydraulic"p"erf6rmance of thb . reactor ~ p core,the following additicnal, areas sh'cG1d .txa addressed: . ,
'. l
- 1. The effect of radial pressure gradients at the exit.of open lattice cores. -
- 2. The effect of radial pressure gradients in the upper plenum.
1
'I 3. The effect of fuel rod bowing, j In additicn.a commitment to perform tests to verify the transient
, analysis methods and codes is required.
I V.C.12 DEGRADED GRID VOLTAGE CONDITIONS - - (8.3) As a result of the Millstene Unit Number 2 low grid voltage occurrence, the staff has developed additional requirements concerning (a) sustairsd degraded voltage conditicns at the offsite power source, and (b) in:e:- action of the offsite and onsite emergency power systems. These ad:ition:I
, requirements are defined in the following staff position.
t .....
! 1. We require that a second level of voltage protection for the e--" e power syst'em be provided and that this second level of voltage pr;-
tection satisfy the follcwing requirements: . l
. a) The selection of' voltage and time set points shall be determined from an analysis of the voltage requirements cf the safety-related loads at all onsite system distribution
.- levels;
. . s b) The voltage protection shall include coincidence logic to preclude spurious trips of the offsite power source; t
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c) The time delay selected shall be based on the' following
- conditions:
.!' (i) .The all'owable time delay, including ma,rjin, shall .
! not exceed the maximum time delaywthat is assumed in ~ ' "
the - SAR ' accident analyses; , - ~. . y , , --- . . . ,.. .. . . I - (ii) The time delay shall minimize the effect of shorti . '
,j duration disturbances from reducing the availability
~
, ' f 'the offsite power source (s); and o
1 -
! (iii) The allowable time duration of a degraded voltage conditicn at all distribution system levels shall not result in failure of safety systems or components; (iv) The voltage sensors shall automatically initiate the disconnection 'of offsite pcwer sources whenever the voltage set point and time delay limits have been exceeded; (v) The voltage sensors shall be designed to satisfy the applicable requirements of IEEE Std. 279-1971 " Criteria
. for Protecticn Systems for Nuclear Power Generating Stations"; and ,
(vi) The Technical Specifications shall include limiting
. conditions for ~ operation, surveillance requirements, trip set points with minimum and maximum limits, and
' allowable values for the second-level voltage protection
~
sensors and associated time delay devices.
- 2. We require that the system design automatically prevent load shedding of the emergency buses once the onsite sources are supplying power to all sequenced loads on the emergency buses.
The design shall also include the capability of the load shedding feature.t'o be automatically reinstated if the onsite source supply breakers are tripped. The autcmatic bypass and reinstatement feature shall be verified during the periodic testing identified in item 3 of this position. . i 1
; ~. 3. We require that the Technical Specificatiens include a test require-ment to demonstrate the full functional operability and independence of the onsite power sources at least once per 18 months during shut-
. . down. The Technical Specifications shall include a requirement for tests: (a) simulating loss of offsite power; (b) simulating loss e
- of offsite power in conjunction with a safety injection actuation -
signal; and (c) simulating interruption and subsequent reconnection of onsite power sources; to their respective buses. l
)
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- 4. The voltage levels at the safety-related buses should be .
optimized for the full load and minimum load conditions th3t are expected throughout the ang.1 p variati' ons of the off site power.4j sourceat?e'd by range,appropriateof voltage adjust-ment of the voltage tap settingr 'of the intervening-transformers. We require that the adequacy of the design in this regard be perified by actual measurement, and by correla' tion.of measured i values with analysis results.
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. l
. i IV.C.13' ASYMMETRIC LOADS ON COliP0MENTS ,
(6.2.1.2) LOCATED WITHIN CONTAIhMENT SUBCOMPARTMENTS. { In the unlikely event of a pipe rupture inside a major cceponent sub-compartment, the initial bicwdown transient would-lead to ressure loadings en both the structure and the enclosed component ( ). Tha staff's generic Category A Task A: tion Plan A-2 is designe to develcp
- generic resolutions for this matter. Our present schedule calls for completing A-2e for PWR's during the first quarter,1979. Pending completion of A-2, the staff is implementing the following program:
s 1. For PWRs at the' CP/PCA stage of review, the staff requires appli-cants to comr.it to address the safety issue as part of their appli-
~
c,ation for an operating license. -
. 2. . For PWRs. at the OL/FDA stage of review, the staff requires case-by-c-analyses, including implementation of any indicated corrective measusres . prior to the issuance of an operating license.
