ML20027A452
| ML20027A452 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek, Callaway, Sterling, 05000484  |
| Issue date: | 11/21/1978 |
| From: | Boyd R Office of Nuclear Reactor Regulation |
| To: | Arthur J, Bryan J, Dienhart A, Koester G KANSAS GAS & ELECTRIC CO., NORTHERN STATES POWER CO., ROCHESTER GAS & ELECTRIC CORP., UNION ELECTRIC CO. |
| References | |
| NUDOCS 7812050104 | |
| Download: ML20027A452 (41) | |
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UNITED STATES y*
-' t NUCLEAR REGULATORY COMMISSION j.
WASHINGTON D.C.20555
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NOV 211978 Dock *dt*Ms. STN 50-482, STN 50-483, STN 50-484 STN 50-485 and STN 50-486 Mr. J. K. Bryan Mr. A. B. Dienhart Engineering & Construction Vice President - Engineering Union Electric Company Northern States Power Company P. O. Box 149 414 Nicollet Mall St. Louis Missouri 63166 Minneapolis, Minnesota 55401 Mr. John Arthur Mr. Glenn L. Koester Chief Engineer Vice President - Operations Rochester Gas & Electric Corporation Kansas Gas & Electric Company 89 East Avenue 201 North Market Street Rochester, New York 14649 Wichita, Kansas 67201 Gentlemen:
SUBJECT:
IMPLEMENTATION OF STAFF REVIEW REQUIREMENTS - SNUPPS PLANTS (CALLAWA PLANT, UNITS 1 AND 2, STERLING POWER PROJECT NUCLEAR UNIT N0.1, TYRONE ENERGY PARK, UNIT 1, WOLF CREEK GENERATING STATION, UNIT N0. 1)
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 Comittee (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 appruval 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 l
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 sumary of RRRC Meeting No. 31 concerning this categoriza-tion is attached as Enclosure 1.) Category 1 matters are those to be applied to applications in accordance with the implementation section I
of the published guide. We have enclosed lists of actions which are i
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 your design conforms, or you should describe an acceptable alternate, or you should demonstrate why confor-mance is not necessary.
Category 3: Conformance or an acceptable alternative is required.
If you do not conform, or do not have an acceptable alternate, l
7812050I68/w/ en staff-approved design revisions will be required.
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. NOV 211978 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 matter 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 encl sures may not be applicable to your application. 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.
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- rely, S
s er S.
oyd, Dire tor Division of Project Manag mt Office of Nuclear Reactor Regulation
Enclosures:
As stated cc: See next page i
2 Mr. J. K. B ryan Mr. A. B. Dienhart Engineering & Construction Vice President - Engineering Union Electric Company Northern States Power Company P. O. Box 149 414 Nicollet Mall St. Louis, Missouri 63166 Minneapolis, Minnesota 55401 Mr. John Arthur Mr. Glenn L. Koester Chief Engineer Vice President - Operations Rochester Gas & Electric Corporation Kansas Gas & Electric Company 89 East Avenue 201 North Market Street Rochester, New York 14649 Wichita, Kansas 67201 cc:
Orange & Rockland Utilities, Inc.
Niagara Mohawk Power Corporation ATTN: Mr. D. H. Barnes, Jr.
ATTN: Mr. G. Rhode, Vice Senior Vice President President - Engineering 75 West Route 59 300 Erie Boulevard, West Spring Valley, New York 10977 Syracuse, New York 13202
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Central Hudson Gas & Electric Kansas City Power & Light Company Corpor5 tion ATTN: Mr. D. T. McPhee -
ATTN: Mr. Charles A. Bolz Vice President Vice President 1330 Baltimore Avenue Engineering & General Svc.
Kansas City, Missouri 66141 284 South Avenue Poughkeepsie, New York 12602 Mr. Nicholac A D;+ rick E AeLuti ve Director, SNUPPS Arvin E. Upton, Esq.
5 Choke Cherry Road LeBoeuf, Lamb, Leiby & MacRae Rockville, Maryland 20850 1757 N Street, N.W.
Washington, D. C.
20036 Mr. James T. Wiglesworth, Esq.
9800 Metcalf Gerald Charnoff, Esq.
Suite 400 Shaw, Pittman, Potts, General Square Center Trowbridge & Madden Overland Park, Kansas 66212 1800 M Street, N.W.
Washington, D. C.
20036 Lex K. Larson, Esq.
Leboeuf, Lamb, Leiby & MacRae Mr. William H. Griffin 1757 N Street, N.W.
