ML19289D123
| ML19289D123 | |
| Person / Time | |
|---|---|
| Site: | Yellow Creek  |
| Issue date: | 02/02/1979 |
| From: | Boyd R Office of Nuclear Reactor Regulation |
| To: | Niav Hughes TENNESSEE VALLEY AUTHORITY |
| References | |
| NUDOCS 7902230034 | |
| Download: ML19289D123 (40) | |
Text
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UNITED STATES y '(
- 4 NUCLEAR REGULATORY COMMisslON
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W %SHINGTON, D. C. 20555 g,p/
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FEB 0 2 1979 Docket Nos. STN 50-566 and STN 50-567 Mr. N. B. Hughes Manager of Power Tennessee Valley Authority 830 Power Building Chattanooga, Tennessee 37201
Dear Mr. Hughes:
SUBJECT:
IMPLEMENTATION OF STAFF REVIEW REQUIREMENTS - YELLOW CREEK NUCLEAR PLANT, UNITS 1 AND 2 - OPERATING LICENSE REVIEW During the last several years we have reviewed and approved several new regulatory guides and branch technical positions or other modifications to existing staff positions. Our practice is that substantive changes in staff positions be considered by the NRC's Regulatory Requirements Review Committee (RRRC) which then recommends a course of action to the Director, Office of Nuclear Reactor Regulation (NRR). The recommended action includes an implementation schedule. The Director's approval then is used by the NRR staff as review guidance on individual licensing matters. Some of these actions will affect your application. This letter is intended to bring you up to date on these changes in staff positions so that you may consider them in your Final Safety Analysis Report (FSAR) preparation.
The RRRC applies a categorization nomenclature to each of its actions.
( A copy of the summary of RRRC Meeting No. 31 concerning this categori:a-tion is attached as Enclosure 1). Category 1 matters are those to be applied to applications in accordance with the implementation section of the published guide. We have enclosed lists of actions which are either Category 2 or Category 3, which are defined as follows:
Category 2: A new position whose applicability is to be determined on a case-by-case basis. You should describe the extent to which ynur design conforms, or you should describe an acceptable alternate, or you should demonstrate why conformance is
't necessary.
Category 3: Conformance or an acceptable alternative is required.
If you do not conform, or do not have an acceptable alternate, then staff-approved design revisions will be required.
790223oo30 A
FEB 0 21979 Mr. N. B. Hughes We believe that providing you with a list of the Category 2 and 3 matters approved to date will be useful in your FSAR preparation, and they will be an essential part of our operating license review. Enclosure 2 is a list of the Category 2 matters. Enclosure 3 is a list of the Category 3 matters.
In addition to the RRRC categories, there also exists an NRR Category 4 list which are those matters not yet reviewed by the RRRC, but which the Director, NRR, has deemed to have sufficient attributes to warrant their being addressed and considered in ongoing reviews. These matters will be treated like Category 2 matters until such time as they are reviewed by the RRRC, and a definite implementation program is developed.
A current list of Category 4 matters is attached (Enclosure 4). These also should be considered in your FSAR.
In some instances the items in the enclosures may not be applicable to your i.pplication. Also, we recognize that your application may, in some instances, already conform to the stated staff positions. In your FSAR you should note such compliance.
If you have any questions please let us know.
- erely,
(
Roger S.
oyd, Di r ter Division of Project Management Office ?f Nuclear Reactor Regulation
Enclosures:
As Stated cc w/ enclosures:
See next page
se Mr. N. B. Hughes Herber't S. Sanger, Jr., Esq.
cc:
General Counsel Tennessee Valley Authority 400 Commerce Avenue E llB 33 Knoxville, Tennessee 37902 Mr. M. J. Burzynski, Licensing Engineer Tennessee Valley Authority 400 Chestnut Street Tower - II Chattanooga, Tennessee 37401 Mr. E. G. Beasley Tennessee Valley Authority W10C131C 400 Commerce Avenue Knoxville, Tennessee 37902 Mr. R. L. Lumpkin, Jr.
CE Power Systems Combustian Engineering, Inc.
1000 Prospect Hill Road Windsor, Connecticut 06095 Ivan W. Smith, Esq., Chairman Atomic Safety and Licensing Board U. S. Nuclear Regulatory Commission Washington, D. C.
20555 Mr. Lester Kornblith
~
Atomic Safety and Licensing Board U. S. Nuclear Regalatory Commission Washington, D.
C.~
20555 Dr. Oscar H. Paris Atomic Safety and Licersing Board U. S. Nuclear Regulatory Commission Washington, D. C.
20555
UNITED STATES NUCLEAR REGULATORY COMMiss..
W ASHINGTON. D. C. 20555 SEP 2 4 G75 Lee V. Gossick Executive Director for Operations REGULATORY REQUIRElENTS REVIEW COMMITTEE MEETIllG N0. 31, JULY 11, 1975 1.
The Committee discussed issues related to the implementation of Regulatory Guides on existing plants and the concerns expressed in the June 24, 1974 memorandum, A. Giambusso to E. G. Case, subject: REGULATORY GUIDE IMPLEMENTATION, and made the following recommendations and observations:
Approval of new Regulatory Guides and approval of revisions a.
of existing guides should move forward expeditiously in order that the provisions of these regulatory guides te available for use as soon as possible in on-going or future staff reviews of license applicaticns.
The Committee noted that over the recent past, the approval of proposed regulatory guider whose content is acceptable for these purposes has experienced significant delays in RRRC review pending the determination of the applicability of the guide to existing plants, often requiring significant staff effort.
To avoit! these delays, the Comnittee concluded that, henceforth, approval of proposed regulatory guides should be uncoupled from the consideration of their backfit applicability.
b.
The implementation section of new regulatory guides should address, in general, only the applicability of the guide to applications in the licensing review process using, in so far as possible, a standard approach of applying the guide to those applications docketed 8 months after the issuance date of the guide for commant.
Exceptions to this general approach will be handled on a case-by-case basis, The regulatory position of each approved proposed guide (or c.
proposed guide revision) will be characterized by the Committee as to its backfitting pot'ential, by placing it in one of three categories:
Category 1 - Clearly forward fit only.
No further staff consideration of possible backfitting is required.
ENCLOSUFE: 1
Lee V. Gossick Category 2 - Further staff consideration of the need for back-fitting appears to be required for certain identified items of the regulatory position--these individual issues are such that existing plants need to be evaluated to detemine their status with regard to these safety issues in order to detemine the need for backfitting.
Category 3 - Clearly backfit.
Existing plants should be evaluated to detemine whether identified ittms of the regulatory position are resolved in accordance with the guide or by some equivalent alternative.