- 3. For BWRs, for which this issue is expected to be of lesser safetv significance, the asymmetric loading conditions will be evcluasid on a case-specific basis prior to the issuance =of an operating lice -
For those cases which analyses are required, we request the perfernanca of a subcompartment, multi-node pressure response analysis of the pressure transient resulting from postulated hot-leg and cold-leg (pump suction and discharge) reactor coolant system pipe ruptures within the reactor cavity, pipe penetrations, and steam genera:cc compartments. Provide similar analyses for the pressurizer surge -
. and spray lines, and other high energy lines located in containman compartments that may be subject to pressurization. Show how the results of these' analyses are used in the design of structures and ,
component supports. - 9 6 Y
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,. 3 - j; i : IV.C.14 CONTAINMENT LEAK TESTING PROGRAM . s j '. (6.2.6)
- j' To avoid difficulties experienced in this area in recent OL re/iec, i
j- the staff.has increased its scope of. inquiry at the CP/PDA stage a?
- g review. For this purpose, the following information with regard t)
- j' the containment leak testing prog'r'aWsficuld be supplied.
! : . ;t Those systems that wili' remain fluid filled for th'e Type A tut h'! a.
- ;j .should be identified and justification given. - ' . , ,
1 .t
'i '
- b. Show the design provf sions that will pennit the personnel cir.
lock door seals and the entire aFlock to b'e tisted.
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,' c. For each penetration.i.e., fluid system piping, instrument, electrical, and equipment and personnel access penerations,
- i, identify the Type B and/or Type C local leak testing that j , , will be done. .
- d. Verify that containment penetrations fitted with expansien bellows will be tested at Pa. Identify any penetraticn fit *si with
, expansion bellows that does not have the design capability 4 -
for Type B testing and provide justification. '
! N a .
i IV.C.15 CONTAINMENT RESPONSE 00E TO' MAIN STEAM LINE - l (6.2.1.4) BREAX AND MSLIV FAILURE , 1
- l In recent CP .an'd" 0L application reviews, the results of analyses for a postulated main steam line break accident (MSLS) for designs utilizing pressurized water reactors with conventi~.ti containments show that the peak calculated containment tempera *.ure
. can exceed for a short time period the environmental qualificatica i temperature-time envelope for safety related instruments and .
- . components. This matter was also discussed in Issue No.1 of
. NUREG-0138 and Issue No. 25 of NUREG-0153. The si'gnifiance of the matter is that it could result in a requirement for requalifying safety-related equipment to higher time-temper::gre envelopes. , l The staff's generic Category A Task Action Plans A-21 and A-24 are t designed to develop generic resolutions for these matters. The
- presentl Portion)y- are scheduled completion and first quarter,1979 dates for A-21 fourth and A-24 (Short quarter,1978, resscc
- ively. Tern Pending completion of A-21 and A-24, some interim guidance will be
- used as detailed below. -
l We have developed and are implementing a plan in which all applicants f:r , construction permits and operating licenses and those already issued c:r.- struction permits must provide information to establish a conserynive 4 temperature-time envelope. * } t
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Therefore, describe and justify the analytical model used to conservatively i determine the maximum containment temperature and pressure for a spectrum of j Incluce postulated main steam line breaks for various reactor - power levels. '
- the following in the discussion. "e v(, - ., ,
j j , .,.
- L (1) Provide single active failure analyses whfch specitically '
identify those safety grade systems and components relied up*on . l 6 to limit the mass and energy release and containment pressute/ ,
)
temperature response. The single failure analyses should include, but not necessarily be limited to:, main steam and
- l
' connected systems isolation; feedwater auxilfary'feedw'ateF, and l i
4 connected systems isolation; feedwater, condensate, and auxiliary feedwater pump trip, and ' auxiliary feedwater run-out centrol system; the loss of or availability of offsite pcuer; diesel failure when loss of offsite power is evaluated; and partial loss of containment cooling systems. (2) Discuss and justify the assumptions made regarding the time at which active contai'nment heat removal systems become effective. (3) Discuss and justify the heat transfer correlatien(s) (e.g., Tagami, Uchida) used to calculate the heat transfer frem the containment i atmosphere to the passive heat sinks, i.nd provide a plot of the i l i heat transfer coefficient versus time for the most severe steam line ) B.i.gak geident B analyzed.