Assistant Attorney General Washington, D.C.
20036 State of Kansas State House Ms. Sharon Morey Topeka, Kansas 66612 California Road, RD #3 Oswego, New York 13126
cc:
Mr. J. E. Birk Assistant to the General Counsel Union Electric Company St. Louis, Missouri 63166 Michael K. McCabe, Esq.
First Assistant Conaission Counsel Peter Peshek, Esq.
Missouri Public Service Commission Public Intervenor P. O. Box 360 Departnent of ;ustice Jef ferson City, Missouri 61510 123 West Washington Avenue Madison, Wisconsin 53702 Professor Dennis J. Tuchler St. l.ouis University Law School Joseph P. Schaeve, Esq.
3700 Lindell Boulevard Bureau of Legal Services St. Louis, Missouri 63101 Departgent of Natural Resources State of Wisconsin 53707 Dr. Vern Starks Route 1, Box 863 Sandra S. Gardebring, Esq.
Ketchikan, Alaska 99901 Minnesota Pollution Control Agency 1935 W. County Road B-2 Williard E. Fantle, III, President
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Roseville, Minensota 55113 Northern Thunder 22-1/2 S. Barstow Barbara J. Willard, Esq.
Eau Claire, Wisconsin 54701 Public Service Commission of, Wisconsin Hill Farms State Office Building Mr. Harold C. Bauer 4802 Sheboygan Avenue Route 1, Box 191 3.43 H ie.s LMkitei n 53707 c:.th 741);, Wii ensin 54764 Mr. Stanley Cider T. K. DeBoer, Director-c/o Durand Postmaster
~ New York State Energy Officew Tyrone, Wisconsin 54736 Empire State Plaza Albany, New York 12223 David J. Newburger, Esq.
Utility Censumers Council of Mo.
John Halzer, Esq.
Suite 503 - 7710 Carondelet Law Offices of Ralph Foster l
Clayton, Missouri 63105 Kansas Gas & Electric Building P. O. Box 208 Dr. William E. Seymour Wichita, Kansas 67201 N. Y. State Atomic Energy Council N. Y. State Dept. of Commerce William H. Ward, Esq.
i 99 Washington Street MACEA Albany, New York 12210 5130 Mission Road Shawnee Mission, Kansas 66205 Ms. Helen M. Kees j
l Rcute 3 Edward Luton, Esq., Chairman Durand, Wisconsin 54736 Atomic Safety and Licensing Board U. S. Nuclear Regulatory Commission Washington, D. C.
20555 Mr. Lester Kornblith, Jr.
Atomic Safety and Licensing Board.
U. S. Nuclear Regulatory Commission Washington, D. C.
20555
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Dr. George C. Anderson Department of Oceanography
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Alan S. Rocenthal, Esq., Chairman s
,3 tomic Safety and Licensing Appeal Coard A
<1). S. !Wclears Regulatory Cormiissica s
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'9dshington,!D. C.
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Dr. John H. Buck 5
1 la Atomic Safety and Licensing Appeal Board s,'
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Washington, D. C.~ 20555
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'N ' ( 'h Michael C. Farrar, Esq.
Atomic Safety and Cicensing Appeal Board i
- 0. S. Nuclear Regulatory Comnission e
a Washington, D. C.
20555
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Ivan W. Smith,.Esq., Chairman s
Atcalc Safety and Licensing Board s
U, S. Nuclear Regulatory Connission Was,hington, D. C.
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3 NUCLEAR REGULATORY, COMMisse.
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i Lee V. Gossick
.Executise Director for Operations REGULATORY REQUIREl1ENTS REVIEW COP 51ITTEE MEETING NO. 31, JULY ll,1975 s
1.
The Committee discussed issues related to the implementation of i
Regulatory Guides or,' existing plants and the concerns expressed
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in the June 24, 1974 memorandum, A. Giambusso to E. G. Case, subject: REGULATORY GUIDE IMPLEMENTATION, and made the feitowing recomendations and observations:
Approval of new Regulatory Gu' ides and approval of revisions a.
of existing guides should move forward expeditiously in order
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that the provisions of these regulatory guides be available
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for use as soon as possible in on-going or future staff reviews of license applications. The Comittee noted that over the recent past, the approval of proposed regula:ory guides whose content is acceptable for these purposes has ' experienced significant delays in RRRC review pending the determination
- l of the applicability of the guide to existing plants, often requiring significant staff effort. To avoid these delays, the Comittee concluded that, henceforth, approval of proposed f
regulatory guides should be uncoupled from the consideration of their backfit applicability.
b.