From time to time, for a specific guide, there will probably be some variation among these categories or even within a category, and these three broad category characterizations will be qualified as required to meet a particular situation.
d.
It is not intended that the Committee categorization appear in the guide itself. The purpose of the categorization is to indicate those items cf the r20alatory position for which the Comittee can make a specific backfit recomendation without additional staff work (Categories 1 and 3), and to indicate those items for which additional staff work is required in order to detemine backfit considerations (Category 2).
e.
The Comittee recommends that for approved guides in Category 2, staff efforts be initiated in parallel with the process leading to publication of the guide in order that specific backfit requirenents for existing plants be detemined within a reasonable period of time after publication of the guide, f.
The Committee observed that more atten-icn needs to be given to the identification of acceptable alternatives to the positions outlined in the guides in order to provide additional options and flexibility to applicants and licensees, with the possible benefits of additional innovation and exploration in the solution of safety issues.
2.
The Committee reviewed the proposed Regulatory Guide 1.XX: THERMAL OVERLOAD PROTECTION FOR MOTORS 0" MOTOR-OPERATED VALVES and recomended approval.
This guide was characterized by the Comittee as Category 1 - no backfitting, with the stipulation that as an appropriate occasion presented itself in conjunction with the review of some particular aspect of existing plants, the thermal overload protection provisions be audited.
ENCLOSURE 1 (CONT'D)
Lee V. Gossick 3.
The Comnittee reviewed the proposed Regulatory Guide 1.XX:
INSTRU iE!!T SPAtiS N1D SETPOItiTS and reco:xnended approval subject to the following comment:
Paragraph 5 of Section C (page 4 of the proposed Guice) should be reworded in light of Committee comments, to the satisfaction of the Director, Office of Standards Devalopr.en t.
This guide was characterized by the Committee as Category 1 - no backfit.
4.
The Comr.ittee reviewed Proposed Regulatory Guide 1.97:
ItiSTRU:'E iTATIO!! FOP. LIG9T ilATER COOLED f:U: LEAR PO*.1ER PLAf;TS TO ASSESS PLN;T CO:iDITIO:;S DU?.I!;G A!;D FOLLO'.111:G Ai ACCIDEfiT and deferred further consideration to a later meeting in order to permit incorporation of recent comments by the Division of Technical Review.
/b Edson G.
ase, Chairran Regulatory Requirements Review Committee ENCLOSURE 1 (CONT'D)
September 15, 1978 CATEGORY 2 MATTERS Document Number Revision Date Title RG 1.27 2
1/76 Ultimate Heat Sink for Nuclear Power Plants RG 1.52 1
7/76 Design, Testing, and Maintenance Criteria for Engineered-Safety-Feature Atmosphere Cleanup System Air Filtration and Adsorption Units of Light Water Cooled Nuclear Power Plants (Revision 2 has been published but the changes from Revision 1 to Revision 2 may, but need not, be considered.
RG 1.59 2
8/77 Design Basis Floods for Nuclear Power Plants RG 1.63 2
7/78 Electric Penetration Assemblies in Containment Structures for Light Water Cooled Nuclear Power Plants RG 1.91 1
2/78 Evaluation of Explosions Postulated to Occur on Transportation Routes Near Nuclear Power Plant Sites RG 1.102 1
9/76 Flood Protection for Nuclear Power Plants RG 1.105 1
11/76 Instrument Setpoints RG 1.108 1
8/77 Periodic Testing of Diesel Generator Units Used as Onsite Electric Power Systems at Nuclear Power Plants RG 1.115 1
7/77 Protection Against Low-Trajectory Turbine Missiles RG 1.117 1
4/78 Tornado Design Classification RG 1.124 1
1/78 Service Limits and Loading Combinations for Class 1 Linear Type Component Supports RG 1.130 0
7/77 Design Limits and Loading Combinations for Class 1 Plate-and Shell-Type Component Supports (Continued)
ENCLOSURE 2
CATEGORY 2 MATTERS (CONT'D)
Continued Document Number Revision Date Title RG 1.137 0
!/78 Fuel Oil Systems for Standby Diesel Generators (Paragraph C.2)
RG 8.8 2
3/77 Information Relevant to Ensuring that Occupational Radiation Exposures at Nuclear Power Stations Will be as Low as is Reasonably Achievable (Nuclear Power Reactors)
OTP ASB Guidelines for Fire Protection for 9.5-1 1
Nuclear Power Plants (See Implementation Section, Section D)
BTP MTEB 5-7 4/77 Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping RG 1.141 0
4/78 Containment Isolation Provisions for Fluid Systems ENCLOSURE 2 (CONT'D)
September 15, 1978 CATEGORY 3 MATTERS Document Number Revision Date Titl' RG 1.99 1
4/77 Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials (Paragraphs C.1 and C.2.
RG 1.101 1
3/77 Emergency Planning'for Nuclear Power Plants RG 1.114 1
11/76 Guidance on Being Operator at the Controls of a Nuclear Power Plant RG l.121 0
8/76 Bases for P!agging Degraded PWR Steam Generator Tubes RG 1.127 1
3/76 Inspection of Water-Control Structures Associated with Nuclear Power Plants RSB 5-1 1
1/78 Branch Technical Position: Design Require-ments of the Residual Heat Removal System RSB 5-2 0
3/78 Branch Technical Position: Reactor Coolant System Overpressurization Protection (Draft copy attached)
RG 1.97 1
8/77 Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident (Paragraph C.3 - with additional guidance on paragraph C.3.d to be provided later)
7/78 Initial Startup Test Program to Demonstrate Remote Shutdown Capability for Water-Cooled Nuclear Power Plants RG 1.50 1
7/78 Maintenance of Water Purity in Boiling Water Reactors At*,achment:
BTP RSB 5-2 (Draft)
ENCLOSURE J
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mi A.
3 a C k C Tc ur.d General Design Criterien 15 cf :;encix A.10 CFR 50, recuires :nat *:ne Peac:ce Ccolan: System and associated auxiliary, control, and protec:icn syste s shall te cesignec witn sufficient rargin to assure :na the design con 0itiens of ne reac:Or ccciant Oressure Ocuncary are not exceeced during any conciticn of ncr=al coeration, inciucing antici:ated coeraticnal occurrences.*
-nticicatec : erationa,, cccurrences, as definec in -::ea:1x - c,.
- 3. u,. -,s
- c. 0,
r are '*:ncse ccnci:icns of normal cceration wnicn are ex:ecte: to Occur one or more times curing :ne life of tre nuclear ;c er unit anc inciuce but are not limitec to icss of cwer to all recirculat' n zum:s, tr1; ping of ne tur:ine generatcr set, isciati n of :ne main c:ncenser, and icss of all offsite ;cwer."