.(4) Specify and justify the t'emperature used in the calculatien of condensing heat transfer to the passive heat sinks; i.e.,
specify whether th'e saturaticn temperature correspcnding to the partial pressure of vapor, or the atmosphere temperature (which may be superheated)was used. f (5) Discuss and justify the analytical model including the ther nodynamic equations used to account for the removal of the condensed mass from the containment atmosphere 'due to condensing heat transfer to the passive heat sinks; (6) Provide a table of the peak values of containment atmosphere temperature and pressure for the spectrum of breek areas and power levels analyzed; (7) For the case which results in the maximum containment atmosphere temperature, graphically show the containment atmosphere temperature, the containment liner temperature, and the containment concrete ' temperature as a function of time. Compare the calculated contain- - ment atmosphere temperature response to the temperature profile
- used in the environmental qualification program for those safety related instruments and mechanical comnonents needed to mitigate the consequences of the assumed main steam line break and effect
- safe reactor shutdown; O
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l j. (8) For the case which results in maximum containmergt atmosphere pressure, graphically show the contair.me W ppessure as ,a . .
- function of time; and -
i s i (9) For the case which results in t.he maximum containment atmosphere-i I: pressure and temperature, provide the mass and energy release J data in tabular form. { In order to demonstrate that safety-related equipment has been adequately
- !. , qualified as described above, provide the following information regard-
; ing its environmental qualification.
- (1) Provide a comprehensive list of equipment required to be operational in the event of a main 'steamline break (MSL3) accident. The list should include, but not necessarily be limited to, the follcuing safety related equipment:
(a) Electrical containment penetrations; i 1
.(b) Pressure transmitters; (c). . Containment iso.lation valves; 4
(d) Electrical powe[ cab'les; l I (e) Electrical instrumentation cable;. and curran: .. .'
~
(f) Level transmitters. Describe the qualification testing that was, or will be,'done on this 2;P.r Include a discussion of the test environment, namely, the temperet'ure, pressure, moisture content, and chemical spray,
.as a function of time.
(2) It is our p'osition that the thermal analysis of safety related equipment which may be exposed to the containment atmosphere l following a main steam line break accident should be based on the following: (a) A condensing heat transfer coefficient based on the recommendatiens in Branch Technical Position CSS 6-1, ..
" Minimum Containment Pressure Model for PWR ECCS Performance . .'-
Evaluation,"should be used. (b) A convective heat thansfer coefficient should be used when . the condensing heat flux is calculated to be less than the l convective heat flux. During the blewdown period it is l appropriate to use a conservatively evaluated forced convection heat transfer correlation. For example, e a
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. Since the Reynolds number is dependent en ve'locity, it is a necessary to evaluate the forced flow currents which will be ~
generated by .the steam generaor blowdown. The CVTR experiments provide limited data in this regard. Convective currents of . l from 10 ft/see to 30 ft/see were measured locally. We recommend that the CVTR test results be extrapolated conservatively to
, obtain forced flow currents to detemine the convective heat
. transfer coefficient during the blowdown period. After the -- -
l blowdown has ceased or been reduced to a negligibly low value,
-l a natural convection heat transfer correlation is acceptable.
(3) For each component where thermal analysis is done in conjunction with an environmental test at a temperature lower than the peak calculated temperature following a main steam line break accident
, compare the test the,rmal response of the component with the accident themal ' analysis of. the component. Provide the basis by which the component themal response was developed frem the environmental ! qualification test program. For instance, graphically show the
- thermocouple data and discuss the themocouple locations, method of attachment, and performance characteristics, or provide a .
detailed discussion of the analytical model used to evaluate the component themal response during the test. This evaluation should
. be perfomed for the potential points 'of failure such as thin cross-sections and temperature sensitive parts where themal stressing, temperature-related degradation, steam or chemical interaction at j elevated temperatures, or other thermal effects could result in the t . failure of the component mechanically or electrically. If the component thermal response comparison results in the prediction of-a more severe themal transient for the accident conditions than for the qualification test, provide justification that the affected component will perform its interded function during a MSLB accident, ,.