TIM implementation section of new regulatory' guides should addrecs, in general, only the applicability of the guide to y
- 4 applications in the licensing review process using, in so far as possible, a standard approach of applying the guide to 3
tho:e applications docketed 8 months after the issuance date of the guide for comment.
Exceptions to this general approach s~
will be handled on a case-by-case basic, The regulatory position of each approved proposed guide (or c.
proposed guide revision) will be characterized by the Comittee as to its backfitting pot'ential, bycplacing it in one of three categories:
Category 1 - Clearly forward fit only. No further staff consideration of possible backfitting is required.
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Category 2 - Further staff consideration of the need for back '
fitting appears to be required for certair, identified items of the regulatory position--these individual issues are such that existing plants need to be evaluated to determine their status with regard to these safety issues in order to determine the need for backfitting.
Category 3 - Clearly backfit.
Existing plants should be evaluated to determine whether identified items 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 Comittee categorization appear in the guide itself. The purpose of the categorization is to indicate those ite'.s of the regulatory 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 determine backfit considerations (Category 2).
e.
The Comittee recomends 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 requirements for existing plants be detemined within a reasonable period of time after publication of the guide.
f.
The Comittee observed that rmre atten-ion 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 Comittee reviewed tne proposed Regulatory Guide 1.XX: THERMAL OVERLOAD PROTECTION FOR MOTORS OM l10 TOR-OPERATED VALVES and recommended 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 themal overload protection provisions be audited.
ENCLOSURE 1 (CONT'D) i
l Lee V. Gossick
- 3.
The Comittee reviewed the proposed Regulatory Guide 1.XX:
INSTRU 4ENT SPANS AND SETPOINTS and recomended approval subject to the following coment:
Paragraph 5 of Section C (page 4 of the proposed Guide) should be reworded in light of Comittee coments, to the satisfaction of the Director, Office of Standards Development.
This guide was characterized by the Comittee as Category 1 - no backfit.
4.
The Comittee reviewed Proposed Regulatory Guide 1.97:
INSTRU:'ENTATIO!! FOR LIGHT llATER COOLED NUCLEAP, POWER PLAf!TS TO ASSESS PLAT;T C0ilDITIOT15 DUP.ING AND FOLLOUIt!G Al1 ACCIDENT and deferred further consideration to a later meeting in order to pemit incorporation of recent conments by the Division of Technical Review.
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Edson G.
ase, Chaiman Regulatory Require.ments Review Committee l
l ENCLOSURE 1 (CONT'D) i
I September 15, 1978 CATEGORY 2 MATTERS Document Number Revision Date Title RG 1.27 2
1/76 Ultimate Heat Sink for Nuclear i
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
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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 Naclear 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)
ENCLOSUFG: 2 1
l CATEGORY 2 MATTERS (CONT'D)
Continued Document Munber
. Revision Date Title RG 1.137 0
1/78 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)
.BTP ASB Guidelines for Fire Protection for
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9.S-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 l
i.
ENCLOSURE 2 (CONT'D)
n.
September 15, 1978 CATEGORY 3 MATTERS Document Number Revision Date Title RG 1.99 1
4/77 Effects of Residual Elements on Predicted Radiation Damage to
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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 1.121 0
8/76 Bases for Plugging Degraded PWR
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Steam Generator Tubes RG 1.127 1
3/78 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 t.
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 1.56 1
7/78 Maintenance of Water Purity in Boiling Water Reactors f
Attachment:
BTP RSB 5-2 (Draft)
ENCLOSURE 3
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BRAftCH TECHNICAL 9051T1011 R58 5-2 OVERPRESSURIZATION PROTECTION OF PRESSURIZED WATER REACTORS WHILE OPERATING AT LOW TEMPERATURES A.
Backaround General Design Criterion 15 of Appendix A,10 CFR 50, requires that "the Reactor Coolant System and associated auxiliary, control, and protection systems shall be designed with sufficient margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during any condition of normal operation, including anticipated operational occurrences."
Anticipated operational occurrences, as defined in Appendix A of 10 CFR 50, are "those conditions of normal operation which are expected to occur one or more times during the life of the nuclear pcwer unit and include but
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are not limited to loss of power to all recirculation pumps, tripping of the turbine generator set, isolation of the main condenser, and loss of all offsite power."
Appendix G of 10 CFR 50 provides the fracture tougnness requirements for reactor pressure vessels under all conditions. To assure that the Accendix G limits of the reactor coolant oressure boundary are not exceeded during any anticipated operational occurrences, Technical Specification pressure-temperature limits are provided for operating the plant.