Ac;ercix G of '.0 C R SC crovices tne fracture : ugncess re:uire ents for reac:ce cressere vessels uncer all c:rc1:1cns.
To assure ina: :ce Accencix G limits of the reac:Or ccolant cressure toundary are not exceecea curing any Onticipated c:erational occurrences, Tec--icai Scecification pressure-tem;erature limits are rovicec f or c;erating One ;lant.
The primary concern of nis position is :na curing startuc anc snutdc*n conciticns at icw tem erature, es;ecially in a aater-sclid ccnditicn,
- ne reactor ccolant system pressure eign exceec tre reactor vessei pressure-tem:erature limita:icns in the technical 5:ecificaricns estaclisnec for Orc ection against brittle fracture. This inadvertent over:ressurizaticn c ui Ce generated Oy any One cf a variety of mai-any incicer,ts nave occurrec in 3:erating v
fun;;icts cr creratcr errors.
plants as cescribed in Reference 1.
Acc1tional ciscussion en :ne backgr0und of this positicn is contained in Reference 1.
ESCL 3 (COST)
UN!!H '
B.
Branch Position 1.
A system should be designed and installed which will prevent exceeding the applicable Technical Specifications and Appendix G limits for the reactor coolant system wnile operation at low terceratures.
The system should be ca:able of relieving pressure during all anticipated overpressurization events at a rate sufficient to satisfy the Tecnnical Speci.ication limits, particularly while the reactor coolant system is in a water-solid condition, 2.
The system must be able to perform its function assuming any single active ccm;cnent failure. Analyses using appropriate calculational technicues must be provicec which demonstrate that the system will provide the recuired pressure relief capacity assuming tne most limitinc sincie active failure.
The cause for initiation of the event, e.g.,' operator error, ccmponent mal function, will not be consicerec as tne single active failure. The analysis should ass,ume tne most limiting allcuacie cceratinc concitions and systems configuration at tne time of tre costulatec cause of :ne overaressure event. All actential everpressur1:ation events must be considered snan establisninc tne worst case event.
Some events may be 1"evented by protective interlocks or by lockinq out power.
't.Te events should be revised on an individual basis.
If the mterlock/ power lod nut is acceptable, it car. lie excluded from tn...ualyses provided the controls to prevent the event ar e m the plant Technical Specifications.
3.
Tne system must meet the design recuire ents of IEEE 279 (see Inciementation).
Tne system may be manually enabled, hcwever, the electrical instru: entation anc control system must provice alarms to alert tne cperator to:
a.
preterly enable the system at the :orrect plant concition curing ccoldc n, b.
indicate if a pressure transient is occurring.
A.
To assure aceraticnal readiness, the over;ressure protection system must be tested in the fciicwing manner:
a.
A test must ce cerformed to assure acerability of the system electror.ics crice to each shutdcwn.
b.
A test for valve c;;crability must, as a minimum be concucted as specified in th ASME Ccde Secticn XI.
c.
Subsecuent to system, valve, or electronics maintenance, a test on that portion (s) of tre system must be performed prior to declaring the system c;erational, ENCL 3 (CONT)
LJiillr1 5.
The system must meet the recuirements of Pegulatory Goice 1.26,
'Ouality Group Classifications and Standards for uater, Steam,
and Radicactive-Waste-Containing Components of Nuclear Power Plants" and Section 111 of the ASME Code.
6.
The overpressure protection system must be designed to function during an Operating Sasis Earthcuake.
It must not ccmcremise the design criteria of any otner safety-grace system with which it would interface, sucn that tne recuirements of Regulatory Guide 1.29, " Seismic Cesign Classification" are met.
7.
Tne overpressure protection system must not ce enc en tne availability of offsite pc*er to perform its functicn.
S.
Overpressure protecticn systems wnich take crecit for an active cem;cnent(s) to mitigate the consecuences of an cverpressurization event must inciuce accitional analyses c:nsicering inadvertent systen initiation / actuation or provice justificaticn to snow that existing analyses bound such an event.
C.
Inclementation The Brancn Tecnnical Position, as specifiec in Section B-will be usec in tne review of all Preliminary Cesign to;reval (PCA), Final Cesign A:creval (PCA), Manufacturing License (ML), Ocerating License (OL), and Constructicn Permit (CP) applications involving plant cesigns incerocrating pressuri:ed water reactors. All ascects of tne position will be a:plicable to all a;clications, including CP applicat;ons utilizing tne replication cction of the Ccamission's standardization rogram, that are cocketed after Marcn 14, 1973. All aspects of tne positicn, with tne exception of easonaole and justified deviations fecm IEEE 279 recuirements, will ce acolicable to CP, CL, ML, PDA, and CCA applications decketed prior to "arcn 14, 1978 tut for which the licensing action has not been c maletad as of Marcn 14, 1978. Holcers of acorcariate PCA's will be inforced by letter that all aspects of tne position with tne excepticn of IEEE 279 will be a;;licable to their a;;rc.ec standard designs and that sucn cesigns should be modified, as necessary, to conform to the ecsition.
Staff accroval of pro csec mccifications can be a: plied for eitner oy application by tne FDA-nolcer on tne PCA-cccket or oy eacn CP applicant referencing tne stancard design on its docket.
The felicwing guicelines may be used, if necessary, to alleviate impacts en licensing schedules for plants involvec in licensing proceecings nearing c mpletien on Marcn 14, 1975:
ENCL 3 (CONT)
_4_
a 1.
Those applicants issueo an OL during the period between March 14, 1978 and a date 12 months thereaf ter may merely comit to meeting tne position prior to OL issuance but shall, by license ccndition, be required to install all required staff-approved enodificacions prior to plant startup following the first scheduled refueling outage.
2.
Those applicants issued an OL beyond March 14, 1979 shall install all recuired staff-approved modifications prior to initial plant startup.
3.
Those applicants issued a CP, PDA, or ML during the period between March 14, 1978 and a date 6 months thereafter may merely ccmit to meeting the cosition but shall, by license condition, be required to amend the acclication, within 6 months of the date of issuance of the CP, PDA, or ML, to include a description of the pro::osed todifications and the bases for their design, and a recuest for staff approval.
4 Those applicants issued a CP, PDA, or ML after Sectember 14, 1978 shall have staff a:: proval of prcposed modifications prior to issuance of the CP, PDA, or ML.
D.
References 1.
NUREG-0138, Staff Discussicn of Fif teen Tecnnical Issues Listed in Attachr:ent to November 3,1976 Memorandum frcm Director, NRR, to NRR Staff.