or provide protection for the cc mponent whch would appropriately . limit the themal effects. t 0 t 6- . p e
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!' IV.C.16 ENVIR0tiMENTAL EFFECT OF PIPE FAILURES.,, ,,, , , . , ,
(3.6.1, , Identify the " break exi:1usion" regionro'f'the ma-iri steam . - 3.6.2) and feedwater lines. Compartments that contain break - exclusion regicns of main steam and feedwater lines and any safety j related equipment in these compartments should be designed to wita-i stand the environmental effects (pressure,- temper,ature,. humidity an" flooding) of a crack with a break area equal to the cross sectionai area of the' break excluded' pipe. I- IV.C .17 DESIGN REQUIREMENTS FOR COOLING WATER ( 5.4.1 ) TO REACTOR C0CLANT PUMPS
- Demonstrate that the reactor coolant system (RCS) pump seal injeco ,
' flow will be automatically maintained for all transients and acci::9 a or that enough time and information are availela to permit l corrective action by an operator. ,
I We have established the following criteria for that portion of the I component cooling water (CCW) system which interf aces with the reacnr coolant pumps to su'pply cooling water to pump seals and bearings during ' normal operation. antigipated transients, and accidents.
- 1. A singl'e' active f ailure in the component cooling water system I shall not result in fuel damage or a breach of the reactor j coolant pressure boundary (RCpB) caused by an extended loss
- I of cooling to one or more pumps. Single active f ailures include operator error, spurious actuation of motor-operated valves, and loss of CCW pumps.
- 2. A pipe crack or other accident (unanticipated occurrence) shall not result in either a breach of the RCPB or excessive fuel damage when an extended loss of cooling to two or more RC pumps occurs. A single active falure shall be considered when evaluating the consequences of this accident. Moderate leakage cracks should be detenained in accordance with Branch Technical Position ASB 3-1.
In order to meet the criteria established above, an NSSS inter-f ace requirement should be imposed on the balance-of-plant CCW ', system that provides cooling water to the RC pump seals and motor and pump bearings, so that the system will meet the following con-ditions: .
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i: -1. That por' tion of the component cooling w4er (CCW) system which 1, supplies co. cling water to the reactop coolant pumps and motors . may be designed to, non-seismic Category I requirements and ,Cuclity l, Group D if it can be demonstrated tha't 'the Peactor, coolant puns j: --- will operate without component cooling water for.at least 30 i' minutes without loss of function or the need for operator pro-
!! tective action. In addition, safety grade instrumentation l including alarms should b,e provided to detect the loss,of -
component cooling water to the reactor coolant pumps and ll motors, and to notify the operator in the control recm. The
- j. entire instrumentation system, including audible and visual alar,as, j should meet the requirements of IEEE Std 279-1971.
i l If it is not demonstrated that the reactor coolant pumps and motors will operate at least 30 minutes without loss of function or oper::T-l' protective action, then the design of the CCW sys tem must meet the '
- - . following requirements:
M
.l 1. Safety grade instrumentation consistent with the criteria for
.; the reactor protection system shall be provided to initiate automatic protection of the plant. ' For this case, the component cooling water supply to the seals and pump and motor bearings. may be designed to non-seismic Ca tegory I requira- .
; m,ents and QJality Group 0; or
- 2. The ccmponent cooling water supply to the pumps and motors ;
shall be capable of withstanding a single active failuru or
, a moderate energy line crack as defined in our Branch 1 Technical Position APCSB 3-1 and be designed to seismic {
n Category-I, ~ Quality Group D a'nd ASME Section III, Class 3 l requirements.
~-
- The" reactor coolant (RC) pumps and motors"kre within the hSSS scope of design. Therefore, in order to demonstrate that an RC pump (
design can operate with loss of component cooling water for at least l 30 minutes without loss of function or the need for operator action, I the following must be provided: l
- 1. A detailed description of the events following the loss of component cooling water to the RC pumps and an analysis demon-strating that no consequences important to safety may result -
I from this event. Include a discussion of the effect that the - loss of cooling water to the seal coolers has on the RC pump l
- seals. Show that.the loss of cooling water does not result '
in a LOCA due to seal failure. . 5 1
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- 2. A detail'ed analysis to show that loss of_ cooling water to -
l1! the RC. pumps and motors will not caupe a loss of the flow f coastdown characteristics or cause seizure of, the pbmps, ;
! assuming no administrative action Ys 'taken." The' response l' should include a detailed description of the calculation '
-jf procedure including: -
I a. The equations used. -
-~ -
.: b. The parameters used in the equations, such as the design I: parameters for t.he motor _ bearings, motor, pump and any
- 1 other equipment entering into the calculations, and -
l
!, material property values- for the oil and metal parts.