The primary concern of this position is that during startup and shutdown conditions at low temperature, especially in a water-solid condition, the reactor coolant system pressure might exceed the reactor vessel pressure-temperature limitations in the Technical Specifications established for protection against brittle fracture. This inadvertent overpressurization could be generated by any one of a variety of mal-functions or coerator errors. Many incider.ts have occurred in operating plants as described in Reference 1.
Adoitional discussion on the background of this position is contained in Reference 1.
ENCL 3 (CONT)
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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 while operation at low temperatures. The system should be capable of relieving-pressure during all anticipated overpressurization events at a rate sufficient to satisfy the Technical Specification 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 component failure. Analyses using appropriate calculational techniques must be provided which demonstrate that the system will provide the required pressure relief capacity assuming the most limiting single active failure. The cause for initiation of the event, e.g., operator error, component malfunction, will not be considered as the single active failure. The analysis should ass,ume the most limiting allowable coerating conditions and systems configuration at the time of the costulateo cause of the overoressure event. All cotential overpressur12ation events must be considered I
when establishino the worst case event.
Some events may be prevented by protective interlocks or by locking out power.
ihese events should be reviewed on an individual basis.
If the interlock / power lockout is acceptable, it car. De exclurferi frnm the analyses provided the controls to prevent the' event are in the plant Technical Specifications.
3.
The system must meet the design requirements of IEEE 279 (see implementa tion). The system may be manually enabled, however, the electrical instrumentation and control system must provide alarms to alert the operator to:
a.
properly enable the system at the correct plant condition during cooldown, b.
indicate if a pressure transient is occurring.
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4.
To assure operational readiness, the overpressure protection system must be tested in the following manner:
a.
A test must be performed to assure operability of the system electronics prior to each shutdown.
b.
A test for valve operability must, as a minimum be conducted as specified in the A5ME Code Section XI.
c.
Subsequent to system, valve, or electronics maintenance, a test on that portion (s) of the system must be performed prior to declaring the system operational.
F.NCT.
'4 (CONT)
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5.
The system must meet the requirements of Regu;atory Guide 1.26',
" Quality Group Classifications and Standards for Water, Steam,
and Radioactive-Waste-Containing Components of Nuclear Power Plants" and Section III of the ASME. Code.
6.
The overpressure protection system must be designed to function during an Operating Basis Earthquake.
It must not compromise the design criteria of any other safety-grade system with which it would interface, such that the requirements of Regulatory Guide 1.29, " Seismic Design Classification" are met.
7.
The overpressure protection system must not depend on the availability of offsite power to perform its function.
8.
Overpressure protection systems which take credit for an active component (s) to mitigate the consequences of an overpressurization event must include additional analyses considering inadvertent system initiation / actuation or provide justification to show that existing analyses bound such an event.
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C.
Imolementation The Branch Technical Position, as specified in Section 8, will be used in the review of all Preliminary Design Approval (PDA), Final Desi Approval (FDA), Manufacturing License (ML), Operating License (0L)gn
, and Construction Permit (CP) applications involving plant designs incorporating pressurized water reactors. All aspects of the position will be applicable to all applications, including CP applications utilizing the replication option of the Conrnission's standardization program, that are docketed after March 14, 1978. All aspects of the position, with the exception of reasonable and justified deviations from IEEE 279 requirements, will be applicable to CP, OL, ML, PDA, and FDA applications docketed prior to March 14, 1978 but for which the licensing action has not been l
completed as of March 14, 1978. Holders of appropriate PDA's will be informed by letter that all aspects of the position with the exception i
of IEEE 279 will be apolicable to their approved standard designs and i
that sucn designs should be modified, as necessary, to conform to the position. Staff approval of proposed modifications can be applied for either by application by the PDA-holder on the PDA-docket or by each CP applicant referencing the standard design on its docket.
The following guidelines may be used, if necessary, to alleviate impacts on licensing schedules for plants involved in licensing proceedings nearing completion on March 14, 1978:
ENCL 3 (CONT)
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1.
Those applicants issued an OL during the period between March 14 1978 and a date 12 months thereafter may merely commit to meeting s
the position' prior to OL issuance but shall, by license condition, be required to install all required staff-approved modificacions prior to plant startup following the first scheduled refueling outage.
2.
Those applicants issued an OL beyond March 14, 1979 shall install all required staff-approved modifications prior to initial plant startup.
3.