ENCL 3 (CONT)
CATEGORY 4 MATTERS A.
Regulatory Guides not categorized Issue Date Number Revision Title 4/74 1.12 1
Instrumentation for Earthquakes 12/75 1.13 1
Spent Fuel Storage Facility Design Basis 8/75 1.14 1
Reactor Coolant Pump Flywheel Integrity 1/75 1.75 l
Physical Independence of Electric Systems 4/74 1.76 0
Design Basis Tornado for Nuclear Power Plants 9/75 1.79 1
Preoperational Testing of Emergency Core Cooling Systems for Pressurized Water Reactors 6/74 1.80 0
Preoperational Testing of Instrument Air Systems 6/74 1.82 0
Sumps for Emergency Core Cooling and Containment Spray Systems 7/75 1.83 1
Inservice Inspection of Pressurized Water Reactor Steam Generator Tubes 11/74 1.89 0
Qualification of Class lE Equipment for Nuclear Power Plants 12/74 1.93 0
Availability of Electric Power Sources 2/76 1.104 0
Overhead Crane Handling Systems for Nuclear Power Plants ENCLOSURE 4 B.
SRP Criteria Impl ementa-Applicable tion Date Branch SRP Section Title 1.
1.1/24/75 MTEB 5.4.2.1 BTP MTEB-5-3,. Monitoring of Secondary Side Water Chemistry in PWR Steam Generators 2.
11/24/75 CSB 6.2.1 BTP CSB-6-1, Minimum 6.2.1A Containment Pressure Model 6.2.lB for PWR ECCS Performance 6.2.1.2 Evaluation 6.2.1.3 6.2.1.4 6.2.1.5 3.
11/24/75 CSB 6.2.5 BTP CSB-6-2, Control of Combustible Gas Concentra-tions in Containment Following a Loss-of-Coolant Accident 4.
11/24/75 CSB 6.2.3 BTP CSB-6-3, Determination of Bypass Leakage Path in Deal Containment Plants 5.
11/24/75 CSB 6.2.4 BTP CSB-6-4, Containment Purging During Normal Plant Operations 6.
11/24/75 ASB 9.1.4 BTP ASB-9.1, Overhead Handling Systems for Nuclear Power Plants 7.
11/24/75 ASB 10.4.9 BTP ASB-10.1, Design Guidelines for Auxiliary Feedwater System Pump Drive and Power Supply Diversity for PWR's 8.
11/24/75 SEB 3.5.3 Procedures for Composite Section Local Damage Prediction (SRP Section 3.5.3, par. II.l.C)
ENCLOSURE 4 (CONT)
Implementa-Applicable tion Date Branch SRP Section Title 9.
11/24/75 SEB 3.7.1 Development of Design Time History for Soil-Structure Interaction Analysis (SRP Section 3.7.1, par. II.2)
- 12. 11/24/75 SEB 3.8.1 Design and Construction of Concrete Containments) SRP Section 3.8.1, par. II)
- 14. 11/24/15 SEB 3.8.3 Structural Design Criteria for Category I Structures Inside Containment (SRP Section 3.8.3, par. II)
- 15. 11/24/75 SEB 3.8.4 Structural Design Criteria for Other Seismic Category I Structures (SRP Section 3.8.4, par. II)
11/24/75 SEB 3.7 Seismic Design Requirements for 11.2 Radwaste Sysems and Their Housing 11.3 Structures (SRP Section 11.2, BTP 11.4 ETSB 11-1, par. B.y)
ENCLOSURE 4 (CONT)
. Implementa-Applicable tion Date Branch SRP Se,ction Title
10/01/75 ASB 10.4.7 Water Hammer for Steam Generators with Preheaters (SRP Section 10.4.7 par. I.2.b) 21.
11/24/75 AB 4.4 Thermal-Hydraulic Stability (SRP Section 4.4, par. II.5)
- 22. 11/24/75 RSB 5.2.5 Intersystem Leakage Detection (SRP Section 5.2.5 par. r!.4) and R.G. 1.45
- 23. 11/24/75 RSB 3.2.2 Main Steam Isolation valve Leakage Control System (SRP Secticn 10.3 par. III.3 and BTP RSB-3.2)
C.
Other Positions Implementa-Applicable tion Date Branch SRP Section Title 1.
12/1/76 SEB 3.5.3 Ductility of Reinforced Concrete and Steel Structural Elements Subjected to Impactive or Impulstve 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 Containment Pool 3.8.2 Dynamics 4
9/01/76 SEB 3.8.4 Air Blast Loads 5.
10/01/76 SEB 3.5.3 Tornado Missile Impact 6.
6/01/77 RSB 6.3 Passive Failures During Long-Term Cooling Following LOCA ENCLOSURE 4 (CONT)
Impl ementa-Applicable tion Date Branch SRP Section Title 7.
9/01/77 RSB 6.3 Control Room Position Indica-tion of Manual (Handwheel) Valves in the ECCS 8.
4/01/77 RSB 15.l.5 Long-Term RecoveFy from Steamline Break: Operator Action to Prevent Overpressurization 9.
12/01/77 RSB 5.4.6 Pump Operability Requirements 5.4.7 6.3
- 10. 3/28/78 RSB 3.5.1 Gravity Missiles, Vessel Seal Ring Missiles Inside Containment
- 11. 1/01/77 AB 4.4 Core Thermal-Hydraulic Analysis
- 12. 1/01 /78 PSB 8.3 Degraded Grid Voltage Conditions
- 13. 6/01/76 CSB 6.2.1.2 Asymmetric Loads on Components Located Within Containment Sub-compartments
- 14. 9/01/77 CSB 6.2.6 Containment Leak Testing Program
- 15. 1/01/77 CSB 6.2.1.4 Containment Response Due to Main Steam Line Break and Failure of MSLIV to Close
- 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 Pumps
- 18. 8/01/76 ASB 10.4.7 Design Guidelines for Water Harmer in Steam Generators with Top Feedring Design (BTP ASB-10.2) 19.
1/01/76 ICSB 3.11 Eraironmental Control Systems -for Safety-Related Equipment ENCLOSURE 4 (CONT)
DESCRIPTION OF POSITIONS IDENTIFIED AS NRR CATEGORY 4 MATTERS IN ENCLOSURE 4, PARAGRAPH C Numbering scheme corresponds to that used in Item C of Enclosure 4.