- c. A discussion of the effects of possible variations in *
;- part dimensions and material properties, such as bearing clearance tolerances and misalignment.
- d. A description of the cooling and lubricating systems (with I e
appropriate figures) associated with the RC pump and motor l
! and their design criteria and standards.-
; c:a .
Information' t.o verify the applicability of the equations e.
-: _' and mate.cial properties chosen for the analysis (i.e.,
ref.erences should be listed, and if empirical relations
, are us'Ed, provide a comparison of their range of appli-
. cation to the range used in the analysis). -
Should an analysis be provided to demonstrate that loss of I component cooling water to the RC pumps and motor assembly is acceptable, we will . require certain modif.icatiens to the plant
.i . Technical Specifications and an RC pump test conducted under i
operating condtf ons and with component cooling water terminated for a specified period of time to verify the analysis.
+
IV.C.18 WATERHAMMERINSTEAMGENERATORSWITHTOPFEEDRINGDEShGN l Events such as damage to the feedwater system piping at Indian Point Unit No. 2, November 13, 1973, and at other plants, could originate as a consequence of uncovering of the feedwater sparger in the steam generator or uncovering of the steam generator - o feedwater inlet no:zles. Subsequent events may in turn lead to 'the - generation of a pressure wave that is propagated thrcugh the pipes and cculd result in unacceptable damage. e 4 w-
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- li l- -For CP/PDA and OL/FDA applications, provide the following for stcam h generators uti14 zing top feed: "c t C -
s,, . t
- l. Prevent or delay water driining from the feedring following a _
,, drop in steam generator water level 'by means such as,J-Tubes; I!
- l 2. Minimize the volume of feedwater pip,ing external to the steam l! -
generator whch could pocket steem usihg the 'shdrteit possible'
!l- (less than seven feet) horizontal run of inlet piping to the o steam generator feedring; and i .
- 3. Perform tests acceptable to the staff to verify that unaccept; le facu-water hammer will not occur using the plant opercting procedur:s for normal and emergency restoration of steam generator water ,
. level folicwing loss of normal feedwater and possible draining cr ,
i the feedring. Provide the procedures for these tests for staff cpprc'/? l
. before conducting.the tests.
Furthermore, we; request. that the following be provided: n] ll a. Describe normal operating occurrences of transients that , could cause the water level in +5e ,: cam generator to . t drop below the'sparger or no- ) cause uncovering and allow steam to enter the sr " e J.feedwater piping.
- b. Describ$ your criteria or sh ef isometric diagrams, the
. routing of the feedwater piping from tbc steam generatcrs outwards to beyond the containment structure up to the outer
- isolation valve and restraint. .
- c. Describe any analysis en the piping system including any
, , forcing functions that will be performed or the results
- of test programs to verify that,either uncoverinq of
'feedwater lines could not occur or that, if it did occur, unacceptable damage such as the experience at the Indian Point Unit No. 2 facility would not result with your design..
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'I IV C.19 E'!VIRONMENTEI. CONTFOL SYSTEMS FOR SAFETTtELATED E0!!IPMENT, . ' -
(3.h) Most plant areas that contain safety relat~ed equipmen't depend on the " continuous operation of environmental control systems to maintain the { environment in those areas within the range of environmentai qualification of the safety related equipment insta11ecf fr'1'those areas. It appears
' that there are no ' requirements for maintaining'those environmental control systems in operation while the plant is shutdown or in hot standby l
. conditions. During periods when these environmental control systems are shutdown, the safety related equipment could be exposed to environmental
. conditions for which it has not been qualified. Therefore, the safety related equipment should be qualified to the extreme environmental conditions that could occur when the control equipment is shutdown or j these environmental control systems should operate continuously to
]-
; maintain the .wironmental conditions within the qualification limits ;
of the safety related,. equipment. ' In the second case an environmental. monitoring system that will alarm when the environmental conditions exceed those for which safety related equipment is qualified shall be provided. This environmental monitoring system shall (1) be of , high quality, (2) be periodically tested and calibrated to verify its continued functioning, (3) be energi:ed from continuous power sources! and (4)
- provide a continuous record of the environmental parameters during
~
.the time the environmental conditions exceed the normal limits. 't he
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