Those applicants issued a CP, POA, or ML during the period between March 14, 1978 and a date 6 months thereafter may merely commit to meeting the position but shall, by license condition, be required to amend the application, within 6 months of the date of issuance of the CP, PDA, or ML, to include a description of the proposed modifications and the bases for their design, and a request for staff approval.
4 Those applicants issued a CP, POA, or ML after Saotember 14, 1978
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shall have staff approval of proposed modificat'ons prior to issuance of the CP, PDA, or ML.
s Q. References 1.
NUREG-0138, Staff Discussion of Fifteen Technical Issues Listed in Attachment to November 3,1976 Memorandum from Director, NRR, to NRR Staff.
l E.
i ENCL 3 (CONT) l
CATEGORY 4 MATTERS A.
Regulate:7 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 I
1/75 1.75 1
Physical Independence of Electric Systems 4/74 1.76 0
Design Basis Tornado for Nuclear Power Plants 9/75 1.79 1
Precperational 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 1E Equipment for Nuclear Power Plants 12/74 1.93 0
Availability of Electric Power Sources 2/76 1.104 0
Overhead Crane Handling fystems for Nuclear Power Plants ENCLOSURE 4
B.
SRP Criteria Implementa-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 Piet Steam Generators 2.
11/24/75 CSB 6.2.1 BTP CSB-6-1, Minimun 6.2.lA Contafrunent Pressure Model 6.2.1B for PWR ECCS Performance 6.2.1.2 Evaluation 6.2.1.3 6.2.1.4 6.2.1.5
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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, Detennination of Bypass Leakage Path in Qual Containment Plants 5.
11/24/75 CSB 6.2.4 BTP CSB-6-4, Containment Purging During Normal Plant Operations 6.
11/24/75 AS ?
9.1.4 BTP ASB-9.1, Overhead Handling ~
Systems for Nuclear Power Plants 7.
11/24/75 ASB 10.4.9 STP 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.1.C)
ENCLOSURE 4 (CONT)
- e e,
Implementa-Appitcable 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 Section3.7.1, 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. 11/24/75 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)
- 16. 11/24/75 SEB 3.8.5 Structural Design Criteria for i
Foundations (SRP Section 3.8.5, par. II)
- 17. 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.v) l ENCLOSURE 4 (CONT) l
4 Implementa-Applicable tion Date Branch SRP Section Title
I
- 20. 10/01/75 ASB 10.4.7 Water Hanner for Steam Generators with Preheaters (SRP Section 10.4.7 par. I.2.b) 21.
11/24/75 AB 4.4 Themal-Hydraulic Stability (SRP Section4.4, par.II.5)
'22.
11/24/75 RSS 5.2.5 Intersystem Leakage Detection (SRP
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Section 5.2.5 par. II.4) and R.G.1.45
- 23. 11/24/75 RSB 3.2.2 Main Steam Isolation Yalve Leakage Control System (SRP Section 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 Elsnents Subjected to Impactive or Impulsive 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-Tem Cooling Following LOCA ENCLOSURE 4 (CONT)
)
Implementa-Applicable tion Date Branch SRP Section Title 7.
9/01/77 R$8 6.3 Control Room Position Indica-tion of Manual (Handwheel) Valves in the ECCS 8.
4/01/77 RSB 15.1.5 Long-Term RecoveFy from Steamline Break: Operator Action to Prevent Overpressurization 9.
12/01/77 R$8 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 Thennal-Hydraulic Analysis
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- 12. 1/01 /78 PSB 3.3 Degraded Grid Voltage Conditions
- 13. 6/01/76 CSB 6.2.1.2 Asymetric Loads on Components Located Within Contairvnent 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 Namer in Steam Generators with Top 1
l.
Feedring Design (BTP ASB-10.2) 19'. 1/01/76 fCSB 3.11-Environmenta' l Control Systems -for Safety-Related Equipment ENCLOSURE 4 (CONT)
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DESCRIPTION OF POSITIONS IDENTIFIED AS NRR CATET,dY 4 MATTERS IN ENCLOSURE 4, PARAGRAPH C Numbering scheme corresponds to that used in Item C of Enclosure 4 1
i i
i ENCLOSURE 4 (CONT) l
C.1 UUCTILITY OF REINFORCED CONCRETE AND STEEL STRUCTURAL ELEMENTS SUBJECTED TO IMPACTIVE OR IMPULSIVE LOADS INTRODUCTION In the evaluation of overs'l response of reinforced concrete structural elemwnts (e.g., missile barriers, columns, slabs, etc.) subjected to impactive or impulsive loads, such as impacts due to missiles, assumption of non-linear response (i.e., ductility ratios greater than unity) of the structural elements is generally acceptable provided that the safety functions of the structural elements and those of safety-related systems and components supported or protected by the elements are maintained.