ENCLOSURE 4 (CONT)
C.1 DUCTILITY OF REINFORCED CONCRETE AND STEEL STRUCTURAL ELEMENTS SUBJECTED TO IMPACTIVE OR IMPULSIVE LOADS INTRODUCTION In the evaluation of overall response of reinforced concrete struttural elements (e.g., missile barriers, columns, slabs, etc.) subjected to icoactive or impulsive loads, such as impacts due to missiles, assumotion of non.-linear response (i.e., ductility ratios greater than unity) of the structural elecents is generally acceptable provided that the safety functions of the structural elements and those of safaty-related systems and components supported or protected by the elements are maintained.
The following sumarizes specific SEB interim positions for review and acceptance of ductility ratios for reinforced concrete and steel structural elements sucjected to impactive and impulsive loads.
SPECIFIC POSITIONS 1.
REINFORCED CONCRETE MEMBERS 1.1 For beams, slabs, and walls where flexure controls design, the pemissible ductility ratio ( u ) under impactive and impulsive loads should be taken as 0.05 for 33
>.005 y
o ~p 10 for p-o'
<.005 u
=
where p and o'are the ratios of tensile and compressive reinforcing as defined in ACI-318-71 Code.
1.2 If use of a ductility ratio greater than 10 (i.e., u> 100) is required to demonstrate design adequacy of structural elecents against impactive or impulsive loads, e.g., missile impact, such a usage should be identified in the plant SAR.
Infomation justifying the use of this relatively high ductility value shall be provided for SEB staff review.
ENCLOSURE 4 (CONT)
- 1.3 For beam-columns, walls, and slabs carrying axial compressian loads and subject to impulsive or impactive loads producing flexure, the permissible ductility ratio in flexure should be as follows:
(a)
When canpression controls the design, as defined by an interaction diagram, the permissible ductility ratio shall be 1.3.
(b)
When 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 Section 1.1.
(c) The permissible 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 structuras elements resisting axial compressive impulsive or impactive loads only, without flexure, the permissible axial ductility ratio shall be 1.3.
1.5 For shear carried by concrote only u = 1.0 For shear carried by concrete and stirrups or bent bars u = 1.3 For shear carried entirely by stirrups u = 3. 0 2.0 STRUCTURAL STEEL MEMBERS 2.1 For flexure compression and shear u
= 10.0 2.2 For columns with slenderness ratio (1/r) equal to or less than 20 u
= 1.3 ENCLOSUIEUE 4 (CONT)
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where 1 = effective length of the member r = the least radius of gyration For columns with slenderness ratio greater than 20 u = 1.0 2.3 For members subjected to tension u =.5 c
where cu= uniform ultimate strain of the material cY = strain at yield of material C.2 RESPONSE SPECTRA IN THE VERTICAL DIRECTION Subsequent to the issuance of Regulatory Guide 1.60, the report
" Statistical Studies of Vertical and Horizontal Earthquake Spectra" was issued in January 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 spectrum for horizontal motion over the entire range of frequencies in the Western United States. According to Regulatory Guide 1.60, the vertical response spectrum is equal to the horizontal response spectrum between 3.5 cps and 33 cps. For the Western United States only, consistent with the latest available data in NUREG-0003, the option of taking the vertical design design response spectrum as 2/3 the horizontal response spectrum over the entire range of frequencies will be accepted.
For other locations, the vertical response spectrum will be the same as that given in Regulatory Guide 1.60.
C.3 BWR MARK III CONTAINMENT POOL DYNAMICS 1.
POOL ShTLL a.
Lubble pressure, bulk swell and froth swell loads, drag pressure and other pool swell loads should be treated as abnormal pressure loads, P. Appropriate load combinations a
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 occurrence.
ENCLOSURE 4 (CONT)
. 2.
SAFETY RELIEF VALVE (SRV) DISCHARGE a.
The SRV loads should be treated as live loads in all load combinations 1.5Pa where a load factor of 1.25 should be applied to the appropriate SRV loads.
b.
A single active failure causing one SRV discharge must be considered in combination with the Design Basis Accident (DBA).
c.
Appropriate multiple SRV discharge should be considered in combination with the Small Break Accident (SBA) and Inter-mediate Break Accident (IBA).
d.
Themal loads due to SRV discharge shculd be treated as T0 for nomal operation and T for accident conditions.
a e.
The suppression pool liner should be designed in accordance with the ASME Boiler and Pressure Yessel Code, Division 1 Subsection NE to resist the SRV negative pressure, considering strength, buckling and low cycle fatigue.
C.4 AIR BLAST LOADS (Pa, Ta, To as defined in ACI 359-740)
The following interim position on air blast loadings on Nuclear Power Plant Structures should be used as guidance in evaluating analyses.
1.
An equivalent static pressure may be used for structural analysis purposes. Ih equivalent static pressure should be obtained from the air blast reflected pressure or the overpressure by multiplying these pressures by a factor of two. Any proposed use of a dynamic load factor less than two should be treated on a case by case basis.
Whether the reflected pressure or the overpressure is to be used for individual structural elements depends on whether an incident blast 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:
U=0+L+B where, U is the strength capacity of a section D is dead load L is live load B is air blast load.
3.
Elastic analysis for air blast is required for concrete structures of new plants. For steel structural elements, and also for rein-forced concrete elements in existing plants, some inelastic response may be permitted with appropriate limits on ductility ratios.
ENCLOSURE 4 (CONT) 4.
Air blast generated ground shock and air blast wind pressure may be ignored. Air blast generated missiles may be important in situations where explosions are postulated to occur in vessels which may fragment.
5.
Overturning and sliding stability should be assessed by multiplying the structure's full projected area by the equivalent static pressure and assuming only the blast side of the structure is loaded. Justification for reducing the average equivalent static pressure on curved surfaces should be considered on a case by case basis.
6.
Internal supporting structures should also be analyzed for the effects of air blast to detemine their ability to carry loads applied directly to exterior panels and slabs. Moreover.in vented structures, interior structures may require analysis even if they do not report exterior structures.
7.
The equivalent static pressure should be considered as potentially acting both inward and outward.
C.5 TORNADO MISSILE PROTECTION As an interim measure,the minimum concrete wall and roof thickness for tornado missile protection will be as follows:
Wall Thickness Roof Thickness Concrete Strength (psi)
(inches)
(inches) 3000 27 24 Region I 4000 24 21 5000 21 18 3000 24 21 Region II 4000 21 18 5000 19 16 3000 21 18 Region III 4000 18 16 5000 16 14 These thicknesses are for protection against local effects only. Designers mu::t establish independently the thickness requirements for overall structura response. Reinforcing steel should satisfy the provisions of Appendix C, ACI 349 (that is,.2* minimum, EWEF). The regions are described in Regulatory Guide 1.76.