The following sumcrizes specific SEB interim positions for review and acceptance of ductility ratios for reinforced concrete and steel structural elements subjected to impactive and impulsive loads.
SPECIFIC POSITIONS I
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 o-o'
> t05
=
u l
0o 1
005 10 for o-o'
=
u where p and o'are the ratios of tensile and compressive reinforcing as defined in ACI-318-71 Code.
l 1.2 If use of a ductility ratio greater than 10 (i.e., u> 1001 is required to demonstrate design adequacy of structural elements against impactive or impulsive loads, e.g., missile impact, such a usage should be identified in the plant SAR.
Information justifying the use of this relatively high ductility value shall be provided for SEB staff review.
ENCLOSURE 4 (CONT) s.--
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1.3 For beam-columns, walls, and slabs carrying axial compression loads and subject to impulsive or in.pactive loads producing flexure, the pemissible ductility ratio in flexure should be as follows:
(a) Een compression controls the design, as defined by "an interaction diagram, the permissible ductility ratio shall be 1.3.
(b) men the compression loads do not exceed 0.1fe'Ag or one-third of that which would produce balanced conditions, which-ever is smaller, the pemissible ductility ratio can be as given in Section 1.1.
l (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 structural elements resisting axial compressive impulsive er impactive loads only, without flexure, the permissible axial ductility ratio shall be 1.3.
1.5 For shear carried by concrete 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 i
2.1 For flexure compression and sheer u
= 10.0 2.2 For colunns with slenderness ratio (1/r) equal to or less than 20 u
= 1.3 ENCLOSURE 4 (CONT)
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ENCLOSURE 4 (CONT)
I 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 v =.5 where tv= unifonn 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
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" 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 i
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 OYNAMICS
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POOL SWELL a.
Subble pressure, bulk swell and froth swell loads, drag pressure and other pool swell loads should d 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 (SRY) 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 SRY loads, b.
A single active failure causing one SRY discharge must be considered in combination with the Design Basis Accident (DBA).
c.
Appropriate multiple SRY discharge should be considered in combination with the Small Break Accident (58A) and Inter-mediate Break Accident (IBA).
d.
Thermal loads due to SRV discharge should be treated as T' for nomal operation and T, for accident conditions.
e.
The suppression pool liner should be designed in accordance with the ASME Boiler and Pressure Vessel Code, Division 1 Subsection NE to resist the SRV negative pressure, considering i
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. The 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.
i 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 j
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 pemitt, with appropriate limits on ductility ratios.
ENCLOSURE 4 (CONT) t
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 determine 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 support exterior structures.
7.
The equivalent static pressure should be considered as potentially
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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
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5000 19 16 3000 21 18 i
Region III 4000 18 16 t
5000 16 14 l
These thicknesses are for protection against local effects only. Designers must establish independently the thief. ness 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.
1 ENCLOSURE 4 (CONT)
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C.6 PASSIVE ECCS FAILURES DURING LONG-TERM COOLING FOLLOWI'lG A LOCA l
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 LOCA should be con-t i
sidered. To mitigate the effects of such leaks, a leak detection system I
having design features and bases as described below should be included in the plant design.
The leak detection system should include detectors and alarms 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 I
isolation of the faulted ECCS line before the performance of more than ene subsystem is degraded. The design bases of the leak detection system should include:
l (1) Identification and justification of the maximum leak rate;
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(2) Maximum allowable time for operator action and justification therefor; (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 l
ECCS train and that the leak can be isolated; and (5) Alarms that corform with the criteria specified for the control room l
alarms and a leak detection system that conforms with the require-ments of IEEE-279, except that the single failure criterion need not be imposed.
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C.7 CONTROL ROOM POSITION INDICATION OF MANUAL (HAN0 WHEEL) VALVES Regulatory Guide 1.47 s ifies, automatic position indication of each l
bypass or deliberately uce:, inoperable condition if the following three conditions are met:
(1) The bynass or fnocerable condition affects a system that is
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designed to perfonn an automatic safety functica.
i ENCLOSURE 4 (CONT) t I
, l (2) The bypass or inoperable condition can reascnably 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) valvt nich could jeopardize the operation of the ECCS, if inavertently left in the wrong position, must have position indication in the control room.
In the PDA extension reviews it is important to confim that standard designs include this design feature.
Most standard designs do but.this matter was probably not specifically addressed in some of the first PDA reviews.
C.8 LONG-TERM RECOVERY FROM STEAM LINE BREAK - OPERATOR ACTION TO PREVENT OVERFRE55URIZATION (PWR) 4 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 low temperature could repressurize the reactor to an unacceptable pressure-temperature region thereby compromising reactor vessel integrity. Anal-yses are required to show that following a main steam line break that (1) no additional fuel failures result from the accident, and (11) the l
I pressures following the initiation of the break will not compromise the integrity of the reactor coolant pressure boundary giving due considera-l 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 punp redesign in some cases, to assure i
that long term perfomance could be met. The following kinds of infor-l mation may be sought on a case-by-case basis where such doubt arises.
a.
Describe the tests performed to demonstrate that the pumps are capable of operating for extended periods under post-LOCA conditions,
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including the effects of debris. Discuss the damage to pump seals caused by debris over an extended period of operation.
ENCLOSURE 4 (CONT) l
~8-b.
Provide detailed diagrams of all water 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 cost-LOCA environmental conditions were factored into the spect'fications and design of the pump.
C.10 GRAVITY MISSILES, VESSEL SEAL RING MISSILES INSIDE CONTAINMENT Safety related systems should be protected against loss of function due to internal missiles from sources such as those associated with pressurized components and rotating equipment. Such sources would include but not ?>e limited to retaining bolts, control rod drive assemblies, the vesse? seal
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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 orientation of the poten-tial missile sources, missile barriers, physical separation of redundant safety 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.
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ENCLOSURE 4 (CONT)
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C.11 CORE THERMAL-HYORAUf.IC ANALYSES i
In evaluating the themal-hydrau11c perfomance 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 bowing.
In addition.a connitment to perform tests to verify the transient analysis methods and codes is required.
C.12 DEGRADED GRIO 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 power systens. These additional requirements are defined in the following staff position.
1.
We require that a second level of voltage protection for the onsite power system be provided and that this Mcond level of voltage pro-taction satisfy the following requirements:
a) The selection of voltage and time set points shall be detemined 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 power source; e
e l
gucLoSURE 4 (CONT) l e
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c) The time delay selected shall be based on the following conditions:
(1)
The allowable time delay, including iargin, 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 condittorsfor operation, surveillance requirements, l
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 l
supplying power to all sequenced loads on the amergency buses.
l 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 i
of offsite power in conjunction with a safety injection actuation i
signal; and (c) simulating interruption and subsequent reconnection of onsite power sources to their respective buses.
l 4
I ENCLDSURE 4 (CONT)
sa i
f :
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 measureent, and by correlation-of measured values with analysis results.
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C.13 ASYMETRIC LOADS ON CONONENTS LOCATED WITHIN CONTAINMENT SU8 COMPARTMENTS
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A In the unlikely event of a pipe rupture imide a major component sub-compartment, the initial blowdown transfee st would lead to ressure loadings on both the structure and the enclosed component The staff's generic Category A Task Action Plan A-2 is designe( ).to develop generic resolutions for this matter. Our preser.t schedule calls for completing A-2 for PWR's during the first quarter,1979. Pending completion of A-2, the staff is implementing the following progrem:
- 1. For PWRs at the CP/PDA stage of review,'the staff requires appli-cants to cosuiit to address the safety issue as part of their appli ~ '
j cation foi an operating license.
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2.
For PWRs at the OL/FDA stage of review, the staff requires case-by-case analyses, including implementation of any, indicated corrective j
measusres prior to the issuance of an operating license.
1,.
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3.
For BWRs, for which this issue is expected to be of lesser safety
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significance, the asynsretric loading conditions will be evaluated on a case-specific basis prior to the issuance of an operating license, s
For those cases which analyses are required, we request the performance of a subcompartment, multi-node pressure response analysis of the pressure transient resulting from postulated hot-lag and cold-leg (pump suction and discharge) reactor coolart system pipe ruptures within the reactor cavity, pipt penetrations, and stari generator compartments. Provide similar analyses for the pressurizer surge and spray lines, and other h)gh energy lines kcated in containment compartments that may be subject to pressurization. Shos how the results of these analyses are used in @he design of structures and component supports.
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l ENCLOSURE 4 (CONT)
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i 12-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 information 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 deston provisions that will permit 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 S 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 penetratien fitted with expansion bellows that does not have the design capability L
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 I
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 higher 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 Tem are first quarter,1979 and fourth quarter,1978, respectively.
l 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 enveltpe.