ENCLOSURE 4 (CONT)
- C.6 PASSIVE ECCS FAILURES DURING LONG-TERM COOLING FOLLOWING A LOCA Passive failures in the ECCS, having leak rates equal to or less than those from the sudden failure of a pump seal and which may occur during the long-term cooling period following a postulated LOCAd ould be con-h sidered. To mitigate the effects of such leaks, a leak detection system having design features and bases as described below should be included in the plant design.
The leak detection system should include detectors and alams which would alert the operator of passive ECCS leaks in sufficient time so that appro-priate diagnostic and corrective actions may be taken on a timely basis.
The diagnostic and corrective actions would include the identification and isolation of the faulted ECCS line before the performance of more than one subsystem is degraded. The design bases of the leak detection system should include:
(1)
Identification and justification of the maximum leak rate; (2) Maximum allowable time for oparator action and justification therefer; (3) Demostration that the leak detection system is sensitive enough to initiate and alarm on a timely basis, i.e., with sufficient lead time to allow the operator to identify and isolate the faulted line before the leak can create undesireable consequences such as flooding of re-dundant equipment.
The minimum time to be considered is 30 minutes; (4) Demonstration that the leak detection system can identify the faulted ECCS train and that the leak can be isolated; and (5) Alarms that conform with the criteria specified for the contr21 room alams and a leak detection system that confoms with the require-ments of IEEE-279, except that the single failure criterion need not be imposed.
C.7 CONTROL ROOM POSITION INDICATION OF MANUAL (HANDWHEEL) VALVES Regulatory Guide 1.47 specifies, automatic position indication of each bypass or deliberately induced inoperaole condition if the following three conditions are met:
(1) The byoass or inoperable condition affects a system that is designed to perfom an automatic safety function.
ENCLOSURE 4 (CONT)
(2) The bypass or inoperable condition can reasonably be expected to occur more frequently than once per year.
(3) The bypass or inoperable condition is expected to occur when the system is nomally required to operate.
Revision one of the Standard Review Plan in Section 6.3 requires confomance with Regulatory Guide 1.47 with the intent being that any manual (handwheel) valve which could jeopardize the operation of the ECCS, if inadvertently left in the wrong position, must have position indication in the control rocm.
In the PDA extension reviews it is important to confim that standard designs include tais design feature.
Most standard designs do but this matter was probably not specifically addressed in some of the first PDA reviews.
C.8 LCNG-TERM RECOVERY FRCN STEAM LINE BREAK - OPERATOR ACTICN TO PREVENT GVERPRE55uRIZATION (PWR)
A steam line break causes cooldown of the primary system, shrinkage of RCS inventory and depletion of pressurizer fluid. Subsequent to plant trip, ECCS actuation, and main steam system isolation, the RCS inven-tory increases and expands, refilling the pressurizer. Without operator action, replenishment of RCS inventory by the ECCS and expansion at icw temperature could repressurize the reactor to an unacceptacle pressure-temperature regien thereby cc= promising reactor vessel integrity. Anal-yses are required to show that following a main steam line break that (i) no additional fuel failures result from the accident, and (ii) the pressures following the initiation of the break will not compromise the integrity of the reactor ccolant pressure boundary giving due considera-tion to the changes in coolant and material temperatures. The analyses should be based on the assumption that operator action will not be taken until ten minutes after initiation of the ECCS.
C.9 PUMP OPERABILITY REQUIREMENTS In some reviews, the staff has found reasonable doubt that some types of engineered safety feature pumps would continue to perfom their safety function in the long term following an accident.
In such instances there has been followup, including ptzr.p redesign in some cases, to assure that long term perfor.ance could be mt. The following kinds of infor-mtion may be sought on a case-by-case basis where such doubt arises.
a.
Describe the tests perfomed to demonstrate that the pumps are capable of operating for extended periods under post-LOCA conditions, including the effects of debris. Discuss the damage to pump seals caused by debris over an extended period of operation.
ENCLOSURE 4 (CONT)
. b.
Provide detailed diagrams of all wate. cooled seals and compo-nents in the pumps.
Provide a description of the composition of the pump shaft c.
seals and the shafts. Provide an evaluation of loss of shaft seals.
d.
Discuss how debris and post-LOCA environmental conditions were factored into the specifications and design of the pump.
C.10 GRAVITY MISSILES, VESSEL SEAL RING MISSILES INSIDE CONTAINMENT Sa fety related systems should be protected against 13ss of function due to internal missile; from sources such as those associr.ted with pressurized components and rotating equipment. Such sources would include but not be limited to retaining bolts, control rod drive assemblies, the vessel seal ring, valve bonnets, and valve stems. A description of the methods used to afford protection against such potential missiles, including the bases therefor, should be provided (e.g., preferential orientativ; of the poten-tial missile sources, missile barriers, physical separativt,f redundant sa fety systems and components). An analysis of the effects of such poten-tial missiles on safety related systems, including metastably supported equipment which could fall upon impingement, should also be provided.
ENCLOSURE 4 (CONT)
-g-C.11 CCRE THERMAL-HYDRAULIC ANALYSES In evaluating the thermal-hydraulic performance of the reactor core the following additional areas should be addressed:
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.
3.
The effect of fuel rod bewing.
In addition,a commitment to perform tests to verify the transient analysis methods and codes is required.
C.12 DEGRADED GRID VOLTAGE CONDITIONS As a result of the Millstone Unit Number 2 low grid voltage occurrence, the staff has developed additional requirements concerning (a) sustained degraded voltage conditions at the offsite power source, and (b) inter-action of the offsite and onsite emergency pcwer systens. These additional requirements are defined in the follcwing staff position.
1.
We require that a second level of voltage protection for the onsite power system be provided and that this second level of voltage pro-tection satisfy the following requirements:
a) The selection of voltage and time set points shall be deter-nined from an analysis of the voltage requirements of the safety-related loads at all onsite system distribution levels; b) The voltage protection shall include coincidence logic to preclude spurious trips of the offsite pcwer source; ESCLOSURE 4 (CONT)
. c) The time delay selected shall be based on the following conditions:
(i)
The allowable time delay, including margin, shall not exceed the maximum time delay that is assumed in the SAR accident analyses; (ii) The time delay shall minimize the effect of short duration disturbances from reducing the availability of the offsite power source (s); and (iii) The allowable time duration of a degraded voltage condition 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 power 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 Protection Systems for Nuclear Power Generating Stations"; and (vi) The Technical Specifications shall include limiting conditionsfor 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 energency buses once the onsite sources are supplying power to all sequenced loads on the energency buses.
The design shall also include the capability of the load shedding feature to be automatically reinstated if the onsite source supply breakers are tripped. The automatic bypass and reinstatement feature shall be verified during the periodic testing identified in Item 3 of this position.