ENCLOSURE 4 (CONT) i
13 Therefore, describe and justify the analytical model used to conservatively I
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; feedwater auxiliary feedwater, and connected systems isolation; feedwater, condensate, and auxiliary feedwater pump trip, and auxiliary feedwater run-out control system; the loss of or availability of offsite power; 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 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 (which to the partial pressure of vapor, or the atmosphere temperature may be superheated)was used.
(5) Discuss and justify the analytical model including the thennedynamic i
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 contairunent 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 shutdown; ENCLOSURE 4 (CONT) 1
(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 isolation 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 ccefficient based on the recommendations in Branch Technical Position CSB 6-1,
" Minimum Containment Pressure Model for PWR ECCS Perfomance Evaluation,"should be used.
(b) A convective heat transfer coefficient should be used when the condensing heat flux is calculated to be less than the convective heat flux. During the bicwdown 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
= empirical constants dependent on geometry and Reynolds No.
Since the Reynolds number is dependent on velocity, it is generated by the steam generaor blowdown.necessary to evaluate provide Ifmited data in this regard.
The CVTR experiments from 10 ft/sec to 30 ft/sec were measured locally. Convective currents of obtain forced flow currents to detemine the convective We recomend transfer coefficient during the blowdown period.
blowdown has ceased or been reduced to a negligibly low value After the 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 themal response of the component with the accident thermal analysis of the component.
qualification test program. component themai respon:e was dev For instance graphically show the of attachmant, and perfomance characteristicsthermocouple da detailed discussion of the analytical model use,d to evaluate the or provide a component themal response during the test.
be performed for the potential points of failure such as thinThis evaluation should cross-sections and temperature sensitive parts where themal stressing, temperature-related degradation, steam or chemical interaction at elevated temperatures failure of the compone,nt mechanically or electrically.or other therma component themal response comparison results in the prediction of If the a more severe themal transient for the accident conditions than for the qualification test, provide justification th&t the affected component will perform its intended function during a MSI.B accident, or provide protection for the component whch would appropriately limit the themai effects.
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e C.16 ENVIRONMENTAL EFFECT OF PIPE FAILURES Identify the " break exclusion" regions of the main steam and fhedwater lines. Compartments that contain break exclusion regions of main steam and feedwater if nes and any safety related equipment in these compartmen?.s should be designed to with-stand the environmental effects (pressure, temperature, humidity and flooding) of a crack with a break area eoual to the cross sectional area of the' break 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 = i+
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corrective action by an operator.
We have established 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 norinal operation, anticipated transients, and accidents.
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- 1. A single active failure in the component cooling water system shall not result in fuel damage or a breach of the reactor I
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 deterinined in accordance with Branch Technical Position ASB 3-1.
In order to meet the criteria established above, an NSSS inter-face 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)
c.
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 minutes without loss of function or the need for operator pro-tactive action. In addition, safety grade instrianentation including alanns should 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 alarms, 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, then the design of the CCW sys tem must meet the t.
. 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 Category I, Quality Group D and ASME Section III, Class 3 requirements.
The reactor coolant (RC) pumps and motor's are within the NSSS scope of design. Therefore, in order to demonstrate that an RC pump design can operate with loss of component cooling water for at least 30 minutes witnout loss of function or the need for operator action, the following must be provided:
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 coolir.g water does not result in a LOCA due to seal failure.
ENCLOSURE 4 (CONT)
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18-2.
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 description 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 att 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.
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d.
A description of the cooling and lubricating systems (with appropriate figures) associated with the RC pump and motor and their design criteria 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 i
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)
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For CP/PDA and OL/FDA applications, provide the following for steem generators utilizing top feed:
1.
Srevent 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 possibla (less than seven feet) horizontal run of inlet piping to the steam generator feedring; and 3.
Perfom tests acceptable to the staff to 'terify that unacceptable feed-water hammer will not occur using the plant operating procedures for normal and emergency restoration of steam generator water level following loss of normal. feedwater and rssible 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 normal operating occurrences of transients that-could cause the water level in the steam generator to dron 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 either uncovering 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.
1 ENCLOSURE 4 (CONT)
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I C.19 ENVIRONMENTAL CONTROL SYSTEMS FOR SAFETY RELATED E0t!!PMENT Most plant areas that contain safety related equipment depend on the continuous operation of environmental controi systems to maintain the environment in those areas within the range of environmental qualification f
of the safety 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
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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 sourcesi and (4) provide a continuous record of the environmental parameters during the time the environmental conditions exceed the nomal limits.
ENCLOSURE 4 (CONT) l
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