3.
We require that the Technical Specifications 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 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.
ENCL SURE 4 (CONT)
. 4.
The voltage levels at the safety-related buses should be optimized for the full load and minimum load conditions that are expected throughout the anticipated range of voltage variations of the offsite power source by appropriate adjust-ment of the voltage tap settings of the intervening transfomers.
We require that the adequacy of the design in this regard be verified by actual measurement, and by correlation-of measured values with analysis results.
C.13 ASYMMETRIC LOADS ON COMPONENTS LOCATED WITHIN CONTAINMENT SUBCOMPARTMENTS In the unlikely event of a pipe rupture inside a major component sub-compartment, the initial blowdown transient would lead to pressure loadings on both the structure and the enclosed component (s The staff's generic Category A Task Action Plan A-2 is designed).to develop generic resolutions for this matter. Our present schedule calls for compiding A-2 for PWR's during the first quarter,1979. Pending complGion of A-2, the staff is implementing the following program:
- 1. For PWRs at the CP/PDA stage of review, the staff requires appli-cants to comit to address the safety issue as part of their appli-cation for an operating ifcense.
2.
For PWRs at the OL/FDA stage of review, the staff requires case-by-case 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 safety significance, the asymetric loading conditions will be evaluated on a case-specific basis prior to the issuance of an operating license.
For those cases which analyses are required, we request the perfomance 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 generator compartments. Provide similar analyses for the pressurizer surge and spray lines, and other high energy lines located in containment compartments that may be subject to pressurization. Show how the results of these analyses are used in the design of structures and component supports.
ENCLOSURE 4 (CONT)
C.14 CONTAINMENT LEAK TESTING PROGRAM To avoid difficulties experienced in this area in recent OL reviews, the staff has increased its scope of inquiry at the CP/PDA stage of review. For this purpose, the following infomation with regard to the containment leak testing program should be supplied.
a.
Those systems that will remain fluid filled for the Type A test should be identified and justification given.
b.
Show the design provisions that will pemit the personnel air-lock door seals and the entire air lock to be tested.
c.
For each penetration,i.e., fluid system piping, instrument, electrical, and equipment and personnel access penerations, identify the Type B and/or Type C local leak testing that will be done.
d.
Verify that containment penetrations fitted with expansion bellows will be tested at Pa.
Identify any penetration fitted with expansion bellows that does not have the design capability for Type B testing and provide justification.
C.15 CONTAINMENT RESPONSE DUE TO MAIN STEAM LINE BREAK AND MSLIV FAILURE In recent CP and OL application reviews, the results of analyses for a postulated main steam line break accident (MSLB) for designs utilizing pressurized water reactors with conventional containments show that the peak calculated containment temperature can exceed for a short time period the environmental qualification 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 signifiance of the matter is that it could result in a requirement for requalifying safety-related equipment to higner time-temperature envelopes.
The staff's generic Category A Task Action Plans A-21 and A-24 are designed to develop generic resolutions for these matters. The presentl Portion)y scheduled completion dates for A-21 and A-24 (Short Term are first quarter,1979 and fcurth quarter,1978, respectively.
Pending completion of A-21 and A-24, some interim guidance will be used as detailed below.
We have developed and are implementing a plan in which all applicants for construction pemits and operating licenses and those already issued con-struction pemits must provide information to establish a conservative temperature-time envelope.
ENCLOSURE 4 (CONT)
Therefore, describe and justify the analytical model used to conservatively determine the maximum containment temperature and pressure for a spectrum of postulated main steam line breaks for various reactor power levels.
Include the following in the discussion.
(1) Provide single active failure analyses which specifically identify those safety grade systems and components relied upon to limit the mass and energy release and containment pressure /
temperature response. The single failure analyses should include, but not necessarily be limited to: main steam and connected systems isolation; feedwatcr auxiliary feedwater, and connected systems isolation; feedwater, condensate, and auxiliary feedwater pump trip, and auxiliary feedwater run-out control system; the loss of cr availability of offsite power; diesel failure when loss of offsite power is evaluated; and partial loss of centainment cooling systems.
(2) Discuss and justify the assumptions made regarding the time at which active containment heat removal systems become effective.
(3) Discuss and justify the heat transfer correlation (s) (e.g., Tagami, Uchida) used to calculate the heat transfer from the containment atmosphere to the passive heat sinks, and provide a plot of the heat transfer coefficient versus time for the most severe steam line break accident analyzed.
(4)
Specify and justify the temperature used in the calculation of condensing heat transfer to the passive heat sinks; i.e.,
specify whether the saturation temperature corresponding to the partial pressure of vapor, or the atmosphere temperature (which may be superheated)was used.
(5) Discuss and justify the analytical model including the thermodynamic 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 break 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 components needed to mitigate the consequences of the assumed main steam line break and effect safe reactor shutcown; ENCLOSURE 4 (CONT)
, (8) For the case which results in maximum containment atmosphere pressure, graphically show the containment pressure as a function of time; and (9) For the case which results in the maximum containment atmosphere pressure and temperature, provide the mass and energy release data in tabular fom.
In order to demonstrate that safety-related equipment has been adequately qualified as described above, provide the following infomation regard-ing its environmental qualification.
(1) Provide a comprehensive list of equipment required to be operational in the event of a main steamline break (MSLB) accident. The list should include, but not necessarily be limited to, the following safety related equipment:
(a) Electrical containment penetrations; (b) Pressure transmitters; (c) Containment isolatica valves; (d) Electrical power cables; (e) Electrical instrumentation cable; and (f) Level transmitters.
Describe the qualification testing that was, or will be, done on this equipment.
Include a discussion of the test environment, namely, the temperature, pressure, moisture content, and chemical spray, as a function of time.
(2)
It is our position that the themal analysis of safety related equipment which may be exposed to the containment atmosphere following a main steam line break accident should be based on the following:
(a) A condensing heat transfer coefficient based on the recomendations in Branch Technical Position CSB 6-1,
" Minimum Containment Pressure Model for PWR ECCS Perfomance Evaluation,"should be used.
(b) A cowective heat transfer coefficient should be used when the condensing heat flux is calculated to be less than the convective heat flux. During the blowdown period it is appropriate to use a conservatively evaluated forced convection heat transfer correlation. For example, ENCLOSURE 4 (CONT)
, Nu = C(Re)
Where Nu = Nusselt No.
Re = Reynolds No.
C
= empiric 31 constants dependent on geometry and Reynolds No.
Since the Reynolds number is dependent on velocity, it is necessary to evaluate the forced flow currents whi3 will be generated by the steam generaar blowdown. The CVTR experiments provide limited data in this regard. Convective currents of from 10 ft/see to 30 ft/sec were measured locally. We recomend 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 blowdown has ceased or been reduced to a negligibly low value, a natural convection heat transfer correlation is acceptable.
(3) For each component where themal 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 themal response of the component with the accident themal analysis of the component. Provide the basis by which the component themal response was developed from the environmental qualification test program. For instance, graphically show the themoccuple data and discuss the themocouple locations, method of attachment, and perfomance characteristics, or provide a detailed discussion of the analytical model used to evaluate the component themal response during the test. This evaluation should be performed 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 elevated temperatures, or other themal effects could result in the failure of the component mechanically or electrically.
If the component themal 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 intended function during a MSLB accident, or provide protection for the component whch would appropriately limit the themal effects.
ENCLOSURE 4 (CONT)
^
. C.16 ENVTRONMENTAL EFFECT OF PIPE FAILURES Identify the " break exclusion" regions of the main steam and feedwater lines. Compartments that contain break exclusion regions of main steam and feedwater lines and any safety related equipment in these compartments should be designed to with-stand the environmental effects (pressure, temperature, humidity and flooding) of a crack with a break area equal to the cross sectional area of the't tak exclude # pipe.
C.17 DESIGN REQUIREMENTS FOR COOLING WATER TO REACTOR COOLANT PUMPS Demonstrate that the reactor coolant system (RCS) pump seal injection flow will be automatically maintained for all transients and accidents or that enough time and information are availahla ta n..
4+
" "'~
corrective action by an operator.
We have establishea the following criteria for that portion of the component cooling water (CCW) system which interfaces with the reactor coolant pumps to supply cooling water to pump seals and bearings during normal operation, anticipated transients, and accidents.
- 1. A single active failure in the component cooling water system shall not result in fuel damage or a breach of the reactor coolant pressure boundary (RCPB) caused by an extended loss of cooling to one or more pumps. Single active failures 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 determined in ivordance 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:
ENCLOSURE 4 (CONT)
. 1.
That portion of the component cooling water (CCW) system which supplies cooling water to the reactor coolant pumps and motors may be designed to non-seismic Category I requirements and Quality Group D if it can be demonstrated that the reactor coolant pumps will operate without component cooling water for at least 30 mint,tes without loss of function or the need for operator pro-tective actio1.
In addition, safety grade instrumentation including alarms shou 1J be provided to detect the loss.of component cooling water to the reactor coolant pumps and motors, and to notify the operator in the control room. The entire instrumentation system, including audible and visual alams, should meet the requirements of IEEE Std 279-1971.
If it is not demonstrated that the reactor coolant pumps and motors will operate at least 30 minutes without loss of function or operator protective action, ther. the design of the CCW sys tem must meet the following requirements:
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 Category I require-ments and Quality Group 0; or 2.
The component cooling water supply to the pumps and motors shall be capable of withstanding a single active failure or a moderate energy line crack as defined in our Branch Technical Position APCSB 3-1 and be designed to seismic Cateoory I, Quality Group D and ASME Section III, Class 3 requirements.
The reactoIcoolant (RC) pumps and motors are within the NSSS scope of design. Therefore, in order to demonstrate that an RC pump design can operate with loss of competent cooling water for at least 30 minutes without loss of functie or the need for operator action, the following must be provion.
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 from this event.
Include a discussion of the effect that the loss of cooling water to the seal coolers has on the RC pump seals. Show that the loss of cooling water does not result in a LOCA due to seal failure.
/
ENCLOSURE 4 (CONT) 2.
A detailed analysis to show that loss of cooling water to the RC pumps and motors will not cause a loss of the flow coastdown characteristics or cause seizure of the pumps, assuming no administrative action is taken. The response should include a detailed descriptf)n of the calculation procedure including:
a.
The equations used.
b.
The parameters used in the equations, such as the design parameters for the motor bearings, motor, pump and any other equipment entering into the calculations, and 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 appropriate figures) associated with the RC pump and motor and their design cr:teria and standards.
e.
Infomation to verify the applicability of the equations and material properties chosen for the analysis (i.e.,
references should be listed, and if empirical relations are used, 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 component cooling water to the RC pumps and motor assembly is acceptable, we will require certain modifications to the plant Technical Specifications and an RC pump test conducted under operating condtions and with component cooling water teminated for a specified period of time to verify the analysis.
C.18 WATER HAMMER IN STEAM GENERATORS WITH T0p FEEDRING DESIGN 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 feedwater inlet nozzles. Subsequent events may in turn lead to the generation of a pressure wave that is propagated through the pipes and could result in unacceptable damage.
ENCLOSURE 4 (CONT)
'. For CP/PDA and OL/FDA applications, provide the following for steam generators utilizing top feed:
1.
Rrevent or delay water draining from the feedring following a drop in steam generator water level by means such as,,J-Tubes; 2.
Minimize the volume of feedwater piping external to the steam generator whch could pocket steam using the shortest possible (less than seven feet) horizontal run of inlet piping to the steam generator feedring; and 3.
Perform tests acceptable to the staff to verify that unacceptable feed-water hamer will not occur using the plant operating procedures for nomal and emergency restoration of steam generator water level following loss of nomal feedwater and possible draining of the feedring. Provide the procedures for these tests for staff approval before conducting the tests.
Furthemore, we request that the following be provided:
a.
Describe nomal operating occurrences of transients that could cause the water level in the steam generator to drop below the sparger or nozzles to cause uncovering and allow steam to enter the sparger and feedwater piping.
b.
Describe your criteria or show by isometric diagrams, the routing of the feedwater piping from the steam generators outwards to beyond the containment structure up to the outer isolation valve and restraint.
c.
Describe any analysis on the piping system including any forcing functions that will be perfomed or the results of test programs to verify that,ef ther uncovering of feedwater lines could not occur or that, if it dia occur, unacceptable damage such as the experience at the Indian Point Unit No. 2 facility would not result with your design.
ENCLO URE 4 (CONT)
S
C.19 ENVIRONMENTAL CONTROL SYSTEMS FOR SAFETY RELATED E0ilIPMENT Most plant areas that contain safety related equipment depend on the continuous operation of environmental control systems to maintain the environment in those areas within the range of environmental qualification of the,e'aty related equipment installed in those areas. It appears that there are no requiremehts for maintaining these environmental control systems in operation while the plant is shutdown or in hot standby 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 these environmental control systems should operate continuously to maintain the environmental 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 energized from continuous power sourcess and (4) provide a continuous record of the environmental parameters during the time the envircnmental conditions exceed the nomal limits.
ENCLOSURE 4 (CONT)